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Copyright 2013 GE Hitachi Nuclear Energy - Americas, LLC - All rights reserved

Safety, constructability,and operational

performance of the ABWRand ESBWR designs

Douglas McDonald

Vice President, Nuclear Power Plant

Sales Middle East and Africa

IAEA Technical Meeting on TechnologyAssessment for Embarking Countries

June 24-28, 2013

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Wilmington, NCUSA

Nuclear Power Plants: ABWR,ESBWR and PRISM

Nuclear Services

Nuclear Fuel Fabrication.BWR and CANDU

CANDU ServicesFuel Engineering and Support

Services

Peterborough, ONCanada

Wilmington, NCYokosuka, Japan

UraniumEnrichment ThirdGenerationTechnology

Wilmington, NCUSA

Tokyo,Japan

GE Hitachi Nuclear Alliance

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Vallecitos USA

Dresden 1 USA

Laguna Verde - Mexico

Tarapur 1&2 India

Dodewaard - Netherlands KKM - Switzerland

Garigliano - Italy

Santa Mara de Garoa - SpainLungmen - Taiwan

K6/K7 - Japan

KRB - Germany

BWRs around the world

84 operating BWRs

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ABWR PRISM

GE Hitachis new reactor portfolio

ESBWR

Operational Gen IIItechnology

Lowest core damagefrequency of anyGeneration III reactor

Extensive operationalexperience since 1996

Licensed in US, Taiwan,and Japan

Evolutionary Gen III+technology

Lowest core damage frequencyof any Generation III+ reactor

Passive cooling for >7 days

without AC power or operatoraction

Lowest projected operations,maintenance, and staffing costs1

25% fewer pumps, valves, andmotors than active safetynuclear plants

Revolutionary technology

with a rich, 40-year heritage

Passive air-cooling with nooperator or mechanical actionsneeded

The answer to the used fueldilemma - can reduce nuclearwaste to ~300-year radiotoxicity2

while providing new electricitygeneration

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1 Claims based on the U.S. DOE commissioned Study of Construction Technologies and Schedules,

O&M Staffing and Cost, and Decommissioning Costs and Funding Requirements for Advanced ReactorDesigns and an ESBWR staffing study performed by a leading independent firm2 To reach the same level of radiotoxicity as natural uranium

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PWRs and BWRs the basics

Typical Pressurized Water Reactor Typical Boiling Water Reactor

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Operation of a BWR

Saturated water/steammixture cooling fuel

Direct cycle (No externalsteam generators)

Water moderator modifiedby steam voids (bubbles)

Saturated Steam

Saturated Water

Subcooled Water

To Turbine

FeedWater

Core

Rx InternalPumps(ABWR)

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The Boiling Water Reactor

99.9% Steam550

F / 288C

420F / 216F

100% Water

BWR Fuel Assembly

- 90 fuel rods encasedin a channel

- 2 water rods- part-length rods- burnable absorbers

Reactor Pressure Vessel

SteamDryer

SteamSeparator

Control Rod Drives

Control Rod Blades

ReactorInternal Pumps (ABWR)

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0

Source: IAEA PRIS Database and 3/2013 EPRI Fuel Reliability Update

A benchmark for operational performance

BWR PWR

100%

90%

80%

89.83%90.96%

1% BWRadvantageprovides 8

additionalmonths ofrevenue over60-yearlifetime

Data represents top quartile for 2002-2012

Zero BWR fuel failures inNorth America

Capacity factors Average U.S. Cycle Length Trends

BWRs 20 months

PWRs 16.7 months

Fuel performanceAverage Outage length

2002-2012 N. American outages includinginspection, maintenance or repair with refueling

35.5days

45 days

BWR

PWR

10 fewer days inBWR outagesin North America

BWR

PWR

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Safety and simplicity

III+ III+

U.S. PWRs2 E-5 (avg.)

U.S. BWRs8 E-6 (avg.)

