bwr 90—the advanced bwr of the 1990s

Nuclear Engineering and Design 180 (1998) 53–66

BWR 90—the advanced BWR of the 1990s

Tor PedersenABB Atom AB, S-721 63, Vasteras, Sweden

Accepted 29 October 1997

Abstract

The future global role of nuclear power will be determined by its ability to provide economical and safe energy.Nuclear power, like any other substantial contributor to the world’s energy needs, must be generated at an acceptablecost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level ofpublic acceptance, which in turn is not necessarily governed by rational assessments of the true safety andenvironmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a‘cautious evolution’; for the next decade the company will largely base its offerings to the market on its ‘evolutionary’light water reactor design, the BWR 90. This design builds closely on the experience from successful construction andoperation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average loadfactor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very bestperforming plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR designdevelopment focuses on meeting requirements from utilities as well as new regulatory requirements. A particularemphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Otherdesign improvements involve: all-digital I and C systems and enhanced human factors engineering to improve workenvironment for operators, optimization of buildings and containment design to decrease construction time and costs,and selection of materials as well as maintenance and operating procedures to even further reduce occupationalradiation exposures. Probabilistic safety assessments and life-cycle cost evaluations have become major tools in thedesign optimization work. The BWR 90 was offered to Finland in the early 1990s, and will now as the first BWRdesign be reviewed by a number of European utilities with respect to its conformance to the European UtilityRequirements (EUR); a specific EUR Volume 3 for the BWR 90 will be the final result. The paper describes someof the unique characteristics of the BWR 90, with emphasis on the features that are most important for achievingimproved economy and enhanced safety. © 1998 Elsevier Science S.A. All rights reserved.

1. Introduction

The BWR 90 standard plant design of ABBAtom represents an ‘evolution’ of the design of itssuccessful predecessor, the BWR 75, with a num-ber of design modifications, improvements andsupplements that address new licensing require-

ments and aim at meeting utility needs for low-ered cost, ease of operation and maintenance,increased investment protection, and enhancedpublic safety.

The BWR 90 design is characterized by the useof internal recirculation pumps, fine motion con-trol rod drives, a prestressed concrete contain-

0029-5493/98/$19.00 © 1998 Elsevier Science S.A. All rights reserved.

PII S0029 -5493 (97 )00297 -5

T. Pedersen / Nuclear Engineering and Design 180 (1998) 53–6654

ment, and extensive redundancy and separation ofsafety-related systems in the same way as in theBWR 75 design that was developed in the 1970s.The modifications, mostly moderate, aim at ac-commodating technological advances, new safetyrequirements and to achieve cost savings.

There is one easily distinguishable departurefrom previous designs, however; the containmentarrangement. In the new concept the connectionsbetween the drywell and the condensation pool inthe wetwell are accomplished in a quite differentway, and design measures to cope with a ‘de-graded core’ accident have been incorporated (byprovision of a core catcher arrangement andfiltered venting for the containment) in order toensure that public and environment will be pro-tected even in the event of a degraded core acci-dent situation.

The BWR 90 originally had two standard sizes,closely corresponding to the BWR 75 sizes, withnominal thermal power of 2350 and 3300 MWth,respectively. These standard sizes have later beensupplemented by a larger unit, with a nominalthermal power of 3800 MWth, taking advantageof the margins that are gained by utilization ofthe new generation of ABB Atom BWR fuel. Thesize described in the following is the 3300 MWthsize.

As noted above, the BWR 90 is not a newreactor concept; it has been developed by makingspecific changes to an established reference design,the Forsmark 3 and Oskarshamn 3 power plants,with a strong emphasis on maintaining ‘provendesign’ features unless changes would yield im-provements and simplifications.

The operating records of the company’s BWRplants show high plant availability and powerproduction reliability, and low occupational radi-ation exposure. Typical energy availability factorsover the last years are 90–95%, and the occupa-tional radiation exposures have been in the orderof 1 manSievert per plant. A basis for suchachievements is a good basic plant design; notonly with respect to systems performance andcomponent reliability, but also a design whichfrom the beginning has taken the needs formaintenance and service into consideration. Theoperating utility obviously has a profound influ-

ence on the plant performance, but even a profi-cient utility will most likely fail to achieve goodresults, if the plant design is not good enough.

A ‘suitable’ plant design involves many differ-ent aspects—the design of various systems, choiceof materials and components, their installation,radiation shielding, accessibility to components,transport routes, proper routing of ventilation air,general building arrangement, etc. The end resultwill always represent a compromise between anumber of concerns, and in this context, a co-op-eration with the Finnish utility TeollisuudenVoima Oy (TVO), with its feedback of practicalexperience, has been of great value for the devel-opment of BWR 90.

