NSTD Introductory Course NSTD Introductory Course New Gen III+ - - PowerPoint PPT Presentation

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NSTD Introductory Course NSTD Introductory Course New Gen III+ - - PowerPoint PPT Presentation

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NSTD Introductory Course NSTD Introductory Course New Gen III+ Reactor New Gen III+ Reactor Pow er Plant Designs Pow er Plant Designs Economic Simplified Boiling Economic Simplified Boiling Water Reactor (ESBWR) Water Reactor (ESBWR)

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SLIDE 1

ESBWR Overview 1 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

New Gen III+ Reactor New Gen III+ Reactor Pow er Plant Designs Pow er Plant Designs Economic Simplified Boiling Economic Simplified Boiling Water Reactor (ESBWR) Water Reactor (ESBWR)

NSTD Introductory Course NSTD Introductory Course

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SLIDE 2

ESBWR Overview 2 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

Disclaimer

Information contained herein is derived exclusively from publicly available documents. The content of this introductory course does not necessarily represent what may be submitted to the Nuclear Regulatory Commission in the form of a license application for a new reactor. ORNL neither endorses this design nor has performed any design reviews to validate design improvements, design margins, or accident

  • probabilities. The intent in compiling this information at

this time is for the express purpose of constructing an internal, introductory course for our own staff.

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SLIDE 3

ESBWR Overview 4 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Key Design Features

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SLIDE 4

ESBWR Overview 5 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

Nuclear Pow er Plant (NPP) Development

Gen II

Large Commercial NPPs Currently in Operation Throughout U.S.

Gen III

Advanced LWRs AP 600(W) ABWR (GE) System 80+ (CE)

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SLIDE 5

ESBWR Overview 6 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

Nuclear Pow er Plant (NPP) Development (cont.)

Transition

Probably Could Be Classed as Gen III+ SBWR

Gen III+

Evolutionary Designs ESBWR (GE) Improved Economics AP 1000 (W) Advanced Safety Features ACR 700 (AECL) Some Passive Design Aspects EPR (AREVA - Framatome ANP) Advanced Containment Design PBMR (South Africa, PBMR Pty. Ltd.) Simplified System Designs

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SLIDE 6

3 GE Energy / Nuclear September 27, 2005

Oyster Creek KRB Dresden 1 ABWR Dresden 2 ESBWR SBWR

BWR Evolution

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SLIDE 7

ESBWR Overview Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

BWR Design Progression

BWR 2-6  ABWR  SBWR  ESBWR

− 35 Domestic (U.S.) operating BWRs − 17 International operating BWRs − 2 International (Japan) operating ABWRs − Provide the current status of its design certification process with the NRC.

BWR Product Line 2/3/4

− Motor Generator Used for Recirculation System Flow Control − High Pressure Coolant Injection (except early BWR 2s - Nine Mile Point 1 and Oyster Creek which used Feedwater Coolant Injection)

BWR Product Line 5/6

− Flow Control Valves for Recirculation System Control − High Pressure Core Spray

Recirculation Systems

− 5 Loops - Nine Mile Point 1 and Oyster Creek − 2 Loops - all others

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SLIDE 8

ESBWR Overview Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

BWR Design Progression (cont.)

Isolation Condenser Systems

− Dresden 2 & 3 − Nine Mile Point 1 − Oyster Creek

Natural Circulation

− Humboldt Bay

Containment

− Mark 1 (23) BWR 2,3 and older BWR 4s

inverted light bulb drywell and torus usually an inerted atmosphere

− Mark II (8) Newer BWR 4s and BWR 5s

frustum of cone called “over-under”

− Mark III (4) BWR 6s

pressure suppression

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SLIDE 9

ESBWR Overview Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ABWR NPPs

Kashiwazaki Units 6 & 7 Located in Japan Expected time to fuel load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 months Actual construction time. . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 6 - 61 months Actual construction time. . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 6 - 61 months Actual time to fuel load . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 6 - 36.5 months . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 7 - 38.3 months Broke ground September 17, 1991 Commercial operation . . . . . . . . . . . . . . . . . . . . . Unit 6 - November 7, 1996 . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 7 - July 2, 1997

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SLIDE 10

ESBWR Overview Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ABWR NPPs (cont.)

