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.

The use of microprocessors for reactor protection is a new approach and required to be proved to meet established safety principles, one of the most impor­tant being fail safe design.

Fail safe design has been achieved to a large extent in the past by making the system dynamic, i. e., contin­ually changing state when healthy and assumed failed if it remains in one state for longer than a preset time.

Microcomputer technology is of itself dynamic in nature and, by the use of careful software control and the use of multiple microprocessors checking them­selves and each other continuously, a system of high reliability has been engineered equivalent to that ob­tained by fail safe design.

A simplified architecture for the reactor trip system is shown in Fig 2.141. The principles are also applied to the ESF systei# but the outputs are used to start the selected pumps and valves provided specifically for safety duties.

The protection system has four identical guardlines, with ‘2 out of 4’ majority decision voting used to de­termine a trip requirement. Figure 2.141 shows only one guard line architecture but for voting purposes it requires to receive the trip status of the other three guard lines. The use of digital processing enables this intercommunication of data between the guard lines to be done with optical data links. This provides a level of isolation between the four guard lines not previously possible with hardwired interconnections. The use of data links to provide information to control systems eliminates any risk of feedback from these non-safety systems causing a malfunction of the protection system.

In magnox fuel elements there is considerable fine structure of the thermal flux shape, radially and axial­ly due to the fact that neutrons are absorbed by the fuel as they enter the rod. There is also a marked distribution of resonance capture rate in U-238. These facts lead to non-uniform isotopic changes within the fuel during burn-up and hence a change in rating fine structure. The flux fine structure itself is influ­enced only little by the isotopic content and remains lairly constant, but the heat generation distribution and neutron balance are affected, and fission product distributions change significantly.

The thermal flux dip radially into the fuel is typi­cally about 30ro from surface to mean. Burn-up in! his tlux shape will result in a relatively larger in­crease in fission rate towards the surface of the fuel, at least for the early part of fuel life while Pu-239 build-up dominates oser the removal of L’-235. A more pronounced effect on fission rate fine structure is, however, produced by the Pu-239 arising from resonance captures in U-238. This increased pluto­
nium production rate occurs in a very thin skin (less than 0.1 mm thick) and can give rise to localised fission rates of about four times the fuel mean in this~very thin layer. This is of little significance for reactor operation because the effect is localised and does not affect clad temperature distributions to first order. It is of interest, however, in considerations of fission product concentrations in studies of activity contamination. The surface activity can be as much as four times the mean fuel activity.

In this section, the general principles of the start-up of a magnox and AGR will be described, from shut­down to full power. Prior to commencing rod with­drawal, pre-start checks may be necessan to ensure that the required systems are available. Gas How is set to the start-up value. Regulating rods are set to the required positions, then bulk rods are withdrawn to establish criticality. At low — power various checks are carried out to ensure that it is safe to increase power.

For the subsequent stages of start-up there are substantial differences between magnox and AGR. On a magnox reactor the core is very demanding whereas the boilers, particularly drum boilers, are very docile. In an AGR the start-up procedure for the once-through boilers largely dictates the reactor start­up and the reactor core itself is relatively undemand­ing from the control engineer’s point of view, partly because moderator temperature changes are small and partly because the control and protection systems are more refined than on a magnox reactor. On a magnox reactor the aim is firstly to raise gas temperature at low ^gas flow to achieve good steam quality to the turbines, then to increase gas flow at constant gas temperature to achieve the required reactor output; ‘no such simple^ generalisation can be made for the AGR. The final stage is trimming for optimum power output.

This section will concentrate on operations asso­ciated with the reactor core. When the reactor is started up there will be many operations to be carried out on other station plant; main plant such a boilers and turbines, condensate and feed systems; auxiliary plant such as pressure vessel cooling (for concrete vessels) and shield cooling (for steel vessels). On reac­tors with once-through boilers, particularly the AGRs, and on reactors with steam-driven gas circulators, the reactor start-up cannot readily be dissociated from the boiler start-up as is the case wdth drum boilers and electrically-driven circulators. However, in this section it is necessary to restrict the discussion to the reactor start-up only.

