Decommissioning is the term used to describe the whole process which follows reactor final shutdown. It includes defuelling, dismantling plant and build­ings, transport of waste materials to authorised dis­posal sites and site clearance.

6.3Decommissioning stages

Three decommissioning stages may be identified:

• Stage I decommissioning relates to the period im­mediately following final shutdown. It is usually assumed that this is a planned operation rather than precipitated by a breakdown or accident, so that some small benefit may be realised by op­timising the management of the last fuel change. Gas-cooled reactor defuelling is a major task that could extend over several years to make best use of flasks, transport and reprocessing services. Dur­ing this time other tasks such as removal of opera­tional solid wastes, sludges and resins and removal of thermal insulation will be carried out using exist­ing plant and facilities by operating or decommis­sioning staff.

• Stage 2 has the objective of dismantling all plant and buildings external to the reactor biological shield or prestressed concrete pressure vessel. The major part of this will be inactive and removed by con­tractors under the strict control of the CHGB. The radioactive, mostly contaminated waste material, will be taken off-site in large packages which com­ply with IAEA Regulation for the Safe Transport of Radioactive Waste for disposal or storage. The residual reactor structure, which contains over 90% of the total radioactive inventory after defuelling, is sealed off until Stage 3 commences

For a magnox station, the radioactive plant re­moved includes the contaminated boilers, main ducts, radwaste and fuelling equipment, fuel ponds, laundries, workshops, etc. This plant will be dismantled using built-in facilities and contract engi­neering plant and equipment under controlled condi­tions, to ensure that safety of the workforce and public is maintained to the standards required by NIL The waste arising, though significant in terms of mass, will not generally require elaborate or remote techniques.

• Stage 3 is concerned with the removal of the reactors themselves and final clearance of the site rendering it safe for future use. This is a very dif­ficult task because the activated reactor structure, which for the early magnox design comprises the graphite core, steel restraint structure and diagrid support, steel pressure vessel and inner 1 m layer of reinforced concrete biological shield, requires the use of specialised remote methods (see Fig 4.13). Suitable techniques are being developed but the detailed design, planning and proving the complete system will be a substantial project. For the late magnox and AGR stations the boilers are within the prestressed concrete vessel and will be removed during Stage 3.

• The time required for each stage of decommission­ing varies for each reactor system and individual site but is likely to take 5 to 7 years.

It is intended that Stage 2 should follow on directly from Stage 1. There is an option between proceeding to Stage 3 immediately, so that decommissioning is completed as a single project in about 15-17 years from shutdown, or to defer Stage 3 up to 100 years.

Handling and movement of all nuclear fuel in the UK is controlled by legislation which is described in detail in Chapter 4. Attached to the station site licence

are conditions which relate specifically to the storage and carriage of nuclear fuel and also to the formula­tion of Station Operating Rules, in which the obligation to carry out the various Criticality Safety Assessments described in the previous section is formally laid down bv the CEGB as an instruction to the appropriate power station manager. The work involved is under­taken by a nominated assessor, who prepares a Cri — ticalitv Safety Submission’ (which includes the ‘Assess­ment’) for subsequent consideration by a Criticality Safety Panel, consisting of the Station Manager and representatives of the CEGB Headquarters Depart­ments and Divisions. If the panel agree to the pro­posals contained within the submission, clearance is reported to the Nuclear Safety Committee and the Station Manager then formally signs a ‘Criticality Safe­ly Certificate’, an intrinsic part of the original Cri­ticality Safety Submission, copies of which will be kept in Station Operating Manuals. He then ensures that the appropriate area(s) of the fuel route at the power station displays a copy of a relevant notice containing all related essential guidance and restrictions to be observed in practice and which will therefore guarantee the avoidance of criticality at all times. The full pro­cedure for gaining criticality safety clearances is set down in a formal document issued by the CEGB.

Radioactive wastes at nuclear power stations arise from the fission reactions of the uranium fuel and associated neutron irradiation of reactor materials, coolant and fuel cladding.

2.6.1 Principles of waste management

The key principles of radioactive waste management

are:

• To minimise waste quantities by careful selection of reactor materials which have a low potential for radionuclide production upon neutron irradiation.

• To minimise waste arisings by appropriate selection of operating procedures at the power station.

• To dilute and disperse low level wastes to the en­vironment under controlled conditions.

• To accumulate wastes unsuitable for immediate dis­posal into the environment in order to take ad­vantage of radioactive decay and immobilisation/

packaging facilities.

