Category Archives: Modern Power Station Practice

Handling restrictions and criticality assessments

Decay heating In practice, the timing of the various irradiated fuel movements and related handling op­erations is rigidly controlled by post-discharge heating considerations. During their residence in the reac­tor, AGR fuel stringers generate powers ranging from about 6.5 MW to typically 4 MW at end of life. These are figures which should be compared with peak mag — nox channel powers of perhaps 270 kW. Coupled with the high irradiation achieved at discharge, the increased ratings experienced by AGR fuel gives rise to very high heating levels following its discharge from the reactor. Referred to as decay heat, this is the energy released by the continued radioactive decay of fission products and heavy isotopes.

There are many reasons why the safe handling of irradiated AGR fuel needs to be linked to decay heating. The implications of stringer or element over­heating arising from various postulated fault condi­tions has been rigorously studied and, together with a general requirement to restrict fuel tempermures at other stages during handling, has led to the s.;ict ap­plication of maximum heating limits to the individual operations. Of all the likely fault conditions which are thought could arise in the fuel route, the most severe is that of a dropped irradiated fuel stringer in which gross damage to some of the tuel elements could be compounded by fuel can melting and the consequent release of radioactive fission products. Al­though such an occurrence, either in the reactor or at any of the fuel route facilities, is considered to be extremely unlikely, it is nevertheless necessary from a safety viewpoint to be able to guarantee that, even if it were to happen, an unacceptable release of radioactivity to the environment would not take place.

In this context the withdrawal of an irradiated stringer is forbidden even from a shut down reactor unless the total stringer decay heating is less than predetermined limits. The absolute values depend upon the available cooling capacity provided by the charge machine during the discharge process. The most com­mon off-load refuelling regime adopted involves com­pliance with a maximum stringer heating of around 40 kW, which in practice results in post reactor shutdown delays of up to 12 or 15 hours before refuelling can begin. A similarly derived limit is de­signed to protect against the consequences of a dropped stringer within a buffer storage decay tube. Other potentially serious implications of overheated fuel have been considered. Within the irradiated fuel dismantling (IFD) cell, for example, dismantling is not permitted unless the total stringer decay heating is less than 40 kW; this ensures that fuel temperatures do not reach levels which would significantly enhance the oxidation rate of any exposed UO2, at possible pin failure sites, to U3O8. UjOg would be produced as a fine powder, therefore constituting a ready source of cell contamination.

Routine estimation of AGR fuel decay heating is a highly complex matter. Since the very high fuel ratings and irradiations produce considerable levels of post-discharge heating, it is important that changes in stringer conditions during service are properly ac­counted for in the calculational process, so that heating levels are neither under nor over estimated. Therefore sophisticated techniques have been developed which allow the retrospective examination of ratings seen over the entire operating histories of individual stringers, prior to their removal from the reactor. By this means, recommended minimum cooling periods can be provided in order to ensure that the various heating limits are not exceeded in practice.

Criticality considerations In discussing the accidental criticality of AGR fuel, reference was made to pes­simisms included within the criticality safety assess­ments and also to allowances made for the possibility of flooding occurring under freak accident conditions. The latter holds special significance for the handling of irradiated fuel since some of the facilities involved are associated, directly or indirectly, with the pre­sence of large quantities of water. In the case of decay tube storage, for example, studies have assumed that the dropped stringer accident, referred to in the pre­vious section, causes a breach of the cooling water jackets with consequential flooding of the fuel and debris, and yet even if this were to happen, the con­clusions are that safety from criticality would be as­sured. No criticality hazard is foreseen in the IFD either, mainly because of the small amount of fuel involved. Provision has been made for deliberately flooding the cell with an emergency supply of boro — nated pondwater in order to provide cooling for the fuel following a postulated dropped stringer accident.

The cooling ponds are of special importance and criticality safety assessments have considered accident situations, such as overturned skips and fuel element oser-stacking. It has been demonstrated that under normal circumstances in which skips are properly loaded with fuel and stored in the pond bays only in the approved manner, criticality cannot be achieved, even if pure water were used. However, in order to provide further safety margins against skip accidents which could conceivably lead to damaged fuel be­coming reassembled in a more reactive array, the pondwater is dosed with a soluble neutron poison — boron, in the form of boric acid. The skip insert, which is welded to the main skip body, is made of boron steel (see Fig 3.55) and, together with good working practices which have evolved from the criticality assess­ment work, therefore provides for additional safety margins.

