Various attempts have been made to identify chemi­cals which will either inhibit the extent of corrosion beyond the required passivation layer, or selectively, inhibit the mechanism of corrosion product release by modifying the structure and adherence of the oxide film. A further possibility is the use of chelating re­agents to continuously and selectively remove active species by dissolution from deposited material and oxide films. Such soluble species could then be re­moved on an ion exchange bed possibly with regenera­tion of the chelating agent.

Both of these approaches are under development and will have to be proved effective and acceptable to the existing complex primary circuit chemistry be­fore they can be recommended for use by reactor operators.

Ail the magnox stations except Wvlfa use cooling t‘■akh about 6 m deep for spent fuel storage in Tip’- tor a minimum ot some three months before Tspateh, this allows the post-irradiation heat and radiation to decay to an acceptable level. Water is a cheap and elfective medium for cooling and shield­ing whilst allowing the handling operations to be observed.

A typical sequence of events is shown diagramma — tically in Fig 2.29, which shows the irradiated fuel handling and storage arrangements at Oldbury. There are essentially three basic routes depicted on the dia­gram; the incoming flask and skip route (stages /-6), the outgoing irradiated fuel route (stages 7-17) and the splitter flask route (stages 18-26).

A brief description of the facilities required and the 26 stages shown on Fig 2.29 is given as follows:

Incoming flask and skip

1 An empty skip arrives inside a road transport flask.

2 The flask is transferred to the storage bay by the flask crane.

3 When required, the flask is transferred to the washdown bay where it is washed, the lid bolts removed and the lid jacked open.

4 The prepared flask is transferred into the dispatch bay of the cooling pond: the lid is removed and returned to the decontamination bay, and the empty skip is removed from the flask and trans­ferred to the pond storage bay by the skip crane.

5 Flask lid seals are inspected, if defective, the lid is decontaminated and transferred to the leak testing bay where seals are renewed.

6 When required, the empty skip is transferred to a pond handling bay.

Outgoing irradiated fuel route

7 Fuel elements and bottles discharged from the reactor arrive in the unloading tray in the cooling pond. From here they are transferred to a storage skip by manipulators.

8 When full, the skip is moved to the pond storage bay for a cooling period.

9 After cooling (for at least 90 days), the skip is returned to the pond handling bay w’here elements are removed to check that the correct cooling period has elapsed.

10 Polyzonal fuel elements are desplittered and placed in a second skip. Herringbone fuel elements are not desplittered and are placed in a separate skip,

11 When a skip is ready to leave the pond, it is transferred to the caesium sampling position in the storage bay.

Fig. 2.29

Irradiated fuel handling and storage arrangements at Oldbury

A Road Transport Flask containing an empty Fuel Element Skip arrives at Oldbury on a Low Loader and is transferred to the Storage Say by the Flask Crane.

When the Flask is to be processed it is transferred to the Washdown Bay for a Pre-Ponding Wash. The Lid Bolts are removed and the Flask Lid is jacked open.

After cleaning and preparation the Flask is transferred to the Despatch Bay in the Cooling Pond by the Flask Crane.

Once the Flask is in the Cooling Pond the Lid is removed and transferred to the Washdown Bay and the empty Skip inside the Flask is removed and transferred to a Storage Bay by the Skip Crane.

The Lid Seals are visually inspected ready for replacement on the Flask. If the Seals are seen to be defective the Lid is decontaminated and transferred to the Leak Testing Bay where the Seals are renewed.

When the empty Skip is required it is transferred to a Handling Bay.

Fuel Elements and Bottles discharged from the Reactor pass through the Unloading Well equipment and arrive in the Unloading Tray in the Cooling Pond. From here they are removed by manipulators and placed in a Storage Skip.

When the Skip is full it is moved to the Storage Bay by the Skip Crane and left to cool.

After the necessary cooling period has elapsed (at least 90 days) the Skip is brought back into the Handling Bay.

Polyzonal Fuel Elements are removed from the Skip, monitored to check that the correct Cooling Period has elapsed, Desplittered and placed in a second Skip. Herringbone Fuel Elements are monitored but not Desplittered and are put in a separate Skip.

Once a Skip is ready to leave the Pond it is transferred to the Caesium Sampling Position in the R1 Storage Bay.

When the Skip is ready to leave the Pond it is checked under the Caesium Sampling Hood and then placed inside an empty Flask in the Despatch Bay.

Once the Skip has been placed in the Flask by the Skip Crane the Flask Crane collects the Flask Lid trom the Washdown Bay. lowers it onto the Flask and then transfers the Flask to the Washdown Bay. The Flask is Decontaminated and the Lid is bolted down. Health Physics check that decontamination is satisfactory then the Flask is transferred to the Leak Testing Bay. The Flask is Leak Tested and Nitrogen purged, if required by the Transport Approval, it is Fluoride Dosed then transferred to the Storage Bay. When a Low Loader is available the Flask is checked tor contamination and then transferred to the Low Loader.
14
15
16
Once clearance has been obtained, the Low Loader leaves Site for either the British Rail Railhead, Berkeley Power Station or Berkeley Nuclear Labs.
17
When a Splitter Flask is full of Magnox debris it is transferred to the Despatch Bay by the Skip Crane and deposited in a Cradle. The Flask Crane now transfers the Splitter Flask and Cradle to the Washdown Bay for Decontamination. After decontamination the Splitter Flask and Cradle are transferred to the Splitter Flask Transporter on the Magnox Debris Corridor. The splitter Flask Transporter transfers the Sputter Flask and Cradle along the Corridor to the Active Waste Oump Loading Bay. At the Active Waste Dump the Splitter Flask is removed from the Transporter and Cradle by the Dump Crane and lowered onto a Magnox Debris Vault Door. The Splitter Flask contents are emptied into the Vault and the Flask is then transferred back onto the Transporter by the Dump Crane. The Splitter Flask Transporter transfers the Splitter Flask and Cradle along the Magnox Debris Corridor to the Flask Crane. The Splitter Flask and Cradle are transferred to the Despatch Bay by the Flask Crane. When required, the Splitter Flask is removed from the Cradle by the Skip Crane and is transferred to a Handling Bay adjacent to a Desplittering Machine.
19 20
21
22
23
24
25
storage arrangements at Oldbury

12 Caesium sampling complete, the skip is placed inside an empty flask in the pond dispatch bay.

Note: At Sizewell A and Hinkley Point A, the skip

is removed from the pond up a shielded ramp before being placed in a flask, but the arrange­ment at Oldbury is more representative.

13 The flask crane collects a flask lid from the decon­tamination bay, lowers it onto the flask and returns the flask and its lid to the decontamination bay (see Fig 2.30).

14 The flask is decontaminated and the lid is bolted down. Health physics check for satisfactory de­contamination and the flask is transferred to the leak testing bay.

15 The flask is leak tested, nitrogen purged and, if necessary, fluoride dosed.

16 The flask is transferred to the storage bay to await road transport and a final contamination check.

17 Transfer to road transport for delivery to re­processing plant.

Splitter flask route

18 Splitter flask full of magnox debris.

19 Full splitter flask deposited on a cradle in the dispatch bay.

20 Flask and cradle transferred to the washdown bay for decontamination.

21 Decontamination flask and cradle transferred to the splitter flask transporter for delivery to the active waste dump loading bay.

