The temperature of the reactor coolant in the cold leg is compensated by a lead-lag circuit, and is then compared with the cold leg reference temperature which is a function of station load. The error signal is then added to the difference between the reactor power and the total turbine power to form an error signal; this then passes to the principal programmer whose main role is to introduce a dead band within which all the dump valves remain closed.

The signal from the principal programmer is di­vided between two individual programmers, one tor each turbine-generator unit, according to the propor­tion of the total loss of load which has been rejected bv each unit. The signals from the indmdual pro­grammers are then used to modulate the sequential openina of the three banks of dump saUes propor — — ional to the partial error signals. Each bank of dump — ibes can be modulated tuily open or closed within 20 seconds.

Each dump vale also has associated with it a trip open setpoint. The valve is tripped to its fully open position when the error signal from the individual proarammer exceeds the trip open setpoint, the valves beins able to move from the fully closed to the fully open position within 3 seconds. The dump valves are subsequently modulated dosed as the reactor control svstem reduces the reactor coolant temperature and the error signal falls below the dead band of the load rejection controller.

If it were not for the build-up of plutonium isotopes reactivity would decrease rapidly as a function of ir­radiation. In a natural uranium reactor the plutonium build-up is in fact sufficient to cause a net increase in reactivity with time over a substantial part of the fuel life. The k« of a magnox lattice shown in Fig

3.4 (a) as a function of time is typical of magnox reactor lattices and varies only little from one station to another due to small differences in the fuel-to — graphite ratio and reactor operating conditions.

The consequence of this k«, variation is that once the reactor is fuelled there is a long interval before any refuelling is needed in order to maintain criti­cality. This is a useful level of flexibility which has been utilised in practice when difficulties in the fuelling route have occurred-

With the enriched fuel of an AGR, the burn-up of U-235 dominates and the reactivity decreases sig­nificantly from the start of irradiation. The variation of k« for typical inner-zone feed fuel of an AGR is shown in Fig 3.4 (b). Also shown is the variation of kco with time for fuel containing burnable poisons designed to reduce the reactivity change and the local rating effects of new fuel.

In a PWR the change of k* with burn-up is even more pronounced, largely because of the higher en­richment and therefore greater importance of burn-up of the U-235. The k® for a 3.1% enriched PWR lattice with constant boron level in the water and at a constant water density of 0.67 gm/cm3 is shown on Fig 3.4 (c). The large change in reactivity resulting from this к я, drop is accommodated partly by refuel­ling, done in batches; for example, one-third of the core may be refuelled at a time, in which case the change in reactivity within a fuel cycle is about one — third that indicated by Fig 3.4 (c) and is counteracted by reductions in boron level. There are limits to the level of boron dosing because of the positive contri-

bution to the temperature coefficient of reactivity contributed by boron density reduction as water tem­perature increases. Burnable poisons may be incor­porated in initial fuel charges as a further reactivity control.

Black rods which spend most of their life substan­tially inserted into a high neutron flux may exhibit a reduction in worth due to partial exhaustion of their neutron capture capability. Consideration of the neutron flux incident on the surface of the rod, the surface area of the rod (to calculate the number of neutrons entering the rod) and the number of boron atoms in the boron steel inserts of a black rod, en­ables an estimate to be made of the time it will take for all the atoms of Boron-10 (the isotope with the high capture cross-section) to each absorb a neutron and thus the boron component to become ineffective. This phenomenon has been experienced at Hunterston

A where, after several years of operation, black rods used for radial flux shaping started to decrease in worth so that they were being progressively further inserted into the core to achieve the desired effect.

This phenomenon does not occur however in the steel component of a black rod or in a grey rod. Consideration of the number of iron atoms in a rod, the capture cross-section of iron and the neutron flux in the rod shows that it will be thousands of years before the neutron capture capability of the steel is exhausted. It is interesting to note that if, for ex­ample, an atom of Fe-56 (the most abundant stabte isotope of iron) absorbs a neutron it becomes Fe-57 which is also stable and equally capable of absorbing a neutron, so that even when all the atoms have each absorbed a neutron the rod is still capable of further neutron absorption. Examination of the Chart of the Nuclides will clearly show this effect.

A control rod may of course become time-expired after a few years for mechanical reasons rather than reactivity reasons.

