IANS Meeting, Idaho Falls, Idaho March 21, 2013

TerraPower and the Traveling Wave Reactor

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TerraPower’s Traveling Wave Reactor

• Provides energy security by using vast stores of depleted uranium

• Improves safety through passive systems

• Reduces costs through fuel cycle simplification and reduced uranium use

• Reduces nuclear waste

• Provides large environmental benefits

• Reduces proliferation risk

• Enables a global nuclear export market

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Profile of TerraPower

• For profit company funded by Bill Gates (Chairman), Mukesh Ambani (Reliance), Vinod Khosla, and other visionary private investors

• A nuclear power innovation company. Will not build, own or operate nuclear plants – no nuclear liability exposure

• Expert management and staff with 300 man-years of experience on fast reactors (e.g., FFTF, EBR –II)

• Over 80 contracts/agreements with national labs, universities, companies, government agencies and expert consultants since 2007

• Have state-of-the-art computer capabilities and proprietary software for core performance simulations

• Enormous data base and access to spent fuel assemblies from previous fast reactors. Owns TWR intellectual property and know-how

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Current Nuclear Fuel Cycle

Uranium mining

and milling

Uranium enrichment Fuel fabrication

Nuclear power

generation

Depleted

uranium

storage

Reprocessing Spent fuel storage

Actinide fuel

fabrication

Long-term

geologic

repository

Conversion to

uranium hexafluoride

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Uranium mining

and milling

Conversion to

uranium hexafluoride

Uranium enrichment Fuel fabrication

Nuclear power

generation

Depleted

uranium

storage

Reprocessing Spent fuel storage Actinide fuel fabrication

Long-term

geologic

repository

TWR Simplified Fuel Cycle

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Depleted Uranium: A Sustainable, Cheap and Secure Source of Fuel

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•Paducah, KY storage facility of 38,000 canisters of depleted uranium (DU)

•Enough fuel to power the US for over 750+ years

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TWR uses a Breed-and-Burn “Equilibrium”

• Point at which only depleted uranium (DU) is needed to sustain criticality

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Initial fissile remnants

Bred fissile material

Fertile material

Burning/burned

fissile material

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The Traveling Wave Reactor

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A Change to Cylindrical Geometry Has Advantages • Fresh fuel is moved into the wave • The burning region remains stationary • Exhausted fuel is moved to outer rings

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Fundamental Physics of a TWR

• Breeds and burns depleted uranium or waste LWR fuel

• Fueled once and can burn for 40+ years without refueling

• Weapons proliferation resistant. Assemblies are shuffled within a sealed vessel

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TWRs Set a High Safety Standard

• Advanced Reactor Design – Safety benefits of pool type sodium-cooled fast reactors

• Inherent safety: automatic shutdown without need for human interaction in event of accident

• System operates at low pressure; less likely to fail • Fukushima type accidents are not possible

– Passive decay heat removal even without offsite or onsite emergency power

• Simplified Fuel Cycle – Reduced safety risk in support activities due to less

mining, enrichment, processing, and waste transportation and storage

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How Does TWR Differ in Station Blackout?

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Typical BWR/PWR TerraPower’s TWR

Coolant water at high pressure

(7MPa for BWR, 15 MPa for PWR

o Possible LOCA w/ rapid

coolant loss

Coolant sodium at atmospheric

pressure

o LOCA not credible

Loop reactor with low thermal

intertia

o Decay heat to boil coolant at

1 atm <2 hours

Pool reactor, large thermal inertia

o Decay heat to boil coolant at

1 atm ~25 hours

Relies on Diesels for backup

power for decay heat removal

Relies on natural air circulation

for decay heat removal

Zr-H2O reaction generates H2 No H2 generation

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Reactor Decay Heat Removal

• The Direct Reactor Auxiliary Cooling System (DRACS) is used for decay heat removal when normal path through the Intermediate Heat Transport System is unavailable

• DRACS is a completely passive, natural convection NaK heat transport loop that transfers heat from primary coolant to ambient air

• Two heat exchangers in each loop

– DRACS (DHX) (Na-to-NaK, placed directly in primary sodium pool)

– Natural Draft HX (NaK-to-air, placed in air stack)

• Multiple loops are employed for redundancy

DHX

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TWRs are More Environmentally Beneficial