22 Splitter flask removed from transporter and cradle and lowered onto a magnox debris vault door by the dump crane.

23 Magnox debris emptied into the vault and the empty flask returned to its cradle and transporter.

24 Splitter flask and cradle transported to flask crane.

25 Splitter flask and cradle transferred to the pond dispatch bay.

26 When required, splitter flask is removed from cradle and transferred to the pond handling bay, next to the desplittering machine.

Whilst ponds provide convenient shielding, careful

and rigorous water treatment and pond management

is necessary if magnox fuel can corrosion and re­

lease of contamination are to be acceptable. The decay heat also has to be removed. Ponds there­fore are provided with water circulating systems em­bodying filters, coolers and water treatment equipment. Contamination is removed by the filter and water treat­ment media and hence there is supplementary shield­ing provision for the exchange and storage of these arisings. Early experience with storage ponds was un­satisfactory, particularly when storage periods were excessive. Can-corrosion occurred and contamination spread from the water surface and from equipment as it was moved in and out of the water. Considerable ingenuity by station and other staff was necessary to cope with the difficulties.

It was these difficulties that led to the adoption of dry spent fuel storage at Wylfa (Fig 2.31). Each dry store cell consists of a bundle of vertical thimble tubes with a carbon dioxide atmosphere in w’hich the spent fuel is placed by the charge machine. The outside of the tubes is cooled by natural convection air. After the decay period the elements are discharged using a separate hoist to a shielded vault where they are de- lugged, placed in skips and dispatched in the trans­port flask. Subsequently, further dry storage has been provided at Wylfa in low head air-cooled vaults.

2.3 Reactor charge/discharge In principle, refuelling consists of coupling the pres­surised FM to the reactor, removal of the spent fuel elements one at a time from the selected channel and subsequent charging of the new elements.

The AGR concept was that uranium dioxide fuel with stainless steel dad would permit the channel gas out­let temperatures (about 650°C) necessary to give the steam conditions for conventional large turbine-gen­erators. The commercial fuel element design permits average channel powers of some 4.5 MW and hence some 300 to 400 channels are necessary to match the reactor power to the heat input required by a 660 MW turbine-generator.

The high gas outlet temperature precludes the use of in-pile refuelling grabs which had been trouble­some at the magnox stations and, furthermore, the change in channel power over its lifetime requires a variable gag to maintain a constant gas outlet tem­perature. The high channel powers and the pitch of the channels make individual channel access feasible and economic. Hence, each fuel channel has an in­sulated and watercooled standpipe above it passing through the pressure vessel top slab. Below the stand­pipe there is a guide tube which passes down through the top dome and into the top of the fuel channel in the core. The fuel stringer within the core consists of eight fuel elements (seven at Dungeness B) with top and bottom fitments and is lifted by a tie bar which passes up the middle of the stack and is con­nected to the plug unit. The plug unit sits on top of the fuel stringer and extends up the guide tube/ standpipe to the pile cap level. The plug unit together with its fuel stringer is called a fuel assembly. This has an overall length of some 23 m and is handled as a single unit. The refuelling grab engages with the top of the plug unit and never enters the pressure vessel.

The basic post-trip heat removal mechanical systems arrangements are shown in a simplified form in Fig 2.114.

The individual mechanical systems, which together with the associated essential electrical system and post — trip sequencing equipment comprise the complete post­trip heat removal system are:

* system

• Main gas circulators and inlet guide vanes.

• Circulator auxiliaries cooling systems.

• Circulator auxiliaries diverse cooling system.

• Decay heat boilers and decay heat boiler feed system.

• Decay heat boiler air cooling system.

• Control system for starting and standby feed pumps.

The X systems are seismically qualified and are de­signed to remain functional after all earthquakes up to the magnitude of the safe shutdown earthquake.

Y system

• Emergency boiler feed and main boilers, including LP vent system.

• Inlet guide vanes emergency drive.

• Reactor seawater system.

• Pressure vessel cooling system.

All the post-trip heat removal systems, except the diesel generators, are provided on a per-reactor basis, the follow’ing plant descriptions are for a single reactor unit.

g 2 1 Forced gas circulation systems (X)

Forced gas circulation is achieved by operating the main gas circulators at 450 r/min (15<?o of normal speed) with the inlet guide vanes (IGV) fully open. This speed reduction is achieved by supplying the motors at a reduced frequency by means of variable frequency converters, which are connected automati­cally by the post-trip sequencing equipment (PTSE); each eas circulator has its own independent frequency converter. For depressurisation faults, the speed of the gas circulators is signalled to increase as a func­tion of the reactor pressure until normal full speed is achieved.

An independent circulator auxiliaries cooling system (CACS) is provided for each quadrant pair of cir­culators to cool the main motors and the lubricating oil systems. For normal operation, heat is rejected from the CACS to the reactor seawater system via two ІОО^о-duty plate heat exchangers; during post­trip operation an additional heat sink to the atmos­phere is available via the circulator auxiliaries diverse cooling system (CADCS). The latter comprises a single loop per reactor with two forced draught air-cooled heat exchangers, cooled by four fans and four pumps circulating the cooling water to the four CACS sys­tems; the capacity of the system is only sufficient for the post-trip heat removal duty and it is shut­down during normal operation.

The reactor coolant system has four reactor coolant pumps (RCPs), one in each of the four reactor cool­ant loops, in the return pipework from the steam gen­erators to the reactor essel. The design Пои through each pump is approximately 6.3 m "

The RCP assembly comprises a vertical sinele-staffp

m. xed now type pump with an overhung &

and a three-stage controlled leakage shaft §$Єа1 .The pump is driven by an above-mounted air cooled three Phase induction motor, solidly coupled to the pump shaft by a removable intermediate spool piece A cut away perspective view of the RCP assemblv is shown in Fig 2.132.

The more important elements of the RCP art — Hp scribed below; e’

• The pump casing is fabricated from a single-piece austenitic stainless steel casting complete with in­tegral suction and discharge nozzles and support lugs. This arrangement eliminates fabrication welds within the casing and hence reduces the in-service inspection of this component to surface examina­tion only, with a consequent minimisation of person­nel radiation exposure.

• The pump is a high specific speed unit and hence the overhung impeller is of mixed flow design with double curvature Francis vanes, which are shrouded along upper and lower sides. The diffuser converts a portion of the velocity head into static pressure and guides the flow out of the radial discharge nozzle. A thermal barrier, attached to the diffuser flange, restricts the transfer of heat from the hot reactor coolant to the pump bearing and seal areas.

• The seal system comprises three seals arranged in series such that reactor coolant leakage to the re­actor building is negligible. The number 1 seal is a controlled-leakage film-riding face seal and the num­ber 2 and 3 seals are rubbing face seals, mounted in a single cartridge type of assembly for ease of maintenance.

• The complete pump and motor shaft assembly runs on three radial bearings, of which two are located in the motor and the third in the pump. The bear­ing within the pump is of the hydrodynamically water-lubricated sleeve type with a self-aligning ca­pability. The bearing material is a carbon matrix impregnated with graphite whilst the journal is stellite on stainless steel. The two radial bearings within the motor are conventional oil-lubricated babbit-on-steel pivoted pad guide bearings.