Solid wastes

Three basic types of solid waste arise at nuclear power stations:

• Low? level radioactivity waste in the form of paper, fabrics, plastics glass and metals.

• Intermediate level waste as ion-exchange resins, sludges and metals.

• Intermediate level waste as fuel cladding components.

The principles of solid waste treatment are to con­centrate and contain the radioactivity by incineration of combustible material and compaction or immo­bilisation of non-combustible waste,

l ow level waste, after compaction or incineration, is disposed of at licensed sites by burial in shallow’ trenches covered with about 1 metre of soil. Typical annual disposals by this route for magnox and AGR are given by Heap and Short [8] and are shown in Table 4.8. PWR waste estimates of this type are given by Passant (91, and are shown in Table 4.9. The radio­nuclides present in the waste are a mixture of activa­tion and fission products, with cobait-60 and caesium-
137 prominent. Ion-exchange resins and sludges from water treatment are stored as slurries in steel or con­crete tanks located in concrete cells, typically 1 metre thick, which act both as radiation shields and second­ary containment should a tank begin to leak. Storage is temporary — until the waste is processed for dis­posal. In the case of steel tanks, a corrosion allowance is made in the choice of metal thickness and if mild steel is used a protective coating is applied. Concrete tanks have a lining of steel or asphaltic paint. Provision is made for chemical dosing of the slurries, if neces­sary, to minimise corrosion of the tanks. All storage facilities have leakage detection arrangements to give early warning of failure of the primary tank.

A limited disposal of sludges to sea in sealed packages has taken place. The area selected for sea disposal is subject to international agreement and is approxi­mately 600 miles from the south-eastern point of England. The packages consist of steel drums contain­ing the sludge which has been immobilised in cement. Similar packages have been designed for ion-exchange resins using a polymer matrix in place of cement.

Metallic wastes, for example valves, pipes and re­actor control machinery, are stored in concrete cells at magnox and AGR power stations. These cells have thick walls which act as radiation shields in addition to providing containment. Like the resin and sludge cells, the walls are typically 1 m thick. The design of the facilities is such as to ensure that the waste is kept dry with provision for the checking of adventi­tious water ingress and removal.

Typical arisings of magnox and AGR sludges and resins are given by Bennett D (1983) and are shown in Table 4.10. PWR resins and sludges are given by Passant [9] and shown in Table 4.11. Fuel cladding wastes arise as the result of the removal of extraneous metal cladding and supports from the fuel before it is dispatched to a reprocessing facility. Removal of this material assists in the packing of the fuel in the transport containers.

Magnox fuel cladding (magnesium/aluminium alloy) is pyrophoric in the finely divided state and requires

TBI. E 4.10

Typical annua! ansings of sludges and resins ar magnox and AGR power smltons




: :iU і: TBq

!0 tir ‘ TBq


C m ‘ *

— 1 — !1V *

* Ааіч(> per m1 procjbk similar іо magnox

TBl E 4.11

Typical annual ansings of sludges and resins at PWR power stations



7 nr-0.3 TBq

23 mV 100 TBq

special safety precautions. Two types of storage cell have been adopted. At some power stations the waste is stored under chemically-dosed water and at other sites is stored dry. The cells are of concrete to pro­vide containment and radiation shielding. In the wet cells, forced ventilation is provided to prevent hydro­gen build-up from the air/water reactive magnox, the air being filtered before discharge. In the dry cells, temperature probes are installed to warn of pos­sible magnox overheating, the probes being coupled to alarms. Forced ventilation is again provided to remove any hydrogen.

AGR power station fuel yields steel and graphite debris, which is accumulated in dry concrete cells, providing containment and radiation shielding. This waste is not reactive and unlike magnox waste special. precautions against fire are not necessary.

