The pond is designed as a water-retaining structure and as a means to provide shielding against the ionising radiation from stored fuel elements. These considera­tions necessitate typical pond external wall thicknesses of 1.0 m of concrete although walls of up to 1.7 m thickness are required where there are working areas immediately adjacent to the external walls.

The ponds are designed to comply with the relevant Code of Practice for water retaining structures [37]. The pond walls and floor are painted with a phenolic resin paint. Water bars are incorporated in all con­struction joints and each joint in the pond base is provided with drainage channels leading to leakage tell-tale sumps. Facilities are provided for pumping out any seepage into the sumps. By these means, accidental loss of radioactivity to the environment is guarded against, and the operators provided with early information on any defects in the pond structure.

In order to avoid undesirable structural stresses it is recommended that the rate of change of pond temperature is kept to a minimum.

• Advice and warning to farmers whose land and livestock may be affected.

• The control of the distribution of milk and agri­cultural and horticultural products.

• The collection of samples of milk and other foodstuffs.

• The provision of alternative food supplies.

• Liaison with health physicists from the CEGB, NRPB and the Department of the Environment.

r Department of the Environment representative

• Advice to local water undertakings on the control of drinking water supplies.

• Liaison with the water authority representative.

• Arrangements for the collection and analysis of water samples.

• Arrangements for the disposal of radioactive waste arising from the accident.

• Liaison with health physicists from the CEGB, NRPB and MAFF.

Before discussing the more detailed physics aspects it would be prudent at this stage to look at the aims of the fuel cycle itself, since there are several of them and not all are compatible. Quite simply, the overall objective must be to ensure that both the capital and the continued running costs of the reactor are kept as low as possible. At the same time, the reactor must provide safe and dependable operation at maxi­mum power throughout its working life. It is worth emphasising that the properly optimised fuel cycle, which encompasses choice of enrichment as well as the formulation of a detailed refuelling strategy, needs to take into account fuel handling machinery and reactor safety as well as the more obvious needs ot reactor optimisation and control.

Throughout reactor life, from first power raising and subsequent refuelling of the initial core (І. Є., the approach to equilibrium) and through the various replacement charges thereafter, it is of prime impor­tance that acceptable axial and radial power distri­butions should be maintained. That is to say the shape of the power distribution should be as flat as possible, both along the fuel stack and across the core in all directions, so that maximum power can be obtained from as much of the fuel as possible. This is assisted in practice by using two or three different fuel enrichments, as will be explained later, but choice of enrichment level must also reflect the need to pro­vide sufficient reactivity at the fuel irradiation dis­charge limit chosen and must ensure adequate shut­down margins throughout reactor life. Therefore the design and operation of a suitable control rod system, w’hich is capable of providing adequate control and shutdown capacity, forms an integral part of the over­all task. Under normal circumstances the fuel enrich­ments for an AGR system are chosen to match a particular discharge irradiation limit, the higher the enrichment, the higher the achievable irradiation.

The complete fuel cycle study examines the highly complex interplay between these issues so that, within the framework of the various constraints on reactor operation, balanced judgements can be made in re­spect of choice of enrichments and overall refuelling strategy (i. e., refuelling rate and sequencing) which will ultimately enable a uniform distribution of re­activity to be maintained throughout life.

The monitoring of oxidation takes the form of com­ponent inspection, the acquisition of material speci-

Fio. 3.71 Prediction of oxide thickness

mens, and the measurement of oxide thickness on components within the reactor or on representative components in autoclaves. (An autoclave is an oven filled with pressurised CO] and operated to simulate reactor conditions.) The reactor and boiler inspections are aimed at confirming the integrity of components by visual, photographic, ultrasonic and television tech­niques. The results of the oxidation assessments are used to direct attention to the areas and components of greatest risk. The use of material samples, which have been oxidised in the reactor environment, allows correlation with laboratory data on the variables which effect time to breakaway and its rate of oxidation. This data is collected for all reactors, to determine the dependency of oxidation rate on such variables as water or hydrogen concentration, operating tempera­ture and materia! composition. From the collected data, predictive equations can be formulated enabling oxide thicknesses for future operation to be estimated. In addition, the continuity of thermocouples and BCD extrapolation pipework, refuelling checks and control rod freedom of movement are all used to monitor for the effects of oxidation during normal operation.

