Advanced boiling water reactor, the next generation, Notas de estudo de Engenharia de Produção
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Advanced boiling water reactor, the next generation, Notas de estudo de Engenharia de Produção

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ADVANCED BOILING WATER REACTOR, THE NEXT GENERATION
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Advanced boiling water reactor, the next generation-status and future - Nuclear Science Symposium and Medical Imaging Conference, 1991., Conference Record of the 1991 IEE

ADVANCED BOILING WATER REACTOR, THE NEXT GENERATION; STATUS AND FUTURE

Steven A. Hucik, GE Nuclear Energy 175 Curtner Ave., San Jose, CA. 95125

Abstract

The ABWR is an advanced light water reactor designed by an international team of engineers and designers to address the utility and public needs for the next generation of power plants. It incorporates major innovations and includes the best from BWR designs in Europe, Japan, and the US.

The major emphasis in the design of the plant has been on improved operability and maintainability. This has led to an overall plant design that is simpler to build, maintain, and operate and at the same time has significantly enhanced safety features including several features that are inherent to BWR designs and ensure "passive" responses to transients and accidents. The ABWR design has several additional features that ensure that offsite doses would be extremely low following a severe accident.

This paper also discusses the other key features of the A B M : internal recirculation pumps, fine motion control rod drives, digital control and instrumentation, multiplexed fiber optic cabling network, pressure suppression containment, structural integration of the containment and reactor building, and advanced turbine/ generator with 52" last stage buckets.

The 1356 MWe ABWR design is being applied as a two unit project by the Tokyo Electric Power Co., Inc. at its Kashiwazaki-Kariwa site in Japan. On May 15, 1991, Japan's Mmistry of International Trade and Industry (MITI) formally announced the granting of the "Establishment Permit" to Tokyo Electric Power Company for constructing two ABWRs at the Kashiwazaki site. This licensing milestone culminates the successful safety review in Japan and clears the way for construction of the two ABWR units. Construction began September, 1991 and commercial operation is scheduled for 1996, with the second unit one year later.

INTRODUCTION

The ABWR design is based on design, construction and operating experience in Japan, USA, and Europe, and was jointly developed by the BWR suppliers, General Electric Company, Hitachi, Ltd., and Toshiba Corporation, as the next generation BWR. The Tokyo Electric Power Co. (TEPCO) has provided leadership and guidance in the establishment of the ABWR and has combined with five other Japanese utilities (Chubu Electric Power Co., Chugoku Electric Power Co., Ho :uriku Electric Power Co., Tohoku Electric Power Co., and Jagm Atomic Power Co.) in participating and providing support for the test and development programs. 0-7803-0513-2/92$0.3.O0 QIEEE

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The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. Conceptual design followed by detailed design engineering of the ABWR has progressed to the point where the Tokyo Electric Power Company, Inc. announced the selection of GE Nuclear Energy, Hitachi Ltd., and Toshiba Corporation to design and construct the two lead ABWRs, Units 6 and 7 at the Kashiwazaki-Kariwa Nuclear Power Station. The three companies form an international joint venture to design the plant and supply equipment.

BWR EVOLUTION

The evolution of the BWR has occurred in two major areas - the reactor system and containment design. This evolution resulted from design enhancements and experience gained from operating reactors (including abnormal occurrences) and testing programs.

Throughout the BWR evolution, there has been an ovemding trend toward simplification and optimization. The ABWR is the result of this progressive design simplification of the BWR and its containment structure. Systematic review of both the technical merits and cost have played a key role in achieving designs that meet all objectives. This thorough evaluation by the designers and subsequent review by the utility sponsor, TEF'CO, has helped keep the effort focused and achieve excellent results.

PLANT DESIGN OBJECTIVES

The major ABWR plant objectives were defined very early in the ABWR development in close cooperation with TEPCO. These overall plant objectives were selected to mainly improve the performance and safety and reduce costs.

The major plant objectives which guided the selection of new technologies and features of the design are:

(1) Enhance plant operability, maneuverability and daily load following capability.