2.6.2 Nuclear characteristics of waste

Nuclear reactors give rise to three principal types of radionuclides:

• Fission products.

• Activation products.

• Actinides.

Fault studies are carried out to investigate the behav­iour of a nuclear power station under abnormal or accident conditions.

The studies, which are carried out during the con­ceptual and design stage, examine the temperature of fuel and any other sensitive components under fault conditions and the effectiveness of the various pro­tective devices in limiting the consequences of the fault. They also examine the capability of the essential plant required to maintain the reactor in a safe state after the shutdown has occurred, and the effective­ness of any long term remedial actions which the operator may take.

The studies form the basis for defining the safety limits to which the plant must be operated, the var­ious settings of the protection equipment and indicate the minimum amount of essential plant which must be available for use post-trip. The object is to ensure that for all credible accident conditions, the risk of a release of radioactive material to the environment is acceptably low.

For the latest stations, the acceptable release for any accidents is related to the estimated frequency of occurrence.

Table 4.13 shows the guidelines for the accidental releases of radioactivity. Releases of greater than 1 ERL are only acceptable at a frequency of the order of 10“fi per year.

Single faults with a calculated frequency lower than about 10“ per year are considered to be sufficiently unlikely not to warrant any detailed study of the con­sequences. These are termed incredible’ or ‘beyond design base’ faults.

For the earlier stations, the approach was less for­malised and the division between ‘credible’ and ‘incre­dible’ faults was based very largely upon engineering judgement, statistical analyses to estimate the frequen­cy of each fault were only rarely carried out. It is a requirement for all faults that there are two separate lines of protection via the safety circuits capable of causing the reactor to be safely shut down. These lines of protection should, if possible, be diverse in terms of the particular parameter being monitored. For example, one line of protection may detect a fault from the transient increase in neutron flux whilst the other detects the fault from the increase in channel gas outlet temperature. If diversity cannot be achieved and there are a large number of sensors, only a few of which are required to detect a fault, it may be possible to claim redundancy. Hence, channel gas out­let thermocouples may input into both lines of pro­tection on the grounds that only, say, three or four out of a few dozen need see the fault to cause the reactor to be shut down in safety. The studies establish which of these lines is the least effective and bases the operating limit on the assumption that the most effective line fails to function.

The types of fault considered for the three designs of reactor to be discussed here are very similar, al­though the inherent differences in the design of the plant result in difference in the treatment and in the consequences of particular events.

Experience with the rules over the years has shown little need to change the underlying principles and this has clearly been reflected by the comparatively low radiation doses received within the CEGB. Amend­ments have of course been made over the years, pri­marily due to changes in legislation and international recommendations. As the rules relate to work in es­tablishments where the Electrical and Mechanical Rules are implemented, account is also taken of the changes made to these.

The proper training of staff for nuclear power station operation has been given full priority by the CEGB from the start of the civil nuclear power programme, with the station manager of each nuclear power sta­tion being directly responsible for ensuring that his staff are adequately trained for their work and for uti­lising both local and national training facilities. At present, with one exception, there are no formal train­ing requirements specified in the statutory nuclear regulations but there is a duty on the CEGB to give instruction to all persons employed, and authorised to be on site, on the radiological risks associated with the plant and its operation, and the precautions and actions in the event of an emergency. The one excep­tion referred to is a requirement for the CEGB to have its nuclear emergency training arrangements ap­proved by the HSE. Such arrangements are formally audited by the Nil site inspectors.

The nuclear training needs of the CEGB are re­garded as quite separate from that required for fossil — fired plants, and the potential hazards of nuclear plants make safety the primary aim of training for operating staff in particular. The initial training of staff for all types of power station does, however, follow a general pattern with high priority being given to thorough training in all areas in line with the Elec­tricity Supply Industry Training Committee (ESITC) recommendations. The policy is continued for the nuclear operations with special emphasis on the radio­logical safety aspects.

The training programmes for CEGB nuclear power station staff cover the existing wide range of gas — cooled reactors and the future requirements for the pressurised water reactor which will be built at Size — well. The level of basic nuclear training is essen­tially the same for all operations engineers, but higher levels of training are plant and station orientated for individual nuclear stations and are divided into ‘on­site’ and ‘off-site’ training. Following the incident at Three Mile Island (USA) in 1979, the CEGB carried out a comprehensive review of nuclear training. The review indicated that, whilst the training was adequate, there was a need for a more structured approach and subsequently a CEGB standard nuclear training spe­cification was issued to all operating locations. The specification was not mandatory, but gave firm guid­ance to the nuclear station managers on the way to attain satisfactory training levels. Following the issue of the training specification, which was fully endorsed by CEGB senior management, each nuclear station produced a station training document setting out its nuclear training arrangements.