• A removable spool, located between the pump shaft and the motor shaft, is provided to facilitate in-service inspection and maintenance of the shaft seal system without removing the motor. This de­sign feature reduces personnel radiation exposure and pump maintenance downtime by minimising the number of operations involved in seal inspec­tion and maintenance. [24] reverse rotation of the pump in an idle reactor coolant loop due to back-flow generated by the operating loops. This situation is encountered dur­ing the sequential start-up of the four RCPs.

10.2.5 Steam generator

The four steam generators provide the physical inter­facing link between the primary pressure boundary of the reactor coolant system and the secondary circuit system. Their function is to transfer heat from the reactor coolant water to the feedwater supplied from the turbine condensate system, thus converting the latter to essentially dry and saturated steam. The steam is then used to drive the turbine-generators.

Each loop of the reactor coolant system contains a recirculation type steam generator, as shown in Fig 2.133. The unit is basically a vertical shell and U-tube heat exchanger, in which heat is transferred from a single-phase fluid at a higher temperature and pres­sure on the tube side, to generate a two-phase steam/ water mixture at lower temperature and pressure on the secondary (or ‘shelf) side.

The reactor coolant water at 155 bar and 323.6°C, is pumped at a constant flow rate through the U — tubes, where its heat is transmitted through the tube walls to the water/steam mixture on the secondary side of the unit. During steady state operation, the steam generated in the secondary side is balanced by the addition of feedwater, thus maintaining a constant fluid inventory and heat content on both secondary and primary sides of the unit.

The primary side consists of the inlet and outlet plena jocated in the hemispherical bottom channel head, and the inside of the U-tubes. Reactor cool­ant flows into one half of the channel head, thence through the U-tubes and back into the other side of the channel head, from where it rejoins the reactor loop.

The secondary side consists of a lower cylindrical shell surrounding the tube bundle and an upper steam drum which contains the moisture separation equip­ment, joined by a transition cone. Feedwater enters through the feedwater nozzle and flows into the feedwater ring. It then spills out from inverted U — tubes located on the feedwater ring and flows down an annular gap formed by the inside of the lower shell and a wrapper barrel which surrounds the U-tube bundle.

The location of the wrapper barrel ensures that the mixture of recirculated water and feedwater enters the U-tube bundle at the level of the tubesheet upper face.

The feedwater is supplied from the plant feedwater system at approximately 56°C, below its saturation temperature. Inside the steam generator, the feed — water is joined by the water recirculating from the moisture separators, producing a feed mixture for the tube bundle that is close to the saturation tern-

perature. Thus, only a small portion of the tube bundle, located just above the tubesheet, functions as a ‘preheater’ to raise the fluid temperature to sat­uration point. The majority of the tube bundle thus operates in the heat transfer efficient nucleate boiling regime.

As the water flows upward through the tube bun­dle, the heat transferred from the reactor coolant raises its temperature to saturation value. Thereafter steam is generated until, at exit from the tube bun­dle, approximately 25% of the water has been con­verted to steam. The steam/water mixture rises into
the upper section where a set of centrifugal moisture separators removes most of the entrained water from the steam. The steam continues to the secondary separators located just below the outlet nozzle, where most of the remaining moisture is removed. The steam quality at exit from the unit is at least 99.75(ro. The entrained water removed by the primary and secondary separators is returned to mix with the in­coming feedwater, the mixture then being recirculated through the tube bundle.

An inherently useful characteristic of the design is that it is relatively insensitive to reductions of tube surface area. The reason for this is the relatively small temperature difference between the primary and sec­ondary side fluids. At the hot end of the steam gen­erator, the temperature difference is approximately 40°C, whilst at the cold end the temperature differ­ence is approximately 8.5°C. This results in an overall logarithmic mean temperature difference of approx­imately 20°C. The effect of this is that if for any reason tubes have to be blanked off, a significant change in surface area can be compensated by a small change in operating temperatures.

Some further details of the major steam generator components are as follows.

The hemispherical channel head is made from low alloy steel and is of forged construction. It is welded to the underside of the tubesheet and is divided inter­nally by a vertical divider plate, made from Inconel, which is welded to the channel head itself and to the underside of the tubesheet. Each chamber has a nozzle which is welded onto the primary loop piping, and an access manway. The manways are provided with bolted covers which are isolated from the re­actor coolant by an Inconel insert diaphragm. The internal surfaces of the channel head are clad with austenitic stainless steel, which provides a corrosion — resistant barrier between the channel head material and the reactor coolant.

The tubesheet is a single forging of low alloy steel, approximately 4190 mm in diameter and 610 mm thick. Together with the U-tubes, it forms the bound­ary between the primary and secondary sides of the steam generator. The outer circumferential portion of the forging is a flat ring into which the vertical supports for the unit connect. Tube holes are drilled through the tube plate into which the U-tubes are inserted. The underside face of the tubesheet is clad with Inconel to provide a corrosion-resistant barrier between the forging and reactor coolant. The Inconel cladding is the same alloy material as used for the tubing, thus enabling the tube-to-tubesheet weld to be made between similar materials.

The tube bundle consists of 5626 U-tubes, fabricated from Inconel 600 material. The tubes are 17.48 mm outside diameter and 1.016 mm wall thickness. The total length of tubing within one steam generator is approximately 100 km. The tubes are given a special thermal treatment after being drawn to size, which improves the corrosion resistance of the material. Additionally, tubes with the smallest radius bends are given a further heat treatment to reduce the resid­ual stresses resulting from the bending operation. The tubes are inserted through, and supported by, the tubesheet and tube support plates. The tube ends are welded to the tubesheet cladding, after which each tube is hydraulically expanded over virtually the full depth of the tubesheet.

The tube bundle is supported by seven tube sup­port plates which are themselves supported and held in position by stay rods and spacer tubes.

The U-tubes pass through quatrefoil-shaped holes in the support plates formed by drilling and broaching. The hole shape consists of four lobes and four flat support lands around each tube. This design of hole shape reduces the tendency to dry-out and, hence, to increase the concentration of corrosive chemicals in the region of the tube and support plate intersection. The support plate material is a ferritic stainless steel.

A flow distribution baffle, made of stainless steel, is located between the bottom tube support plate and tubesheet. Its purpose is to promote cross-flow in the tubesheet region, thus reducing the potential for tube dry-out in this region. The tubes pass through the baffle in either individual round clearance holes or through the circular central cut-out. In the U-bend region, anti-vibration bars are provided to stiffen the bundle and restrain any vibration tendency of the tubes.

The secondary side pressure boundary consists of the lower and upper shells, the transition cone and the top head. The lower and upper shells are constructed from low alloy steel forged into cylinders of requi­site diameter. The transition cone between lower and upper shells is also of forged construction. The ellip­soidal head is either a single forging or made from two shaped plates. The shell contains several access openings consisting of manways (upper shell) and handholes (upper and lower shells). These allow ac­cess for the maintenance of moisture separation equip­ment, visual inspection and sludge lancing operations at the top surface of the tubesheet. In addition to the main feed and steam nozzles, various other small nozzles are provided for water level instrumentation.