Some of the special features of the BWR 90 arereviewed briefly below. The description reflectsthe design which was offered to Finland as one ofthe contenders for the fifth Finnish nuclear powerplant project.

2. Description of the nuclear systems

2.1. Primary circuit and its main characteristics

The general reactor pressure vessel arrangementis very much the same as in the Forsmark 3 andOskarshamn 3 plants in Sweden. The steam andfeedwater lines connected to the upper portion ofthe vessel and with housings for the recirculationpump motors integrated with the bottom of thepressure vessel.

The recirculation system is based on the use ofinternal glandless pumps which provides meansfor an accurate control of the reactor power, andeliminates large break LOCAs below the corelevel.

The ‘dried’ steam is conveyed from the RPV tothe turbine plant through four steam lines. Thesteam lines connect to nozzles with built in ‘flowlimiters’, evenly distributed along the vessel cir-cumference; own medium operated isolationvalves are provided on the inside and outside ofthe containment wall.

The feedwater lines enter the containment viatwo lines, each with inner and outer isolationvalves, splitting up into four lines adjacent to the

T. Pedersen / Nuclear Engineering and Design 180 (1998) 53–66 55

RPV for connection to four nozzles, at ‘mid-height’ of the vessel. The nozzles and the internalremovable feedwater distributors are of a specialABB Atom design that ensures a ‘thermal sleeve’protection against the ‘cold’ feedwater for theRPV wall, and an efficient distribution into thedowncomer.

The RPV is provided with a pressure reliefsystem which consists of 16 safety (relief) valvesconnected evenly onto the four steam lines, withblowdown pipes leading down into the condensa-tion pool. The valves are own medium operatedvalves.

In comparison with its predecessor, the primarycircuit of the BWR 90 incorporates improvementsrelated to the design of the reactor pressure vesselproper, to the reactor core and the reactor inter-nals, to the recirculation pumps and their controlsystems, to reactor auxiliary systems, and to theinstallation and arrangement of the reactor pri-mary system in the containment.

2.2. Reactor core and fuel design

The reactor core is a typical ABB Atom BWRcore, made up of 700 fuel assemblies of theSVEA-100 type. The SVEA fuel assemblies incor-porate 5×5 subassemblies, with thin fuel rods(�9 mm in diameter), with an internal cruciformwater gap between them. This water gap signifi-cantly improves moderation and reduces localpower and burnup peaking factors. Advancedutilization of burnable absorber material (Gd2O3),axially and radially graded, in the fuel makes itpossible to achieve good axial and radial powerdistribution with low peaking factors, and goodoperating margins.

The increased operating margins with theSVEA-100 type fuel can be used to increase aver-age core power, to improve total neutron econ-omy, or for a combination thereof, and improvedthermal-hydraulic stability. For the BWR 90, aportion of the increased margins has been takeninto account to raise the power level of the reac-tor.

The control rod blades and control rod drivesfor the BWR 90 are of a well-proven design. Thecruciform rod is based on solid steel blades that

are welded together. Holes filled with B4C asneutron absorber are drilled horizontally in theblades. In the top of the rod, the absorber consistsof Hafnium which makes the rod tip more ‘grey’and provides for a long service life.

The control rod drives (CRDs) utilize two sepa-rate drive mechanisms, one electro-mechanicaland one hydraulic. The former is used for normal,continuous fine motion of the control rod—forburnup compensation or for adjustment of thepower distribution—whereas the latter is used forrapid control rod insertion (scram). The controlrods are divided up into scram groups; each groupis equipped with its own scram module, consistingof a scram tank, piping and valve. A total of 18such scram groups are provided, comprising ofeight to ten rods. The scram signal also initiates arapid run-back of the recirculation pumps and acontinuous insertion of all rods by the electro-me-chanical drives, as a back-up to the hydraulicinsertion. The latter function is also initiated byan independent scram back-up circuit of the reac-tor protection system.

The diversified means of control rod actuationand insertion (together with a generous reactorpressure relief capacity) provide, in combinationwith a capability of rapid reduction in the recircu-lation flow rate (recirculation pump run-back), anadequate countermeasure against ATWS (antici-pated transient without scram).