• Uses depleted uranium or waste from LWR

• Greatly reduced uranium mining

• Significantly less enrichment needed; none later

• No reprocessing facilities required

• At least 7X less high level waste relative to LWR

• Waste retained in the reactor; delayed external storage for up to 40 years

• Waste disposal footprint smaller and permanent

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TerraPower’s TWR Program

Integrated world class expertise and design innovations to maximize TWR performance: • Targeted design innovations

• Systematic qualification

• Irradiation test programs

• Leading calculation tools

• International suppliers

• National laboratories and universities

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The TerraPower Team:

Innovative Collaborative Approach

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TerraPower Global Team

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NEA RIAR

Kobe Steel

Y-12 NFS Michigan

INL

PNNL

LANL

UNLV

TAMU Florida DOE

ANL TerraPower HQ

KAERI MIT

Burns & Roe

ANL – Argonne Nat’l Lab

DOE – US Dept.of Energy

INL – Idaho Nat’l Lab

Florida – University of Florida

KAERI – Korea Atomic Energy Research Inst.

LANL – Los Alamos Nat’l Lab

MIT – Massachusetts Inst. of Technology

Michigan – University of Michigan

NEA – National Energy Administration, China

NFS – Nuclear Fuel Services, Inc.

PNNL – Pacific Northwest Nat’l Lab

RIAR – Research Inst. of Atomic Reactors, Russia

TAMU – Texas A&M University

UNLV – University of Nevada, Las Vegas

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TWR-P Project

• TWR prototype plant: – First electricity producing TWR –

Startup about 2023 – Confirms TWR design, verifies

shuffling strategies – Demonstrates key plant equipment

and verifies that models agree with operational performance

– Provides bases for 600 & 1150 MWe TWR plants

– Last step of fuel and material qualification

• Design features included for additional testing & development – Accommodates lead test fuel

assemblies – Refueling capability for post

irradiation fuel examinations – First-of-a-kind instrumentation, maintenance considerations

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TerraPower’s TWR-P Plant

TWR-D Nuclear Island

Containment

Dome

Reactor & Guard

Vessel

Reactor Core & Core

Support structurePrimary Sodium

Pump (2)

Intermediate

Heat Exchangers

(4)

Upper Internal

Structure

Thermal Shield

Secondary Sodium

Pipes and Guard pipes

Large and Small

Rotating plugs

Equipment Hatch

In Vessel Fuel

Handling Machine

Reactor Head

TWR-P Nuclear Island

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TerraPower’s TWR-P Plant

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No reproduction or distribution without express

written permission of TerraPower, LLC

Build on Applicable Fast Reactor Experience

• In the U.S., over 219,000 metallic fuel pins were irradiated in DOE’s own EBR-II and FFTF reactors

• Successful experience with components and systems used at FFTF (400MWt) and EBR-II (62.5 MWt) forms the basis for the design of the TWR

• The TWR design takes advantage of FFTF and EBR-II experience and incorporates the proven features of those sodium fast reactors into the design of the TWR

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Advanced Reactor Modeling Interface (ARMI)

Software: • MC**2 • REBUS/DIF3D • MCNPXT/CINDER • SUPERENERGY • ALCHEMY • XTVIEW • SAS4A/SASSYS-1 • ARMI

Will run Monte Carlo simulations of 110,000 zones, each with 3400 nuclides, out for 60 years, and receive results in 1 day.

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Core Support Structure – Fuel Design Integrated Modeling

No reproduction or distribution without express written permission of TerraPower

Pin Models

Pin-Duct Interaction

Models

Single Assembly Models

Full Core Models

Seismic Models &

Core Restraint System

Core Neutronics & Thermal Hydraulics

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Computational Fluid Dynamics (CFD) Result in Design and Testing Efficiencies

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Fuel Assembly Models – Duct dilation

– Fuel rod swelling

– Coolant thermal mixing

– Duct wall shaping or roughness

– Peak clad temperature

Technology Development: Focus on Fuels and Materials

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Priorities – Fuels and Materials

• High burnup, metal fuel needed

– At least 30% peak for TWR

– Data limit is 20% (in EBR-II)

• High neutron dose needed

– Approximately 500 dpa peak for TWR

– Data limit is 200 dpa (in FFTF)