The initial effect of the xenon change is to reinforce the power change, but on a timescale of hours the effect of the power increase is to increase the xenon production rate which contributes a negative change in reactivity (see Section 2 of this chapter). At first sight this may appear to be a stabilising effect, reinforcing the effects of neutron leakage and the negative tem­perature coefficients of reactivity. However, due to the delay between the change of power and the consequent change in xenon production rate, and provided this effect is of a magnitude which can dominate over the destabilising effect of the positive temperature coeffi­cients of reactivity, the increase in power is terminated and the power is driven downwards.

The destabilising effects of the positive temperature coefficients of reactivity now reinforce the effect of the change in xenon production rate and the power continues downwards until the decrease in power shows an effect in decreasing xenon production rate and the trend is reversed a second time. Thus an oscillation is set up. An oscillation has been observed at Berkeley in a test carried out during the early months of op­eration, at a core irradiation of approximately 900 MWd/t. The oscillation is shown in Fig 3.28, mani­fested as a change in fuel channel gas outlet tempera­ture. This was an oscillation in one of the higher modes, as described in the next Section 5.6.4.

Detection of faulty fuel elements may be achieved by receipt of high count alarms or high differential counts. Once any of these signals are received, the rate-of-rise of count rate must be investigated from the analysis of previous readings.

The level of the signal and the rate at which the signal is rising dictates the type of action that the op­erator takes. This varies from placing the channel on single sampling, where the count rate doubles in about 24 hours to immediate reactor trip where the count rate is rising and doubling at a rate of one hour or less, or the count exceeds an absolute limit set usually by the Operating Rules.

The absolute high limits and the maximum rate — of-rise of counts is specified for each BCD system and related to the course of action that the operator must take.

Each course of action for an identified abnormal set of readings on the BCD system will be contained in the Plant Operating Instructions.

In normal operation, frequent checks are carried out on the equipment to ascertain that it is perform­ing correctly and, at less frequent intervals, to ensure its sensitivity to fuel leaks is adequate. The latter may be done by using purposely-contaminated fuel or by the use of probes contaminated to give a measure of sensitivity.

The operating temperatures typical of water reactors are not high in terms of homologous temperatures (<0.4 Tm) so that thermal creep is relatively slow. Furthermore, the creep strength is considerably en­hanced by the presence of dissolved oxygen and hy­drogen present as hydride. In practice, however, the in-reactor deformation seems to be primarily radia­tion, induced. In-pile creep tests show that the creep rate (e) is almost athermal and approximately linearly dependent on the fast neutron flux (Ф) and on the — stress (or) (Wnod and Watkins, 1971 [18]). The me­chanisms werg assessed by Hesketh (1968) [19] who attributed the enhanced in-pile creep to yielding creep [4]. Diameter measurements on PWR fuel rods by Franklin (1982) [20] produced an equation of the form t = к Ф0-6 a1 5.

In ‘fair agreement’ with the results of in-pile creep experiments (k is a constant). In this study, ovalisa­tion of the clad and ridge formation due to the clad taking on the (hour-glassed) shape of the fuel pellets occurred during the second reactor cycle; subsequent (diametral) clad shape changes were controlled by the pellet stack.

In the early days of PWR, problems were encoun­tered with collapse of the clad into large interpellet gaps. These had formed by accumulation of smaller gaps caused by fuel densification (Roberts et al, 1977 [21]). Whilst few failures were produced, such an effect is obviously undesirable and steps were taken to prevent its re-occurrence. These were the use of a higher starting density for the fuel and internal pre — pressurisation of the pin (to about 10 atmospheres, which roughly doubles at operating temperature) to

reduce the driving force for collapse (Frost, 1982 [22]). Experience has shown that these can prevent clad col­lapse into large inter-pellet gaps and, as shown by Franklin’s results, clad/fuel contact is delayed until some way into the second cycle. At the ends of the fuel rod, of course, the plenums always consist of lengths of unsupported tube. In this case clad col­lapse does not occur because both the neutron flux

and the temperature are low, hence there is little scope for either irradiation or thermal creep.

