In order to reduce to a minimum the radiation expo­sure to operators performing inspection of the steam

generator primary side components, sophisticated re­motely operated robotic devices are provided. Inspec­tion of the channel head welds and nozzle penetrations, and the steam generator tube bundle are all performed without recourse to man access into the channel head. The equipment is introduced into the channel head following removal of the manway access covers. The various inspections required are carried out using tools and effectors, attached to the robotic arm, designed for specific inspection tasks.

Ultrasonic inspection of the steam generator second­ary side shell welds and nozzle welds is performed us­ing remotely operated track-mounted equipment, which is controlled from a low radiation area.

1.2.1 Operation

The object of a nuclear power station once it has been constructed and commissioned is to be a source of reliable and economic electrical power. For it to re­coup the high capital cost it must run on base load at its full output with good availability exceeding 80%. Base load assists in high availability with less wear and tear than that experienced in plants which cycle in output to meet load demands. The availability achiev­able will depend to some extent on the design of the plant, the design of the pressure vessel, the pressure and the reactor and steam cycles function, the degree of flexibility of refuelling (i. e., whether power reduc­tion is necessary, as is currently the case for AGRs) and the flexibility of access to reactors and boilers, a factor which will offset the time required to com­plete the biennial shutdowns. The loss of availability due to these factors is predictable once experience has been gained, and by that experience systems and methods of work have been developed to keep outage time to the minimum consistent with an acceptable standard of performance.

Less controllable factors may have a bearing on availability:

• Breakdown of plant which may affect reactor op­eration, safety of plant, levels of generation or could cause an environmental hazard, or a breach of the operating rules. Breakdown must to a large extent be controlled by maintenance philosophy and the degree of routine maintenance carried out.

• Restrictions on output may be caused by a number of reasons ranging from imposed limitations on reactor operation to plant restrictions caused, for instance, by circulating water temperature limits. [28] of a reactor. The success of handling a fault condi­tion is dependent on the skill of the operator, and the factors which bear on this include his experience and training. Base load working poses problems for operators to gain real experience in the manoeuvr­ing of reactors operations and plant response.

• Human error in the form of ill-conceived operation or mistake could lead to loss of generation. Reactor safety, however, is never in doubt with human error, since reactors are subject to an extensive system of protective devices which prevent the operator run­ning it in an unsafe condition. The degree of human error is conditioned also by experience and training.

There are a number of activities with which the op­erator is primarily concerned in operating a nuclear site. The extent of his involvement in some of them is dependent on the time of day or week, i. e., is in and out of working hours. The following represents the principal areas of his involvement:

• Operation of plant to produce electricity must be the foremost duty of the operator, with due regard to safety and security of the electrical system. In carrying out this duty he must have due recognition of the site licence and the operating rules provided under that licence.

• Refuelling is an important job in securing the full potential of a nuclear reactor. Equilibrium fuel rates must be achieved to establish a fairly homogenous performance across the zones of the reactor. The pattern of refuelling is determined by the station physics department to achieve good trim.

• Preparation of work for maintenance, inspection, test or other reason is carried out by the operations department. This may involve isolation and the is­sue of safety documents which may contain health physics precautions.

• Health physics requirements in carrying out work are part of a Senior Authorised Person’s (Nuclear Radiation) duties. That role is always performed by the operations department and engineering staff are certified to do this following an assessment test.

• Discharge to the environment of both chemical and radioactive waste is under the direct control of the operators, but both the health physics and chem­istry departments have a major responsibility in this area. Chemical waste results mainly from the opera­tion of water treatment plants. However, radioactive waste may be gaseous, liquid, or solid and results principally from reactor operation and irradiated fuel storage. Limits are placed on the licensee under the appropriate legislation.

• Emergencies need to be dealt with. These may range from malfunction and breakdown of plant to release of radioactive material due to the fracture of the primary containment of a reactor pressure vessel. In the latter case an emergency plan exists which must be provided under conditions of the site licence.

• Security of the site and stores containing radioactive material including new fuel. The former is of more significance during the silent hours when the Charge Engineer carries the major responsibility.

