To ensure that those items included in the mainte­nance schedule have a thorough overhaul and are done in a consistent way at each prescribed interval, maintenance instructions for the craftsmen are pre­pared. These will detail the stages of work required during its overhaul to ensure that all items of main­tenance are covered. The maintenance instruction will carry a Quality Assurance procedure that records the checks to be carried out at each stage of the work. ~ The maintenance instructions may be vetted by the Nuclear Installations Inspectorate to satisfy them that the work is of an adequate nature. All the work car­ried out under the maintenance schedule has to be fully recorded and submitted to the Nil by the end of each biennial overhaul.

While a reactor is operating at steady power the chain reaction, described in Chapter 1, Section 4, is just sus­tained and the multiplication constant kerr is exactly equal to unity. If there is any change in any of the items which make up the value of keff, the value of keff will deviate from unity and the reactor power will change. For example, a control rod may be moved up or down; reactor gas temperature may change due to a change in feedwater flow into a boiler, or gas circulator speed may change thereby upsetting the balance of temperature within the reactor core, and changes in temperature lead to changes in keff via the temperature coefficients described in Section 3 of this chapter.

Reactor kinetics is concerned with deviations of keff from unity and the effect that the deviation has on the neutron population within the reactor; if the neutron population changes, the rate of fission will change, so the reactor power will also change. Reactor kinetics is concerned with the magnitude of such changes and the way in which the neutron population varies with time.

2.4 Prompt neutrons

In the fission process, most of the resultant neutrons appear within 10“ 14 s of the fission occurring; these neutrons are called ‘prompt neutrons’.

In Chapter 1, Section 6 the relationship between the multiplication constant к and neutron life cycle was described. Let us consider how the neutron po­pulation caries with time, by following the ecents as the neutrons go through their life cycle. At the be­ginning of the cycle, suppose the thermal neutron population in the reactor core amounts to nt neu­trons, let us call this the first generation. These П| neutrons will be absorbed in the various reactor materials (some will escape from the core), some which are absorbed in the fuel will cause fission and the fast neutrons produced in fission will be slowed down to thermal energies (some fast neutrons will escape from the core, some will be absorbed during the slowing- down phase). When one round of the neutron life cycle has been completed there will be a new thermal neutron population of ny neutrons, let us call this the second generation. The number of neutrons in the second generation may be expressed in terms of the number in the first generation and the multiplication constant:

ny = keff x щ

The change in neutron population from one generation to the next, fin, is then:

fin = П2 — щ = (kerr — 1) x щ (3.10)

The term (kelf — 1) is the deviation of kerr from unity, and is given the symbol Дк:

Дк = keff — 1

So for Equation (3.10) we can write:

fin = Дк x n] (3.11)

In a power reactor the value of kerr does not deviate markedly from the value unity. While the reactor is at power the value of к0rr lies in the range 0.998 to 1.002, so the deviation Дк is very close in value to the reactivity q which is defined as:

kef: ~ 1

e = —————

к ett

Therefore, for all practical purposes, Дк is a close approximation to reactivity, and it is common prac­tice to refer to л к as reactivity.

The timescale over which the change in neutron population on occurs is the time taken to go exactly once round the neutron lifecycle. This is called the ‘mean neutron lifetime’, 1*. and its value depends on the reactor system. The mean neutron lifetime is made up of two components, ‘slowing-down time’ and ‘diffusion time’:

• ‘Slowing-down time’ is the time from production of a fast neutron by fission to the neutron be­coming a thermal neutron.

• ‘Diffusion time’ is the time the thermal neutron spends diffusing through the reactor materials be­fore it is absorbed.

Typical values of slowing-down time and diffusion time in a graphite-moderated reactor are approximate­ly 0.33 ms and 0.66 ms respectively, adding to give a mean neutron lifetime 1* of approximately 1 ms. The time from absorption of a neutron in a fissile nucleus, through the fission process to the fast neu­trons being released, is several orders of magnitude smaller than the slowing-down and diffusion times, so is negligible.