Gro w(h

Just as with uranium (Section 10.1.2 of this chapter), fast neutron irradiation of zirconium alloy cladding results in growth because the vacancies and intersti-

. js which are produced haw different preferences for^sink positions. If (be metal trains were perfectly randomly oriented, this would haw little effect other [han the” generation of high internal stresses. In prac — of course, any metallurgical treatment will result I j” s’ome kind of texture and, although this can be — patrolled to some extent by the fabrication proce­dure in Zircalos tubes г he ‘exture is usuallv sucii dial fast neutron irradiation tends ю make the tubes I oncer and imperceptibly thinner. Other factors which uTect erowth are prior cold work and temperature (Adamson, Tucker and Fileris, 1982 [23]) and, since the process produces rod length changes it is appa­rent that the presence of cross-rod flux and tempera­ture sradients will also be capable of producing dif­ferential length changes which will lead to rod bowing.

Other, less important, contributions to bowing are possible from non-uniform, radiation-induced stress re­laxation (Montgomery, Mayer and French, 1977 [24]). To date rod bowing seems to have caused few failures and this may be due to the very firm built-in support offered by most spacer designs and the fairly short unsupported rod lengths (typically 0.5 m).

The use of high nickel alloys in a PWR (stainless steel, inconel 600) is responsible for the Co-58 pro­duction from Ni-58 during early reactor life. In con­trast, Co-59 present in hard facing alloys and as an impurity in other materials determines the Co-60 pro­duction later in life. The relative contribution of each of these sources to the Co-60 production depends upon operational aspects such as valve seat lapping and upon the established metal corrosion and corro­sion product release rates. There is therefore an ad­vantage in minimising the Co impurity level in inconel and stainless steel, and developing Co-free or very low Co hard facing alloys. Typical Co specification for inconel and stainless steel is 0.1% with inconel available at 0.05%Co.

Table 1.21 shows that since approximately 73% of the surface area is inconel compared to 7% for stain­less steel, and 20% for zircaloy, there is a clear in­centive to pursue lower inconel Co levels, provided the cost benefit analysis is favourable.

Similarly, although only 0.04% of the surface area is hard facing alloy, a typical Co level for stellite is 66% and some material is located in the vicinity of
the core (control rod drive mechanisms) and in valves. Direct release of Co-60 species is consequently pos­sible by wear and corrosion mechanisms. There is therefore a clear long term benefit in developing low or Co-free hard facing alloys, and this is being ad­dressed worldwide. However, apart from the need to resolve questions of availability and fabrication it is possible that a Co-free alloy may wear more exten­sively and require more maintenance. This increased maintenance dose will, to some extent offset the bene­fits of a low system Co inventory.

All the BCD systems installed, from the earliest mag­nox power stations such as Bradwell up to the last at Wyifa, were engineered from the same basic design concept developed for Calder Hall A (commissioned in 1956).

However, with the construction of large magnox reactors, the twin requirements of improving fuel sur­veillance and rationalising the pipework system (par­ticularly within the pressure vessel envelope) resulted in major design development on BCD components, such as standardised precipitators and improved high temperature selector valves. This greatly improved the BCD system overall performance. Examination of the BCD system currently in use has highlighted two dif­ferent design concepts, mainly involving the location of the primary selector valves:

• Positioning of selector valves on or adjacent to standpipe penetrations. These have been taken out horizontally and housed below the pile cap floor, outside the main primary circuit.

• Positioning of selector valves within the primary circuit above the reactor core. These have been taken out horizontally through a special side-mounted penetration.

1.7.5 Typical BCD layout The Wyifa BCD system has been chosen to illustrate generally the design intent of a typical BCD system installation. Like most of the BCD systems Wyifa has three independent operating facilities, each with its own performance regarding sensitivity and scanning frequency. Only the compressor unit is common to all three BCD systems. The complete BCD system pipework arrangement can be seen in the diagrammatic arrange­ment shown in Fig 2.27. The three facilities are com­monly known as bulk, group and single.
hAwMiFE
r SSQN
2 77 -‘OURS З‘0 M:4JTES 32 SECONDS 9.8 SECONDS 3 0 SECONDS?:SECONDS ■ 4 SECONDS 4′ SECONDS ‘tSECONDS ‘ ’ SECONDS ’ SECOND	‘ 7 8 ViNUTES 1 5 4 MINUSES 2’4 M’NJTES ■ 4 MINUTES 8C SECONDS ShOPT ShOpt 32 MiNuTcS 95 UNITES Є6 SECONDS Short Short
Fig, 2.25 Relevant krypton and xenon isotopes

The AGR fuel assembly designs for Dungeness B, Hinkley Point Bt Hunterston В, Hartlepool/Heysham l and Heysham 2/Torness reactors are developments of the basic type proven in the Windscale advanced gas cooled reactor (WAGR) prototype operation since 1963.