1.2.2 Maintenance

The operation of a nuclear power station results in the need for maintenance to keep plant and equipment in a aood operational and reliable condition, with the aim of good availability and proper performance. There are four areas of maintenance which have a direct bearing on the performance of a nuclear site; mechanical, electrical, instrumentation, and civil. Any industry requires such input, but in the case of a nuclear site and the operation of nuclear reactors each has a significant role to play. All are involved in some way with the protection of the reactor sys­tem. In some parts of the plant and equipment a high degree of maintenance and standard of workmanship is called for, and the most significant of these are listed in the maintenance schedule under the conditions of the site licence. One of the major problems for a maintenance engineer is that of obsolescence, where a worn out component is no longer available. In these cases the equivalent component has to be treated as a modification and to undergo the assessment and test­ing called for by the site licence.

Maintenance work generally covers the following areas:

• Routine maintenance of all plant and equipment to a frequency determined by good sense and ex­perience and conditioned by the maintenance sche­dule under the conditions of the site licence.

• Routine testing of systems that have a safety sig­nificance, one of the most important areas being that of reactor protection. The principal department involved is that of instrument maintenance and in particular the w’ork involving safety lines (also referred to as guard lines).

• Biennial overhauls of reactor plant including inter­nal inspection of reactors and boilers. There is a site licence condition requiring a reactor to be inspected, at the most, two years following its consent to be started up. At the end of the two years the consent to run the reactor is revoked. The biennial overhaul also includes testing of plant including NDT of essential components of the re­actor pressure vessels. [29]

• New work is always a requirement at stations some time during their lifetime. This may include work on a new installation or the total replacement of an existing system. In the latter, obsolescence plays a significant role where the replacement of a sys­tem may involve the introduction of new techno­logy which can be used to update and improve the system.

• Subsequent changes in legislation may require a change in technology and if the change cannot be accommodated within the existing design, major mo­difications or new systems may be necessary.

2.2.3 Fast feedback effects

The oxygen atoms present in the oxide fuel used in AGRs act as an efficient moderator. Consequently fuel temperature effects are more complex than for the magnox reactors. At start-of-life there is no Pu — 239 present and the dominant feedback mechanism is U-238 resonance broadening with small contributions from U-235 non-l/v absorptions and the axial fine structure effect. The coefficient is typically — l.4 mN/°C (the Dungeness В design has a different core layout to the other AGRs and has a somewhat ne­gative fuel temperature coefficient). As irradiation proceeds, the Pu-239 concentration builds up giving an increasingly positive contribution to the coefficient. The magnitude of the Pu-239 effect is enrichment dependent; as enrichment increases so fissions in Pu — 239 become relatively less important and the contri­bution due to this mechanism falls. At 18 GWd/t the temperature coefficient of inner zone initial fuel (1.162 w/o enrichment) is +0.5 mN/°C and for inner zone feed fuel (2.012 w/o enrichment) is -0.38 mN/ °С in the Hinkley Point В design. Refuelling main­tains the overall fuel coefficient negative at a typical value of -0.8 mN/°C. Currently the AGRs are mov­ing to higher discharge irradiation fuel (higher en­richment) — this will give somewhat more negative fuel coefficients for a given fuel burn-up but the fuel cycle equilibrium value is largely unchanged by virtue of the higher mean core irradiation.

The inner graphite sleeve of the AGR responds on a time scale comparable with that of the fuel (although the magnitude of the temperature changes is considerably smaller than for the fuel). The im­portant mechanisms here are U-235 non-l/v absorp — lions and Pu-239 thermal resonance with small contributions from radial fine structure effects and e-135 non-1 ‘v absorption resulting in a small nega­tive coefficient initially (typically -0.1 mV°C) in­creasing to about 1 mN.°C at stringer discharge with a fuel cycle equilibrium value of about 0.4 mN/°C.

The limiting factors which determine the maximum power and temperatures at which the reactor can be operated, and the optimisation of the reactor for maximum electrical generation within those limits, are covered elsewhere in this volume. This section is con­cerned with deviations from steady power, i. e., the kinetic behaviour of reactor parameters. Consider the formula:

Power = gas flow x specific heat x temperature rise

The temperature rise is from reactor inlet to outlet. Although this formula may not be suitable for accu­rate calculations, it is adequate for the purposes of this discussion. In this section we shall consider:

• Changes in power at constant gas flow and its ef­fect on temperatures.

• Chances in gas flow at nominally constant tem­perature and its effect on reactor power.

• Changes in temperature at constant gas flow and its effect on reactor power,

Note that a change in power does not affect gas flow unless the change is so large that blowers which are steam-driven, either directly (Oldbury and Dun­geness A) or indirectly via auxiliary turbine-generators (Bradwell and Hinkley Point.4), are affected by re­ductions in steam flow.