It is expected that any serious component failure in the charge pan or core areas would affect some or all of these systems.

To control oxidation, de-rating of the CEGB re­actors was carried out in 1968 by limiting the gas out­let temperatures to 360°C. On the grounds of safety alone there is no real problem since assessments are for one year predictions only. A new base-line can be established each year at the annual inspection. The condition of the inspected reactor is taken as indica­tive of the running reactor.

The branch is responsible for reviewing the safety and gaining acceptance by the regulatory authorities of the safety of operating plant, including any modifications made to it. The branch comprises Assessment, Inspec­tion and Health Physics Sections, with the latter two aiso providing a service to the Nuclear Safety Develop­ment Branch.

Independent assessment of the nuclear safety of the CEGB’s reactors is a major part of the department’s work. It starts from the first consideration of a nuclear project, covers the design stages and continues throughout the working life of the plant. The magnox and AGR assessment section is responsible for the assessment of the CEGB’s operating gas-cooled reac­tors, A major part of this work is the assessment of the pressure circuit integrity. An example on the mag­nox reactors was the assessment of the safety case for the gas duct bellows defects, discovered at three of the earlier nuclear stations. Another example is the CO2 oxidation problem and the remedial action taken to control the effects. Having considered and agreed the safety argument for a particular problem with either the designers or the operators, or both, the Depart­ment will then take the lead in presenting the CEGB’s case to the Nil, supported as necessary by the location representatives.

Safety assessment of nuclear reactors covers a very wide area and in order to give both designers and assessors some guidance on the safety targets to be
achieved, the department has produced CEGB Design Safety Criteria which give targets to be met by the design.

Further information is given on design safety cri­teria later in this chapter.

The Health Physics Section is responsible for the independent consideration of the radiological safety of all health physics aspects of the CEGB’s opera­tions and specifically looks at radioactive waste aris — ings, accumulation and discharge of radioactive waste, transport of nuclear fuel, and radiological protection and control. It is responsible for obtaining the neces­sary authorisation for discharge from the authorising ministries and for obtaining the relevant approvals for the transport of irradiated fuel.

The inspection section has inspectors at each nu­clear licensed site who have full and free access to all parts of the site subject to normal safety limita­tions, and to all personnel on the site. They carry out independent, unbiased, and informed scrutiny of all nuclear safety related activities on the site and provide a positive link between the station and headquarters. Each site inspector has a duty to advise the station manager and his staff on any matter affecting the nuclear safety of the plant, and to report back as necessary through a formal reporting system to head­quarters. In particular, the site inspector has a duty to advise and inform the station staff of any pending implications on regulatory matters and the implemen­tation of their responsibilities under the site licence and associated regulatory requirements. The inspectors carry out their duties as far as possible in a suppor­tive and co-operative manner and invariably receive the stations’ support in fulfilling their role.

The results of various epidemiological studies have been incorporated into a rationalised approach pub­lished by the International Commission on Radiolo­gical Protection as ICRP Publication 26 [6]. In this the Commission concludes that the mortality risk factor for radiation-induced cancers is about 102 per sievert for uniform whole-body exposure and on this basis the annual acceptable whole-body dose is 50 mSv. For non-uniform exposure the total risk to the organs irradiated should not exceed the whole-body risk.

Paradoxically, despite this being a rationalised ap­proach, it would be possible, using risk calculations, for a single irradiated organ to receive a relatively large annual dose. For example, the thyroid of a ra­diation worker could receive a dose of 1.7 Sv/year, but at such doses it is thought that non-stochastic effects would become important so a blanket of 0.5 Sv. for all organs except the eye (0.15 Sv) is suggested.

ICRP’s criterion for occupational exposure is that the average risk to those employed in radiation work should be no greater than the average risk to those employed in a moderately safe ‘conventional’ industry. By considering average risk, while setting a maximum annual dose equivalent limit of 50 mSv, ICRP are able _to assume an annual average dose equivalent received of orie-tenth of the limit (5 mSv). The best estimate of total mortality risk from all stochastic effects is taken to be about 1 in 700 per mSv re­ceived per year — of exposure. So that the risk to a radiation worker receiving the average annual expo­sure of 5 mSv would be about 1 in 14 000, which is of the same order of magnitude as the average risk in a moderately safe ‘conventional’ industry, taken to be less than 1 in 10 000 per annum. The average CEGB nuclear power worker in fact receives about

1.5 mSv per annum, equivalent to a annual risk of about 1 in 30 000.