(2) Increase plant safety and operating margins. (3) Improve plant availability and capacity factor. (4) Reduce occupational radiation exposure. (5) Reduce plant capital and operating costs. (9 Reduce plant construction schedule.

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ABWR FEATURES

The ABWR design represents the integration of eight years of conceptual development and design along with an extensive confiiatory test program.

Increased Plant Output and Turbine Design

The ABWR plant is designed for a rated thermal output of 3926 MWth which provides for an electrical output in excess of 1350 MWe. In order to improve plant efficiency, performance and economy, the turbine design incorporates a 52-inch last stage bucket design. Combined moisture separator/reheaters remove moisture and reheat the steam in two stages. Also to help increase plant output and reduce cost, the design has incorporated the concept of both high pressure and low pressure pumped-up drains. Rather than cascading the heater drains back to the condenser, the pumpup drain system takes advantage of this waste heat and injects it back into the feedwater ahead of the heaters. This concept has increased the generator output nearly 5 MWe, has helped to reduce the capacity of the condensate polishers, and has also reduced the size requirements for both the high and low pressure heater areas. The overall design has made optimal use of these design improvements to maximize plant output and reduce cost.

Improved Core and Fuel Design

The ABWR core and fuel design goal was improved operating efficiency, operability, and fuel economy. This was accomplished primarily by utilizing PCI-resistant (barrier) fuel, axially zoned enrichment of the fuel, control cell core design, and increased core flow capability. The use of minimum shuffle fuel loading schemes reduced refueling times, while fuel burnup increased to higher values provides for longer continuous operating cycle capability and improved fuel cycle costs.

The axially zoned fuel, with higher enrichment and less gadolinia (Gd-absorber) content in the upper half of the fuel rods, allows the axial power distribution to be kept uniform throughout the operating cycle. This feature assures a higher thermal margin that together with the other core design features results in improved fuel integrity, plant capacity factor, and operational flexibility. The axially zoned fuel eliminates shallow control rods which control the axial power shape and the control rods are only used to control reactivity.

The core design employs the control cell core concept successfully applied to many of the operating plants. In this design all of the control rods are fully withdrawn throughout an operating cycle except for those in the control cells. Each control cell consists of four depleted fuel bundles surrounding a control rod. Only these control cell rods move to control reactivity. This minimizes the operator's tasks of manipulating control rods during the cycle to control reactivity or for power distribution shaping. This design also improves capacity factor since the control cell eliminates the need for rod sequence exchanges. The flat hot excess reactivity minimizes rod adjustments during the cycle.

The capability for excess core flow above rated of greater than 11% provides for several benefits. Daily load following from 100% to 70% to 100% power (in a 14-1-8-1 hour cycle) is easy using core flow rate adjustment and no control rod movement. For both maximum use of the excess flow and slight control rod adjustment, load following of 100%-50%-100% is easily attainable. In addition, the excess flow capacity allows for spectral shift operation to provide additional bumup with all rods out to increase operational flexibility, extend operation, and reduce fuel cycle costs.

Reactor Vessel Incorporating Internal Pumps

The most dramatic change in the ABWR from previous BWR designs is the elimination of the extemal loops and the incorporation of internal pumps for reactor coolant recirculation. The reactor pressure vessel (RFV), and core internals have been optimized for the internal pump concept. As shown in Figure 1, all large pipe nozzles to the vessel below the top of the active fuel are eliminated. This alone improves the safety performance during a postulated Loss of Coolant Accident (LOCA) and allows for decreased ECCS capacity.

SPRAY mHEAD VENT

Figure 1. RPV and Internals 1378

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The RPV is 7.1 m in diameter and 21 m in height. The reactor vessel height and total volume have been minimized, which has resulted in reduced volume requirements for the containment and reactor building @/El). In-service inspection (ISI) has been reduced due to the incorporation of internal pumps and elimination of the recirculation pipe nozzles and reduced amount of vessel welding during vessel fabrication. The RPV has a single forged ring for the internal pump mountirlg nozzles and the conical support skirt. Forged rings are also utilized for the core and upper regions of the vessel shell sections. The elimination of the extemal recirculation piping and the use of vessel forged rings has resulted in over a 50% reduction in the weld requirement for the primary system pressure boundary.