2.3. Primary components

2.3.1. Reactor pressure 6esselThe general reactor pressure vessel arrangement

is the same as in the Forsmark 3 and Oskarshamn3 plants; with steam and feedwater lines con-nected to the upper portion of the vessel and withthe recirculation pump motor housings integratedwith the pressure vessel at the lower portion. Thesteel vessel proper has been modified slightly,however. The cylindrical portion is made up ofcylindrical forgings in the same way as in theForsmark 3 and Oskarshamn 3 plants; this elimi-nates the longitudinal welds. The bottom portionis redesigned in such a way that large sections ofit can be made by forging; the number of welds isreduced significantly. This reduction in number of


welds is important since it reduces the amount ofin-service inspection to be carried out during therefuelling outage. The reactor vessel height is 20.9m and the width is 6.4 m.

2.3.2. Reactor internalsThe moderator tank and the core support plate

arrangement correspond closely to the BWR 75design; this applies also to the moderator tankcover. The steam separator units on top of thecover have been improved—as well as the steamdriers in the upper portion of the vessel—in orderto ensure low moisture content in the steam at theincreased power output level; the basic arrange-ment of the units is just the same as in previousplants.

2.3.3. Reactor recirculation pumpsThe recirculation system is based on the use of

internal glandless pumps driven by wet, asyn-chronous motors, supplied individually with ‘vari-able frequency–variable voltage’ power fromfrequency converters. This type of pumps hasbeen operating reliably in ABB Atom reactors(for more than three million operating hours)since 19781.

In BWR plants, the reactor power is easilycontrolled by means of the recirculation pumpflow rate. Normally, an upper level of reactorpower is established by means of control rodmanoeuvring until a certain control rod pattern inthe core has been attained, and then adjustmentsof the recirculation flow rate are utilized to con-trol the power level. A BWR is characterized bythe presence of void in the core coolant duringnormal operation, and this yields a strong feed-back of coolant flow rate; an increased flow rateresults in a decreased void content and a subse-quent increase in reactor power. Therefore, theinternal pumps provide means for rapid and accu-rate power control in the high power (or normaloperating) range, and they are also advantageousfor load following purposes.

2.4. Reactor auxiliary systems

A shut-down cooling system with one highpressure and one low pressure loop is providedfor the ‘normal decay heat removal’ functionwhen the reactor is shut down to cold conditions.A reactor water cleanup system (RWCU), with aradial flow type of deep bed filters, heat exchang-ers, and pumps, draws water from the shut-downcooling system nozzles and returns it as purgeflows through the control rod drives or dischargesdirectly into the vessel.

Other auxiliary systems serve to cool and cleanthe water in the condensation pool in the contain-ment wetwell and the water in the reactor serviceand spent fuel storage pools on top of the con-tainment structure.

The main development objective related to thereactor auxiliary systems was to evaluate possiblesimplification of their design in order to achievecost reductions and more straight-forward opera-tion. The reactor water cleanup system can betaken as an example on this review. In previousplants, a certain flow rate of reactor water, apercentage of the full power feedwater flow rate,was continuously passed through the RWCUfilters, and a forced flow mode (at twice the flowrate) was initiated when needed. In BWR 90, theRWCU operation is controlled by the waterchemistry in the reactor; during normal full poweroperation cleanup needs are limited and only asmall reactor water flow is passed through theRWCU, but whenever measurements show aneed, the RWCU is taken into operation at fullcapacity. This reduces the heat losses, etc. andtherefore yields ‘cost reductions’.

3. Description of turbine generator plant systems

3.1. Turbine generator plant

The power conversion process is depicted onFig. 1. The turbine plant design of the BWR 90 issimilar to that of the modern reference plantsForsmark 3 and Oskarshamn 3. The nominalpower output of the turbine unit will be 1100–1240 MWe depending on the site conditions, in

1 Such internal pumps are adopted also by other BWRvendors, in the ABWR plants.


Fig. 1. BWR 90—the power conversion process.

particular with respect to the temperature of thecooling water.

The saturated steam from the reactor vessel isconveyed to the admission valves of the highpressure cylinder via the four steam lines. Afterexpansion through the HP unit, the steam passesthrough a steam moisture separator unit and asteam reheater, on its way to the admission valvesof the three low pressure turbine cylinders.

A full-capacity steam bypass system is providedto enable dumping the full nominal steam flowdirectly to the main turbine condenser in the eventof certain disturbances, in order to avoid pressuresurges, and corresponding power peaks, in thereactor.

The generator is a two-pole type turbo genera-tor, designed for continuous operation with hy-drogen as the cooling medium for the rotor andwater as the cooling medium for the stator wind-ings. Its rotor is directly coupled to the turbine.The electric power is transmitted to the externalgrid via individual, isolated air-cooled generatorbuses incorporating a generator breaker, and amain transformer.