Dry solid wastes have isotopic compositions related largely to neutron activation products with the radio­nuclides Co-60, Mn-54 and Fe-59 being prominent. The isotopic composition before disposal is time dependent in view of the long storage times. Ion exchange resins contain predominantly Cs-137 and sludges are often a mixture of activation products, fission products and actinides.

National policies on disposal of radioactive wastes are under review. Environmental concern over the dumping of solid radioactive waste at sea led to in­dustrial action by the transport unions in 1983. This followed a resolution calling for a voluntary suspen­sion ot sea dumping at the 1983 meeting of the London Dumping Convention (LDC), pending the comple­tion ot an LDC scientific review. As a result the government suspended sea dumping and agreed to establish an independent review of the subject in conjunction with the Trades Union Council (TUC). This review by Holliday [10] was published in late 1984 and had the principal recommendation that the

dumping of radioactive waste at the North East At­lantic dump site should not be resumed until the cur­rent international reviews, and a comparison of sea dumping with land-based alternatives, had been com­pleted. These international reviews, due for completion in 1985, are the Ad-hoc Scientific Review — tor the LDC, the Nuclear Energy Agency (NEA) Site Suita­bility review under the Organisation of Economic Co­operation and Development (OECD) and the Review of the International Atomic Energy Agency (IAEA) Definition and Recommendations for the LDC.

A Radioactive Waste Management Advisory Com­mittee (RWMAC), established by the government, is also examining issues relating to an overall policy for the management of radioactive wastes and its fifth report to government was published in 1984. RWMAC monitors the activities of the Nuclear Industry Radio­active Waste Executive (NIREX), an organisation set up in 1982 by the principal organisations that pro­duce radioactive waste to manage the disposal of most solid low level and intermediate level radioactive waste. The activities of NIREX include the identification of land sites potentially suitable for the disposal of low and intermediate level wastes and managing the sub­sequent work.

Reactivity faults

Reactivity faults occur if there is an uncontrolled in­crease in the reactivity of the reactor either over the entire core, or locally. Such an increase inevitably gives power and hence temperature transients which must be limited by the protective equipment to less than the melting point of magnox. Indeed, because the strength of magnox decreases substantially at a few degrees centigrade less than 640°C and because full coolant mass flow conditions usually exist, the acceptable maximum temperature is reduced to avoid distortion of the fins with a consequential reduction of heat transfer coefficient.

The reactivity increases because the control rods, either as groups of rods or locally as a few rods, run out in an uncontrolled fashion. The reactivity release rate depends upon the worth of the rods running out, and their speed of withdrawal. Both of these aspects are limited by design, by restricting the size of rod groups and the speed of withdrawal and by using in­terlocks to prevent the main rod groups being moved together. There are two types of reactivity faults to be considered, those where there is a uniform increase of reactivity over the core, symmetric faults, and those where only a few rods run out, asymmetric faults. These need to be considered separately using slightly different techniques because the protection for each 4s organised in a different way. In each case, however, it is necessary to demonstrate that the probability of exceeding the limiting clad temperature is acceptably low. As in the case of the depressurisation faults, this is done by considering the magnitude of the uncertainties in the manufacture of the fuel, the heat transfer properties and the model used in the study. For those faults where the gas circuit remains intact, the acceptable probabilities that any fuel temperature will reach the limit in the event of the fault occurring is 1 in 1000.

Fuel transport

Transport requirements in respect of nuclear fuel may be summarised as:

• Delivery of new fuel to power stations and return to BNFL if it fails to pass final inspection, e. g., due to damage during handling.

• Return of irradiated fuel for reprocessing at BNFL.

• Consignment of small quantities of irradiated fuel for post-irradiation examination (PIE) at labora­tories.

• Consignment of PIE debris from laboratories to the fuel reprocessing plant.

• Return of discharged fuel flasks to power stations for further use.

The new fuel, which is only slightly radioactive, is transported by road in strong industrial containers (magnox fuel) or in Type A containers (AGR fuel), being treated with care because uranium is a chemical­ly toxic material and because even slightly damaged fuel is unsuitable for reactor use and costly to replace.

In the case of new AGR fuel, the stringent require­ments of criticality control necessitate that the amount of uranium in each package, and the number of pack­ages per vehicle are restricted. The criticality aspects of the packaging are subject to competent authority approval.