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TWR Prototype fuel assembly height in comparison with ALMR, FFTF and EBR-II

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ALMR

Shield – 0.4m

Nosepiece – 0.33m

Core – 2.0m

plenum – 2.3m

socket– 0.3m

5.4

m

socket– 0.3m

plenum – 1.88m

Core – 1.35m

Shield – 1m

Nosepiece – 0.33m

5.0

m

TWR

FFTF

Socket– 0.3m

Plenum – 1.4m

Core – 0.9m

Shield+orifice – 0.8 m

Nosepiece – 0.33m

3.6

5m

Socket– 0.4m

Plenum– 1.4m

Blanket– 0.46m

Blanket– 0.46m

Core – 0.36m 2.3

3m

Nosepiece – 0.6m

EBR-II

Scale-up from FFTF to TWR prototype roughly comparable successful step from EBR-II to FFTF

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Metal Fuel Clad Breach Safety Advantage

Test of fuel operation with breached cladding

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Oxide fuel 9% burnup Large enlargement of breach site due to generation of low density sodium-fuel reaction products, loss of fuel

Metal fuel 12% burnup, after 169 days No indication of breach size increase Metal fuel compatible with sodium

Even if cladding fails and continues operation at power, no fuel is lost to coolant! For severe accidents with fuel melting , fuel is dispersed in sodium pool and reactivity reduced Very small release of radioactivity due to absence of driving forces (sodium pressure ~1atm)

Chang, Nucl.Eng.Tech., Vol 39, No.3, 2007

No reproduction or distribution without express

written permission of TerraPower, LLC

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TWR Qualified Fuel

Michigan (ion irradiations)

CSM, PNNL (welding)

TerraPower (metallography and

analysis)

LANL/PNNL (material)

Florida (material prep)

Kobe (fabrication)

RIAR (neutron

irradiations) NFS (fabrication)

Y-12 and INL (U metal fab)

TerraPower (fab development,

modeling)

TAMU (advanced fuel studies, FCCI)

UNLV (fuel alloy, FCCI)

INL (fuel data, PIE) -------------------

ATR Irradiations

Supplies product Supplies results

MCE (PT’s)

TWR-P (irradiation)

Testing Activities and Organizations

BWTS (material)

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HT9 Material Development

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HT9 shows strong resistance to radiation damage, but the TWR reactor will push the limits of cladding performance,

where swelling and creep become significant

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Fast Reactors Materials Performance

Poor grain structure (heat 3)

Good grain structure (heat 4)

-0.2

0

0.2

0.4

0.6

0.8

1

1.2

1.4

0 100 200

Swe

llin

g (v

olu

me

%)

Irradiation Dose (dpa)

Heat 1

Heat 2

Heat 3

Heat 5

HT9, Heat 1

HT9, Heat 2

HT9, Heat 3

HT9, Heat 4

?

Swelling rate

Examination of archive HT9 material reveals significant differences in grain structure, helps

explain the variability in material swelling performance after irradiation

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HT9 Selection Strategy

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HT9 - 1

HT9 - 2

HT9 - 3

HT9 - 4

Kobe fabricates HT9 Tensile testing to estimate thermal creep performance

Ion irradiations to determine swelling performance

Extractions to examine carbide microstructure

Each component of the test program leads to a selection of final HT9 heat treatment

HT9 – TWR-P

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Swelling in HT9

0

0.5

1

1.5

2

2.5

3

1 2 3 4 5 6 7 8

Swel

ling

(% V

ol.)

@ 4

40°

C

HT9 Heat Treatment #

Expected to be worse heat treatment

Heavy ion irradiation of HT9 heat treatments to understand relative effects on swelling

Expected to be best heat treatment

Validation of an understanding of microstructure

and swelling

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• TerraPower HT9 supply – Kobe optimized plate and tube – INL tube for weld study

• TerraPower MCE Fabrication – Phase 1: Weld Study and plate

– Phase 2: Kobe tube material

• Irradiated Samples – LANL FFTF ACO-3 Duct – PNNL FFTF MOTA

• Shipping – LANL Shipper, Edlow Logisitics

Barrier Coatings

INL HT9 Tube

BWTS Container

Pressurized Tube Package

BOR-60 Materials Irradiation Test

INL HT9 Tube

Kobe HT9

Currently on schedule to ship by end of Apr. 2013 to support a Jul-Aug 2013 insertion in BOR-60

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1 – capsule with gas-filled wall for thermal insulation; 2 –

container with a single-layer wall; 3 – samples; 4 – unflowing

sodium; 7 – capsule with single-layer wall; 8 – 30 fuel rod heater.