The organisation of training varies from station to station, but all stations have a full-time training officer/ engineer with clearly defined responsibilities and strong formal links with senior management. An essential part of the training arrangements is the preparation of comprehensive and up to date records, these are compiled at each station and there is steady progress towards more centralisation and computerisation.

On-site training is a most important part of the training pattern, ranging from site familiarisation, with its application to all people on site, to the more de­tailed and technical training for operations engineers. The more specific ‘on-job’ training forms a most im­portant part of the operations engineers’ education and is carried out in a properly structured manner with formal assessment on completion. One most important element in the on-site training programme is the emer­gency training. In order to maintain a high level of preparedness to cope with any possible emergency on site, a comprehensive on-site training programme is fulfilled at each nuclear station; this includes periodic training in first aid and fire fighting, the use of breath­ing apparatus, and assembly point procedures. Spe­cialist training is also given on such duties as damage control and rescue measures, VHF radio operation, and off-site data plotting and health physics control room duties. Although such training is now required by the site licence it has been an integral part of the CEGB emergency training measures from 1960. In order to maintain this high level of training and to ensure that all shift staffs are suitably exercised, it is the practice for each station to hold five full rehearsals per year based on a postulated serious incident on the plant, and designed to test every facet of the com­prehensive emergency arrangements.

An essential part of the training of nuclear power station operations staff is off-site training. This is mainly carried out at the CEGB’s National Nuclear Power Training Centre (NPTC) at Oldbury, which provides a mixture of theoretical and simulator-based training on a wide range of nuclear related subjects. The NPTC is administered directly by the CEGB and staffed almost exclusively by scientific and engineer­ing staff with experience of nuclear station operation. The NPTC also provides a forum for short seminars and conferences on nuclear subjects of current interest to specialist groups within the industry, a typical se­lection being asymmetric fault studies, probabilistic risk analysis, reactor physics, and gas and waterside chemistry. Its principal role, however, is to provide three essential training functions to implement the CEGB’s nuclear training policy, these are:

• The training of operational staff following their initial training in a nuclear station.

• The revision training of experienced operations staff from nuclear stations.

• The training of non-operational staff from the sta­tions, divisions and headquarters departments.

The NPTC training courses are basically divided into three areas, namely, introductory, operational and plant familiarisation.

Introduction to nuclear power

This is a four week course with emphasis on basic nuclear technology; the syllabus includes nuclear and reactor physics, reactor kinetics, reactor heat trans­fer, health physics and reactor chemistry. The lecture periods are supplemented by practical work in the laboratory and visits to an operating nuclear power station.

Operational training

Separate courses are provided for magnox and AGR staff following completion of the introduction course and the period of plant familiarisation. This phase of training concentrates principally on the opera­tional aspects of the engineers’ responsibilities and, as with the introduction course, is aimed at obtaining a thorough understanding of the dynamics of the plant with the primary objective of ensuring safe op­eration, and the secondary but desirable objective of improved commercial performance. At this stage an increasing proportion of the training is given by ex­perienced operating engineers, thus allowing the en­gineer under training the opportunity of discussing current operational procedures and the problems likely to be met on the plant. The lectures aim to develop a proper understanding of basic principles rather than the development of mathematical equations and their solution.

After a further period of training at the station the operators then attend plant-specific simulator-based operation courses at NPTC.

The present life of magnox fuel is limited to 5000- 6000 MWd/t channel average irradiation or dwell time of 11 years, whichever is the shorter. For fuel cycle purposes, the reactor is divided into three radial zones:

• Flattened zone — irradiation limited.

• Unflattened zone — irradiation or’reactivity limited.

• Unflattened zone — dwell time limited.

In the case of the flattened zone, the rate of fuel changing is at a maximum and uniform over the whole zone.

The unflattened zone will have fuel changing rates which will vary with the radius from the rate em­ployed in the flattened zone at its inner radius to the rate in the dwell zone at its outer radius. For practical purposes, it is usual to sub-divide these zones into smaller radial regions such that the average channel irradiation rate in each of the smaller regions may be regarded as constant giving a uniform rate of re­fuelling. The outermost zone of the unflattened region is usually known as the dwell zone and contains fuel w^hose life is limited metallurgical^ to 11 years. The channel content of this zone is of the order of a few — percent at the most of the total core. A typical fuel cycle for the whole reactor is illustrated by Fig 3.38.