APR14002 E-6

APWR1.2 E-6

EPR2.8 E-7

AP10002.4 E-7

ABWR1.6 E-7

ESBWR1.7 E-8

PRAofCore

DamageFrequ

ency

References: Plant licensing DCDs and publically available information

Note: PRA of CDF is represented in at-power internal events (per year)Note: NSSS diagrams are for visualization purposes only

Generation IIIGeneration II

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Responses needed to maintain core cooling

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EPR and

GEN II WaterOperatorAction ElectricPower

72 HRS.

>7 days

ABWR

AP1000

ESBWR

~36 HRS.*

*ABWR DCD credits water addition at 8 HRS.References: AP1000: US DCD rev. 18 Section 8.5.2.1, EPR: US DCD Rev. 1 Section 8.4

Responses to extended loss of all AC power

Gen III+ passive plants allow fora much longer coping time

Decay heat level impacts urgency

DECAY HEAT

30 MIN.

24 HRS.30 MIN. 2 HRS.

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Advanced Boiling Water Reactor

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ABWR Design Objectives

Improved operability

Improved capacity factor12-24 month fuel cycle

~95% on a 10 year rolling average

Improved safety and reliabilityNo core uncovery during design basis

accidents

Reduced occupational exposure

Reduced costsPredictable Construction Time and Costs

Operations and Maintenance (O&M)

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Emergency Core Cooling System

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HPCF

RCICADSLPF

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Key ABWR differentiators for extreme events

Separate and passive

containment venting toprevent hydrogenexplosion

Reactor depressurizationcapability for >7 days due tobattery segregation andpneumatic controls

Seismic AC independent water

injection into core

Containmentheat removal

Core cooling

shutdown

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ABWR Station Blackout prevention and

mitigation3 x 100% nominal safety

divisions

Emergency Diesel Generators 3 located in Reactor Building

Each has a 7-day fuel tank that isburied in a concrete vault outsidethe Reactor Building

Combustion Turbine Generator

Air-cooled Service Water notneeded

Safety-related batteries arelocated in the ControlBuilding - just below the

Main Control RoomAC Independent Water

Addition (ACIWA) System Hard-piped connections to reactor

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In Operation

Under Construction

4 Units

4 Units

Hamaoka-5 ABWR

Kashiwazaki-Kariwa 6/7 ABWR

: BWR Power Plant Site

Japan

Taiwan

Under ConstructionCOD TBD

Ohma ABWR

Shimane-3 ABWRShika-2 ABWR

Japan

Higashidori-1

Kaminoseki-1

COD 1996/1997

COD 2006

COD 2005

Lungmen-1/2 ABWR

Under ConstructionCOD 2014 (estimated)

Under ConstructionCOD TBD

Recent experience and project status

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Construction lessons learned:

Efficient, repeatable execution model

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ABWR modularization proven in JapanRoof Truss Steels

RCCV Top Slab

RCCV liner

Central Mat

Base Mat HCU Room Offgas Equipment Lower Condenser Block

T-G Pedestal Piping Unit

Upper Condenser

Condensate Demin. Piping

Condensate DemineralizerUpper Drywell Module

RPV Pedestal

MSIV/CV RWCU Reheat Exchanger

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Detailed engineering before on-site work

Walk-through simulation

Full 3D CAD design

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ModularizationProven experience in operating Gen III plants

Central Mat RCCV RebarsTop Slab

RCCV liner Roof Truss Steel

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Building Commissioning

RI : Rock Inspection BC:Start of Basemat Construction FL : Fuel Loading CO: Commercial Operation

ABWR#1CO:1996/11

Kashiwazaki-Kariwa-6

ABWR#2CO:1997/7

Kashiwazaki-Kariwa-7

ABWR#3CO:2005/1Hamaoka-5)

ABWR#4CO:2006/3

Shika-2)

BC FL

39.5M 8.8M3.5M 48.3M

43.5M1M 10.5M 54.0M

44.5M 10.5M2.5M 55.0M

RI

38M2M 49.2M11.2M

COD

BC CO

Predictability of Schedule

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North Anna 3 ESBWR

Copyright 2013 GE Hitachi Nuclear Energy - Americas, LLC - All rights reserved 22

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Economic Simplified Boiling Water Reactor

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Parameter

Core Thermal Power Output Plant Net Electrical Output(1)

Reactor Operating Pressure Feedwater Temperature(2)

RPV Diameter Height

Reactor Recirculation Fuel

Control blades

4500 MWt1520 MWe7.17 MPa (1040 psia)216C (420F)

7.1 meters (23.3 feet)27.6 meters (90.5 feet)Natural Circulation1132 fuel bundlesShortened length of 3m269 Fine Motion ControlRod Drives (FMCRDs)

(1) Typical (site dependent)

(2) Nominal Rated Operation

ESBWR: Economy of Scaleand Simpler Design

Key plant / reactor characteristics

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ABWR to ESBWR evolution: Nuclear Island