Flasks being returned after the removal of their load of irradiated fuel cannot be described as empty as they retain some residual radioactivity, possibly above the Type A limits. They are therefore referred to as ‘discharged flasks’, and are moved as Type В packages with similar pre-despatch precautions as those applied to loaded flasks.

Emergency duties and responsibilities

Overall command of the site emergency organisation is vested in the Emergency Controller who will be either the station manager or a nominated deputy. However, it is probable that the first indication of an accident will be received by the shift charge engineer (or shift manager) who will act as emergency control­ler until relieved by a more senior member of staff. The emergency controller is responsible for the coordination of on site actions, all aspects of plant recovery and immediate post-accident off-site activities aimed at protecting the public. The emergency con­troller is supported by a team of specialists, available on a rota basis, who will give advice and assistance on the health physics, reactor physics, operational and administrative aspects of the emergency situation. The emergency controller is responsible for:

• The assessment of any potentially hazardous situa­tion on the site.

• The issuing of appropriate warnings and notifi­cations both on the site and to relevant off-site organisations.

• The mobilisation of personnel and equipment to deal with the situation. Actions necessary to control the situation on the site.

• Radiological surveys both on, and initially off, the site.

• The assessment of any hazard to members of the public and the issue of appropriate advice for their protection.

• The initial provision of information for the con­trol of drinking water, milk and other agricultural products.

• The provision of information for statements to the public and the news media.

• The compilation of a record of events for future study.

On the declaration of a site incident or emergency condition all personnel will assemble at pre-arranged
muster points for roll-call and will then disperse to their emergency posts. An emergency control centre will be established at a suitable designated location, e. g., the conference room in the administration building. Operations staff will safeguard running plant and will try to minimise the effects of the accident. If the emergency declaration is made by the shift charge engineer, the main control room will serve as the emergency control centre until the duty emergency controller assumes his duties.

In the event of an emergency standby or alert, an operational support centre (see Section 6.2 of this chapter) will be established at a designated location some 5 to 20 miles from the site. Responsibility for the coordination and control of off-site activities, and for liaison with external individuals and organisations having responsibilities in an emergency, will be as­sumed by a senior manager at the operational support centre once that centre has been set up.

New fuel

The magnox fuel element is essentially a cylindrical bar of natural uranium encased in a magnesium al­loy can. Both the uranium and the can material have additional trace elements to provide the required met­allurgical properties. The bar is about 25 mm diameter by 750 mm in length (the Berkeley and Hunterston elements are about half this length) and a weight of 10 11 kg. It is ‘grooved’ and during manufacture the can is pressurised down into the bar. This procedure is employed to provide intimate bar-to-can contact and to present ratcheting between the bar and can, due to thermal changes during operation. The can is dosed by end caps which are screwed and seal welded into the ends after the bar has been inserted. A ther­mal disc of sintered alumina is positioned between the bar and end cap to reduce the transfer of heat to the end fittings. The latter are screwed into the can to provide a cup at the top end of the element and a cone at its lower end, thus providing a suitable means ot stacking the elements in a channel. The top end
fitting is designed to accept the jaws of a grab used for handling procedures. In most designs this fitting is provided with a link which is sprung, allowing it to press against the channel wall thus reducing the tendency of the element to rattle or rotate within the channel. The method of stacking the fuel in the Berkeley reactors differs from the Hunterston A reactors in that instead of the lower dements sup­porting those above, each element is individually sup­ported. The Berkeley element is contained in a struc­ture made up of graphite struts and steel bridges so that the weight of element above is taken by the struts. The Hunterston A element is contained in a graphite sleeve which supports the weight. This design has the additional advantage that the graphite ad­jacent to the fuel in the channel (which experiences greater radiation damage) is replaced when the fuel is exchanged.