During normal power operation, the steam flowto the condenser amounts to about 60% of thetotal steam flow, but the condenser system isdesigned to accommodate the full steam flow for a

limited time period; the steam flow shall be re-duced to 60% within 20 s to avoid a reactor tripdue to too high condenser pressure.

The condenser is cooled by the circulating seawater system which typically incorporates threeelectrically driven pumps; loss of one pump willcall for a power reduction, but will not yield aturbine trip in the short term.

3.2. Condensate and feedwater systems

The condensate is pumped forward to thedeaerator (or the feedwater tank) through lowpressure heaters and a condensate clean-up systemwith ion exchange filters by means of three 50%condensate pumps.

The feedwater system consists of the main feedpumps, two high pressure feedwater heaters, andassociated piping. There are three 50% electricalmotor driven main feed pumps, drawing from thedeaerator (the feedwater tank). Drainage from thehigh pressure heaters is routed to the deaerator.The power supplies to the FW pumps are utilizingstatic converter units which eliminate the largeinrush currents at direct-on-line starting, permit-ting reduction of the requirements on ‘voltagestability’ (or rather short circuit strength) of theauxiliary power supply system busbars. Feedwater


Fig. 2. Plant instrumentation and control system structure.

flow control is achieved by adjusting feed pumpspeed and the feedwater flow control valves.

4. Instrumentation and control systems

4.1. Design concept, including control room

Modern process control and communicationtechnology is applied to the BWR 90—its controland instrumentation systems are mainly based onmicro-computers. Communication with the con-trol room or the process systems is realized bymeans of distributed functional processors. Thus,the protection and control system configuration ischaracterized by decentralization. The arrange-ment satisfies the requirements of redundancy andphysical separation. It includes intelligent self-monitoring of protective circuits. The structure ofthe instrumentation and control systems of theBWR 90 plant is depicted in Fig. 2.

The use of serial communication links guaran-tees interference-free performance and reduces ca-bling. Using hardware modules and basicsoftware from a standard industrial digital systemwhich is available on the market, minimizes

maintenance and the necessary stock of spareparts. The arrangement will also tend to improveavailability, since components can be replacedquickly and simply by using the engineering toolsincluded in the standard system. A very importantaspect is that the application software is alsostandardized to simple program functions.

The decentralized configuration, combined withthe use of isolation devices, reduces the safetyconcern of a damaged control room. If the con-trol room should become unavailable, the operat-ing personnel may supervise the process from aseparate emergency monitoring centre. The con-cept allows substantial reduction of space, andhas resulted in savings in terms of reduced build-ing volumes.

The man-machine communication in the con-trol room is facilitated by a consistent use ofvideo display units (VDUs), keyboards, and dis-play maps. The main control room contains sev-eral work positions, each equipped with a numberof VDUs. Typically, one VDU will display a totalview of the process in interest, another willprovide a list of alarms, and a third VDU willdisplay a diagram with sufficient detail to facili-tate operator action.


Fig. 3. Main control room arrangement.

This arrangement is supplemented with a spe-cial overview panel, on which an ‘overview’ ofplant functions and status is provided by conven-tional instruments as well as computer-basedVDU displays (VDU projections or EL displays).

The overview presentation shows the main pro-cess in the form of a flow diagram and indicatesthe status (normal, disturbed or failed) of variousplant functions in correspondence with the oper-ating instructions for the plant. It is visible to alloperators in the control room, as indicated onFig. 3.

The status of safety-related systems and func-tions is presented in a similar way, in accordancewith the organization of the Emergency OperationProcedures (EOP). The parameters of immediateinterest in a disturbance situation are presented ina direct form. This means that the reactor pres-sure vessel with in- and outflow connections, to-gether with neutron flux, water level, and reactorpressure, as well as control rods fully in (or not),are displayed directly. Other safety functions areindicated as normal, disturbed or failed in a simi-lar way as for the plant overview, with detailedinformation at the reactor operator’s desk.

In this context it can be noted that the digitalRPS (Reactor Protection System) is diversified forevents with a frequency of more than once perreactor life time. Diversity can include hardwiredas well as other digital technology.

The main computer has the task of collectinginformation from the process control systems, andit communicates with the distributed micro-com-puters via serial links. The main computer com-piles information and generates reports, such asdaily and weekly operation reports, reports ofperiodic testing, actual status reports, and distur-bance reports. During normal plant operation, themain computer will present occurrences on VDUdisplays in the control room and in a special‘observation room’.

