The new fuel route for Oldbury is illustrated in Fig

2.28. Each stage in the process is numbered from I

to 10 and is described as follows:

1 New fuel elements arrive by road transport in self-stacking steel boxes, each box containing 20 fuel elements. Thermocouple fuel elements arrive on site packed individually in separate boxes. The boxes are unloaded from the road transport vehi­cle and deposited in the buffer fuel store by a fork-lift truck.

2 Boxes are stacked in the buffer fuel store by a fork-lift truck.

3 When required, boxes are transferred to the fuel lift by the fork-lift truck.

4 The fuel lift transports the boxes to the main fuel stores on the seventh or ninth floor where they are stacked in designated bays by a fork-lift truck.

5 When new fuel elements are required, boxes are transferred to an unloading position adjacent to the transfer chute of the reactor (1 or 2) fuel preparation room on the seventh floor.

6 New fuel elements are removed from the boxes and placed in the transfer chute, the individual

ел го

1-ю. 2.28 New fuel route M Oldbury

polythene wrappers are split and the elements passed down the chute into the appropriate reactor fuel preparation room. Empty boxes are returned to the buffer fuel store to await transport.

7 ln the fuel preparation room, each fuel element is placed on a rubberised-hair pallet and passed along a stainless steel inspection bench for check­ins Defective elements are rejected or repaired.

s Satisfactory standard fuel elements are placed in a hand trolley for transport to the fuel loading room. Top thermocouple elements, absorber ele­ments, channel gags and supports are also taken to the fuel loading room after inspection.

9 Standard fuel elements are removed one at a time from the hand trolley by a jib hoist and lowered into separate compartments in a new fuel maga­zine. Top thermocouple elements are passed via a transfer chute to the upper maintenance room for attachment to the standpipe assemblies. Ab­sorber elements, channel gags and supports are also loaded into the new fuel magazine.

10 When new fuel elements, etc., are required, the fuelling machine is connected to a loading tube located between the operating floor and the new fuel magazine. The fuelling machine then removes items one at a time from the magazine and transfers them (via the loading tube) to its new fuel storage tubes.

The charge/discharge process at the reactor is con­sidered in Section 2.3 of this chapter, but it finishes with new fuel in the reactor channel and the ‘spent’ irradiated elements in the fuelling machine (FM) ma­gazine. The irradiated elements are discharged to the irradiated fuel storage pond via the irradiated fuel disposal facility (IFDF).

In addition to fuel, the FM handles activated re­actor components such as flattening bars, channel shock absorbers and gags. The IFDF has equipment to divert such components to dry storage vaults (while preventing the discharge of elements to the vaults). Ai Hunterston A the fuel element graphite sleeves also go to vaults after crushing. The IFDF has equip­ment to insert and seal elements for post-irradiation examination (including failed fuel) in dry bottles be­fore discharge to the pond.

Hardwired systems such as those using relays or
laddies, when combined in complex systems, reach the limit of attainability.

It is possible to meet the logic switching require­ments with computer-based systems. The advantages of these over the hardwired system become apparent when there is a large number of inputs to the pro­tection system, when the logic is complex and when plant transmitters, e. g., in-core flux detectors, have to be shared between protection, indication and control. Such systems can provide more precise monitoring of plant status, including departure from nucleate boiling for local core power density in PWRs, and also perform some trip functions. Such systems can provide greater plant operating flexibility when re­quired to work in a load-following mode and improve availability and safety. In such cases an integrated computer-based scheme offers a more attractive solu­tion than a very extensive and complex set of hard­wired logic. Furthermore, the computer-based scheme facilitates the provision of comprehensive on-line checking of transducers and logic which is difficult to provide with hardwired schemes,

clutch supplies

The integrity requirements of the automatic pro­tection systems for nuclear reactors are very high and, as in the case of the interlocks, the presence of soft­ware ereailv complicates the assessment of reliability. However, it is important to note that the programs tor interlocks and protection systems can be made relatively simple and they are fixed. This contrasts with data processing software which is more complex (see Volume F, Chapter 7),

Reference has already been made to a photogra­phic module that has been developed as part of the Triumph system. This module uses cartridge loading with a 35 mm film cassette and a combined view — tinder, TV camera head and shutter assembly which enables the image to be identified and focussed. A range ot lenses can be accommodated in the design. A tilting mirror allows sideways viewing with a limited forward view, lighting is provided for TV viewing and a tlash for photography. This arrangement is a con­siderable improvement on earlier photographic devices where it was not possible to view and focus the image be to re making the exposure.