The secondary lower shell contains the tube bundle wrapper and the upper shell contains the moisture separation equipment. The carbon steel wrapper as­sembly encloses the tube bundle and forms the inner surface of the downcomer annulus. A conical transi­tion piece welded to its upper end provides a transi­tion to the assembly barrels of the primary moisture

separators.

f!,e main feedwater nozzle located at one side of the upper shell connects to an internal distribution rinc Feedwater is discharged from the ring through j. nozzles located on its top side. The design of the j nozzle, inlet ring and J-nozzIes inhibits drain — — a 0f the ring if the water level should drop below it’thereby reducing the possibility of water hammer effects.

The steam outlet nozzle at the top of the unit is pro­vided with a flow restrictor which is designed to limit steam flow in the unlikely event of a break in the main steam line. It consists of seven Inconel venturi inserts, installed in holes in a low alloy steel forging.

The moisture separation equipment located within the upper shell region comprises two stages. The sixteen first stage separators are located directly above the tube bundle and consist of 508 mm diameter by 3048 mm high assemblies containing static swirl vanes. These vanes impart rotary motion to the steam-water mix­ture causing the heavier water to be thrown outward by centrifugal force and diverted to a concentric — downcomer barrel. The water returns to the recir­culating water plenum.

The second stage separators consist of banks of contoured vanes contained in a housing. The contours in the vanes produce multiple changes in the flow direction, thus causing removal of the remaining en-> trained water from the steam.

A steam generator blowdown system is provided to draw a small water flow continuously from each steam generator, just above the tubesheet, to purify it and return it to the condensate system, or reject it as etfluent. The purpose of this system is to prevent excessive concentration of non-volatile impurities due to continuous evaporation in the steam generator.

The steam generator rests on four vertical steel col­umns with spherical plain bearings at top and bottom. This arrangement allows for the thermal expansion of the coolant loop to be accommodated by lateral movement of the unit. The support columns are at­tached at four locations 90 degrees apart on the outer circumferential ring of the tubesheet forging, in addition to vertical support, lateral support is provided at tubesheet level and just below the shell transition cone. These lateral supports consist of steel bumpers and dampers. The bumpers are provided with clearances to allow for normal thermal expan­sion. In the event of earthquake or pipe rupture.

however, their function is to prevent excessive move­ment of the steam generator. The dampers provided at the upper support location are designed to allow slow thermal movements but also prevent any exces­sive sudden movement by earthquake or pipe rupture.

10.2.6 Reactor coolant pressure control system

During normal station operation and followina fault conditions, the reactor coolant system experiences tem­perature changes which cause the volume of the cool­ant to increase or decrease, which in turn affects the pressure. If the pressure were to rise unchecked, the design limits of the pressure parts might be ex­ceeded; an uncontrolled pressure reduction would cause boiling in the system. The purpose of the pressure control system, of which the pressuriser is the major component, is to maintain the pressure within pre­scribed limits during normal operations and in fault conditions, to prevent design limits being exceeded.

The pressuriser is a vertical cylindrical vessel fa­bricated from carbon steel and clad internally with austenitic stainless steel. It has an overall height of approximately 16 m, an internal volume of 51 m which is filled partly with water and partly with steam in thermal equilibrium with each other. The steam space in the pressuriser acts as a buffer to limit the magnitude of pressure changes which result from changes in the volume of the reactor coolant. Such volume changes cause a flow of water into or out of the pressuriser through the surge line, which connects the bottom of the pressuriser to the hot leg of one of the reactor coolant loops. Relatively small changes in pressure are counteracted by the use of either elec­trical immersion heaters or a water spray in the pres — suriser, both of which are under the control of the station control system. The 78 heaters, with a total power of 1800 kW, are located in vertical sheaths which pass through the bottom head of the pres­suriser in three concentric rings around the surge line nozzle. A general view of the pressuriser is shown in Fig 2.134.

During normal operation a small quantity of water is sprayed continuously through the spray nozzle at the top of the pressuriser, being supplied via two control valves from the cold legs of two of the loops. This small flow avoids temperature fluctuations and ensures that the boron concentration of the water in the pressuriser does not differ significantly from that in the remainder of the reactor coolant. Should the pressure rise, the spray control valves will open to increase the flow of spray water at 294°C into the steam at 345°C, thus condensing some of the steam and reducing the pressure. Conversely, should the pressure fall, power to the electrical heaters will be increased from the normal steady state value required to offset the effect of the small spray flow and heat losses from the pressuriser. This additional heating will generate steam which will increase the pressure.

There are two sets of safety valves connected to the top of the pressuriser which prevent overpres — surisation of the reactor coolant system under all normal fault conditions. The first set of valves con­sists of three tandem pairs of pilot-operated safety relief valves (POSRV) produced by the French com­pany SEBIM. In each pair, the valves are connected in series; the upstream valve being the normally closed relief valve, and the downstream valve being the nor­

mally open isolation valve. The second set of safe­ty valves consists of two spring-loaded safety valves (SRV), which both act as a diverse (back-up) relief system for frequent faults and also contribute to the total system relief capacity needed for extremely un­likely faults, which have a requirement for a large relief capacity.

The POSRVs are primarily self-actuating valves op­erating under the action of system pressure alone, but

they can also be actuated by an electrical signal to a solenoid control valve in the pilot unit. Each POSRV consists of two basic parts; the valve itself and the

ilot unit which is mounted remote from the valve. The pressuriser pressure is sensed by a small bore pipework connection between the pressuriser and the

pilot unit.

When the pressuriser rises to the relief valve open­ing set pressure, a small valve in the pilot unit opens and depressurises the space above the valve’s actuating piston, thereby allowing the system pressure acting on the underside of the valve disc to open the valve. Following sufficient relief from the system, the pres­sure in the pressuriser will fall to below the relief valve closing set point and another small valve in the pilot unit will open and apply pressuriser pressure to the top of the relief valve’s actuating piston, thereby closing the valve.

Should the relief valve fail to close, the pressuriser pressure will continue to fall until the closure set point of the downstream isolation valve is reached, whereupon it will close and terminate the discharge.

This automatic isolation feature reduces the risk of coolant loss due to the POSRV sticking open. The three pairs of POSRVs have staggered set pressures, the lowest being 162 bar, ensuring that the number of valves which open under any particular fault tran­sient is minimised. The two SRVs have a higher set point (172 bar) and are only required to open under extreme fault conditions.

The electrical opening facility of the POSRVs is used to provide cold overpressure protection to the reactor pressure vessel at or near cold shutdown con­ditions. The tendency for reduced toughness of the reactor vessel material at low temperature is exacer­bated by neutron irradiation during station life. Hence, pressure relief is required at pressures below the nor­mal operating set pressures when the vessel material is at a low temperature. To provide protection against faults involving a pressure increase at low tempera­ture, the reactor protection system calculates the maxi­mum permissible pressure at the measured reactor coolant temperature and opens the POSRVs if this pressure is exceeded.

All the safety valves discharge into a pipeline which is connected to the pressuriser relief tank located at a low level in the reactor building. This tank is kept partly filled with coid water and the discharge pipe­line is connected to a submerged sparger pipe so that the discharged steam is condensed in the water and contained within the tank, thereby preventing any discharge to the reactor building. However, bursting discs are fitted to this tank to protect it against over — pressurisation should the steam discharge be excessive.