The diffusion time of 0.66 ms may seem surpris­ingly long, since it makes up two-thirds of the neutron lifetime and is unproductive. It is a consequence of two main factors. First, there are many more mod­erator atoms in the core than fuel atoms, the ratio being approximately 100 : I (see Chapter 1, Section 6), so there are more opportunities for the neutrons to collide with the graphite atoms than with fuel atoms, Second, the neutron capture cross-section of graphite is very low (see Chapter 1, Section 5), so it is to be expected that the neutrons will be able to exist in the graphite for a relatively long time before being ab­sorbed. In a PWR, by comparison, the ratio of mod­erator atoms to fuel atoms is much lower, approxi­mately 4:1, and the neutron capture cross-section of water is higher than that of graphite, so the diffusion time in a PWR is shorter than in a magnox reactor or an AGR. The slowing-down time in a PWR is also shorter than in a magnox or an AGR, because it requires fewer collisions to slow down a neutron in a hydrogenous moderator than in a graphite moderator. Thus the mean neutron lifetime 1* in a PWR is shorter than in a magnox or an AGR, a typical value of 1* for a PWR is between 10-i and 10~4 s. It will be seen later that the time-dependent behaviour of the neutron population in a power reactor is most strongly influenced by factors other than mean neu­tron lifetime, so these differences between PWR and graphite-moderated reactors are of little consequence in practice.

The change in neutron population fin takes place over a time interval fit which, in the case w:e have considered in Equation (3.11), is the mean neutron lifetime I*. Thus the rate of change of neutron popu­lation, from Equation (3.11), is given by:

which give rise to delayed neutrons (in the case of Br-87, this is 2%) and divided by the total number of neutrons released per fission (because the fission yield is per fission, but delayed neutron yield is to be expressed as per neutron). Thus ue ha e:

2,6’ro x 2го

debased neutron yield = ———————- = 0.02 PP

2.438

• ‘Mean life’ is related to the halflife of the pre­cursor. Mean life is similar in concept to a time constant in other branches of science and engineer­ing which deal with exponential behaviour, see Fig 3.17, Mean life is related to halflife by the factor log,, 4 as follows:

halflife = mean life x logn 4 (3.14)

Br-87 has a halflife of 55 s, so the mean life of the delayed neutrons is 79 s. Note that the term ‘mean life’ does not refer to the life of the neutron, it refers to the delay between the fission occurring and the delayed neutron being formed.

There are over 50 fission products which can give rise to delayed neutrons in this way. For convenience they are grouped initially into six groups, each group characterised by a yield and a mean life (Table 3.1). Table 3.1 shows that Br-87 accounts for virtually all the delayed neutrons in group 1.

It is convenient to combine the six groups into one

average group as follows:

6

total yield, 3 = ^ d, = 0.68^0

і = і

The mean life is determined by weighted averaging:

6

З-, x ті

…. i = i

mean lire, г = ————— = P 9 s

6

Although delayed neutrons account for only 0.68 °"o of the total neutron population in the reactor, their mean life of 12.9 s is very much longer than the mean neutron lifetime of prompt neutrons, 1 ms, so the delayed neutrons have a very marked effect on reactor kinetics. An ‘effective neutron lifetime’, le, may be calculated:

U = 1* + (£ — Дк) x r (3.15)

In this calculation, care must be taken to express 3 and Дк as ‘per unit’ values. For example, calculate. the effective neutron lifetime in a reactor with a reactivity of +-50 mN:

le = 1* + (3 — дк) x t

= 0.001 + (0.0068 — 0.0005) x 12.9 = 0.082 s = 82 ms

The effective neutron lifetime is applied to the neu­tron kinetics equation as follows;

Compare the equation for prompt neutrons only, Equation (3.13). Equations (3.13) and (3.16) are shown in Fig 3.18 for a reactivity of +50 mN. The marked effect of the delayed neutrons is clearly demonstrated.

In reality, the time-dependent behaviour of the, neutron population is a combination of the prompt and delayed neutrons, Consider a reactor operating at steady power, then suddenly being subjected to a

step change in reactivity of +200 mN. The effect is as shown in Fig 3.19. The neutron population in­creases rapidly at first, as the prompt neutrons re­spond quickly to the change in reactivity. The prompt rise is followed by a steady increase, as the delayed neutrons respond slowly. Figure 3.19 also shows the effect of a -200 mN step in reactivity, namely a prompt drop followed by a steady decay.

An interesting result is obtained from Equation (3.15) if a reactivity of +680 mN is inserted. In this case the effective neutron lifetime le becomes equal to the mean neutron lifetime 1[32], i. e., 1 ms. This is known as ‘prompt criticality’, and the result is that the reactor will diverge as though the delayed neutrons did not exist. In power reactors it is of academic interest only, for the following reasons:

• There is no credible means by which a net reactivity

of ^680 mN can be created.