They comprise a plug unit and fuel stringer joined together by a central tie bar or coupling to form a

long flexible assembly.

The upper plug unit incorporates a closure unit, gag actuator mechanism, gag and shields and is described in detail in Section 6,3 of this chapter.

As originally proposed, the lower section of the first civil fuel assembly design comprised a number of 36-pin fuel elements of the floating pin type, develop­ments of prototype WAGR 9 and 18 pin configura­tions. However, early developments of this design led to the adoption of a fixed pin element arrangement in which the fuel was attached to the support grid and located radially at central and upper positions by braces mounted from the graphite sleeves which form the coolant channel (see Fig 2.72).

Both improved cluster performance and vibrational stability were achieved by fixing the fuel to the grid and reducing the inter-element gap. Increases in fuel length and reductions in inter-sleeve leakage and ab­sorber content were also possible.

Improvements in fuel pin endurance were also found to be possible when hollow fuel pellets were used in place of the earlier solid design, by assisting in the problem of accommodating fission product gases in providing increased pin voidage and eliminating the hot central core region associated with solid fuel. The use of fuel cans manufactured from 25cold-worked tubing followed by assembly annealing was also found to be beneficial.

Requirements of on load refuelling have necessi­tated the incorporation of charge/discharge and in­pile stabilising features and the use of element anti­gapping devices.

More recently a Stage 2 fuel element has been developed adopting a single thicker graphite sleeve and integrally spot-welded grid and braces, which with additional strength is more impact resistant providing prospects of full power on-load refuelling (see Fig 2.73).

The need for operational control on the radioactive contamination of boilers and ancillary reactor parts has also involved the development and installation of inertial collectors in stringers in order to filter gas — borne particular matter from the coolant.

Each station has active maintenance facilities where plus units and other reactor assemblies (such as con­trol rod assemblies) handled by the fuelling machine can be serviced. Glove boxes with appropriate shield­ing are used at the higher levels, with remote handling tools at the lowrer levels such as at the neutron scatter plug where activity is high. Some sub-assemblies such as closure units or control rod actuators may be re­moved for detail maintenance in special workshops for contaminated components.

The other facilities include a test and training faci­lity which embodies a full scale replication of the reactor charge path for use with the fuelling machine. Fuelling machine maintenance facilities include pro­vision for nose unit, grab and hoist servicing. Pro­vision is made for operations such as fuel element bottle inspection and test before re-use. Recovery equipment for visual inspection of the reactor charge path and removal of damaged components is pro­vided either as part of the fuelling machine or as a separate unit.

8 AGR post-trip heat removal systems and essential electrical supplies

Comprehensive protection systems are provided to shutdown the reactor under fault conditions (see Sec­tion 9 of this chapter), but these actions alone are insufficient to ensure that the reactor remains in a safe long term shutdown state, since heat continues to be generated in the core from the radioactive decay of fission products. This shutdown decay heat is an inevitable consequence of the nuclear fission process und it is entirely outside the control of the operators, wen after a normal controlled shutdown. In order to prevent overheating of the fuel, and the consequential damage to the reactor structure and primary pressure circuit components which could result in a release of radioactivity, it is necessary to maintain a reactor ‘-ooling function following shutdown of the reactor.

The post-trip cooling function is achieved in prin­ciple in a similar manner to that employed to cool the reactor core vvhen at power, by:

• Circulating sufficient coolant gas to transfer the fission decay heat from the fuel to the boilers.

• Providing sufficient feedwater to the boiler systems to remove this heat from the primary circuit and reject it to the environment.

Whilst cooling the reactor core, the plant must be operated in such a manner that the maximum compo­nent temperatures and stresses are maintained within design limits to avoid unacceptable plant damage. This cooling function is effected through the combined operation of a number of plant systems. Some of these are normally operational during power operation of the reactor, whilst others are provided specifically for post-trip heat removal duties, and are shutdown during normal power operation. Automatic sequencing equipment is provided to carry out the large number of operations required to terminate the normal power operational cooling system and to establish the post­trip heat removal systems in service.

The same principles of post-trip heat removal are applicable to all AGR power stations. However, the plant systems differ considerably in detail between stations and Heysham 2 is described in depth, being the latest station and thus most representative of current CEGB practice.