Type 1

T:360˚C; 400˚C

Type 4

Т: 450˚C; 550˚C;

625˚C

BOR-60 Irradiation Program: Two Capsule Types

• Two capsule types to achieve desired temperature uncertainty

• 2 Type 1 Irradiation Rigs – 360˚C and 400˚C – Gas-filled wall and stagnant Na – Cavity: 450L x 30/8.6D mm

• 3 Type 4 Irradiation Rigs – 450˚C, 550˚C, 625˚C – Fuel Pre-Heater and flowing Na – Cavity: 280L x 38.6/8.6D mm

Test Insertion in ATR – June/July ATR Outage

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Advanced Clad Development

January February March April May June

January February March

Final Casting Development

Cast Enriched Specimens

Fuel TMT

Receive Clad

Finish Fuel Slugs

Ship to

ATR

Assemble Experiment

♦ 5% Burnup by Mid-2014 ♦ Initial PIE - 2015

Engineering Development

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TWR-P Engineering Design Overview

Priorities • Neutronics and reactor performance • Reactor safety analysis, with code V&V • Pin venting and associated support systems • Design requirements refinement • Reactor enclosure and internals design • Main Heat Transport optimization • PRA Development • I&C • Testing

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TerraPower Prototype Reactor (TWR-P)

Containment

Dome

Reactor & Guard

Vessel

Reactor Core & Core

Support structure Primary Sodium

Pump (2)

Intermediate

Heat Exchangers

(4)

Upper Internal

Structure

Thermal Shield

Large and Small

Rotating plugs

In Vessel Fuel

Handling Machine

Reactor Head

Secondary Sodium

Pipes & Guard pipes

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Key Suppliers

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Supply Chain Development:

• Curtiss-Wright (Primary, Intermediate pumps)

• Babcock and Wilcox (CRDMs)

• GE (IHX, Steam Generator)

• Doosan (TG, Condenser, steam cycle comp)

• Invensys (I&C)

• Fike (pin vent)

• Toshiba (Sodium Test Facilities, components)

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Key Suppliers

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Design and Analysis: • Burns & Roe (A/E work- GAs, PFDs, calcs ) • AECOM (P&IDs, analysis) • ARES and Becht (Seismic) • Creative Engineers, Inc (sodium systems) • Cryogenic Consulting Services (cover gas) • O’Donnell Engineering (T-H analysis) • Vista Engineering (FE analysis) • Brayton Energy (IHX design and analysis) • Merrick (trade studies) • Cameron Group, CBCG (Sodium Fast Reactor expertise) • Alden (flow testing) • MIT, INL, OSU (core analysis and scaled testing analysis)

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Fuel & Core Mechanical Design

Priorities

• Fuel duct behavior and core restraint system

• Fuel rod and duct fabrication

• Fuel slugs and protective liner application

• Supplier development

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TWR Core View

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Fuel Assembly • Fuel Rods • Duct • End Fittings

Loaded into the TWR core • Control Rods • Shield & Reflector

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Licensing Strategy

• TWR-P design and safety analysis proceeding with anticipated regulatory requirements – Based on current US LWR regulatory requirements

– Adapted for unique characteristics of sodium fast reactors (SFRs)

• Regulatory Requirements – Based on US 10 Code Federal Regulations Part 50

– General Design Criteria for TWR-P completed

• Regulatory Compliance Approach – Based on NUREG-0800, Standard Review Plan (SRP)

– Adapted to reflect characteristics of SFR and TWR-P

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TWR-P SAR Project Plan

1. Introduction and Interfaces

2. Site Characteristics and Site Parameters

3. Design of Structures, Components, Equipment, and Systems

4. Reactor

5. Reactor Coolant System and Connected Systems

6. Engineered Safety Features

7. Instrumentation and Controls

8. Electric Power

9. Auxiliary Systems

10. Steam and Power Conversion System

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11. Radioactive Waste Management