A photographic camera, having the same dimen­sions as the Hartlepool Hevsham / boiler tube view — mg camera described earlier, has been developed for providing a more detailed inspection of the boiler tube finning and support clips. Specially prepared 8 mm film is used and about 300 photographs can be taken before the camera has to be removed for re­loading. In this case, as the geometry of the area to be photographed relative to the camera is known, the camera lens can be focussed remotely without the need for a viewfinder.

Future devices

In the rapidly developing area of viewing devices it is clear that the inspection cameras developed for routine inspection will be considerably improved over the life of a particular station. The use of colour, stereo imaging, photogrammetry (where it is possible to estimate the dimensions of an object by measuring the dimensions on one or more photographs of that object), laser scanning cameras, holography, compact solid state TV cameras (CCDs) and fibre optics are examples of some of the techniques currently being applied to non-routine inspections that could also be applied to routine visual inspection.

9.12.1 Reactor protection

The AGR reactor protection system has a great diver­sity and variety of sensing measurements as shown in a typical tripping schedule (Table 2.15) and double ‘2 out of 4’ is the dominant redundant logic. Sector temperature protection is based on a selection of highest temperatures as arranged by temperature mo­nitoring units, which each take the ‘highest out of 16’ thermocouples for connection to auto-resetting temperature trip amplifiers. The linear power excess neutron flux protection also uses auto-resetting trip amplifiers. Three protection parameters are concerned with boilers and boiler feedwater, and the requirement tor virtually pure water demands the measurement o! extremely low levels of conductivity. A feature of the AGR safety circuits is the use of standard inter­posing trip units for the physical measurements of gas absolute pressure, circulator speed and inlet guide vane position, circulator voltage and feedw-ater con­ductivity and condenser level.

The sensors that convert the plant operating para­meter. e. g., temperature, neutron flux, etc., are of the same type used for measurement purposes.

The Laddie logic switching device described in Section y.4.5 of this chapter is used in AGR protection s> stems.

9.12.2 Quadrant protection

A high integrity protection system is provided on each quadrant using hard-wired electronic equipment and ‘2 out of 3’ relay logic. This system is designed to:

• Protect the gas baffle against over-pressurisation.

• Protect the circulators against lubrication system faults.

• Protect the plant against damage due to mismatch between gas flow and feed flow.

On detection of fault conditions the equipment trips the gas circulators and the feed to the quadrant. The circulators are connected to a low frequency drive as they run down in speed.

The auto control system design is intended to cope with the transient caused by a quadrant trip and main­tain the reactor on-load, albeit with reduced output. Operation with more than one quadrant out of ser­vice is not permitted, so output signals from the quadrant protection equipment feed into a ‘2 out of 4’ trip in the reactor safety system. Using this route, the quadrant protection system provides protection against certain reactor-based faults.

9.12.3 Essential plant protection equipment

In addition to the quadrant protection equipment, a separate set of microprocessor-based protection equip­ment is provided of the same type as used for the PTSE. This equipment, in general, provides additional plant protection rather than direct ‘safety’ protection.

11.1.5 Protection principles

The reactor protection system is designed to achieve the following actions to ensure safe conditions after an incident or reactor fault:

• Trip all rod cluster assemblies (RCCA) into the core to shut the reactor down and thereby provide immediate short term reactivity control.

• Initiate or put into a state of readiness a range of plant specifically provided to maintain the re­actor shut down and ensure decay heat removal. This plant is collectively referred to as the engi­neered safety features (ESF).

Both the above safety actions have to be achieved with an extremely high reliability, necessitating a design which includes considerable redundancy to meet the station risk targets.

Redundancy, however, cannot provide the sole means of achieving the very low failure rates required, since the possibility of common mode failure of identical lines of protection has to be considered.