The system described is generally similar to that of ^estinghouse PWR units in the USA, but significant improvements hae been made for Sizewell B. The, irst improvement is in the use of SEBUM POSRVs v hіch have been developed by Eiectricite de France

especially for use in French PWR stations. After a detailed assessment of all potential suitable valves, the SEBIM type was selected as being the best quali­fied for this particularly arduous duty of both steam and water relief. The second major improvement con­cerns the fabrication details and materials properties of the pressuriser vessel, which have been enhanced by the use of SA 508 class 3 forgings for the three cylindrical shells and for the top and bottom heads. The nozzles on these heads are integrally forged, there­by eliminating the head-to-nozzle welds required using normal fabrication techniques. Longitudinal shell welds are also eliminated.

In order to give added assurance that the pressuriser vessel will not suffer disruptive failure in service, both the forgings and the welds are subject to redundant and diverse non-destructive inspection during fabrica­tion as well as further inspections during the plant lifetime.

10.2.7 Chemical and volume control system (CVCS)

The chemical and volume control system is one of the principal auxiliary systems connected to the reactor coolant circuit and is essential to normal operation of the reactor. It comprises mechanical plant and equip­ment located mainly on the lower floors of the aux­iliary building.

Careful control of the primary coolant chemistry in the PWR is necessary because of the potentially severe corrosive effect of the high temperature water environment on the materials of construction of the reactor circuit and on the fuel. Corrosion rates must be minimised for two main reasons:

„ • To prevent undue degradation in the structural in-
tegrity of the primary circuit pressure boundary.

• To limit the rate of formation of radioactive ‘crud’ which, when deposited around the circuit, contributes to the radioactive dose received by plant operators.

Despite careful chemistry control there will inevitably be some corrosion. Moreover, small unavoidable leak­ages of fission products through the fuel cladding and radiolysis of the coolant will also give rise to gaseous, liquid and solid impurities which must not be allowed to build up in the primary coolant. To control these impurities, the chemical and volume control system uses demineralisation and filtration processes to clean up primary coolant circulating in a bypass loop con­nected to the primary circuit.

Initial filling of the circuit with very high purity demineralised water, to which only boric acid and the necessary pH correction agent (lithium hydroxide) are added, is followed by rigorous degassing procedures to achieve a very low level of dissolved impurities,

particularly oxygen, prior to reaching operating tem­perature. Thereafter, chemistry control is exercised via the CVCS, a simplified arrangement of which is shown in Fig 2.135.

The main flow path of the CVCS consists of a bypass loop through which is circulated a small, more or less constant, fraction of the primary coolant flow extracted or letdown from one of the cross-over legs. After processing as required, the coolant is returned to one of the cold legs as a charging flow. The CVCS also provides the principal method of controlling the amount of coolant in the primary circuit by regulating the charging flow, as required by the reactor make-up control system, so as to maintain the correct level of liquid in the pressuriser.

Since the ion-exchange processes used by the CVCS for purification take place at low temperature (60°C or less) it is necessary to cool the letdown flow prior to processing. To reduce the amount of useful heat energy lost, the first part of the required temperature reduc­tion (to about 150°C) is achieved regeneratively. Further cooling and two stages of pressure reduction
then bring the letdown flow to suitable conditions for purification.

Filtration and demineralisation are used to remove suspended (particulate) and dissolved impurities re­spectively from the letdown flow. A 2-stage filtering process down to 5 fim is followed by passage through mixed-bed and, if required, cation bed ion-exchange resin beds. The filters and demineralisers, which ac­cumulate high concentrations of radioactive corrosion products and other contaminants, are heavily shielded and provided with the means of remotely changing spent filter elements and resins, which are disposed of via the radioactive waste management systems.

After filtration and demineralisation, the letdown flow reaches the volume control tank (VCT). This tank of 7 m3 capacity is maintained partly full of coolant under a continuously fed blanket of hydrogen, at about 5 bar overpressure. At this low pressure point in the system, gaseous impurities in the coolant come out of solution and mix with the hydrogen which transports them to the radioactive waste man­agement plant. The partial pressure of hydrogen in

the blanket causes hydrogen to dissolve in the reactor coolant and this maintains the dissolved oxygen content at an extremely low level because of recombination.

primary coolant is taken from the VCT by one of the two centrifugal charging pumps and supplied con­tinuously to the reactor circuit at a rate controlled either automatically or manually by the operator. The wo pumps are 10-siage horizontal machines driven pv ‘ ‘ kV electric motors supplied with power from redundant switchboards of the station essential elec­trical system. One pump only is normally in use, supplying 7.5 kg s of charging flow to the reactor t 155 bar. Most of this coolant flow reaches the circuit via the regenerative heater exchanger, in which it cools (and is heated by) the letdown flow. However, the charging pumps also supply flow via a suitable filter to the shaft seals of the reactor coolant pumps, and some of this flow also reaches the reactor circuit, the balance that leaks off being returned to the suc­tions of the charging pumps.

During normal plant operation at steady load, the net charging and letdown flows and their boron con­centrations are identical. As burn-up of the fuel pro­ceeds and its reactivity is steadily reduced, a reduction in primary coolant boron level is needed (from about 700 ppm boron at the beginning of a core cycle to about 50 ppm at the end) to maintain the optimum control rod configuration. Thus, from time to time the CVCS is used to dilute the primary coolant with fresh coolant having a lower dissolved boric acid "Content.

This is achieved by mixing demineralised water with the flow leaving the volume control tank; the excess of incoming borated coolant is diverted away to the boron recycle system (BRS) for possible re-use. Con­versely, when reactor shutdown for refuelling is re­quired, the primary coolant must be borated to 2000 ppm and a supply of concentrated boric acid solu — — tion is fed to the charging pumps, the relatively dilute letdown How being diverted away to the BRS. Other chemical additives, such as lithium hydroxide (for pH adjustment) and hydrazine (for oxygen scavenging during reactor start-up), are also introduced from a separate mixing facility. These operations involve an element of manual control.

In the event of small leakages from the primary ‘.ircuit, the reduction in inventory that would result is olivet by an increase in the charging flowrate. The charging pumps can each provide sufficient flow to make up the loss of coolant from any anticipated operational leakage, or from a single instrument line connected to the primary circuit in the unlikely event oi its rupture. The charging flow can also be increased! o compensate for shrinkage of the coolant due to temperature changes. This make-up role of the CVCS! S important in assuring reactor safety following faults involving reduction in primary coolant inventory; a larger external tank of borated water known as the ■ etueliing water storage tank (RWST) provides a se — ‘-nre supply of coolant to the charging pumps if

needed, In the very unlikely event of total failure of this route, a completely separate pair of positive dis­placement emergency charging pumps, driven by steam turbines and drawing supplies from a dedicated stor­age tank, provide a diverse and redundant alternative means of RCP seal injection and borated water make­up.

10.2.8 Control rod drive mechanism

Control of the nuclear chain reaction in the core is achieved by the movement of neutron absorbing rods arranged in clusters, called the rod cluster control as­semblies (RCCA), and by varying the concentration of boric acid in the primary coolant using the CVCS as just described. Each of the 53 RCCAs consists of an assembly of 24 individual rods which are attached to a central hub. These individual rods run in the thimble tubes of selected fuel assemblies. When they are withdrawn from the core, the RCCAs are sup­ported by the guide tubes which are part of the reac­tor internal structures. Each RCCA is held, withdrawn or inserted by a control rod drive mechanism (CRDM), which utilises latches to engage in grooves in a drive rod which is connected to the central hub of the RCCA. A CRDM is illustrated in Fig 2.136.