The reactor vessel has been designed to permit maximum IS1 of welds with automatic equipment. This will help minimize manpower and reduce radiation exposure. Other features of the ABWR RPV design include main steam outlet nozzles containing a reduced diameter throat and diffuser which is used to measbre steam flow. This also acts to restrict flow and reduce loads on the reactor internals and reduces the containment loads during postulated LOCA. The steam dryers and separators are of the improved lower pressure drop design developed for B W 6 . This lower pressure drop contributes to increased stability margins and lower pump power costs.

The ABWR incorporates ten internal recirculation pumps (see Figure 2) located at the bottom inside of the RPV. This simplifies the nuclear boiler system and allows for compact space requirements in the RCCV and R/B. Elimination of the extemal recirculation loops has had many

SHROUD SUPPORT

PUMP IMPELLER

\ I l l ~ c REACTOR VESSEL PURGE WATER INLET

COOLING WATER OUTLET

MOTOR CASING

PUMP SHAFT

~ ROTOR SHAFT

STATOR

_ _ S H A F T COUPLING STUD

THRUST BEARING

COOLING WATER INLET

Figure 2. Reactor Intemal Pump (RIP)

advantages. Key advantages have been the reduction in containment radiation level by over 50% compared to current plants, lower pumping power requirements. The excess flow provided by the pump design has enhanced plant operation and allows for full power operation with one pump out of service.

The internal pump is a wet motor design with no shaft seals. This provides increased reliability and reduced maintenance requirements and hence, reduced occupational radiation exposure. These internal pumps have a smaller rotating inertia, and coupled with the solid-state variable frequency power supply, can respond quickly to grid load transients and operator demands. These pumps are now accumulating plant operating experience in several European BWRs. Improved designs have also been tested in Japan in testing programs now underway.

Fine Motion Control Rod Drives (FMCRD)

The ABWR incorporates the electric-hydraulic FMCRD, which provides electric fine rod motion during normal operation and hydraulic pressure for scram insertion. Reduced maintenance with reduced radiation exposure is a feature of the new drive. Integral shootout steel built into the FMCRD replaces the external beam supports of the current BWRs and improves maintainability and reduces radiation exposure.

The drive mechanism operates to allow fine motion (18 mm step size) provided by the ball screw nut and shaft driven by the electric motor during normal operation. The electric motor also provides increased reliability through diverse rod motion to the hydraulic scram and acts as a backup with motor run-in following scram. The fine motion capability allows for small power changes and easier rod movement for bumup reactivity compensation at rated power. It also reduces the stress on the fuel and enhances fuel rod integrity. Ganged rod motion (simultaneous driving of a group of up to 26 rods) and automated control significantly improves startup time and power maneuvering capabilities for load following.

The ABWR FMCRD design has been improved from the European design by reducing the length and diameter, and by adding the fast scram function. Other refinements in the ball-screw assembly and seal designs have led to less maintenance requirements. Other design features include: (1) a two-section design for the CRD housing with the wearing parts concentrated in the shaft seal housing section in order to provide ease of maintenance time and radiation exposure. (2) The FMCRD scram water is discharged directly into the reactor vessel eliminating the scram discharge volume and associated valves and piping. This reduces radiation exposure and eliminates a potential source of common mode failure for the scram function. (3) The drive maintains a continuous purge into the reactor thus eliminating the accumulation of radioactive crud in the drive and reduces exposure. (4) Continuous full-in indication. (5) Dual safety grade separation switches to detect rod uncoupling and a new bayonet coupling to help eliminate the control rod drop accident. (6) Ganged hydraulic control units (HCU) with two CRDs per HCU. (7) A brake mechanism to prevent rapid wind down of the screw

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prevents rod ejection. Figure 3 illustrates the key components of the FMCRD.