Lungmen Units 1 & 2 Located in Taiwan Expected construction time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 months Delayed up to 2005 at 57% complete Reactor installed. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Unit 1 - March 2005 Expected operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Unit 1 - July 2006 . . . . . . . . . . . . . . . . . . . . . . . . . . . . Unit 2 - July 2007

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SLIDE 11

18 GE Energy / Nuclear September 27, 2005

zero 18 9 9 Safety system pumps zero 3 3 2 Safety diesel generator zero 10 2(large) 2(large) Recirculation pumps 4500/1580 3926/1350 3900/1360 3293/1098 Power (MWt/GrossMWe) 3E-8 1E-7 1E-6 1E-5 Core damage freq./yr ~ 130 160 150 115 Safety Bldg Vol (m3/MWe) 27.7/7.1 21.1/7.1 21.8/6.4 21.9/6.4 Vessel height/dia. (m) 185/LP 50 3.7 764 BWR/4-Mk I (Browns Ferry 3) 193/LP 54.2 3.7 800 BWR/6-Mk III (Grand Gulf) 269/FM 205/FM Number of CRDs/type 54 51 Power density (kw/l) 3.0 3.7 Active Fuel Height (m) 1132 872 Fuel Bundles (number) ESBWR ABWR Parameter

Optimized Parameters for ESBWR

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SLIDE 12

17 GE Energy / Nuclear September 27, 2005

Made non-safety grade Reactor Building Service Water (Safety Grade) And Plant Service Water (Safety Grade) Replaced pumps with accumulators SLC –2 pumps Replaced with IC heat exchangers RCIC Eliminated – only 2 non-safety grade diesels Safety Grade Diesel Generators (3 each) Non-safety, combined with cleanup system Residual Heat Removal (3 each) Utilize passive and stored energy LPFL (3 each) Eliminated need for ECCS pumps HPCF System (2 each) Eliminated Recirculation System + support systems

ESBWR ABWR

What’s different about ESBWR

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SLIDE 13
  • 94 -

BWR Containment Comparison

30 42 39 9 42 44 50 LOCA Pressure (psig) 40 55 45 15 45 62 50 Design Pressure (psig) 0.43 0.3 0.5 1.6 0.5 0.4 2.5 Drywell and wetwell volume (ft3 X 106) Yes Yes Yes Yes Yes Yes No Pressure Suppression ESBWR SBWR ABWR Mark III Mark II Mark I Dry Characteristic

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SLIDE 14

ESBWR Overview 7 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Plant Licensing Status

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SLIDE 15

ESBWR Overview 7 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Design Certification

  • Accepted for docketing by the NRC in December 2005.
  • Final Design Approval (FDA) is expected in December 2009.
  • Design Certification expected in December 2010.

Utility Activities

  • The consortium, NuStart, is expected to apply for a construction/operating

license (COL) for an ESBWR for Entergy Nuclear at its Grand Gulf Site in late 2007 or early 2008.

  • Dominion will be ready to apply for a COL for an ESBWR at its North

Anna Site in September 2007.

  • Entergy Nuclear will apply for a COL for an ESBWR at its River Bend Site

in the first half of 2008.

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SLIDE 16

ESBWR Overview 8 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Plant Overview

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SLIDE 17
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SLIDE 18

ar02-19

Safety Systems Inside Containment Envelope

All Pipes/Valves Inside Containment High Elevation Gravity Drain Pools Raised Suppression Pool Decay Heat HX’s Above Drywell

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SLIDE 19

ESBWR Overview 9 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Core and Vessel Design

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SLIDE 20

19 GE Energy / Nuclear September 27, 2005

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SLIDE 21
  • 9 -

Water Rods Lower Tie Plate Debris Filter Part Length Fuel Rods Interactive Channel

ESBWR Fuel Assembly

  • Same cross-sectional

dimensions as ABWR

  • Active Fuel Length:

ABWR = 144 inches ESBWR = 120 inches Upper Tie Plate Zircaloy Ferrule Spacers

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SLIDE 22

Page 3

Performance PS0206-1

ESBWR Normal Operation

  • No recirculation pumps – total

reliance on natural circulation

  • Significant natural circulation

flow exists in all BWR’s

  • For a given core power, there is a

corresponding natural circulation flow

  • ESBWR uses enhanced design

features to increase the flow compared to standard BWR’s

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SLIDE 23

ESBWR Overview 10 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Important Systems

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SLIDE 24

ESBWR Overview 11 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Control Rod Drive System (CRDS)