1

1

1

Fuel and Aux Pool Cooling equivalent designs2

Reactor Water Cleanup System equivalent designs

2 2

3

3

3 Suppression Pool Cooling & Cleanup System equivalent capability

7 High Pressure Core Flooder replaced by HP CRD makeup

4 Residual Heat Removal System equivalent for shutdown cooling

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4

7

5

Standby Liquid Control System simplified design

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8 Reactor Core Isolation Cooling replaced by Isolation Condenser

8

8

6 Hydraulic Control Unit equivalent design

7

6

9 Residual Heat Removal Containment Spray replaced by PCCS

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9

Safety Relief Valves Diversified by Depressurization Valves

Systems are Equivalent or Simplified

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ABWR ESBWR

10

25

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ESBWR modularization based on ABWRRoof Truss Steels

RCCV Top Slab

RCCV liner

Central Mat

Base Mat HCU Room Offgas Equipment Lower Condenser Block

T-G Pedestal Piping Unit

Upper Condenser

Condensate Demin. Piping

Condensate DemineralizerUpper Drywell Module

RPV Pedestal

MSIV RWCU Reheat Exchanger

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Passive safety/natural circulation Increased volume of water in the vessel

Increased driving head

Chimney, taller vessel

Reduced flow restrictions Open downcomer

Shorter core

Significant reduction in components Pumps, motors, controls, Heat Exchangers

Power Changes with FeedwaterTemperature and Control Rod DrivesMinimal impact on maintenance

Natural Circulation

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ESBWR Passive Safety Systems

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ESBWR LOCA response

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Isolation Condenser System

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passivedecay heat

removal

Fully passive only requires gravity to

function and starts automatically(fails in-service if DC power is lost)

4 separate systems in reinforced concretevaults

Limits reactor pressure (no SRV lifts) andtemperature and conserves waterinventory following containment isolation

Steam (heat) rises from reactor to thecondenser pool, condenses, then gravitypulls the cool water down into the reactor(closed-loop)

Removes heat fromcontainment

Core cooling

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Simple refill actions even in the worst conditions

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PCCS heat

exchanger test

IC/PCCPOOLS

DW2

GDCSPO

WW2

WETWELLWW1

RPV

DRYWELLDW1

GIST facility

IsolationCondenser Testing

BDLBREAK

LOWERDRYWELL

HORIZONTALVENT (1 OF 2)

GDCSINJECTION LINE(1OF 4)

UPPERDRYWELL

VACUUMBREAKER(GDLBTESTSONLY)

MSLBREAK

REACTORPRESSURE

VESSEL

DEPRESSURIZATION LINE(1OF 2)

WETWELL

Panda Full HeightContainment Test facility

WW to DWVacuum Breaker

ESBWR Proven innovationdrywell to wetwell

vacuum breaker test

Depressurization Valve

test

BiMACtesting

fuel modified

GNF2

natural circulationproven at Dodewaard

FMCRDsfrom ABWR

Copyright 2013 GE Hitachi

Nuclear Energy International, LLC- All rights reserved 35

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Operations

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ABWR and ESBWR state-of-the-art operations

Fully Digital ControlSystem

Fewer components, No drift,

less power and heat Fault tolerance control Four division safety

redundancy Automated operation

Surveillance testing greatlyreduced

Improved Man-MachineInterface

Large mimic displays

Prioritized alarms Flat panel controls

minimize hard switches Human factored displays

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ABWR and ESBWR offer substantial

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ABWR and ESBWR offer substantial

improvements in O&M

Key component redundancy Maintenance flexibilityOperational transients

Simplifications in design Safety, operations, and reliability

O&M costs

Improvements in plant maintenance Easier operations, greater reliability

Maintenance cost and dose

Simpler to operate Safety and reliability

Operator actions and transients

Lower radiation exposure Outage efficiency and FME reduction

Occupational dose and rad waste costs

Passive safety (ESBWR) Safety and plant simplificationMaintenance costs and dose

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Copyright 2013 GE Hitachi Nuclear Energy - Americas, LLC - All rights reservedSource: An ESBWR staffing study performed by a leading independent firm

Best in-class O&M

ESBWR requiressignificantly fewerplant personnel thanany other GenerationIII/III+ design.

A direct reflection of theESBWRs simpler design

Allows for a higherpercentage of localworkforce

Fewer ex-pats results indirect cost savings

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ABWR

ESBWR

Safe.

Simple.

Smart.

1 Based on the industry standard measure of reactor safety - core damage frequency2 Claims based on the U.S. DOE commissioned Study of Construction Technologies and Schedules, O&MStaffing and Cost, and Decommissioning Costs and Funding Requirements for Advanced ReactorDesigns and an ESBWR staffing study performed by a leading independent firm

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