Fjg. 3.19 Effect on the reactor of a step change
in reactivity

of neutron population. Not only will this tell him how quickly the reactor power is changing, but it will also indicate the magnitude of the disturbing influence, i. e., the deviation of kerf front unity. We have seen above that the neutron population changes exponentially under the action of a given value of reactivity дк. Clearly the simple rate of change of neutron population 6n/St will change in value as the exponential proceeds, see Equation (3.12) and Fig 3.19, so a meter showing this parameter would be of limited value. In dealing with an exponential change it is convenient to take the logarithm of the func­tion, for example, take the logarithm of Equation (3.16):

logn = (n/no) = Дк/1е x t (3.17)

Then the differential of Equation (3.17) gives a rate
of change which is constant for a given value of Дк:

4.3 Practical operation

4.3.1 Rate of change instrument

It is useful for the reactor operator to have an in­dication on the control desk of the rate of change

o/6t [logn (n/no)] = Дк/Іе (3.18)

By this method, a useful rate of change indication can be presented to the operator. Note that it is directly proportional to the deviation Дк. The units

of this rate of change parameter are not immediately obvious, and a slightly different approach yields a more useful unit as follows.

Returning to Equation (3.16), consider the time taken for the neutron population to increase from its initial value n(> at time zero to a value e x n0 uhere e is the base of natural logarithms (e = 2." 18, etc.). In Equation (3.16). set n to the alue e x n, j and determine the time T to reach this value:

e x no = n0 exp( Д к 1L, x t) therefore T = le /Д к

This time period for the neutron population to change from some initial value to e times that initial value is called the ‘reactor period’;

reactor period, T = 1е/Дк (3.19)

Note that the reactor period is the inverse of the rate of change parameter given by Equation (3.18). Equation (3,16) can now be rewritten:

n = no exp(t/T) (3.20)

A Tate of change’ meter on the reactor control desk which derives its signal from the differential of the logarithm of the neutron population (more precisely, of the neutron flux, see Section 6 of this chapter) can now be calibrated in units of reactor period. A ariation on reactor period is ‘doubling time’, i. e., the time taken for the neutron population to double. From Equation (3.20), determine the time taken for the neutron population to double from its initial value n0 to 2 x no:

2 x no = no exp(t/T) (3.21)

therefore doubling time = reactor period x Iogn2

Thus the principle of our Tate of change’ meter is the same, but the scaling is different. Doubling time is a more convenient unit because it is easier to

understand.

One of the properties of a pure exponential is that us doubling time is constant for a given value of Дк, ‘■e., the reactor will take the same time to double from 1 MW to 2 MW, 2 MW to 4 MW, 4 MW to 8 MW, etc., so for a constant value of Дк the dou­bling time meter will settle to a steady reading during the transient. Figure 3.20 shows a scale for a meter to display doubling time or reactor period. For com­parison, Fig 3,20 also shows corresponding values ot the rate of change parameter k/le given by Equation (3.18), it also shows the corresponding val — aes of reactivity дк.

It should be noted that the numerical work neces­sary to produce the above results for the effect of delayed neutrons has been carried out using an aver­age group of delayed neutrons with a yield of 0.68ro and a mean life of 12.9 s. The true behaviour of a power reactor is more complex, with many delayed neutron groups as mentioned in Section 4.2 of this chapter. The method of combining the delayed neu­trons into one average group is adequate for many purposes, particularly when studies are confined to small values of the deviation Дк. Table 3.2 gives a comparison of doubling times derived from a ri­gorous treatment of delayed neutrons and doubling times derived from one average group using Equations (3.15), (3.19) and (3.21).

Comparison of the single average group model with a 6-group model (see Table 3.1) is given in Fig 3.21 for a step change in reactivity of — 100 mN, similar to that shown in Fig 3.19. It can be seen in Fig 3.21 that there is very little difference between the two models.

0 25 50

ГМЕ SECONDS

Fig. 3.21 Comparison of 6-group and single average
group models

Reactor gas inlet temperature can be assumed, for the time being, to remain constant. If a change in gas flow occurs, for example, because of a blower/circulator failure, an increase in fuel temperature will occur. The negative fuel temperature coefficient of reactivity will cause reactivity to become negative so neutron power will decrease, this is an advantage because it helps to arrest the rise in fuel temperature. The increase in fuel temperature is transmitted to an increase in channel gas temperature and, where an auto control system is fitted and in use, the regulating rods will be driven in by the control system. This action will cause neutron power to be further de­creased until channel gas outlet temperatures are re­stored to the demanded value, and at this stage the reduction in reactor power will be just sufficient to match the reduction in gas flow.

Auto control systems on channel gas outlet tem­peratures are normally designed to cope with faults such as blower/circulator failure, so that there is little required of the reactor control engineer in the short term (as far as reactor parameters are concerned, although he has several operations to carry out on the failed blower/circulator and its associated boiler) other than to check the correct operation of the auto control system. This is the case on stations with six or eight blowers/circulators per reactor, where the loss of gas How amounts to one-sixth or one-eighth of the total, and on AGRs where the characteristics of the reactor and the more sophisticated control and protection systems allow rapid rates of change of reactivity. However, on magnox reactors with four blowers per reactor and on those with once-through boilers, prompt action by the reactor control engi­neer to assist the auto control system by insertion of manually-operated rods is probably required. Such action is also helpful on the six — and eight-blower sta­tions; on stations without auto control or where the auto control is not in use then prompt operator action is essential to prevent an undue rise in temperature.