There are two types of cans, polyzonal or helical finning and herringbone finning (Fig 3,42). The poly­zonal can is produced by an extrusion process and the helix formed by twisting the can which is stif­fened by a splitter and brace assembly or cage. The herringbone can is also an extrusion followed by ma­chining of the tugs and fins. The polyzonal can has to some extent been superseded by the herringbone design since the latter has greater resistance to fin waving during operation. With increased dwell times in the reactor, the thermal changes cause the finning to assume a wavy characteristic (multiple longitudinal bowing) with a permanent set, some fins touching and

IriG. 3.42 Comparison of the polyzonal (helical) and herringbone fuel cans

resulting in a reduced heat transfer efficiency. Since the herringbone fin length is less than that of the polyzonal element, the fin waving tendency is reduced. It is usual for reactors to be wholly charged with either polyzonal or herringbone fuel. The exception is Oldbury where a mixed channel is employed with three polyzonal elements occupying the bottom chan­nel positions with herringbone elements on top of them. It will be noted that the helix of adjacent quad­rants of the herringbone element are such that rota­tional forces resulting from the flow of the coolant gas cancel out. However, it does mean that radially adjacent lugs are at differing temperatures.


Many types of manipulators have been developed by the CEGB and there has been increasing use made of computers for control and display purposes. The ‘snake’ is a_multi-link manipulator in which the in­dividual units are motorised to enable the whole to be manoeuvred through a tortuous route. Computer control is employed to guide each link through the same path as the lead link and to reverse the process on withdrawal. Prompt programs are used extensively to assist the operator to recall critical sequences of operations and to provide data and work procedures.

Main provisions

Four primarv types of package (i. e., container or pack­aging, together with contents) are defined: Excepted,

Industrial, Type A and Type u. In a graded approach the design requirements and, where appropriate, stand­ard tests and performance criteria are specified. They become more stringent as the hazard represented by the contents increases. Thus there are no specific per­formance standards for the lowest group excepted packages, whereas Type A packages are intended to withstand normal transport conditions and Type В packages to survive very severe accidents.

For Type A packages the contents are restricted so that in the event of an accident which damaged the container, the resulting radiation dose to a person at the scene of the accident would be unlikely to ex­ceed the appropriate annual dose limit for a radiation worker.

In general, application of standard dosimetric mod­els to a given radionuclide results in two limits for the amount of that radionuclide in a Type A package. The first limit (Ai) is based on the external radiation exposure due to loss of shielding, and if the radio­nuclide was non-dispersible or ‘special form’, e. g., a sealed source, then this limit would apply. The second limit is based on release of some of the radioactivity in a dispersible form, and the resulting radiation dose by inhalation, ingestion, skin contamination and other routes.

The lower of the two limits is known as the Ai value and it represents the contents limits for a Type A package with contents in dispersible form.

A2 values are tabulated for most radionuclides and they represent a common basis of hazard potential. There are rules for determining the effective A2 value for mixtures of nuclides, and the general rules for package content and leakage limits are expressed in multiples of sub-multiples of A2 values.

The contents of excepted and industrial packaging are limited such that they would present no greater hazard in the event of an accident, than the Type A package. The activity allowed in an excepted package is severely limited in recognition of the fact that this package is not expected to retain its contents after an accident. The contents of the three grades of industrial package are restricted to low specific activity (LSA) materials or to surface contaminated objects (SCO), whose potential hazard is limited by the re­latively low toxicity of the inaccessible nature of their radioactivity content.

For Type В packages no direct limit on radioactive content is specified. However, a series of tests are prescribed which simulate the damage which could be sustained in a very severe accident. These include a drop test from 9 metres onto an unyielding target, and a punch test in which the package is dropped from 1 metre onto a rigid bar. These tests are carried out in the order and attitudes which would maximise the consequences of a subsequent thermal test which involves the complete engulfment of the package for half an hour in a hydrocarbon fuel/air fire with an average flame temperature of at least 800°C. A maxi­
mum leakage rate of radioactivity following these tests is specified, and also a maximum external radiation dose rate. These requirements, together with those for normal transport and consideration of heat dis­sipation requirements, limit indirectly the radioactive contents of Type В packages.

In specifying design requirements for Type В pack­ages, the Regulations prescribe standard conditions of ambient temperature and solar isolation. However, relaxation from these conditions is allowed, provided that the use of the package is appropriately restricted to wnthin a specific country or between specified coun­tries. Packages which are approved on this basis are designated as Type В (M), as distinct from Type В (U) packages which meet the worldwide standards.