5.3,1 Pre-start checks

The extent of pre-start checks to be carried out prior to start-up will depend on the circumstances of the shutdown: clearly if a reactor has been shut down

for only a short period of time following a reactor trip, having previously operated satisfactorily, then fewer checks will be necessary than if the reactor has been shut down for several weeks for extensive overhaul work.

Prior to any start-up, checks will be carried out to ensure that the mininum plant required for safe op­eration at power is available and that safety circuits are healthy. Minimum requirements are set out in Operating Rules. A checklist is provided in Operating Instructions to assist the operating staff. Minimum plant requirements include; appropriate reactor in­strumentation, reactor gas pressure, the number of boilers to be available and their associated gas flow from running circulators, availability of water flow into boilers and means of disposing of steam generated, availability of auxiliary systems such as gas circula­tor lubricating oil, circulator shaft gas seal, cooling water, back-up and essential supplies. Safety circuits are checked to ensure that the correct protective de­vices, appropriate to the shutdown and start-up con­ditions, are in service. Where the start-up follows a longer shutdown period in which maintenance has been carried out, then the maintained equipment may be subject to functional checks prior to start-up, parti­cularly if the equipment is associated with the safety systems on the reactor.

The prime consideration when the reactor is shut down is to ensure that the fuel and graphite are ade­quately cooled. A considerable amount of heat is contained within the graphite (70 MWh if 2000 t of graphite change by 100°C), also heat will continue to be generated in the fuel after shutdown because of:

• The release of delayed neutrons capable of causing fission.

• The decay of fission products.

• The decay of heavy isoropes.

The delayed neutrons are exhausted within minutes, but some fission products and heavy isotopes have long halflives so the total afterheat, called ‘decay heat’ or ‘fission product heat’, remains significant for many days. Figure 3.27 shows how this heat decays with time.

Fig. 3.27 Decay heat after shutdown
This curve shows the amount of heat which continues
to be generated in the fuel after shutdown. Note that
the curve is plotted on a logarithmic timescale to
enable many days to be shown without loss of detail
in the first few minutes. Decay heat is expressed as a
percentage of the reactor power before shutdown; thus
for an AGR which has been operating at 1500 MW
the decay heat one minute after shutdown is about З^о
or 45 MW. This curve is for an AGR which has been
operating at steady full power for several days before
the shutdown, so that the fission products and heavy
isotopes have built up to equilibrium concentrations.

If the reactor remains pressurised with a full charge of CO:, the core is easily kept cool with small amounts of gas circulation and boiler feedwater flow. The amount of cooling applied wall depend on the core temperature to be achieved, for example, if the re­actor is shut down for access it will have to be much cooler than if it is shortly to be restarted. If extensive maintenance is to be carried out, as during an over­haul, and cooling circuits are required to be shut down for maintenance, attention is paid to the Operating

Rules which specify the minimum plant which must be available for core cooling.

Forced circulation of reactor gas is not always re­quired. Natural circulation is adequate at some stations under certain conditions, particularly some magnox stations where the boilers are sufficiently high relative to the core to establish an adequate thermal syphon, As the shutdown proceeds and the fission product heating decays, the probability of natural circulation being adequate improves.

In some respects a shutdown reactor is more diffi­cult to control than a reactor at steady full power. Because of the high load factors of CEGB reactors, shutdown is an infrequent event, so the reactor con­trol engineer is most accustomed to power operation. Many less familiar constraints apply when the reactor is shut down, the transient behaviour of the plant is very different when shut down, and many unusual and non-standard operations can be carried out on a shutdown reactor. Great care is taken when control rods are removed for access or maintenance, particu­larly if adjacent rods are removed, in order to main­tain the shutdown margin of reactivity. Where the reactor control engineer does not have direct control of operations within the reactor core, strict adminis­trative systems are set up and maintained to assure reactor safety while shut down and when the reactor is restarted.

The reactor design is optimised for full power op­eration, so particular care is necessary while shut down to ensure that plant limits are not exceeded. For example, the fuel sheath in a magnox reactor is less ductile when cool than it is at operating temperatures, so care is taken in control of reactor cooling to re­strict the rate of change of temperature. Another example is that the steel pressure vessel is more sus­ceptible to brittle fracture when cool, so restrictions are necessary in operations at standpipes to ensure that the vessel is not subjected to impact, and these restrictions are written into the Operating Rules.