4.2. Reactor protection and other safety systems

The reactor protection system (RPS) and theother safety-related control systems are in thesame way as for the BWR 75 built up in afour-division concept; process monitoring, signaltreatment and conditioning take place in fourindependent channels (or divisions). Trip func-tions are generally generated in two-out-of-four


Fig. 4. BWR 90—single line diagram.

coincidence logics in each individual ‘outgoing’division for all RPS functions, except for theemergency core coolant (ECC) injection; the lattercircuit operates without the coincidence logic in a‘one-out-of-one’ mode within each division sincean initiation of an ECC division would not signifi-cantly impair the operation of the reactor.

The most visible difference compared with thepredecessor is the deployment of digitized technol-ogy; the control equipment structure is based onutilization of a range of standardized micro-com-puter units. The basic software of these units aregenerally standardized to simple program functionsin such a way that process functions can be realisedeasily; the designer can easily transfer the desiredprocess flow logic to the digitized system. Oneadvantage of this approach is that the standardizedmicro-computer units can be thoroughly testedbefore delivery to the plant site. Another advantageof the digitized approach is that the control systemssignals are fully ‘normalized’ to low energy signals;the risk of voltage surges in equipment in thecontrol room area is reduced significantly.

5. Electrical systems

5.1. Operational power supply systems

The single line diagram of the electrical power

systems is shown on Fig. 4.In the BWR 90, the ratings of some of the

major plant loads have been reduced by designchanges in process systems, and the main feedwa-ter pumps in the turbine plant have been providedwith static power supply converters. Modernswitchgear components, having higher short cir-cuit current ratings, are also now available, andconsequently a significant simplification of thestructure of the auxiliary power supply systemshas been made possible.

Another visible feature is the simplification atthe DC distribution level; DC distributions atseveral voltage levels for power supply to varioustypes of control equipment have been replaced bypower supply from battery-backed AC distribu-tion, using distributed AC/DC converters for thesupply to the different types of equipment.

5.2. Safety-related systems

The electrical power systems for safety-relatedobjects are strictly divided into four independentand physically separated sub-divisions—a princi-ple that is implemented in the operating BWR 75plants and maintained in the BWR 90. Thismeans that there are four sets of diesel-backedbusbar systems, four diesel generators, four AC/DC-DC/AC converters with intermediate batter-


ies, having sufficient capacity to ensure the powersupply to safety-related equipment for 2 h, andfour battery-backed AC busbar systems—allwithout interconnections.

The structure has been simplified comparedwith the predecessor; the medium voltage busbarsystems have been eliminated and the diesel gener-ators operate at a low voltage level (660 V) fol-lowing a reduction in size of the pumps of the lowpressure coolant injection system, and a numberof battery systems and corresponding DC distri-butions have been replaced by local AC/DC con-verters supplied with power from thebattery-backed AC busbars.

6. Operating characteristics

A schematic overview of the interrelationshipbetween the reactor control systems with respectto operation is shown in Fig. 5.

The internal recirculation pumps provide meansfor rapid and accurate power control and areadvantageous also for load following purposes.The BWR 90 plant is characterized by a capabil-ity to accept a 10% step change in power with anequivalent time constant of down to 5 s, andramp load changes of 20% per min is accepted. Inthe high power range, between 70 and 100% ofnominal power, daily variations with the abovechange rate can be accommodated without restric-

tions; for wider power variations, the extendedrange is achieved by control rod pattern adjust-ments—at a rate of change of 1–2% per min.Daily load following in a 100–40–100% cyclewith (1-) 2 h ramps can be accommodated.

The plant power control system is providedwith a frequency control mode in which a certainportion of the output (typically 92%) is assignedfor participation in the grid frequency controlstrategy—with an adjustable deadband and anadjustable amplification factor.

With respect to operating characteristics it mayalso be noted that the plant is designed to with-stand a full load rejection without being tripped;the plant will shift to house load operation, beingprepared for a return to normal operation. Theplant is further designed to avoid a reactor trip inthe event of turbine trips, as long as the turbinecondenser remains available for steam dumping,and to withstand certain grid voltage disturbances(voltage drops due to short circuits and otherelectrical faults on the grid) without being discon-nected from the grid.

7. Safety concept

7.1. Safety requirements and design philosophy

The engineered safety systems in BWR 90 arecharacterized by their consistent division into fourredundant and physically separated subsystems,as illustrated by Fig. 6. This concept that wasintroduced already in the TVO I and II plants inFinland and further developed in the Forsmark 3and Oskarshamn 3 plants in Sweden, has beenreconfirmed as constituting an optimal arrange-ment with respect to safety, layout and maintain-ability.