The actuating mechanism and drive rod of each CRDM are completely enclosed within a pressure hous­ing which is attached to a penetration in the reactor pressure vessel closure head. Each head penetration consists of a plain Inconel tube welded to the vessel head with an externally-threaded stainless steel section welded to the top of the tube.

JEach CRDM pressure housing comprises two forg- inp manufactured from type 304 austenitic stainless steel. The lower forging is the latch housing which encloses the drive rod when the RCCA is withdrawn from the care. The top of the rod travel housing is closed with a threaded plug containing a small vent valve. All joints in the pressure housings are threaded, with a canopy seal weld over each threaded joint to provide a hermetically sealed pressure bound­ary. The latches are magnetically actuated and there are thus no penetrations through the CRDM pressure boundaries.

The tops of the CRDMs are restrained by sleeves which fit over the tops of the pressure housings and locate in the holes in a targe steel plate, termed the missile shield, which is supported by four legs attached to the reactor vessel head. During normal pow-er operation of the reactor, the missile shield is attached to the sides of the refuelling cavity by tie rods which restrain the whole assembly in the event of a seismic event. It also serves to prevent any mis­siles generated in the unlikely event of a CRDM hous­ing failure from impacting vulnerable items inside the reactor building.

Each drive rod and its RCCA is held, raised and lowered by the latch mechanism which is located inside

the latch housing. Actuation is effected by a stack of three electrical coils which surround the outside of the latch housing. A latch mechanism consists of two sets of three gripper latches, operated by armatures which move when their corresponding electrical coils are energised with a direct current. The structural parts of the latch mechanism are made from type 304 austenitic stainless steel, whilst the armatures are made from type 410 martensitic stainless steel.

The lower fixed set of latches normally holds the drive rod; to raise the drive rod a sequence of DC pulses is applied to the three actuating coils. This causes the upper movable set of latches to grip the drive rod; the lower latches are then withdrawn and the upper latches raised sufficiently to allow the lower latches to grip the groove below that previously held. The upper latches are then withdrawn and pushed downwards by a spring, and at the end of this se­
quence the drive rod is raised by approximately 16 mm. The control system has a maximum operating speed of 72 such cycles per minute. Rod insertion is essentially the reverse of the sequence described above.

The 53 control rods are arranged in 6 shutdown banks and 3 control banks. The shutdown banks are withdrawn completely prior to criticality, whilst the three shutdown banks are withdrawn sequentially with a predetermined overlap to control the reactor power. When the reactor is tripped, the protection systems open circuit-breakers in the electrical supply to the rod control system, causing the coils to be de-energised on all 53 of the CRDMs. The weight of the drive rods forces the latches to withdraw as the rods fall into the reactor under gravity. A dashpot action between the RCCAs and the bottom part of the fuel assembly thimbles causes the RCCAs to decelerate towards the bottom of their fall, which takes up to two seconds.

The position of each control rod is measured by a column of detector coils which surround its rod travel housing. These coils detect the presence of the drive rod by its effect upon their electrical impedence; the top of the drive rod is between these two coils which have a significant difference in their AC im­pedance. The measurements of rod position are used by the reactor protection system to provide the core limits protection function, and also to actuate the emergency boration system in the unlikely event of failure of sufficient rods to insert fully following a reactor trip. It is also used to check that all of the rods within a bank are inside their alignment tolerance.

When the reactor vessel head is removed, the drive rods and RCCAs are left behind. Using a special tool, the drive rods are decoupled from the RCCAs and removed with the reactor upper internals, leaving the RCCAs behind in their respective fuel assemblies.

The CRDMs used for Sizewell В are identical to those used in many Westinghouse plants, this design having evolved from the original design produced by Westinghouse. Considerable operational experience has been gained with this design of CRDM which has proved to be extremely reliable. Experimental life test­ing of these mechanisms has demonstrated a design lite in excess of 2.5 million steps before wear effects would potentially start to prevent the correct stepping action. The CRDMs in some of the control banks may exceed this limit if Sizewell В is operated in a load-following or frequency regulating mode, but re­placement of latch mechanisms would not present any significant problems.

The equipment mounted on the vessel head forms an integrated head-package which is lifted off with Tie head to facilitate refuelling. This equipment in­cludes the CRDMs, the missile shield and lifting rig, together with the CRDM cabling and connectors and ‘he cooling system shroud, through which air is drawn to cool the CRDM coils. The head package for Size — w. l В has been specifically designed to minimise the amount of equipment which must be removed during refuelling operations, as this reduces both time and personnel radiation exposures. One major constraint imposed upon the design of the package was the maximum diameter which will still permit the use of the complete ring type of multistud tensioner to re­move and replace the reactor vessel flange bolting, It is considered that the optimum design has been de­veloped for minimising both the radiation exposure and the time taken to remove and replace the reactor vessel head.

The main purpose of the Act is to protect the public from ionising radiations from radioactive waste by way of use, accumulation or disposal. It also makes provisions for the discharge of radioactive materials by way of effluents, and provides for a National Disposal Service of radioactive waste not disposable by local means.

The CEGB is exempted from those parts of the Act that deal with use where the location is controlled under the Nuclear Installations Act. Mobile radioactive apparatus is not exempted since the Board may wish to use such equipment from location, to location.

Before disposing of radioactive wastes on or from a site, authorisations must be obtained from the Secretary of State for the Environment and from the Minister of Agriculture, Fisheries and Food.~These ministers will consult local authorities, river authori­ties, water authorities and other bodies similar to those which they believe proper to consult. Public and local authorities may not take into account the radio­active content of waste when exercising their powers in relation to nuisance, pollution and waste since it would be unreasonable for those subject to the 1960 Act to be also accountable to the local and public authorities for the same thing.

Where a site is the subject of a Nuclear Site Li­cence then it is not subject to this Act for the storage of waste prior to its disposal.

When neutrons are scattered by light nuclei there is a significant transfer of energy between the nucleus and the neutron, this being the moderation process. Consequently, when neutrons which are already ther — malised are scattered by light nuclei whose energy is higher than that of the neutrons, the kinetic energy ot the neutron is increased; the neutron energy spec­trum is ‘hardened’. Heavy nuclei (e. g., U-238) can aUo harden the spectrum but with a greatly reduced eiticiency. The effect arises from moderator and can temperature changes in magnox reactors. In addition, !n AGRs, iuel temperature changes contribute to spec­trum hardening as a result of collisions with the oxy — v’en atoms of the oxide fuel.

The absorption cross-section of many materials varies as the inverse of the neutron velocity (1/v). Since, for a given (constant) cross-section, the probability of interaction increases linearly with velo­city, the absorption probability is constant with neutron velocity. Graphite and steel are typical 1/v absorbers and their behaviour is unaffected by tem­perature. Certain materials, however, show significant deviation from this general rule; in the reactor context U-235, Pu-239 and Xe-135 are important non-1 v absorbers. Departures from 1/v behaviour are usually due to the presence of resonances.

U-235

The absorption cross-section for U-235 falls off faster than 1 /v. Any hardening of the thermal spectrum causes a reduction in the reaction rate relative to the overall absorptions in the lattice giving a negative contribution to the temperature coefficient.