Finally, control of the reactor while shut down is as required to prepare the reactor for restarting. For example, minimum temperatures must be achieved before starting up, as outlined in Section 5.3 of this chapter, so it is disadvantageous to overcool the reactor. The plant required for start-up is made available as the need arises.

Scope

The function of the primary coolant is to remove heat from the reactor core and transfer it to water tube boilers in the associated gas circuits. The controlling limits are specified in the station Operating Rules. At Berkeley, for instance, the operator is required to monitor and control the primary coolant with respect to the following parameters:

Gas purity Pressure Mass flow Temperature Reserve levels

In connection with gas purity and for both magnox reactors and AGRs there are sources of water that include release of water already absorbed within the reactor core, the radiolytic breakdown of methane in AGRs, the radiolytic/thermal breakdown of oil enter­ing the system and boiler leaks allowing water into the coolant circuit.

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 (BCD) system.

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 Tables 3.7 and 3.8, i. e., around 10 vpm for magnox and 30 vpm for AGRs and measurements are required within ±5% accuracy of ±2 vpm, whichever is the greater. Response time is an important factor because the boiler leak has to be detected before a significant quantity of water has entered the reactor.

In the case of magnox reactors, the prime reasons for the limit is the reduction in magnox ignition temperature with increasing moisture. The normal Operating Rule limit for magnox stations is 500 ppm by weight of water in the reactor coolant.

The relatively low power density and temperature con­ditions of the magnox design results in a high capital cost plant with moderate thermal efficiency (around ЗО^о). Additionally, ‘standard’ 660 MW turbine-gen­erators used in coal and oil fired stations could not be used.

The AGR operates at a higher temperature, pres­sure and power density, 640°C, 40 bar, and 3 MW/t, giving a compact core, 40<Fo efficiency and enabling the 660 MW sets to be used, one per reactor.

Because of these more onerous operating condi­tions, the fuel is ceramic uranium dioxide and slightly enriched (about 2.5%) to overcome the neutron ab­sorption of the stainless steel cladding that replaces the relatively low melting point magnox and to give originally 18 000 MWd/t and now a higher fuel burn — up. Oxidation of graphite by CO: is unacceptably high at temperatures above 600°C; hence, after leaving the gas circulators, half the coolant flow is initially directed to the top of the core and then downwards through interspaces in the graphite, rejoining the nor­mal coolant flow up through the core. This re-entrant gas flow limits the graphite temperatures to values at which thermal oxidation is acceptable.

As stated in the previous section the AGR pressure vessel is of pre-stressed concrete similar to the later magnox stations and encompasses the core, boilers and gas circulators. Typically the 20 x 20 m cylindrical vessel has 5 m thick walls and the internal surface has a 19 mm water cooled gas-tight steel liner. The pre — stressing and post-tensioning is applied by about 3600 multilayer helical steel tendons passing through mild steel tubes within the vessel wall.

To obtain maximum advantage of the established design of 660 MW turbine-generator units, the AGR was developed to operate at a higher temperature (640°C), pressure (40 bar) and power density (3MW/ m3). This resulted in a compact core, unit construc­tion and an efficiency of 40% (see Table 1.8 and Fig 1.32).

The enriched UO2 fuel pellets, 14 mm diameter, 5 mm bore, are in 1 m length stainless steel tubes which have a roughened external surface to improve heat transfer. Thirty-six of these fuel pins, supported in three concentric rings by stainless steel grids, are in a 190 mm internal diameter graphite sleeve. The as­sembly is called a fuel element and each fuel channel contains eight such elements, stacked vertically. A tie bar passes centrally through the eight elements and takes their weight during handling operations.

The AGR is designed for on-load refuelling. Each of the 300 or so channels has its own guide tube and standpipe through which the complete fuel stringer of eight fuel elements and associated fittings above and below, is handled as a single unit. The full 1500 MW(th), 660 MW(e), output requires a refuelling rate of about two channels per week.

Reactivity is controlled by about 80 stainless steel absorber rods, over half of which have inserts of boron-stainless steel alloy to give coarse control.