12. Radiation Protection

13. Conduct of Operations

14. Initial Test Program and ITAAC-Design Certification

15. Transient and Accident Analysis

16. Technical Specifications

17. Quality Assurance

18. Human Factors Engineering

19. Severe Accident

• Project tasks for each SAR chapter are fairly independent – Tasks can be performed in parallel

– Tasks can be performed efficiently by multiple organizations

• Project structure based on SAR Chapters

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US LWR Regulatory Requirements

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Safety Analysis Report (SAR) Regulatory Requirement (1)

No. Title No. Comment Scope Limit

15.0 Transient and Accident Analyses 10 CFR 20

15.0 Transient and Accident Analyses 10 CFR 50.46

15.0 Transient and Accident Analyses 10 CFR 100

15.0 Transient and Accident Analyses GDC 2 As it relates to the seismic design of structures, systems, and components (SSCs) whose failure could cause an unacceptable reduction in the capability of the residual heat removal system.

15.0 Transient and Accident Analyses GDC 4 As it relates to the requirement that SSCs important to safety be designed to accommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated onditions, including such effects as pipe whip and jet impingement accident.

15.0 Transient and Accident Analyses GDC 5 As it relates to the requirement that any sharing among nuclear power units of SSCs important to safety will not significantly impair their safety function.

15.0 Transient and Accident Analyses GDC 10 As it relates to the RCS being designed with appropriate margin to ensure that specified acceptable fuel design limits are not exceeded during normal operations including AOOs.

15.0 Transient and Accident Analyses GDC 13 As it relates to instrumentation and controls provided to monitor variables over anticipated ranges for normal operations, for AOOs, and for accident conditions.

15.0 Transient and Accident Analyses GDC 15 As it relates to the RCS and its associated auxiliaries being designed with appropriate margin to ensure that the pressure boundary will not be breached during normal operations, including AOOs.

15.0 Transient and Accident Analyses GDC 17 As it relates to the requirement that an onsite and offsite electric power system be provided to permit the functioning of SSCs important to safety. The safety function for each system (assuming the other system is not working) shall be to provide sufficient capacity and capability to ensure that the acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded during an AOO and that core cooling, containment integrity, and other vital functions are maintained in the event of an accident.

(1) Capture Regulatory Requirements from the SRP that corresponds to the SAR section.

Scope Limit identifies requirement that is only applicable for a specific type of application (CP, OL, ESP, DC, COL)

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Testing Development

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Two categories can be identified: 1. Technology Development Testing 2. Component Testing

A Technology Readiness Assessment (TRA) was performed on the Nuclear Island systems

• Identified areas where technology development and specialized component testing was needed

• Became the basis for a comprehensive Test Matrix

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Technology Readiness Levels

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2014 Preliminary Design

2012 Pre-conceptual Design

2016 PSAR

2018 FSAR

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Technology Development and Testing Flowchart

Test Programs

Design Documentation

General Requirements

General Design Criteria

NUREG 0800

Functional Requirements Document

Verification Methods

Design & Drawings

Testing

Calculations & Analysis

Test Matrix

Technology Testing

Proof of Fabrication

Qualification tests

Model V&V

System Design Documents

Equipment Specifications

Procurement Specifications

Technology Readiness

Assessments

TDT&Q Manual

Computational Mechanics

Fuel and Materials Development

Component Testing and Qualification

Test results

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I&C Development

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Worked with Invensys to produce: • Phase I: Control System Architecture • Phase II: Licensing Approach

Next steps: • Work with vendor to create Engineering Simulator • Develop TerraPower I&C software design documents • Create infrastructure for Control System V&V program • Benchmark against previous SFR Control approaches

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• The Traveling Wave Reactor (TWR) is a new Generation IV nuclear power plant developed by TerraPower designed to: – Enable a new regime of nuclear safety standards

– Be cost competitive by using huge stores of depleted uranium waste as fuel, eliminating long term enrichment needs, completely eliminating reprocessing and using simplified final waste disposal

– Be more environmentally friendly with the creation of significantly less waste

– Greatly increase nuclear proliferation resistance and therefore enable freedom to export

– Enhance energy security by requiring significantly less natural uranium and a less complex supply chain

The TWR: Summary of Benefits

The TWR is based on proven technology and can be

licensed and ready to operate in 10 years

Thank you