It should be noted that starting with a virgin re­actor, equilibrium is not reached until the whole of the first charge has been replaced. Under the present regime, this will be 11 years. However, equilibrium conditions approximate by the time the flattened zone and about half of the unflattened zone have been refuelled.

From the point of view of the mechanics of the fuel cycle, an increase in the target channel irradia­tion simply requires the same channels in a zone to be evenly spread over a greater length of time. Re­ferring to Fig 3.38, the flattened zone is shown as completing its cycle in five years. A further cycle is started bv returning to the channels first discharged at the start of the original e>cle and continuing se­quentially through the channel for another foe vearv. It’, however, conditions enabled an increase in target irradiation to give an equivalent dwell time of six years then the same channels would be discharged over a six year period instead of five years. There would be a resultant decrease in refuelling rate. Such a change is illustrated in Fig 3.38 and would normally be reflected in similar changes in the refuelling of the unflattened zones. As shown, changes may take place at any time and are not restricted to the end point of a cycle.

Fuel cost accounting for AGRs is performed within the ADOS suite by a specially provided computer pro­gram which conducts its calculations on a channel — by-channel basis. At regular intervals and for each channel, the cost of the heat generated is computed from a knowledge of the ‘fuel worth’ at the time (initially set to the replacement fuel cost at the time of loading to the reactor), the irradiation accrued within the period and the proportion of stringer life remaining before discharge. The total reactor heat cost is obtained by summating the individual channel fig­ures. In this way the program enables precise records of reactor and individual channel heat costs to be maintained within the ADOS suite.

This part comprises just the one regulation and re­quires the level of radiation in controlled and super­vised areas to be monitored. Section 4 of this chapter discusses how these aspects are applied within the CEGB.

Part 7: Assessments and Notifications The first three regulations of this part require that the employer assesses the consequences of any po­tential radiation hazard which may arise, produce a report of such and, where appropriate, prepare a con­tingency plan. This latter aspect is dealt with at nu­clear licensed sites by the site’s emergency plan. The Safety Rules (Radiological) also contain actions which must be carried out in the event of maloccurrences with radioactive material.

Regulation 28 contains the formal investigation re­quirement for doses in excess of three-tenths of the dose limit.

Regulations 29 and 30 deal with investigations and notifications in the case of exposures in excess of the dose limits and the subsequent dose limitation scheme in such cases.

Regulation 31 requires that radiation incidents be notified to the appropriate authority, when applicable.

Part 8: Safety of Articles and Equipment

This part deals with the safety aspects of equipment

involved with radiation, which may include such items

as X-ray sets and shielding devices. The requirement

is that the requirements of other regulations can be

fulfilled.

Loss of coolant accidents (LOCAs) w-hich are ana­lysed, range from very small leaks from the reactor coolant system boundary to the complete severance of the largest pipe (discharge area approximately 1 m:) in the primary coolant recirculation loop. The system response varies considerably with the size of leak. For the larger breaks, the primary coolant ra­pidly blows down and depressurises. The water coolant changes to steam and, although the reactor is shut­down (loss of moderator ensures this, even if the control rods failed to insert), the decay heat is suffi­cient to increase fuel and clad temperatures. The temperature rise is arrested by the injection of highly borated water from accumulators (storage vessels) and by pumps.

Although only some fuel clad failures are expected to occur, it is conservatively assumed that all will fail and release their fission products to the containment. The extent of the release of these to atmosphere de­pends on the leak rate from the now pressurised containment and on whether other plant failures are assumed to occur, including any which provide a di­rect or indirect leak path to atmosphere. The large. LOCA imposes increased loads on the reactor inter­nals, increased stresses on the pressure vessel and con-‘ tainment structures. Assessment of the ability of these structures to withstand the imposed temperature and pressure transients forms part of the safety case.

Small LOCAs include not only pipe breaks of lim­ited aperture but also, for example, inadvertent open­ing of relief valves and their possible failure to reseat. The accident results in a relatively slow depressuri — sation which is mitigated by the automatic operation of high pressure injection pumps. Again, although few clad failures are expected, for the purpose of radiological consequence analysis all are assumed to tail unless the leak rate is within the capacity of the make-up water supply.