So far we have considered only the first few min­utes of the transient. As time proceeds any remaining mismatch between the actual and desired power will cause the moderator temperature in a magnox reactor to change. Moderator temperature changes slowly, so an auto control system can readily cope with such changes. If the mismatch is such that reactor tem­peratures are too high, control rods can readily be inserted to restore the desired temperatures, if how­ever the moderator temperature is too low, for ex­ample, because of over-insertion of manually operated control rods, problems can arise either where the auto regulating rods are running only lightly inserted into the core or where no auto control is available. In these cases the rate of reactivity release from control rod withdrawal is small, and the reactor may tend to shut down on ‘temperature poisoning’. Further action which may be taken by the reactor control engineer to recover the situation is described in Section 5.5.4 of this chapter. Good operator training minimises the risks of such a problem occurring, particularly as the high reliability of plant in recent years has meant that such faults are rare and reactor control engineers may not experience such a fault on the plant for many years.

If the reduction in gas flow and reactor power are sustained for a period of hours, changes in Xe-135 concentration must be considered. As neutron power is reduced the xenon concentration increases (see Sec­tion 2 of this chapter), adding negative reactivity to the core. The rate of change of reactivity produced by this effect is slow and therefore compensation is easily achieved initially by rod withdrawal. However the total reactivity change, reached about six hours after the start of the transient, will be large if the neutron power change is large, and there may be insufficient rod worth available to offset the change in xenon worth and maintain criticality on the re­actor. Further action which may be taken by the reactor control engineer to provide additional posi­tive reactivity to offset the xenon change is described in Section 5.5.4 of this chapter. If sufficient positive reactivity cannot be found to offset the increase in xenon worth, the reactor will tend to shut down on ‘xenon poisoning’.

So far it has been assumed that a rise in reactor gas outlet temperature above the pre-fault value can­not be tolerated. However, if the pre-fault value is below the limits set by the Operating Rules, for ex­ample, in a magnox reactor if the reactor is operating to corrosion control limits which are more restric­tive than the Operating Rule limits, then a short-term increase in reactor gas outlet temperature up to the limits set by the Operating Rules is tolerable. In this case if it is found that the reactor is shutting down due to ‘poisoning out’, the reactor control engineer may consider it worthwhile to further reduce gas flow by adjustment on the healthy gas circuits, thereby inducing an increase in moderator temperature. This option is not available on an AGR where the mod­erator temperature is held constant by the re-entrant gas flow. It is a difficult operation because of the risk of tripping on excess temperature and because two variables are being changed at the same time, also because the control rods are only lightly inserted into the core so they will be of little value in minute — to-minute control. However it is worth trying as a last resort if the reactor can be kept safely at power.

The design of British reactor systems incorporates the feature that the reactor operator is not part of the safety chain for reactor incidents. This philosophy places reliance on safety systems to detect unaccept­able conditions and to take corrective action within defined time limits.

Such safety mechanisms need to demonstrate their operation at defined intervals to give confidence in their reliability. Principally the safety mechanisms that receive the most attention are those associated with temperature protection. The following items come gen­erally under the heading of safety systems:

• Fuel temperature trips.

• Flux measurement trips (absolute and rate of change).

• Counter trip (absolute and rate of change).

• Barometric devices.

• Loss of reactor pressure — both rate of change and over/under pressure.

• Standpipe closures.

• Dome differential pressure (AGRs).

Many of these systems are tested on a frequent basis, usually monthly, and the results witnessed and as­sessed by an insurance inspector. The results of the tests are certified and the certificates signed by the Station Manager, Production Manager and Engineer­ing Manager.

The tests normally use injection methods to simu­late the source term. The final interruption of the output relay or guard line is prevented so as not to trip the reactor at each test. The remaining part of the chain is tested normally during the reactor’s biennial overhaul to prove its reliability.

Barometric devices are normally tested either in a rig or, in the case of boron balls, during the outages. Similarly, although not included in the above list, control rod drop tests are done during shut down. The test for control rods includes a measurement of its characteristic drop velocity over its range of insertion.

The list includes reactor standpipe pressure closures and although these cannot be strictly interpreted as safety devices, they are considered strategic enough to give a minor loss of coolant defect should they fail. They are therefore closely inspected following the repressurising of a reactor and after each refuelling operation to ensure a gas-tight seal and correct op­erating position.