For the purpose of controlling the radiation dose to transport workers and members of the public under normal transport conditions, appropriate segregation of packages is required. A limiting annual dose for individuals in each group is indicated as a guide to the minimum requirements for segregation distances, and to allowable dose rates in regularly occupied areas.

Segregation of packages from undeveloped photo­graphic film is also a requirement.

To facilitate control, packages are categorised and labelled on the basis of the radiation’dose rates at the surface and of the package and at 1 metre from the surface (see Table 4.12). The latter dose rate (in mSv/h x 100) is known as the ‘transport index’. This, infor­mation is given on the package label. Loading controls include limits on the sum of the transport indices of packages carried in a single conveyance. Limits are also placed on the radiation levels at the external surfaces of transport vehicles, and at 2 metres from the surface.

Limits are also specified for non-fixed contamina­tion on the external surfaces of packages. For packages
other than ‘Excepted packages’ these limits are 4 Bq/cm2 (10 ~4 Сі/cm2) of beta/gamma emitters and low toxicity alpha emitters, and 0.4 Bq/cm2 (10 ~5 Сі/ cm2) for all other alpha emitters. The corresponding limits for Excepted packages are a factor of ten lower. In the case of packages containing fissile material, the hazard of nuclear criticality has to be considered in addition to the basic radiation and contamination control aspects. In order to ensure that the risk of criticality under normal and accident conditions is made practically non-existent, appropriate design re­quirements and tests are described for fissile packages.

Interaction between packages is allowed for by de­fining a nuclear criticality transport index, and by limiting the sum of the transport indices of packages transported together. The transport index is always the higher of the dose rate and criticality transport indices.

The Lungs

The lungs are important as inhalation represents one of the easiest ways for radioactive material to enter the body. The purpose of the lungs is to provide an efficient system for exchanging carbon dioxide for oxygen in the red cells of the bloodstream. This takes place in the deep lung through the walls of alveoli which are tiny cavities formed into compact groups so that a very large surface area is enclosed in a small volume. The total surface area of the lungs is about the same as that of a tennis court!

Particles inhaled and deposited in the lungs will have a residence time dependent upon the region in which they are deposited and also on their size. The larger the particles and the higher in the respira­tory tract they fall, the greater the probability that they will be rapidly removed by the action of hairlike structures (cilia) which carry mucous, secreted by cells in the lung, to a point where it can be swallowed. Small particles which find their way into the deep lung, where cilia do not exist, have long residence times (hundreds of days), although many such parti­cles are engulfed by white blood cells and then re­moved by coughing or via the lymphatic system.

The Skin

The cells of the surface layer of the skin are con­tinuously lost by abrasion. The parent cells below the surface are sensitive to radiation damage because of their high rate of division. The lower layers of skin comprise connective tissue with a good blood supply and nerve network. The outer layers are composed of several layers of cells. In the basal layer are found nucleated cells capable of division. As they grow up to replace ceils removed from the outermost layer they lose their typical ‘round’ shape and nucleus and become simply ‘dead’, flat cells whose job it is to form a protective covering over their parent cells. The first effect of irradiation of the skin, at about 4 Sv, is reddening or erythema. If a large number of basal cells are killed an inflamed wound ultimately results which will be very slow to heal. The dose required for this is about Ю Sv.

The term radiation burn is sometimes used to draw a parallel with thermal burns. The time course in the two cases is, however, very different and whereas a thermal blister appears within a few minutes the ana­logous radiation damage may take more than Ю days to evolve.

Land gamma radiation measurements

The area round the station may be divided into an Inner Zone and an Outer Zone. The inner zone is an annulus with the station at its centre and of radii from 1 to 2 km, while the outer zone is an annulus concentric with the first, of radii from 2 to 10 km. There is an additional Emergency Zone which com­prises an annulus of radii from 10 to 15 km. These zones are illustrated in Fig 4.10.

If this whole area around the station is further divided into quadrants, then there are three fixed



Fig. 4.10 Inner and outer zones for land gamma radiation measurements

measurement sites per quadrant in the land sector of each zone so far as is geographically possible. Currently, routine measurements are made at each site in the inner and outer zones every three months, and at each emergency site every twelve months.