Although some of the factors, particularly power and temperature, are interactive, giving rise to com­plex dynamic behaviour of the reactor parameters, in this discussion we shall endeavour as far as possi­ble to consider the effect of changing one variable at a time. It is hoped that this will aid the under­standing of the principles.

Scope

Specific types of reactor, e. g., magnox and AGR, and specific examples of the two types are provided with a wide variety of C and l systems to suit their par­ticular needs.

Magnox Reactors

Flux scanning Equipment, described in Chapter 2, is provided to enable the axial neutron flux distri­bution to be determined at a number of points across the reactor core. The frequency of flux scans is stated in the Station Operating Instructions and is typically six months to two years. For example, at Hinkley Point A power station with magnox reactors, flux scanning is normally carried out once per year follow­ing the return to power of each of the two reactors from its biennial overhaul: this is the only time when it is possible to cover the complete operational range of control and rod positions.

Delayed neutron detection The status of delayed neu­tron monitoring equipment on CEGB magnox stations is:

(a) Installed, active, and normally capable of tripping the reactors at:

Berkeley Bradwell Dungeness A Hinkley Point A Sizewell A

Due to operational problems and electrical inter­ference effects the equipment may be vetoed under conditions which may include refuelling, electric welding in certain areas, local lightning storms and equipment faults.

(b) Installed but vetoed, hence never capable of trip­ping reactors at Trawsfynydd.

(c) Not installed in concrete PV reactors, i. e., Oldbury and Wylfa.

(d) Equipment installed as (a) and (b) above is vir­tually identical in all cases.

AGRs

Liner leakage and penetration sampling Liner leakage detection and penetration sampling equipment is pro­vided to confirm the absence of significant leakage of primary coolant.

In the case of Heysham 2, the liner leakage holes are normally capped at their upper end and open to the safety shutdown room at their lower end. When required the portable liner leakage sampling trolley is connected to each hole in turn. The upper end of a hole will be uncapped manually and a pump will draw air down the hole and through the analyser. The measured CO; concentration on a 0 to lOO^o scale may be multiplied by the measured flow — to yield the leakage rate.

The trolley is connected via flexible couplings and discharges to the H and V extraction ducts.

The sampling holes can be fitted with additional pipework to provide suitable disposal of leaking gas if a significant leak were to be detected. The trolley is equipped with flow and inlet pressure indication. The portable system is designated Safety Class 3 and QA level 4.

The CO2 in air monitoring system allows detec­tion of leakage which may occur prior to more severe failure of the monitored parts which are pressurised with CO;, thus giving the operators a chance to reduce the economic and radiological consequences. Alarms are raised in the CCR only, unless a sample also serves a personnel protection role.

The installed CO;-in-air monitoring 12-point ana­lyser samples the safety shutdown room H and V extraction and would alarm a major leakage.

Seismic instrumentation Instruments are provided to fulfil the following functions:

• To raise an alarm in the CCR and E1C to warn the operator that a seismic event in excess of the operator shutdown earthquake (OSE) {0.05 g) has occurred.

• To allow the operating staff to assess the situation following a seismic event.

• To provide information for any subsequent analysis of a seismic event.

This section does not apply to the seismic switches used to inhibit secondary shutdown action which are part of the safety system. All seismic instrumentation discussed here performs a monitoring function, alert­ing the operator when some investigation is required. It does not initiate automatic actions.

The seismic instrumentation has no direct safety role. Failure of the instrumentation does not affect reactor safety, generation or the ability to shut down. It does have a role in assessing the integrity of plant following an earthquake and is thus assigned Safety Class 4 and QA level 4.

The equipment is designated Seismic Category A. This is both to give assurance that the information on the earthquake is available and to ensure that those items located in the safety room do not collapse onto safety equipment.

Because of the effect it has upon the physical proper­ties of the fuel, it is useful to describe (briefly) the main effects of neutron irradiation. When a neutron is absorbed by a U-235 atom, the nucleus becomes unstable and splits to produce two fission fragments or fission products, 2 to 3 neutrons, у rays and a considerable amount of energy (210 MeV per fission) which is equivalent to the mass imbalance between the original atom which was fissioned and its pro­ducts. About 80% of this energy is manifested in the kinetic energy of the fragments. These are highly ion­ised and speed through the fuel crystal lattice dis­sipating their energy as heat and causing considerable disruption to the orderly array of fuel ions along the track, in the process creating vacancies and intersti­tial atoms.