Experiments w*ere carried out during the early devel­opment of AGR to measure G(-CH4) (the number of molecules of methane destroyed per 100 eV of energy adsorbed) both in small capsules in various radiation by test facilities and in the Windscale AGR. The bypass plant for the early AGR were based on this data, which showed that the rate of methane destruction was dependent on the radiation dose ab­sorbed, the methane concentration, the inverse of the carbon monoxide concentration and a small depend­ence on the inverse of the water concentration. In addition it was found that a small background meth­ane concentration was formed even if no methane was added. It was also demonstrated that G(-CH4) depended on the presence or absence of graphite. This work was extended during the early operation of Hinkley Point В AGR and the major controlling para­meters confirmed. The resulting relationships are given by:

G(-CH4) = N84 |{CH4) — (СН4)ь exp <-2518.T)]/(CO) For in-pore gas

G(-CH4) = 3232[(CH4) — <CH4)b exp (-2467/T)J/{(CO)

* 64.9 (H:0)]

where (CH4), (CH4)t>, (CO) and (EGO) are in kg/ kg СО?, (СН4)ь is the background methane concentration at the given CO and H;0 concentration in the absence of methane addition, and T is the temperature K.

To calculate the overall methane destruction rate the irradiated gas in the reactor is divided into volumes of isodose and isotemperature and the individual methane destruction rates integrated using

dCH4/dt = 5.976 x КГ-‘ x MDG <-CH4) kg h

where M is the mass of CO:, kg D is the dose rate, w g

The equivalent mass of carbon monoxide and water formed, required for the design of the recombination unit and drier, is then calculated from CH4 + 3CO; 4CO + 2H20.

Chemical and volume control system (CVCS) The CVCS with its main subsystems, the boron thermal regeneration system (BTRS) and boron recycle system (BRS), supports the RCS in the following areas:

• Control of RCS inventory.

• Maintenance of chemical specification in terms of lithium, boron, hydrogen and anion/cation concen­trations.

• Addition or removal of boron for reactivity control.

• Control of activity in terms of corrosion product (soluble and particulate) and fission product removal.

• Recycling boron as appropriate.

The CVCS can be regarded as a chemical control loop and is shown diagrammatically in Fig 1.64. The flow rate is controlled by two normal letdown ori­fices (each rated at 17 mVh although much higher on some plant) and the excess letdown orifice (typi­cally 10 mVh), which reduce system pressure to the CVCS and may be used in any combination. For a typical coolant volume of 240 m3 and a core flow of

18.5 x 103 kg/s, the CVCS flow would be between 0.025% (normal) and 0.040% (maximum), which re­cycles the coolant inventory in 14.2 h.

The CVCS letdown flow is first treated for ionic impurities (fission and corrosion products) by a flush — able mixed bed demineraliser, with the cation resin (40%) in the (Li-7)+ form and the anion resin (60%) in the hydroxyl (OH)- form. The cation resin removes impurity cations as soluble and active species and the Li-7 form is used rather than H-t — form to minimise any pH change. The anion resin is converted into the borate form in operation, implying a small loss of borate from the system initially, and the borate sub­sequently exchanges for impurity anions. The mixed bed demineraliser is designed to reduce the concen­tration of isotopes (except Rb and Cs) by a factor of 10, and is sized to remove fission products and corro­sion products from one fuel cycle with 1% defective fuel. The resin beds also act as crude mechanical filters for particular material. There are two mixed bed demineralisers with one normally on standby and, if necessary, a resin charge can be replaced with the plant on load.

A flushable cation resin bed in the hydrogen form is located downstream of the mixed bed deminerali­sers and is used intermittently on the letdown flow to control the Li-7 concentration. The separate cation unit is sized on the Li-7 production in the core (from the B-10 (n, a) Li-7 reaction) for base-load operation. It also has sufficient capacity to maintain Cs-137 con­centration in the coolant below 0.03 /iCi/cc with 1% defective fuel. The unit is designed to reduce all cation impurity concentrations by a factor of 10. Downstream of the ion exchange units is a filter de­signed to collect resin fines and particulates of 25 fim or greater.

Control of gaseous fission products, including tri­tium, and maintenance of the dissolved hydrogen lev­el is carried out in the volume control tank of the CVCS. The primary coolant is sprayed into a hydro­gen atmosphere at 1.03 bar in order to strip fission products (noble gases such as krypton and xenon) and produce a 25-50 cc (STP) Нг/kg H2O concentration in the circulating primary coolant. Waste H2 contain­ing fission products is passed to the gaseous waste processing system (GWPS).