Pu-239

The absorption cross-section of Pu-239 shows a broad resonance at 0.29 ev. This is higher than the mean energy of the thermal neutrons in a gas cooled re­actor in normal conditions. Any hardening of the neutron spectrum gives an increased number of ab­sorptions relative to U-235. Pu-239 has a larger value of i} (the number of neutrons per absorption) than U-235, giving a positive contribution to the tempera­ture coefficient.

Xe-135

This is a parasitic absorber in nuclear reactors and has a strong absorption resonance at 0.08 ev — hardening the neutron spectrum pushes neutrons out of the resonance giving a positive temperature effect.

Radial fine structure

The flux in a fuel pin is depressed due to absorption. If the neutron spectrum hardens, neutron scattering increases (constant cross-section) whereas absorption remains constant (1/v cross-section) thus tending to flatten the flux. This gives relatively less absorption in the moderator and more reactions in the fuel, i. e., a positive feedback effect.

A controlled shutdown occurs when it is dear that the reactor must be shut down but the severity of the problem does not require an instantaneous shut­down.

Magnox

On a magnox reactor the ideal procedure is in three parts. First, temperature is reduced by a few degrees to ensure that in the subsequent stages of the shut­down, when the uniformity of temperature distribu­tion across the core may be disturbed, Operating Rule temperature limits are not exceeded locally. Second, gas flow is reduced at constant reactor gas outlet temperature to achieve a substantial reduction in re­actor power. By maintaining high temperatures dur­ing the power reduction, the main turbines can be kept on load (their load being reduced to match the reduction in reactor power) to dispose of the reactor heat. Third, with gas flow at a low value, the control rods are run into the core to complete the shutdown.

This is the ideal procedure, but for various reasons it rarely occurs. The initial temperature reduction may not be necessary if the reactor is operating already several degrees below the Operating Rule temperature limits, for example, if it is operating to corrosion control limits which are more restrictive than Operat­ing Rule limits. Indeed it may be highly inadvisable to attempt a temperature reduction; if the bulk rods and sector rods are only lightly inserted into the re­actor core, for example, because of operation at high mean fuel irradiation, there may be inadequate over­ride capability to offset the negative reactivity change arising from the overall positive temperature coeffi­cient of reactivity.

Xenon override is also considered. As neutron power is reduced, particularly during the gas flow re­duction where the majority of the power loss occurs, the Xe-135 concentration will increase. This will in­troduce additional negative reactivity which must be balanced by control rod withdrawal, and where this is limited by the circumstances outlined in the previous paragraph then the shutdown will become uncontrolled and the reactor will shut itself down. However, xenon changes occur over a longer timescale, so unless the shutdown is prolonged it is unlikely to be a problem.

A useful source of positive reactivity for tempera­ture and xenon override is described in Section 5.5 °f this chapter, namely increase in reactor gas inlet temperature, but this is only used for power reduc­tions where the aim is to keep the reactor on load.

If the objective is to shut down the reactor com­pletely, there is tittle to be gained from a controlled shutdown (rather than a trip) except the advantage in being able to perform the necessary operations on the boilers, turbines and various auxiliary plant on a less-rushed timescale. As far as the reactor is concerned, it may as well be tripped, in fact this is preferable to an uncontrolled shutdown as would occur if the reactor ‘poisoned out’ on temperature or xenon. The usual procedure is to effect a tempera­ture reduction of a few degrees, then to trip the reactor w’hen the operating staff on the plant are prepared.

A controlled shutdown in which control is main­tained throughout is an advantage if it is has been initiated by a plant deficiency or abnormality which, if sustained, requires a reactor shutdown, but which may be rectified on a timescale that will permit the reactor power to be restored. Such a situation may occur, for example, if the burst can detection (BCD) system became inoperative; in this situation the Op­erating Rules require that the reactor be shut down within a specified period of time (typically half an hour), so the shutdown is started in accordance with the Operating Rules while emergency maintenance is carried out in an attempt to restore the BCD system to service.

AGR

With sufficient care an AGR can be reduced to about 30-40% power, as is current practice for refuelling, and it could then be tripped from that power; how­ever, only in very unusual circumstances would the reactor be shut down in this way.

On an AGR the conclusion on shutdown proce­dure is the same as on a magnox reactor, i. e., the reactor may as well be tripped as soon as the operat­ing staff are prepared. However the reasons for it differ. Temperature poisoning is not a problem, first because the bulk moderator temperature is main­tained largely constant, second because there is more reactivity worth invested in the grey rods and it is unlikely that they will be only lightly inserted into the core. Xenon build-up occurs but, as mentioned for magnox reactors, this is on a longer timescale. The principal reason for preferring a reactor trip is the limited regime in which the boilers must be operated, the difficulties in maintaining operation within the allowable regime and the consequences for the plant of operating outside that regime.

Scope

An operating rule limit on coolant gas pressure is set, based on vessel integrity at specified temperatures:

• Excess coolant pressure is a trip parameter indi­cative of overfilling or severe boiler tube failure, the boiler operating at a higher pressure than the reactor.

• Rate of fall of pressure is a reactor trip parameter on magnox stations indicative of a depressurisation, for example, caused by a failure of the pressure vessel or circuit.

• Low coolant pressure is also a trip parameter.

• Rate of rise of coolant pressure is drawn to the attention of the operator as an alarm and he may decide to trip the reactor manually.

Typical trip settings are given in Tables 3.5 and 3.6. Magnox reactors

For reactors with steel pressure vessels, a coolant pressure limit based on vessel temperature will be set such that an adequate margin on vessel integrity is maintained if a shock loading was to occur. This limit is normally in the form of a graph of vessel pres­sure against minimum vessel temperatures, where safe conditions are to one side of a line on the graph.

Pressure Accurate measurement of the reactor gas circuit pressure is provided to enable the operator to comply with the station’s Operating Rules. In the case of Berkeley, a digital display of reactor gas pres­sure is prominently positioned on a panel in the CCR with individual gas circuit pressures indicated on the eight boiler/blower panels. The digital display unit also has high and low pressure alarm features to at­tract the operator’s attention to significant changes from the desired level. A recorder situated in the CCR annex also provides information about trends in pressure changes and is a reliable back-up to the CCR indications. An extremely accurate bourdon tube pressure gauge, which is directly connected to the pressure vessel and situated in the reactor blower/ boiler control room, is used by the plant attendant when changes in gas pressure are required for opera­tional purposes. This is a useful instrument as it is independent of all electrical supplies and therefore the operator can verify the reactor pressure at any time without difficulty. A high reactor gas pressure trip system is installed which is part of the safety

circuits and also a warning alarm set just below the

trip lev cl.

Level of reserves The Berkeley power station Operat­ing Rules require a minimum level of 61 t of CO: to be kept in reserve for reactor coolant duty. The station’s storage capacity of 213 t of CO: is far in excess of the minimum required, but for operational requirements, this higher level is maintained as far as is practically possible. There are no direct indications in the CCR of CO: stored, but stock control is administered by the CCR supervisor who orders direct from the supplier, A daily record of CO; stocks taken from direct-reading indicators on the storage tanks is kept by the CCR supervisor and CO; usage is con­trolled by the operator. This method of stock control

T xBt. E 3 6

Typical AGR (ripping schedule

T fil t 3.6 icontd)
Typical AGR tripping schedule

S’oies

1 Trip >eiting is a function of steam pressure

2 Trip is based on integral level switches with fixed setting

3 {,) = quiescent

ihen enables the operator to have a reasonable assess­ment of CO: reserves at any time and is able to en­sure that minimum levels are maintained.