Local heat generation rate (or rating) varies as the isotopic content changes, for three reasons. Firstly the heat produced per fission in plutonium is greater than in uranium, secondly the macroscopic fission cross-section varies as the number densities of the fissionable isotopes change, and lastly the neutron flux shape itself changes. In flux shape effects there are two principal categories called ‘macroscopic’ and ‘fine structure’. The macroscopic flux shape is the overall shape which, for example, in a uniform cylindrical reactor would be given in one energy group form as a cosine in the axial direction and J0 Bessel function in the radial direction. The effect of reductions in fission cross-section and the build up of absorber oc­cur more strongly in the high flux regions. This results in some flattening of the flux shape which combined with burn-out of the fission cross-section leads to more pronounced flattening in the rating shape.

The fine structure flux shape has components in the radial direction, through fuel, clad and moderator and also in the axial direction, from fuel into end gaps. This fine structure shape is basically a thermal neutron flux dip into the fuel which is the more highly absorbing component of the lattice cell. Be­cause of the burn-up of fissile isotopes the rate fine structure lends to reduce with irradiation, but in some cases the differential build-up of plutonium can coun­teract the fine structure burn-out. Particularly signi­ficant fine structure effects are considered below.

Also called sector rods, auto rods, fine rods, zone rods. They are grey rods, a magnox reactor contain­ing typically 10-30 and an AGR typically 35-45; Dungeness В is the exception amongst the AGRs, having five zone rods. The principal function of these rods is the fine control of neutron power and hence reactor gas temperatures. They are also used for xenon override, particularly in AGRs where all the oernde capability is invested in regulating rods; at Dungeness В the five zone rods are supplemented by 16 trim rods. In magnox reactors the override capability of the regulating rods is limited because of their limited worth. For control purposes the reactor core is divided into a number of sectors or zones and the regulating rods are associated with the control zones in which they are situated. Auto control systems are often associated with regulating rods, the rod positions be — ine adjusted automatically to maintain constant fuel channel gas outlet temperatures in their associated zones (see Chapter 2).

Trim rods

A regulating rod is most effective when it is inserted about halfway into the reactor core, as mentioned in Section 5.2.1 of this chapter. Therefore, associated with the regulating rods there may be trim rods, grey or black, on manual control; trim rods are inserted or withdrawn as necessary to maintain the regulating rods in their optimum range of travel. On some magnox reactors the trim function may be provided by bulk rods rather than by dedicated trim rods, for example, limited out-of-line running may be permitted with the bulk rods. Other ways of achieving the trim function are given in Section 5.5 of this chapter.

This section discusses operational factors, implications and limitations of the control and instrumentation (C and 1) equipment described in Chapter 2, the op­erational information that it provides and how it is used for safety and protection purposes.

From an operational viewpoint it is important that adequate information is provided to keep the opera­tor fully aware of current plant conditions, with the ability to rapidly detect and control any change from the required state. Information is displayed in a logi­cal pattern in order to bring important parameters to the operator’s attention. Sufficient redundancy is built into the system to give independent informa­tion regardless of equipment failures, at the same time avoiding excessive data which may cause con­fusion. The information available is sufficient to give full control during normal operation so that essential parameters can be monitored during fault conditions.

Two aspects are important in the context of op­erations:

• Instrumentation accuracy limitations Whenever safety cases are made which require automatic or manual actions based on instrumentation it is nor­mal that any instrumentation errors or limitations are allowed for in the safety case, and any limits set will contain these allowances. The operator does not therefore have to consider instrumentation errors or limitations during normal operation. This may however lead to pessimisms and resultant op­erational penalties.

• Availability of instrumentation It is not practical to fully instrument all items and therefore assump­tions need to be made on items remote from instrumentation. This is particularly important with respect to the large number of fuel channels in magnox reactors, only a proportion of which are instrumented with fuel element or channel gas outlet thermocouples. Allowances for non-instrumentation channels are made for thermocouple variability, channel gag settings, channel, power, etc. The re­actor will normally be controlled on the basis of these pessimistic postulated temperatures.