For the emergency cooling systems, this meansthat the four subsystems are located in their ownbays, adjacent to the reactor containment andsurrounded by thick concrete walls. The physicalseparation is maintained all the way to the ulti-mate heat sink. The individual compartments forsafety-related subsystems and components consti-tute separate fire areas and fire cells. The contain-Fig. 5. BWR 90—reactor control systems.


Fig. 6. BWR 90—general configuration of ECCS.

ment which is of pressure suppression type, isinerted by means of nitrogen during operation.

As in the case of the emergency cooling sys-tems, the safety-related auxiliary electrical powersupply equipment is divided into four independentand physically separated parts, or subdivisions,and the reactor protection system operates on atwo-out-of-four logic for signal transmission andactuation.

The four safety-related, standby power dieselgenerators with their ancillaries are installed intwo diesel buildings, located at opposite sides ofthe reactor building; this provides a high degree ofphysical protection with respect to external im-pacts, e.g. against a crashing aircraft. These build-ings also house safety-related auxiliary powersupply and control equipment, as well as pumpsand heat exchangers for safety-related coolingsystems.

The capacities of the emergency core coolingsystems suffice to provide water under all postu-lated pipe break conditions. This statement is alsovalid assuming that only two of the four redun-dant subsystems are operable; one is out of opera-tion due to maintenance, etc. and one ispostulated to fail (the single failure criterion).

In addition to deterministic analyses, and simu-lations performed with thermal-hydraulic com-puter codes, a level 1 PSA study has beenperformed for the BWR 90. It was adapted to theoff-site electrical power grid conditions at the

Olkiluoto nuclear power station in Finland and torecent data on common cause failures obtainedfrom research work sponsored by the SwedishNuclear Power Inspectorate (SKI). The PSAstudy addresses internal events only since previousstudies have demonstrated that external events donot contribute significantly to the core damageprobability. The PSA shows that LOCA eventsleading to core melt are extremely unlikely whichis typical for BWRs with internal recirculationpumps.

7.2. Safety systems and features

With respect to design for safety, an importantpoint of discussion in the nuclear community inrecent years relates to the concepts of redundancy,diversity, and passivity. All the three concepts areassociated with pros and cons.

It is necessary to strike a balance among thesedesign aspects and to implement that balance inspecific designs. Since BWR 90 is based mainly ontechnology used in operating plants, the balancehere leans towards redundancy, separation anddiversity. Inherent safety features such as dis-tinctly negative reactivity coefficients for the reac-tor core during all operating conditions,utilisation of heat-resistant materials and equip-ment with low fire risk, and avoiding high fireloads in safety areas of the plant, represent tradi-


tional features of nuclear safety engineering, andremain cornerstones for the reactor designs.

A high degree of automation has been pursued,as in the BWR 75 design; no operator action isneeded within 30 min of a disturbance that couldthreaten safety barriers.

7.3. Se6ere accidents (beyond-design-basisaccidents)

When the design review of the BWR 90 wasinitiated, regulatory developments indicated aneed to strengthen the capability of the reactorcontainment to withstand the effects of a coremelt accident.

The essential features of the BWR 90 contain-ment (cf. Fig. 7) that have been introduced toachieve an enhanced containment capability toconfine radioactive products and protect the envi-ronment against large releases of radioactive mat-ter, also in the event of a degraded core accident,are:1. The blow-down of steam to the suppression

pool passes through vertical concrete path-ways to horizontal openings between drywelland wetwell.

2. The relief pipes from the safety/relief valvesare drawn into the suppression pool via thelower drywell rather than penetrating the dry-well–wetwell intermediate floor.

3. A pool is provided at the bottom section ofthe lower drywell for the purpose of collectingand confining fuel melt debris. The pool isprovisionally assumed to be permanently filledwith water to enhance passive safety.

These arrangements improve the reliability ofthe pressure–suppression system and reduce theprobability of containment leakage during an ac-cident. In addition, the containment vessel can bevented, manually or passively through a rupturedisc, to the stack through a filter system, installedin the reactor building. This filter is similar to thefiltered venting systems installed at all nuclearpower plants in Sweden. Arrangements are alsomade to enable filling the containment with waterto the level of the top of the core, in order toestablish a final stable state following a severeaccident involving core damage; this water is sup-plied to the containment spray system and theproviding system uses a completely independentwater source and power supply.

8. Plant layout

8.1. Buildings and structures, including plot plan

The plant and buildings of the BWR 90 are laidout and designed to satisfy aspects of safety,maintenance and communication in a balancedway. The layout is strongly influenced by safetyrequirements, in particular the physical separationof safety-related equipment.