10.6.1 Magnox reactors

In magnox reactors the primary processes which have to be optimised are:

• Carbon deposition on fuel pins or boiler surfaces.

• Graphite corrosion.

• Steel corrosion. The major gas composition varia­bles are carbon monoxide and hydrogen (plus water), with a third parameter, temperature, poten­tially available.

Reactor experience has shown that carbon deposition can be avoided by operating at a carbon monoxide concentration not greater than 1.5 v/o together with a hydrogen concentration greater than 25 vpm. Graph­ite corrosion, in the context of a 30-40 year life, is not life-limiting for the early magnox reactors due to the low coolant gas pressures and power densities and hence low graphite oxidation rates. Steel corro­sion rates decrease as the hydrogen concentration decreases. Hence, for the early (low pressure) magnox reactors the maximum reactor life is attained by mini­mising the hydrogen concentration, subject to the 25 vpm deposition limit.

For the later magnox reactors, which operate at a higher gas pressure, graphite corrosion becomes im­portant in the context of a 30-40 year life. The cool­ant composition therefore has to be balanced between reduced hydrogen concentration, leading to reduced steel corrosion with increased graphite corrosion, and vice versa. Temperature also has a marked effect on steel corrosion rates, e. g., a rise from 360° to 370°C increases steel corrosion rates by 60^0 but has a ne­gligible effect on graphite corrosion rates. Hence, for any given mode of reactor operation, the reactor gas outlet temperature can have a marked effect on the optimised hydrogen concentration, and in a complete system optimisation, the temperature itself would be considered a variable.

10.6.2 Advanced gas cooled reactors

In AGRs, the primary processes that have to be con­sidered in coolant optimisation are:

• Carbon deposition.

• Graphite corrosion.

• Steel corrosion.

• Coolant plant capability.

In most circumstances steel corrosion is not coolant — limiting except for the requirement to avoid high re­lative humidities in low temperature parts of the circuit which could give rise to corrosion of wetted surfaces. AGRs typically operate at 41 bar CO2 pres­sure with a coldest metal temperature of typically 30°C leading to a maximum acceptable water concen­tration of 600 vpm.

The remaining three parameters can be optimised using the concept of ‘The Coolant Window’. This in­volves plotting the boundary of acceptability for each of the three processes on a graph of carbon monoxide against methane concentration. An example is shown in Fig 1.52 where line A indicates the boundary be­tween acceptable and not acceptable carbon deposition on fuel pins, line В shows the coolant compositions required to achieve design core life and line C indicates the limit imposed at 600 vpm H2O by the bypass plant. The particular item in the bypass plant which normally defines the limit is the drier bed rather than the recombination unit and hence the limit is depend­ent on the acceptable water concentration. This is shown by line D which is for the lower water con­centration of 350 vpm, indicating the reduced range of coolant compositions available.

Operation of the coolant within the blue area of Fig 1.52 will thus ensure reactor design life within the other known coolant-related reactor restraints.

Fin deformation At the fuel element design stage, rie tests were carried out to ensure that the can fins would not deform by the action of gas forces and the final designs were approved on this basis. However, after approximately four years’ irradiation HTA fuel elements at Bradwelf showed significant fin deforma­tion. This ted to a 10% power reduction to compen­sate for the resulting increase in can temperatures.

Soon afterwards, during routine post irradiation examination, some Dungeness A fuel elements were seen to be severely deformed at a channel average irradiation of 1000 MWd/t. The deformation was such that it was doubted if the design life of 3600 MWd/t could be achieved.

Investigation showed that the fins were deformed by the growth of oxide layers following a magnesium/ gas interaction process. As a result the fins were be­ing deformed without an externally applied stress and this was a situation that had not been considered during fuel element development.

Following the reactor and laboratory studies various methods of eliminating the problem were devised, for example:

• Reduce operating temperatures.

• Clean up of the coolant.

• Change to a herringbone can design.

lietore a final decision was reached, all operating tem­peratures were reduced following the steel oxidation problem. This effectively eliminated fin deformation.

ture accident is an important life-limiting feature of the magnox fuel element.

Cranium bar swelling is determined by recovering from the reactor fuel elements containing bars that were specially measured prior to irradiation (the mea-