The CVCS contains two boric acid tanks for stor­age of sufficient 4-4.5 wt % boric acid solution for refuelling, plus safe shutdown from full power opera­tion with a new core and most of the control rods not inserted. A batching tank is provided for making up this solution and the contents of the boric acid tanks will be oxygenated due to equilibration with air. However, flow into the primary circuit is small compared with the coolant inventory, and the com­bination of oxygen stripping in the VCT, dissolved hydrogen and radiolytic reactions will ensure that the RCS does not become significantly oxygenated by boron injection.

The CVCS also provides preparation and injec­tion facilities for LiOH solutions for pH control, and N2H4 solutions for chemical de-oxygenation prior to

start up. It is noted that N2H4 addition leads to the formation of NH3 and under these conditions the CVCS demineraliser beds would be bypassed in order to avoid adsorption of NH4+ and release of Li + .

Reactor make-up water system (RMWS)

This system supports the RCS chemistry control by providing a source of high quality water to maintain coolant inventory. The system receives water from the degasifier tank of the demineralised water supply sys­tem, and also from other recycle sources in the bo­ron recycle system. The water is stored in a large tank at atmospheric pressure, covered by a movable diaphragm to prevent egress or ingress of gases and protected against freezing by heaters. The water may
be recycled through a degasifier and demineraliser to meet the specification given in Table 1.26 and if ne­cessary can be purified by re-evaporation.

Access to the core is via standpipes on the reactor pressure envelope. When not in use the standpipes contain shield plugs which provide a pressure seal and preserve the biological shielding. Standpipes are also necessary for control rod mechanisms. In mag­nox reactors the ratio of fuel channels to control rods is approximately 16:1. Because of the difficul­ties experienced at Calder Hall where the same stand­pipes were used for both control rods and refuelling, the early magnox stations had separate refuelling standpipes. At Hunterston A, the refuelling stand­pipes were placed below the core for bottom refuel­ling but this led to a costly increase in the height of the reactor complex. To minimise the number, each refuelling standpipe must serve a number of channels with the greatest number being 64 at Berkeley. Be­fore refuelling, a chute with a parallel motion dog­leg must be inserted. The FM can rotate this chute and move the dog-leg to align with the desired chan­nel (Fig 2.32). Because the fuel elements are too long to negotiate the dog-leg angles, the hoist has to be stopped and the leg moved in or out as appropriate whenever the grab is engaged.

ing requires reliable first time engagement. There are two independent sets of three jaws. Should one set rest on the fuel element sprung arm spider, the second set operates correctly. The operating plunger motion is controlled by the interrupter which gives alternately long and short strokes. The grab is suspended on a wire rope formed around a stainless steel spiral in­corporating an insulated conductor which serves the locking solenoid and provides a ‘jaws engaged’ signal.

In both magnox reactors and AGRs there are sources of water that include release of water already absorbed within the reactor graphite core and, in the case of the AGR, the radioiytic breakdown of methane. The radiolytic/thermal breakdown of oil entering the sys­tem and boiler leaks that allow water into the coolant circuit are causes of an increase in moisture level that are indicative of plant faults.

Gas driers are installed and blowdown is used to control moisture concentrations below that which might cause corrosion of reactor internals, or affect the performance of the failed fuel detection system discussed in Section 1.7 of this chapter.

Measurement of the bulk moisture level is neces­sary to check on the performance of the moisture control systems and to provide data used in calcula­tions of moderator integrity. Typical concentrations are quoted in Table 2.12, i. e., around 10 vpm for magnox reactors. Measurements are required within

±5% accuracy or ±2 vpm, whichever is the greater, and response time is a secondary consideration.

A number of different methods of water in CO2 measurement are available and have been used. No single method is universally applicable and develop­ment and testing continues. The principal online meth­ods are infra-red gas analysers, electrolytic, impedance, piezoelectric and dew-point hygrometers:

Infra-red gas analysers The magnox reactors were initially equipped with these devices, which depend on the principle that water vapour absorbs quite strongly in the infra-red region of the electromagnetic spec­trum near wavelengths of 1.87 and 2.7 дт, and use is made of this for the quantitative measurements of moisture in gases. Conventional infra-red absorption measuring techniques apply, the extinction being pro­portional to the partial pressure of moisture within the sample cell. Cross-sensitivity effects can usually be overcome using interference filters.

Infra-red hygrometers require calibration, but once calibrated are capable of achieving relative accuracies in the order of ±2 vpm in the range of 0-100 vpm. Moisture concentrations up to 4000 vpm can be mea­sured, whilst below 100 vpm sensitivity can be im­proved by operating the cell at pressure.

Principal disadvantages of this type of hygrometer are its relative complexity and need for high sample flow rates (several litres/minute) to achieve a reason­able response time. This is because of the long path — length cells required for moisture measurement.