It is salutary to compare this with other risks we run in coming to work each day:

Combination of the conventional accident and radia­tion risks gives an overall annual risk of death for power station work of 1 in 12 000 which is within the 1 in 10 000 figure for safe industries. This com­bined risk is still well below those which we volun­tarily accept as part of our everyday lives.

A major design intent of all nuclear power stations is that the operator should not be faced with the need to carry out urgent actions or to make rapid decisions in order to prevent a nuclear incident. The term operator in this context may be the station staff cor­porately, a station department or an individual opera­tor who may or may not be in the central control room.

On the latest stations, the safety case makes no demand on the operator beyond the monitoring of the state of plant for 30 minutes following the onset of any fault. He may, however, within this time, need to take some simple actions to prevent economic damage to plant. On the earliest stations, operator action of a straightforward nature is very occasion­ally called for within 15 to 20 minutes, to prevent a hazardous situation arising. For slowly-developing faults, operator action may be claimed as protection against the possibility of a release on a time scale of many hours.

It is essential that during normal operation and under fault conditions the operator is kept aware of any difficulties or deficiencies in the plant and to this end a comprehensive alarm system is provided. The alarms are divided into two groups, urgent, which need immediate attention and, non-urgent, which may be accepted and the fault dealt with when convenient. The AGR stations have equipment which automati­cally takes care of all operations required immediate­ly post-trip to ensure that the reactor is adequately cooled even if the grid supply fails or there is dam­age to plant. A display is provided which enables the operator to follow the post-trip sequence of plant initiation and operation, and he is only required to intervene in the very unlikely event that adequate cooling is not established. Although this intervention is not claimed in the safety case the option exists.

The operator can initiate a reactor trip if the nor­mal protection equipment fails, or if he perceives a situation developing which might lead to a hazardous condition. This operator action is not claimed in the safety case except for very slowly developing faults but is a prudent option.

Operator action is required in the long term fol­lowing an accident to alleviate the consequences and to restore the situation to as near normal as possible. In the event of a release of activity, he is required to seek the advice and assistance of organisations outside the station and to monitor the extent of any release. He is expected to take appropriate action as best he may to deal with any on-site emergency, such as a major fire, until outside help arrives.

The reliance on the automatic protection equip­ment is only valid if the station is operated within the limits assumed by the safety case. These limits which also include the trip settings of the safety cir­cuits and the availability of essential plant required post-trip, are clearly defined in the Operating Rules and Identified Operating Instructions. It is mandatory upon the operator to ensure that these rules and instructions are followed at all times.

At 10 years after shutdown, the total activity will be dominated by the 2.7 year half-life, X-ray emitting isotope iron-55 arising from activation of reactor in-core steel components. This will decay and within a few years the total activity, which is important to waste disposal, will be due to carbon-14 in the graphite cores of gas-cooled reactors and nickel-63, half-life 92 years, present as a trace element in carbon steel and a necessary alloy in stainless steels, especially in PWR pressure vessels and internal core structures. Cobalt-60, half-life 5.7 years, also a trace element in carbon and stainless steel, gives rise to a gamma dose rate far too high to permit manual access for disman­tling the reactor structures. Early Stage 3 therefore, requires the use of remote methods to cut, handle and package active wastes. The option to defer Stage 3 for up to 100 years offers advantages because it is expected that dose rates will decay sufficiently to allow access to set up and maintain automatic equip­ment (see Fig 4.14). This would simplify dismantling, speed-up work and reduce costs, but account must also be taken of the long surveillance and maintenance of the residual buildings. Because of other very long — lived isotopes such as nicel-59 and nicobium-94, no further benefit will accrue from longer delay.

The CEGB will remain responsible under the terms of the site licence, for all aspects of safety until Stage 3 decommissioning is complete and the Nil has de­clared the site to be free from harmful radiation. Thus CEGB is bound to exercise strict management control over the whole decommissioning project, including the handling and transport of all waste to designated dis­posal sites.

Radioactive wastes arising from Stages 2 and 3 will be low or intermediate level. The intention is to limit the requirements for cutting material during disman­tling operations or subsequent conditioning to the minimum necessary to reduce potential radioactive doses and costs. Large reinforced-concrete containers or boxes will be used, therefore, to carry such wastes to suitable disposal or storage facilities.