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SLIDE 25
  • 18 -

Control Rod Drive System

New features added

CORE REACTOR VESSEL SUCTION FILTERS CONDENSATE STORAGE TANK FROM CONDENSATE AND FEEDWATER CRD PUMPS TEST LINE MIN FLOW LINE FW RWCU/SDC

FE FE

INJECTION VALVES ACCUMULATOR CRDs HCUs CHARGING HEADER

FE RO

DRIVE WATER FILTERS

RWCU/SDC PUMPS PURGE WATER CONTROL VALVES PURGE HEADER DRYWELL Second pump starts: Low header press. Low water level 2 Bypass valves open: Low water level 2 *Valves close: Low water level 2

* *

Valves open: Low water level 2

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SLIDE 26

ESBWR Overview 12 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Isolation Condenser System (ICS)

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SLIDE 27
  • 32 -

FOUR TRAINS OF

  • 30 MWt HEAT CAPACITY EACH

CORE REACTOR VESSEL

DRYER

25A RO RO RO 20A MO MO MO NMO 350A DPV STUB LINE SUPPRESSION POOL MAIN STEAM LINE 200A 200A 200A 20A 200A MO MO NO ISOLATION CONDENSER IC/PCC POOL 300A ATMOSPHERIC VENT

TRAIN A SHOWN

NMO MO DRYWELL

ESBWR Isolation Condenser System - Schematic Diagram

LOOPS A, B, C ONLY LOOPS B, C, D ONLY

NMO = Nitrogen-Motor Operated Valve Without Accumulator NO = Nitrogen-Operated Valve Without Accumulator MO = Motor-Operated Valve

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SLIDE 28
  • 30 -

Isolation Condenser Simplified

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SLIDE 29

ESBWR Overview 13 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Standby Liquid Control System (SLCS)

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SLIDE 30

25 / GE / April 5, 2005

Standby Liquid Control System

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SLIDE 31

ESBWR Overview 14 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Reactor Water Cleanup (RWCU) / Shutdow n Cooling (SDC) System

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SLIDE 32

6 / GE / April 5, 2005

Reactor Water Cleanup (RWCU)

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SLIDE 33

ESBWR Overview 15 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Safety Systems

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SLIDE 34

ESBWR Overview 16 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Gravity-Driven Cooling System (GDCS)

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SLIDE 35
  • 37 -

ESBWR Gravity-Driven Cooling System - Schematic Diagram

CORE

DIVISION A SHOWN TYP DIV B, C, D

REACTOR VESSEL GDCS POOL WW/GDCS POOL VENTPIPE TEMP STRAINER GDCS SUMP DRYWELL TM TM TM TM A 150A 150A 150A INJECTION LINE SUPPRESSION POOL WETWELL AIRSPACE BIASED-OPEN SWING-CHECK VALVE EQUALIZING LINE A INJECTION SQUIB-VALVE 200A (UPPER DRYWELL ANNULUS) (OMITTED FROM

  • DIV. D)

SUPPRESSION POOL WETWELL AIRSPACE DELUGE LINE = TORQUE MOTOR (MAGNETIC-COUPLED) = WETWELL AIRSPACE WW

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SLIDE 36

ESBWR Overview 17 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Passive Containment Cooling System (PCCS)

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SLIDE 37

6 / GE / April 5, 2005

Passive Containment Cooling System

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SLIDE 38
  • 40 -

Passive Containment Cooling Simplified

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SLIDE 39

ESBWR Overview 18 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Depressurization

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SLIDE 40

15 / GE / April 5, 2005

MSIV, SRV and DPV Arrangement

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SLIDE 41

ESBWR Overview 19 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Containment Design

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SLIDE 42
  • 60 -

ESBWR Containment System - Schematic Diagram

LEAK DETECTION (TYP OF 10)

Burst Diaphragm Primary COPS (opens at severe accident pressure) Containment Boundary Design Basis Accident Additional Volume Available For Primary COPS

Wetwell Airspace

PCCS Hx’s (Typ of 4) DW/WW LOCA Vertical Ventpipes (Typ of 10) Reactor Vessel Core Suppression Pool GDCS Injection Line Suction End GDCS Pool Sump (Typ of 4) Bolted Access Hatch WW/GDCS Pool Ventpipe (Typ of 3) WW/GDCS Pool (Typ of 3) Upper Drywell VB Vacuum Breaker (Typ of 3) Spill Overflow Lines Equipment Hatch Corium splash Shield Personnel Hatch Undervessel Work Platform Equipment Hatch