BLADE COUPLING

- ~ LABYRINTH SEAL

REACTOR VESSEL

BALL SCREW ~~

GUIDE TUBE

SCRAM LINE INLET

SEPARATION SENSING REED SWITCH

DRAIN LINE

BRAKE

MOTOR

Figure 3. Fine Motion Control Rod Drive (FMCRD)

Safety and Auxiliary Systems

The ECCS and Residual Heat Removal System (RHR) along with the other auxiliary systems were reviewed to simplify and optimize their design. The selected ABWR design incorporates three redundant and independent divisions of ECCS and containment heat removal. The RPV, with its election of the external loops and no large pipe nozzles in the core region, allows for a reduced capacity ECCS. Yet, the fuel remains covered for the full spectrum of postulated LOCAs including a single failure. The ECCS network has each of the three divisions having one high pressure and one low pressure inventory makeup system. The high pressure configuration consists of two motor driven High Pressure Core Flooders each with its own independent sparger discharging inside the shroud over the core. It also consists of the Reactor Core Isolation Cooling System (RCIC) which has been upgraded to a safety system. The RCIC has the dual function of providing high pressure ECCS flow following a postulated LOCA, and also provides reactor cooling inventory control for reactor isolation transients. The RCIC, with its steam turbine driven power, also provides a diverse makeup source during loss of all

A-C power events. The lower pressure ECCS for the ABWR utilizes the three RHR pumps in the post- LOCA core cooling mode. These pumps provide Low Pressure Flooding and are labeled LPFL. The ECCS pumps provide core makeup over the full pressure range. For small LOCAs that do not depressurize the vessel when high pressure makeup is unavailable, an Automatic Depressurization System (ADS) actuates to vent steam through the safety relief valves to the suppression pool, depressurizes the vessel to allow the LPFL pumps to provide core coolant.

The RHR system has a dual role of providing cooling for normal shutdown and also provide core and containment cooling during LOCA. The ABWR RHR systems have been improved such that core and suppression pool cooling are achieved simultaneously since in the core cooling mode, the flow from the suppression pool passes through the heat exchanger in each of the three divisions of the RHR.

As a result of these enhancements in the ECCS network and RHR, there is a substantial increase in the safety performance margin of the ABWR over earlier BWRs. This has been confirmed by the preliminary probabilistic risk assessment (PRA) for the ABWR which shows that the ABWR is a factor of at least 10 better than BWRS and 6 in avoiding possible core damage from degraded events.

Rationalizations have also been made in other auxiliary systems. For example, the main steam leakage control system has been deleted. The Combustible Gas Control System utilizes an inerting system to reduce oxygen concentration in the primary containment and also uses portable recombiner units to be shared at the site for combustible gas control following postulated degraded core accidents. The Standby Gas Treatment System flowrate was reduced due to recent favorable leak rate data for containments and the number of passive filter trains was reduced from two to one.

Advanced Control and Instrumentation Systems

Digital technology and multiplexed fiber optic signal transmission technology have been combined in the ABWR design to integrate control and data acquisition of both the reactor and turbine plants. For the feedwater, recirculation, and pressure control systems, triplicated fault tolerant digital control is utilized. A redundant design for the rod control and information system is used. These systems feature automatic self-test and built-in calibration and trip test to improve reliability. Microprocessor-based digital monitoring and control (DMC) has been implemented for most principal NSSS and Balance of Plant monitoring and control functions. ABWR DMC technical advantages include self-test, automatic calibration, user interactive front panels, full multiplex system compatibility, standardization of the man-machine interface, and where possible, use of common circuit cards. Plant startup and shutdown operations, TIP operation and nuclear instrumentation gain adjustments have been automated to reduce operator error and reduce plant startup time. Technological advances in the ABWR nuclear instrumentation areas include fixed widerange in-core monitors with a

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PLANT dNlT LEVEL

SVSTEN L E JEL

BOP CONTROL I T Y P C A L I

SENSOR' ACTUATOR LEVEL

DRIV IRS

- I ' REPRESENTSGNE OF FOUR SAFETY DIVISIONS

Figufe 4 The Multiplexing Concept

period based trip design to replace the current source and intermediate range monitors and eliminate range switching during startup.