Regulatory authorities now require more attention to be given to station design to identify the features needed for operation and other features specifically introduced that could assist decommissioning when that time comes. This includes:

• Layout and outline plans for decommissioning.

• Selection and control of primary circuit materials.

• Maintenance of lifetime records.

• Surface treatment and decontamination.

• Recovery from design basis accidents.

Proposals for decommissioning magnox nuclear power stations have been prepared over the last five years and are up-dated as required. These provide a basis on which detailed engineering plans can be built-up when the time is nearer to hand. Basic information on the physical layout of the reactor structure, materials and operational history has been used to calculate a radio­active inventory of activated and contaminated ma­terial. This identifies the mass and isotopic contents including important data on long level trace elements which have been checked by measurement of samples

when the opportunity is presented. From this, pre­dictions may be made of the radioactivity decay and requirement for engineering plant, shielding, waste management and transport.

The major task at Stage 2 of dismantling and re­moving the low level contaminated boilers has been studied. It has shown that suitable lifting methods with capability of well over 1000 t could be used to transfer these boilers as single items to a suitable multi-wheeled vehicle and thence onto a barge for transport to a designated land or sea disposal facility. Alternatively the boilers may cut down into pieces and moved by road/rail in packages weighing up to 100 t or so. These methods have been proved by similar tasks undertaken in the offshore oil industry.

The expected approach to Stage 3 is that access to the reactor will be gained from a shielded, ventilated temporary containment through the pile cap. Up to some 80 years after shutdown, remote methods will be essential to cut and remove material and to effect routine plant modifications, maintenance and recovery. This will be controlled from outside the containment using TV, sound and other navigational control sys­tems. Wastes will be lifted out in skips and taken through an airlock to the waste assayence packaging facility using the existing irradiated final route.

It is proposed that the 2000 t of graphite per re­actor will be lifted out using a multi-fingered grab mounted on a small commercially available tracked vehicle. Removal of the 2500 t or so of steel will be divided between abrasive disc cutting of the core sup­port and restraint structure, and hot cutting of the pressure vessel itself. Methods for removing the inner activated layer of the biological shield include impact hammers, concrete saws, explosives and high pressure water jets. All these methods, including designs for ventilation system and filters, are being developed and will be demonstrated initially in the complete dis­mantling of the Windscale AGR.

Decommissioning of PWRs has been reviewed, mak­ing use of extensive studies and development work already being undertaken in USA and Western Europe.

Costs have been assessed and for operating reactors financial provision is made each year to build ade­quate funds to meet their costs.

Appendix A

Under the terms of the Fuel Supply Agreement with BNFL, AGR fuel is ordered twice a year for delivery to site some 15-21 months later. Although this task is performed centrally and is the responsibility of the CEGB’s Nuclear Operations Support Group (NOSG), individual stations estimate their anticipated needs by relating declared generation capability in the months ahead to the fuel cycle. The AGR can only continue to generate maximum power for two or three months without refuelling and therefore, as insurance against any interruptions which may occur to steady fuel supply, an extra ‘contingency allowance’ is built into the fuel order which guarantees a few months’ extra operational needs.

Road transport is used for the delivery of AGR fuel elements from BNFL’s Springfields works to the power station, whereupon the new fuel boxes (each containing 8 elements) are moved to the fuel store in a carefully controlled manner, according to the re­quirements of the relevant Criticality Certificates. As we have seen, the different parts of the fuel stores will each be covered by the requirements of Criticality Certificates, which wilt specify the necessary restric­tions and types of operation permitted as a conse­quence of Criticality Safety Assessments relating to the appropriate area. The certificates will state, for example, whether or not stacking of fuel boxes may take place and to what degree.

In the fissioning of uranium, about one hundred dif­ferent radioactive nuclides are produced. The princi­pal fission products and their half-lives are shown in Table 4.1. In the radioactive decay process beta particles and gamma rays are emitted from most of the nuclides.

Activation products

The materials of reactor construction, the uranium fuel cladding and the reactor coolant are subject to neutron bombardment in the fission process. As a consequence, radionuclides are formed by neutron cap­ture and typical nuclides and their half-lives are shown in Table 4.2. Radioactive decay gives rise to emission of beta particles and gamma rays in general. The radioactive quantities of activation products that arise are considerably less than the fission products.