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SLIDE 43

ESBWR Overview 20 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Additional Systems

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SLIDE 44

ESBWR Overview 21 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Pow er Conversion System (PCS)

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SLIDE 45

Page 5 September 27, 2005

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SLIDE 46

ESBWR Overview 22 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Instrumentation and Control (I& C)

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SLIDE 47

30 / GE / October 3, 2005

Summary of ESBWR I&C Characteristics

  • ESBWR's digital I&C design is based on the same digital I&C framework, design,

and hardware/software platforms of ABWR. The ABWR digital I&C design has been in operation and in construction (with hardware/software in fabrication/testing). – proven system and hardware/software designs.

  • Automation implemented same as ABWR
  • Minimized hardwired cables same as ABWR
  • Digital Remote Shutdown System capable of full plant control and enhances EOP

utilization

  • Enhanced “diverse protection and actuation” capability in compliance to BTP

HICB - 19

  • Fixed in-core gamma thermometer AFIP to replace the TIP system

– simplified operation and reduced personnel radiation dosage.

  • eliminated TIP containment penetrations
  • The ESBWR I&C design will comply with updated or newly developed regulatory

requirements such as BTP-14, BTP-19, as well as RG1.152.

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SLIDE 48

ESBWR Overview 23 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Electrical Distribution

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SLIDE 49

Page 19 September 27, 2005

Standby On-site AC Power Supply

  • Consists of two 15 MVA independent diesels coupled to 6.9 kV AC

generators, the DG auxiliary systems, fuel storage and transfer systems and associated local instruments and controls.

  • Each DG supplies non-safety AC power to it’s associated PIP busses on loss
  • f voltage for plant investment protection.
  • On PIP bus undervoltage the DG starts, accelerates with in 1 minute.
  • Major loads are tripped from the 6.9 kV PIP busses.
  • DG will connect to the PIP busses when incoming preferred and alternate

preferred source breakers have been tripped.

  • Large motor loads are then reapplied sequentially and automatically after

DG power source breaker closes.

  • The DG is capable of being fully loaded within 600 seconds.
  • DG operation is not required to ensure nuclear safety – only investment

protection

PIP = Plant Investment Protection

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SLIDE 50

ESBWR Overview 24 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Accident Analysis

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SLIDE 51

ESBWR Overview 25 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Probabilistic Risk Assessment (PRA)

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SLIDE 52

3 GE Energy / Nuclear September 29, 2005

Document Relationships

ESBWR PRA NEDC-33201P ESBWR DCD Chapter 19

Scope Methods Goals Design Requirements SAMDA

SAMDA = Severe Accident Mitigation Design Alternatives

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SLIDE 53

36 GE Energy / Nuclear September 29, 2005

Breakdown By Initiating Event

Large Steam LOCA 3.2% Transient 0.4% IORV 0.4% Medium Liquid LOCA 0.9% Loss of Feedwater 38.0% Loss of Power 56.8% RWCU Line Break 0.1% Feedwater Line Break 0.1% Loss of Condenser 0.1%

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SLIDE 54

37 GE Energy / Nuclear September 29, 2005

Breakdown By Accident Class

Containment Overpressure 8% Low Pressure Core Damage 90% High Pressure Core Damage 1% ATWS 1% Containment Bypass << 1%

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SLIDE 55

ESBWR Overview 26 Predecisional – Information compiled from public sources – Internal ORNL use only March 2006

ESBWR

Design Control Document (DCD) Accident Analyses

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SLIDE 56

50 GE Energy / Nuclear September 29, 2005

ESBWR Severe Accident Treatment – Work Structure

Analysis of CDF-DS parameters Characterization of Safety Systems Risk- Significant Core Melt Scenarios CDF-Dominant Sequences (DS) Source Terms

Level 3

Threats to Containment Integrity

Conditional Containment Failure Probability (CCFP) Containment Strength Applications of Risk-Oriented Accident Analysis Methodology (ROAAM) (DCH, FCI, CCI, H2, Bypass, etc…)

Plant Design PSA Level 1

SAT CRSS Containment Event Trees (CET) MAAP-ESBWR Calculations of Accident Progression Containment and Safety Systems

Plant Damage State