The use of this equipment can be related to the various phases of operation of a nuclear power station and three of these (start-up, on-load and shutdown) are summarised in Table 3.4 which refers to a typical magnox station.

For the purposes of this chapter, control and in­strumentation aspects during operation are described under the following headings:

• Maintaining safe shutdown.

• Start-up.

• Raising power.

• Operation at power.

• Shutdown action.

• Long term measurements and monitoring.

Safety considerations are paramount in all phases but some phases hase special requirements for measure­ments to be made. During start-up and raising power, minute-to-minute information is required but in other phases different situations exist. When operating at power there is a need for information for optimisa­tion, but this is interrupted by fuelling operations and random incidents such as fuel element failures and boiler leaks. These require rapid action and suit­ably prompt information.

Thus, some of the equipment is used for one phase, some for more than one phase and some is used at all times. On some stations, regular inspection of the sector temperature panels and equipment in the annexes, such as the safety lines and recorders, are required to maintain a full appreciation of the re­actor operating condition. With these factors taken into account, it will be shown that the operator has the necessary information to cope with any given sit­uation including fault and post fault conditions.

An important part of the C and I system is the indications and alarm equipment provided on the cen­tral control room desks and panels, supported by the data processing system, constituting a part of the C and I system which is vital to station operation at all times. {Details of these are given in Volume F, Chapters 6 and 7.)

The neutron flux instrumentation is particularly important in all phases and provides:

• Indications to the operator of the reactor power and doubling-time from sub-critical to full power.

• Alarm signals to the operator giving warning of trip levels being approached.

• Trip signals to the reactor safety system to protect the reactor if trip levels are exceeded.

• Control signals, e. g., doubling-time freeze to the control system, this forming part of the protection function.

Neutron flux measuring channels have to cover the nine decades between shutdown power of a few watts to full power of around 1600 MW (thermal) for 660 MW (electrical) unit output. The limitation of some channels makes it necessary to cover the nine decades with a series of overlapping channels, as indicated in Fig 3.33.

As the reactor power is raised, the channel cover­ing the low level reaches its trip level and, in order to prevent a reactor trip and allow the next over­lapping channel to take over, it is necessary to ‘veto’ or immobilise the trip. This need to veto channels is indicated to the operator by an alarm set below the trip level and the vetoeing procedure is an im­portant operational aspect. In particular it is most important that the veto is removed and the trip re­instated when reactor power is reduced again.

The detailed arrangements vary depending on the type of reactor and the characteristics of the neutron flux density detectors, for example, the need to with­draw detectors at certain power levels.

Surveys are usually done by one of two methods:

• Loss of pressure.

• Reduction in the concentration of an injected tracer.

Loss of pressure is normally conducted over a 24- hour period to give a fair degree of accuracy. All activities such as fuel loading, blowdown of the vessel for purity and any other activity that may give a loss or increase of coolant is suspended during this period. Reactor conditions, in particular reactor temperatures, are stabilised as much as possible so that correction at the end of the test to the results is minimised. The loss of reactor pressure over the prescribed period is used to calculate the gas losses.

A more accurate method to assess losses is to use tracer techniques where a known quantity of a gas such as helium is injected into the pressure circuit; the resulting concentration or dilution of the tracer is measured against the total quantity of coolant pre­sent. This gives a more accurate measure than the pressure drop method but the accuracy of either meth­od is dependent upon the length of time the test can be sustained. Usually, 24 hours is sufficient to give an accuracy around lO^o.

With a graphite moderator, a liquid or gas must be used as the coolant. Although there are water cooled graphite-moderated reactors, e. g., the Soviet Union’s RBMK series of power stations, of which Chernobyl is one, only gas cooled reactors will be referred to here.

Whilst the United States and Canada pioneered, respectively, the light and heavy water moderated de­signs, France and the United Kingdom undertook the early development of the graphite moderated reactor, selecting carbon dioxide as the coolant because of its relative chemical inertness and low neutron activa­tion. France abandoned this approach in favour of an extensive PWR programme. The UK continued to be heavily committed to gas cooled reactors in the form, initially, of magnox and subsequently the advanced gas cooled reactor. Both designs are described in de­tail in other chapters and thus only an indication’is given here.