Coulometric (electrolytic) The moisture in sample gas flowing through an electrolytic cell is absorbed on a film of partially-hydrated phosphoric anhydride in contact with platinum electrodes. A DC voltage greater than the decomposition potential of water is applied to the electrodes and the absorbed water is quantitatively electrolysed. The cell current produced on electrolysis is directly proportional to the rate of water absorption (Faraday’s law), and for a given flow rate the concentration of moisture in the gas may be derived on an absolute basis.

Coulometric analysers are continuous and, for most general purposes, may be regarded as providing an absolute measure of moisture concentration provided certain precautions are taken, particularly the careful control of the sample flow rate. With such precau­tions, commercially available instrumentation can be capable of accuracies better than ±5% reading or ±2 vpm, whichever is the greater.

When used for the measurement of very low mois­ture levels, for example, in high temperature helium — cooled reactors, the current is switched off for a defined period to allow moisture time to collect, thus making the device more sensitive.

Coulometric analysers have found extensive use in the CEGB, particularly for reactor coolant gas mo­nitoring. Main disadvantages, however, include the following:

• They are limited in use to moisture concentrations below 3000 vpm.

• Cells have a finite life, dependent upon operating conditions, before regeneration becomes necessary. Ceil failure is not always immediately apparent unless regular checks on performance are carried out.

• At high hydrogen (and oxygen) concentrations, platinum electrode cells suffer from recombination effects. The use of gold or rhodium electrodes has proved to be only of limited success in overcoming this.

Impedance type In this device the AC impedance of the aluminium-oxide insultant between two elec­trodes is measured, this impedance being a function of water in CO2 concentration. Limitations of this system include poor time response at low levels of water, temperature-dependent drift and memory ef­fects, but a new silicon-oxide type is claimed to have a much superior performance. These types are not absolute and require calibration.

Piezolectric hygrometers These devices rely upon the change in weight and consequent change in resonant frequency of a crystal coated with a hygroscopic film. These changes are detected on alternate exposure to a dry gas and the sample gas, and the changes in resonant frequency are processed to provide a direct readout of moisture concentration. The device is not absolute, prior calibration is necessary and periodic single point checks are required during service.

The principal advantages of this type of hygrometer are its wide range of operation and fast speed of response; in practice, the latter is limited by the flow switching cycle, which is usually 30 s. This type has potential as a differential analyser suitable for boiler leak detection or location. Early indications are that differentials of better than 5 vpm against a back­ground of 300 vpm may be determinable. The per­formance improves still further at lower background moisture concentration.

Mirror type dew-point hygrometers The dew or frost point of a gas may be determined directly by ob­serving the temperature at which formation of a dew occurs on a mirror surface cooled by a semiconduc­tor device. The dew is detected optically and control is effected by an automatic instrumentation system. High accuracy is attainable with commercially avail­able equipment, and this has led to its adoption as transfer or intermediate standards because the mea­surement is absolute.

Ab0Ve -40°C, accuracies attainable can be better than 0.2°С, but at lower dew points errors increase largely as a result of the difficulty of achieving ade­quate" dew formation.

The main disadvantages with respect to this type of h>urometer are its relatively high cost and the fact, tia[ h is vulnerable to interference from condensables other than moisture in the gas, for example, relatively minor concentrations of oil in the gas being sampled can give rise to problems.

As with the viewing devices, developments in the technology associated with the remote control and ar­ticulation of manipulator arms will be applied to inspection manipulators, particularly if the inspection requirements change after the inspection routes for a specific reactor have been fixed.

Storage, testing, training and maintenance

As ISI equipment is only used for a short period eery year, facilities have been provided for equipment storage, testing and maintenance, and also to allow for operator training. Such a facility may also be used to rehearse special manoeuvres that are not part of the routine reactor inspections.

(PWR)

10 Pressurised water reactor and associated systems

Пн* CEGB pressurised water reactor (PWR) power stations will, as Гаг as practicable, be to a common design. This approach confers advantages in minimis­ing the cost of redesign and development work and in avoiding the need to repeat safety analyses. However, for the purpose of the following discussion, attention will be focussed on the Sizewell В station to be built immediately alongside the existing A (magnox) station.

Sizewell В is rated at 1245 MW(e) gross and incor­porates a 3425 MW(t) 4-loop nuclear steam supply system (NSSS) supplying twin 3000 r/min turbine- generators. The NSSS is based on a well-established

Westinghouse design of considerable maturity and proven performance.