With respect to building layout and arrange-ment ABB Atom has traditionally favoured aco-ordinated and compact building complex; thenumber of doors and transport openings, releasepoints, transport routes, etc. can be kept low andsupervision becomes easier.

The general arrangement of the buildings isdepicted in Fig. 8. It is characterized by a divisioninto an essentially nuclear and safety-related por-tion, consisiting of the reactor building and thediesel buildings, and a more conventional portionthat comprises the turbo-generator and auxiliarysystems of the plant. The ‘conventional’ part isseparated from the former by a wide communica-tion area. This arrangement is advantageous whenFig. 7. Severe accident mitigation features.


Fig. 8. General layout, plot plan.

Nevertheless, BWR 90, like previous plants, ischaracterized by a fairly spacious layout, whichensures easy access to the plant components. Theinstallation and ventilation principles are main-tained and the material specifications even morestringent; hence, low occupational exposures (1manSievert and lower) are anticipated also for theBWR 90 plants.

8.3. Containment

The primary system, the reactor coolant pres-sure boundary, and important ancillary systemsare enclosed in the primary containment, which isa cylindrical prestressed concrete structure withan embedded steel liner—as in all previous ABBAtom plants. The containment vessel, includingthe pressure–suppression system and other inter-nal structural parts as well as the pools above thecontainment, forms a monolithic unit and is stati-cally free from the surrounding reactor building,except for the common foundation slab. A steeldome provides a ‘removable’ closure of the shaftabove the reactor pressure vessel. (Fig. 9)

The primary containment is of pressure–sup-presion type, with two major compartments—adrywell and a wetwell. The drywell represents thevolume that surrounds the RPV, with an upper

building the plant as well as during plant opera-tion, since the conventional part does not interferewith the nuclear part.

The building arrangement is also characterizedby a system of communication routes connectedto the wide communication area for personneland equipment, between and inside buildings, thatserves to facilitate maintenance, inspection andrepair work by ensuring good accessibility toplant equipment.

The safety-related portions of the building com-plex, the reactor building with the reactor con-tainment, the adjacent diesel buildings, and thecontrol building, are designed to withstand theeffects of earthquakes. The standard nuclear is-land is designed to sustain a ‘safe shutdown earth-quake’ (SSE) of 0.25 g; higher SSE levels can beaccommodated by further strengthening of struc-tures and some design modifications.

8.2. Reactor building

The reactor building encloses the primary con-tainment with the primary circuit and houses allprimary process and service systems for the reac-tor, such as handling equipment for fuel and maincomponents, fuel pools, reactor water cleanupsystem and engineered safety systems.

In comparison with previous plants, a substan-tial reduction of building volumes has beenachieved, implying a significant cost reduction. Fig. 9. Containment.


portion (basically, extending from a level flushwith the bottom of the core and upwards) anda lower portion located below the RPV (andbelow the core). The wetwell is separated fromthe drywell by a partition floor and a cylindri-cal wall; the lower portion of this separated vol-ume is filled with water—the condensationpool, whereas the upper portion serves as a gascompression chamber. In the event of drywellpressurization, eg. due to a LOCA inside thecontainment, drywell atmosphere together withsteam will be pushed into the condensation poolvia a horizontal passage arrangement throughthe separating wall; non-condensibles will collectin the gas compression chamber whereas thesteam will condense in the pool water.

The blowdown pipes from the safety (relief)valves in the pressure relief system are routedthrough these horizontal passages, leaving thepartition floor without penetrations; the proba-bility of a degraded pressure suppression func-tion has been reduced to a very low level.

The containment design of BWR 90 incorpo-rates also some features that aim at protectionof the public and the environment against majorreleases of radioactive material even in severeaccident situations involving core degradationand core damages. To this end the containmenthas been provided with an overpressure protec-tion system which by means of rupture disksautomatically, and in an entirely passive way,will relieve excessive pressure to the stack via afilter system; this will prevent serious land con-tamination also in such very unlikely situations.Besides, the central, lowermost portion of thelower drywell has been made as a pool (with orwithout water during operation) with cooledsurfaces; this volume serves to collect, confineand cool possible molten debris from the reac-tor in such accident situations.

8.4. Turbine building

The turbine building of the BWR 90 is di-rectly adjacent to the reactor building, and themain turbine generator shaft points away fromthe reactor proper. The HP turbine is located atthe reactor building end, and the generator at

the other; isolated phase buses lead from thegenerator to the generator circuit breaker andfrom there to the main (step-up) transformer,with branching offs to two three-winding planttransformers for the auxiliary power supply.The main transformer and the plant transform-ers are located immediately outside the turbinebuilding.