These two types of point defect may remain as single entities (at low temperatures), they may agglo­merate with defects of similar type to produce line defects (at intermediate temperatures) or recombine to recreate a perfect lattice site (at high temperature). For all but the highest temperatures where recombina­tion is rapid, the increased concentration of lattice defects is likely to have a marked effect upon the diffusion coefficient.

Returning to the fission fragments, these finally come to rest at a distance some 6 microns or so from the original fission site. These fission products may now themselves undergo further radioactive decay; they will also have an effect upon the chemistry of the fuel and may act as neutron absorbers or poisons. In addition they will produce fuel swelling and may release fission gases into the pin or rod; such effects are important for the performance, endurance and safety of nuclear fuel.

The Reactor Coolant System (RCS) of a typical four loop 1200 MW(e) PWR consists of the reactor core within its pressure vessel, four coolant loops with pumps, the primary side of the steam generators and a pressuriser. In addition there are a number of aux­iliary and safety systems which can interface with the RCS and these are discussed later. The RCS is filled with high quality demineralised water (the primary or reactor coolant) under chemical control and is re­quired to fulfil the following functions:

•

Transfer of heat from the reactor core to the primary side of the steam generators,

• Containment of the coolant under operating tem­perature and pressure, accommodation of coolant volume changes and limitation of the leakage of coolant and radioactivity.

• Moderation of the high energy neutrons to thermal

energies.

• Provision of a means for the soluble neutron ab­sorber to give reactivity control, in addition to the control rods.

• Maintenance of a high level of corrosion protection, including the suppression of radiolysis of water.

The latter two requirements are met by imposing an PCS chemistry regime that maintains a suitable pH condition to minimise corrosion, allows independent control of the soluble neutron absorber, suppresses
radiolysis and permits purification to remove undesir­able anions and cations. In addition it is desirable that the RCS chemistry regime minimises the release and transport of corrosion products, as discussed later.

With the exception of Berkeley and the UKAEA re­actors, whose vessels are cylindrical, the vessels at all other stations are spherical in shape with diameters between 15.24 m and 21.33 m and thicknesses between 76 mm and 101 mm. Two types of supports have been used to transfer the weight of the vessel and its contents to the reactor foundations; rollers and fixed cylindrical skirts. Both types of support have to cater for the expansion of the vessel due to gas pressure and temperature, and at the same time ensure that it is supported in a stable manner. Inside the vessels are supports for the diagrid and core, which transmit the load vertically to vessel supports. The internal sup-

Fig. 2.8 Typical duel arrangements for pressure
circuits

ports are similar to the external ones, e. g., either
discrete supports or a cylindrical skirt (see Fig 2.9).

The vessels have a number of penetrations of vary­ing size comprising the inlet and outlet gas ducts and

smaller ‘standpipes’ through which the reactor is re­fuelled or which contain control rods.

The vessels are made from fully-killed, grain-refined mild steels. In the choice and specification of the steel a number of factors were taken into account. In addition to having to withstand the environmental con­ditions of coolant gas temperature and pressure, and exposure to nuclear radiation, it had to be suitable for site fabrication. A low brittle/ductile transition temperature was specified. This was to reduce the risk of weld cracking during fabrication and to ensure that the vessels always remain in a ductile condition after irradiation, which embrittles the steel. The low transi­tion temperature requirement does not give a steel which_has a good creep resistance. The vessels had, therefore, to be kept below 350°C, the temperature above which creep becomes significant for the type of steel which had been chosen. Nearly all the vessels are insulated with stainless steel plain and dimple foil elements on the inner surfaces which see the hot re­actor gas. The exceptions are the vessels at Berkeley, whose internal surfaces are cooled by reactor inlet gas, and the Bradwell vessels which have, in part, been made from a steel which has the required creep properties.

Although the vessels had been designed to meet the basic code requirements, extensive stress analyses were carried out on the vessel features. Care was taken in the fabrication and inspection of the vessels and it is these areas where additional fabrication trials and inspections to those required by the standards were specified. These included plate pressing trials to de­monstrate that the process did not impair the proper­ties of the steel, ultrasonic examination of the plate for laminar defects and the proving of all welding procedures by welding trials.

On the completion of welding, the vessels were given a stress relief at 620°C, and a pressure test to

1.3 times the design pressure. All the vessels with the exception of those at Dungeness A, which were hy-

draulically tested, were given a pneumatic test. During the tests extensive strain gauge measurements were taken to give information about, and confidence in,

the design.