The use of multiplexed fiber optic data transmission in the ABWR is another new feature applying advanced technology. The use of fiber optic multiplexing reduces the amount of cabling and cable pulling time during construction. This also reduces the overall cost of the control and instrumentation area. Multiplexing easily provides a high degree of redundancy in the control system, and also improves maintainability. The ABWR combines multiplexing communications capability with intelligence to perform logic and control functions for interfacing systems by employing the DMC in the multiplexing system. This multiplexing system has been applied to the plant protection and engineered safeguard systems as well as nuclear steam supply and balance of plant control systems. The conceptual structure for application of this digital and optical transmission technology is shown in Figure 4.

Containment and Reactor Building Design

For the ABWR the cylindrical RCCV, with its pressure suppression concept, was selected as the reference design of the primary containment vessel. The concrete walls of the RCCV are integrated with the reactor building and form a major structural part of this building. The annular top slab of the drywell is also integrated with the upper pool girders that run across the building and have direct connection with the

1.

building's outer walls. The pressure retaining concrete wall of the RCCV is lined with leak-tight steel plate. The cylindrical design, a simple shaped concept same as the Mark I11 drywell design, allows for easier and faster construction.

The upper drywell encloses the reactor, and the process lines and valves of the reactor coolant system. The lower drywell. located under the reactor vessel is the space for installation and maintenance of the internal pump motors and their heat exchangers, and the control rod drives. Piping and cables are arranged inside and lead out of this space. Personnel and equipment access are provided for by hatches in the upper drywell, and through the tunnels in the lower drywell.

The wetwell provides an air space and a pool to suppress the steam from the postulated LOCA. Multiple horizontal vents derived from the Mark I11 containment design, discharge the vessel blowdown steam water mixture and the air from the drywell to the wetwell pool. The steam is condensed, and the fission products are scrubbed and retained in the pool.

The ABWR design represents a very significant reactor building volume and cost reduction. The reactor building volume has been reduced to approximately 167,000 cubic meters which has both reduced construction cost and provided a schedule saving of 2-V2 months.

The structure integration scheme discussed earlier takes advantage of both the RCCV and reactor building to carry dynamic and shear loads, and hence, reduce overall size

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and thickness of the supporting walls. The reactor building has been separated into three mas to provide separation for the three division configuration of the safety systems. The reactor pedestal has also been revised to support the drywell diaphragm floor, connect the access tunnels to the undervessel area, contain the horizontal vent system, and provide connecting vents between the lower and upper drywell. The design also allows for fabrication in the shop applying modular construction techniques. Figure 5 represents the RCCV and reactor building design concept.

Figure 5. RCCV and Reactor Building

SUMMARY

The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies for an improved BWR. The description of the key ABWR features and their advantages and performance improvements demonstrate the success of the integrated plant design. The overall technical evaluation shows the superiority in terms of performance characteristics and economics that the ABWR design has achieved. The many features were seen to provide improvement in schedules, reduced maintenance requirements, contributed to less radiation exposure, and still provided plant operation improvements and increased safety. Significant cost reduction in both capital and operating costs were achieved by systematically optimizing and simplifying the design while retaining the existing benefits and penormance improvements.

The confirmatory test and development program has been highly successful in confirming the technical feasibility and reliability of the new components and technology. This has provided considerable support for the design in terms of test data and licensing.

The ABWR design in Japan is currently in the licensing review process at the KashiWazaki-Kariwa Nuclear Power site. Construction for the first unit (K-6) began in

September, 1991 with commercial operation in 1996. The second unit (K-7) will follow one year later.

The ABWR program in the US has been initiated to obtain regulatory certification of the ABWR as a world-class standard plant-

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