8.5. Other buildings

8.5.1. Diesel buildingsThe diesel buildings are located at the corners

of the reactor building and structurally inte-grated with it to enhance the protection againstthe effects of earthquakes. The diesel buildingscontain most of the equipment of the safety-re-lated systems that is located outside the reactorbuilding; they house the four divisions of dieselgenerators and busbars of the diesel-backedauxiliary power supply system, the AC/DC–DC/AC converters, batteries and busbars of thebattery-backed power supply system, the safety-related control equipment, pumps, valves andheat exchangers for the intermediate closedcooling systems, and pumps and valves for theservice water system. The innermost part of thebuildings (as seen from the front of the reactorbuilding complex) is provided with a strength-ened structure to protect against a crashing air-craft; this strengthening and the location on twosides of the reactor building will prevent acrashing aircraft from damaging more than onedivision.

8.6. Control building

The control building which is located at oneside of the reactor building, with one of thediesel buildings in between, houses the maincontrol room and its ancillaries, the centre forhandling work permits for maintenance activitiesin the plant, and the entrance controls for ac-cess to the controlled and uncontrolled areas ofthe plant. It is the main entrance to the plantbuilding complex and contains a reception deskand dressing rooms for personnel with lockersand showers, etc.


9. Project status

The development of the BWR 90 started in1986 as a review of the ‘lessons learned’ fromprevious plant projects; in particular, from design-ing and commissioning the Forsmark 3 and Os-karshamn 3 nuclear power plants in Sweden. Theconceptual design, and most of the basic designhas been completed, and it was offered to Fin-land, as one of the contenders for the fifth Finnishnuclear power plant project.

Current design and engineering activites arebeing focused on detailed analyses of severe acci-dent sequences, studies of the possibilities forfurther improvement of the containment, proce-dures for the verification of digitized controlequipment, and optimization of plant construc-tion activities.

With respect to licensing, it may be noted thata design certification process does not exist inSweden, and reference is therefore made to theclose relationship with the BWR 75 design, and tothe licensing discussions that have taken placewith STUK, the Finnish licensing authority, inconnection with the ‘Finland V’ project.

Construction can build directly on the experi-ence from previous projects, and the constructionactivities have been analyzed by the team of civilengineering, installation and commissioning su-pervisory personnel that built and commissionedthe Oskarshamn 3 Nuclear Power Plant in 57months from the first pouring of structural con-crete to start commercial operation; for the BWR90 plant the construction time is expected to besimilar or shorter.

It may also be noted that an agreement hasbeen signed between ABB Atom and a number of

the utilities in the EUR (European Utility Re-quirements) group to perform an assessment ofhow the BWR 90 design meets the specificationsof the EUR; this study is scheduled for comple-tion in about 1 years time.

10. Note added in proof

The author recommends the following refer-ences for further information: Rastas andSundqvist (1990), Tiren (1990), Lonnerberg andPedersen (1993, 1994a,b), Jonsson and Pedersen(1994), Ivung and Tiren (1996).

References

Ivung, B., Tiren, I., 1996. BWR 90, The next generationnuclear power plant. Proc. Int. Semin. on New GenerationNuclear Power Plants, Warsaw, Poland, 25–27 September1996.

Jonsson, N.O., Pedersen, T., 1994. BWR 90, the advancedevolutionary BWR; some safety aspects of the design.Proc. ARS ’94 Int. Topical Meeting on Advanced ReactorsSafety, vol. 2, ANS, pp. 643–650.

Lonnerberg, B., and Pedersen, T., 1993. In: Peterson, P.F.(Ed.), BWR 90: an evolutionary ABWR plant for the nextdecade(s). Proc. 2nd Nuclear Engineering Joint Conference1993, vol. 2, ASME, New York, pp. 633–638.

Lonnerberg, B., Pedersen, T., 1994a. Le BWR 90, un reacteursophistique a eau bouillante. RGN Revue Generale Nucle-aire, No. 6, pp. 496–500.

Lonnerberg, B., Pedersen, T., 1994b. BWR 90: A Sophisti-cated Boiling Water Reactor. RGN International Edition,vol. B, pp. 60–64.

Tiren, I., 1990. BWR 90—an advanced nuclear power plantfor Finland. ATS Ydintekniikka (Finland) 19 (1), 9–20.

Rastas, A., and Sundqvist, C., 1990. Advanced LWRs: AFinnish-Swedish Proposal. ENC ’90 Trans. vol. 1, VerlTU8 V Rheinl, Koln, pp. 526–540.

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bwr 90—the advanced bwr of the 1990s

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