To prevent brittie failure when in service, the ves­sels are maintained at a temperature which will ensure that the steel remains ductile when under pressure. The minimum operating temperature is based on Robertson crack-arrest tests with an allowance being made for increase in transition temperature due to irradiation and thermal ageing. Charpy and tensile specimens have been placed in the vessels to monitor the change in the steel properties throughout the re­actor lives. Thermocouples are fixed to the vessels external surfaces to monitor their temperatures.

1.4.2 Gas ducts, bellows units, cascade corners, isolating valves and boiler shells

The main ducts through which the coolant gas flows from the reactor pressure vessel to the boilers and then back to the pressure vessel, are arranged in radial planes around the pressure vessel.

Flexibility must be built into the rigidly constructed ducts to accommodate dimensional changes due to variations in temperature and pressure. Bellows units are normally fitted in the duct circuits to act as hinges. Between any two fixed points, e. g., the re­actor pressure vessel and the boiler, a minimum of three bellows units are required. The exceptions to this are the hot ducts at Sizewell A which have been de­signed to be flexible.

Where the ducts change direction, right angled cascade bends have generally been used. Inside the bends are aerodynamically-shaped vanes to guide the gas round the bends and keep the gas pressure drop to a minimum. At some locations, lobster-back bends have been used.

An isolating valve is fitted in each reactor inlet and outlet duct and bypass circuit situated as close to the reactor as shielding and activity levels permit.

Each gas circuit has a bypass duct fitted between the gas circulator outlet and the boiler. This bypass duct provides operational flexibility during the start­up of a gas circulator and its associated boiler when the reactor and other boilers are on load.

The ducts, valves, etc., are supported on constant­load hangers which are designed to take the weight of the duct and to accommodate its expansion.

Neutron detectors based on the principles described in Sections 5.2.5 to 5.2.7 of this chapter, were de­veloped by the United Kingdom Atomic Energy Au­thority (UKAEA). These detectors can be operated with electronic circuitry that either counts individual pulses of ionisation or measures the mean DC caused by the ionisation. The latter method is used at the upper power range and the counters are used for the lower power levels.

The electronic units used with these detectors are:

• Pulse counters Pulse amplifiers, discriminators, logarithmic pulse counters, DC output amplifiers alarm and trip units. Doubling time units and run — on polarising units.

• Mean current ion chambers Logarithmic DC amplifiers, alarm and trip units. Doubling time units and polarising units. Linear DC amplifiers, linear and emergency channel, power deviation channels and shutdown units.

• Campbell system Combined pulse counting and DC units.

This basic method has not changed in principle since the first magnox reactors but there have been many developments to improve both detectors and their cabling and associated electronic equipment, the whole channel being treated as a compatible system. This
work has taken place at the Atomic Energy Establish­ment, Winfrith, in conjunction with UK industry.

This development has provided detectors capable of working at higher temperatures and neutron fluxes, and systems with less susceptibility to electromagne­tic interference. The electronic circuitry, matching the detectors, uses semiconductor technology giving high performance combined with good reliability.

Segregation in the form of separation by distance or barriers is used to isolate the redundant parts of the ‘Miem, so that mechanical or fire damage will not reduce protection to an unacceptable level.

A ke exchange system is used for the trip unit and guard line cubicles to prevent access to more than onc un’t in a redundant group at a given time, and to nummise the possibility of human error causing loss vT protection or spurious tripping of the reactor.

Power supplies

The power supply arrangements are a critical aspect of the system design and must be reliable and re­dundant in a way that is consistent with the equip­ment that they supply. High reliability supplies are discussed in Volume F, Chapter 8.

5.4.2 Equipment practice

Overall reactor protection system The total protection system comprises:

• Detectors, e. g., thermocouples or neutron flux detectors.

• Electronic ‘trip amplifiers’.

• Switching logic, e. g., relays or laddies.

• High power logic to insert control rods.

The concept of high reliability and fail-safe pervades all these systems.

Trip units — basic circuit techniques

The individual trip channels are designed for very low probability of dangerous failure so far as practicable. Open circuits in connections from sensors, in trip am­plifiers and in trip outputs, generally produce a safe (trip) condition. Trip amplifiers are normally designed
using dynamic signal fail-safe techniques in which loss of signal in the amplifier results in a trip output.

Some types of trip units, notably those used for temperature tripping, are specially designed and en­gineered to accept thermocouple signals directly or through DC amplifiers.

Other trip units use logic circuits that respond to the outputs or instrumentation of the same type as that provided for indication. Examples are neutron flux protection and gas circulator protection.