Shutdown action — coolant pressure There are auto­matic reactor trips on high pressure and high rate- ol-ehange of pressure.

To cater for maximum credible depressurisation faults on steel pressure vessel stations, emergency shut­down devices (ESD) are installed. These shutdown devices are individually tripped on high rate of change of pressure, independently of the guard lines. The rapid shutdown is to guard against coincident core distortion caused by the depressurisation fault.

Tvpical values Berkeley Hinktey Роті.-1 Oldbury

Normal operating pressure, bar ’.’9 12."5 25.17

High pressure trip, bar 7.93 13.38 26.34

High oP’6i, bar/min 0.68948 2.07 4.%

{The above are illustrative only, the pressures may be measured at different parts of the gas circuits.)

AGRs

Reactor gas pressure Apart from maintaining a cool­ant pressure to permit adequate fuel cooling, there are also important safety implications:

• Reactor pressure failing at higher than the usual rate to be expected from normal vessel leakage implies some breach in the primary circuit pressure bound­ary. The assessment is facilitated by an on-line cal­culation of mass of gas in the vessel, based on measurements of reactor pressure and temperature.

• High reactor pressure implies that either there is a boiler leak or the CCb admission valve is not closed when it should be. This prompts operator action to isolate the leaking boiler or close the CO2 valve,

• It is necessary to ensure that the reactor is not overfilled prior to the start of power operation.

Pipework is installed so that reactor gas pressure can be measured at four points under the vessel top slab, one in each quadrant above the boiler annulus. The pipework is 17.2 mm o. d. and 3.2 mm thick stainless steel. The pipework for each quadrant is routed through the quadrant instrument penetration at the + 2.3 m level and thence to the quadrant outer re­actor gas instrumentation racks.

For quadrants A and D there is a transducer, giving a 4-20 mA signal for a range 0-60 bar with an accuracy of 0.5%. For quadrant В there is a low pressure range transducer giving a 4-20 mA signal for a range 0-6 bar with an accuracy of 0.5%, The pipe­work from quadrant C is not used, being isolated by a valve in the primary isolation valve rack at the +3.9 m level and capped in the quadrant outer gas instru­ment rack.

The signals from quadrants A, В and D are used

as follows:

Indication-alarm Quadrant signal used

CCR unit desk high range indication A

CCR unit desk ІО’л range indication В

CCR diree(-wire facia low pressure alarm (via A, D

alarm amplifier set at 32 bar) (grouped)

CCR direct-wire facia high pressure alarm A. D

(via alarm amplifier set at 42.5 bar) (grouped)

CR post-trip mimic analogue indication D

CCR post-tnp mimic low pressure digital signal A, D

LIC panel indicator В

НІС post-trip mimic analogue indication A

Valves are provided to enable the transducers and alarm amplifiers to be calibrated and tested while the reactor is at power.

All signals are also fed to the data processing system.

The pipework, transducers, alarm amplifiers and ca­bling are aseismic. Gas pressure reactor trip signals are not derived from this equipment.

Gas baffle dome pressure differential This is impor­tant for safety in three respects:

• High gas baffle dome pressure differential indicates the approach towards the limit of design pressure dif­ferentia! plus the margin allowed in the design code.

• Reduced limits on gas baffle dome pressure differ­ential must be observed during refuelling to ensure that partially withdrawn fuel sleeves are not sub­jected to excessive tensile stress due to internal pres — surisation or buffeting due to excess gas cross-flow.

• Low gas baffle dome pressure differential may in­dicate that the gas flow over the fuel elements is insufficient for adequate cooling and, in particular, the flow may not be sufficient to give proper CGO thermocouple response and hence may render re­actor protection from CGO temperature signals in­effective.

In the case of Heysham 2, the pressure differential across the gas baffle dome is measured at four points, one in each quadrant of the reactor.

The pipework is 17.2 mm o. d. and 3.2 mm thick stainless steel, and is used solely for differential pres­sure measurement; no gas sampling flows are drawn down the pipework since this would affect the differ­ential pressure measurement. The pipework from each measuring point is routed through a different pene­tration for each channel, i. e., the relevant quadrant instrument penetration at the +2.3 m level.

There are four transducers measuring the pressure differential, one in each quadrant outer reactor gas instrumentation rack, approximately 6.5 m above the instrument penetration and about 3.5 m from the ves­sel wall. The accuracy is 0.5% full-scale with a range of 0-4 bar and they are designed for operation at up to 60 bar static pressure.

Each channel provides a 4-20 mA signal. The sig­nals from quadrants A and C are sent to the data processing system. Those from quadrants В and D are fed to a 2-way switch on the unit desk in the CCR so that either may be displayed on the conventional indicator on the desk.

Additionally, alarm amplifiers on quadrant В and D channels generate alarms for high gas baffle dome differential pressure {at least 2.70 bar). These are grouped to give a single direct-wired facia alarm in the CCR and are individually repeated to the data processing system. Valves are provided to enable the transducers and alarm amplifiers to be calibrated and tested while the reactor is at power.

The data processing system derives alarms from the analogue signals from quadrants A and C in the event of abnormally high and low values.

The pipework, transducers, alarm amplifiers and ca­bling are aseismic.

The magnox reactor (Table 1.8 and Fig 1.31) uses a graphite moderator, carbon dioxide coolant and metal­lic natural uranium fuel clad in a magnesium alloy known as magnox. The core is cylindrical in shape and constructed of several hundred tonnes of graphite blocks which is penetrated by vertical 100 mm dia­meter channels in which the fuel elements, typically seven or eight in number, are stacked one on top of another. To achieve the required rate of heat transfer from fuel to coolant, the upward flowing CO2 is at pressure and the magnox cladding has an extended surface in the form of fins along its length. The fuel rod is about 25 mm diameter and 1 m in length.

Nine commercial twin-reactor magnox power sta­tions have been built in the UK. In the early design the core was enclosed in a spherical steel pressure vessel and connected by large diameter ducts to external boilers. However, as the design coolant pressure was progressively increased from 9 to 19 bar, a significant change was introduced for the last two of the type: the steel vessel was replaced by a cylindrical pre­stressed concrete pressure vessel. This enclosed the core, the pressure circuit, boilers and gas circulators. The concrete pressure vessel was also adopted for the AGRs and is referred to in the next Section 9.4.2.

The power density of a magnox core is low, 0.8 MW/m3, and thus the core is comparatively very large with several thousand channels on a half metre

pitch. Thermal output ranges from a few hundred MW(th) per reactor for the earlier stations to nearly 2 GW(th) for the latest. Refuelling, a fuel element at a time with average burn-up of 5500 MWd/t, is on-load and requires a sophisticated charge/discharge machine. Access to a channel is via a standpipe from the charge face above the core, each standpipe serv­ing a number of fuel channels. At full load three or four channels need to be refuelled daily.

Reactivity control is by means of mild steel and boron steel control rods. Also flux flattening absorb­ers may be loaded and removed by the refuelling machine.