The layout of the NSSS, the arrangement of the plant buildings and the design of most of the aux­iliary plant and systems are based on the standardised nuclear unit power plant system (SNUPPS) design produced by Bechtel Power Corporation in associa­tion with L’S utilities. Two stations (Callaway and Wolf Creek, which first raised power in 1984 and 1985 respectively) have been built in the United States to the SNUPPS design, which is itself based on the Bechtel Generic Design that has been adopted for eleven PWR units in the US and other countries.

The PWR system utilises an indirect heat transfer cycle, in which the primary or reactor coolant is used to heat the secondary fluid which, in turn, drives the turbine-generators. The NSSS comprises the reactor, the reactor coolant sy-vm and a number of auxiliary and safety systems. The reactor coolant system (RCS) consists of the reactor in its pressure vessel and four cooling loops, connecting the reactor vessel to the steam generators (boilers) which form the interface with the secondary circuit. The NSSS is located wholly within the reactor building or primary containment. The turbine-generators, located in the separate turbine house, receive nearly-dry saturated steam from the steam generators through pipes which penetrate the reactor building wall.

The reactor vessel and the cooling loops are filled with demineralised natural water which is pumped through the reactor, over the fuel in the core and through the steam generator tubes. The water takes heat from the core and gives it up, in the steam gen­erators, to the secondary (feed) water surrounding the tubes. The feedwater is boiled to produce saturated steam at about 285°C and 72 bar. The primary water, which is slightly hotter (about 325°C at reactor exit), is held at a pressure of 155 bar to prevent boiling. This pressure is controlled by a control system and a part-full vertical surge tank (the pressuriser) connected to one of the loops.

The reactor itself comprises the core containing the fuel assemblies, together with components which support and locate them and direct the primary cool­ant flow over the fuel, as well as guide the control rods which are used to alter the power output. The reactor pressure vessel provides a high integrity en­velope to confine the reactor and its coolant, facili­tates the entry and exit of coolant, and allows access to the fuel from time to time for its replacement. The vessel is manufactured of carbon steel, clad inter­nally with stainless steel, and is housed in a vertical cylindrical cavity at the bottom of the reactor build­ing. This cavity is surrounded by a massive rein­forced concrete wall which supports the weight of the vessel and provides shielding against the intense radiation emitted from the core.

Outside this primary shield wail are the four com­partments containing the coolant loops. Each loop comprises horizontal hot leg and cold leg pipework, welded to flanged nozzles near the top of the reactor vessel, which take coolant from above and supply it to the bottom of the core respectively. The hot leg is welded to the steam generator inlet and the cold leg to the outlet of the reactor coolant pump. A U-shaped crossover leg completes the loop by connec­ting the steam generator outlet to the reactor coolant pump inlet.

The steam generators comprise vertical shell and inverted U-tube heat exchangers, the bundle of U — tubes connecting the inlet and outlet sides of the pri­mary w-ater channel head. The heat transfer surface of the U-tubes is situated well above the core, be­cause of the arrangement of the primary circuit com­ponents, and this helps to ensure adequate natural circulation of primary coolant to remove decay heat from the fuel w-hen the reactor coolant pumps are not running.

The loop compartments and the pressuriser cell are surrounded by the secondary shield wall, outside which is an outer annular zone containing auxiliary equipment. All high energy pipework and equipment is contained as far as practicable inside the second­ary shield wall. The outer annulus experiences lower temperatures and radiation levels during power op­eration and personnel access is permissible for limited periods.

Principal items of auxiliary equipment inside the reactor building include the following:

• The fuel handling equipment, which transports new and irradiated fuel and core components between the core and the fuel transfer tube and between the reactor and fuel buildings.

• The polar crane for handling the reactor vessel head, its muiti-stud tensioner and other major com­ponents, as necessary for refuelling and maintenance operations.

• Heating, ventilating and air conditioning (HVAC) equipment.

• The accumulators and recirculating sumps of the emergency core cooling system (ECCS).

• The spray headers and supply pipework of the containment spray system.

• The emergency boration system equipment.

• The incoming and outgoing lines, ducts and cables to and from reactor and ancillary equipment that penetrates the containment boundary, with isolating valves or dampers as appropriate.

All these auxiliary systems and equipment, together with other features of the power station, are described in the following sections. After a description of the layout of the site and of the major buildings, together with a discussion of measures adopted to control the spread of radioactive contamination, the major re­actor coolant system components are discussed in some detail, dealing in particular with those aspects which ensure that the necessary very high level of integrity is achieved. As well as the auxiliary items listed above, the various control and protection sys — icrns which contribute to the safe operation of the station and the avoidance or mitigation of faults are

described.

The design philosophy with regard to removal of post-trip heat and assurance of power supply integrity js of particular importance for the Sizewell В design and layout, and this is discussed in some detail.