1 Nuclear Energy Division Materials for Generation IV Nuclear Reactors Cargese, Sept. 24 – Oct. 6, 2007 Outlook on Gen IV Nuclear Systems and related Materials

1Nuclear Energy Division Materials for Generation IV Nuclear ReactorsCargese, Sept. 24 – Oct. 6, 2007

Outlook on Gen IV Nuclear Systems and related Materials R&D Challenges

- Goals for innovative reactor systems- Requirements for structural materials: generic and specific

- Synergies, crosscutting R&D areas and modelling- Significance of international collaboration

Frank Carré and Pascal Yvon

CEA – Nuclear Energy [email protected]


Sustainable Nuclear Energy Technology Platform (SNE-TP)

SNE-TP Objectives & Organization

Kick-off meeting : September 21, 2007

Nuclear Fission Technology Platform: SRA and Platform Operation

LWR

Safety & Economics

V/HTR

Process Heat, Electricity & H2

Fast Systems With Closed Fuel Cycles

Sustainability

Innovative Materials and Fuels

Simulations and Experiments:

Reactor Design, Safety, Materials and Fuels

Training and R&D Infrastructures

Geological Disposal Technologies, design, safety assessment

Strategic Research Agenda; Platform Operation

Interactions between NFTP & Interactions between NFTP & other other TPsTPs, initiatives, etc, initiatives, etc

Waste Management

(CARD)

EU High Level Group on Nuclear Safety & Security

TSO:Mirror Group

Technical Safety

Organisations


New goals for sustainable nuclear energyNew goals for sustainable nuclear energy

Continuous progress: Economically competitive Safe and reliable

Membersof the Generation IV International

Forum

Membersof the Generation IV International

Forum

USAUSA

ArgentinaArgentina

BrazilBrazil

CanadaCanada

FranceFrance

JapanJapan

South AfricaSouth Africa

UnitedUnitedKingdomKingdom

South KoreaSouth Korea

SwitzerlandSwitzerland

EUEU

Systems marketable from Systems marketable from 20402040 onwardsonwards

TrueTrue potential for new potential for newApplications:Applications: Hydrogen,Syn-fuel, Desalinated water,Process heat

Internationally shared R&DInternationally shared R&D

Break-throughs:Natural resources conservationWaste minimisationProliferation resistance

Generation IV Nuclear Systems

ChinaChina RussiaRussia

A closed fuel cycle


Generation IV Forum: selection of six nuclear systems

Sodium Fast Reactor

Lead Fast Reactor

Molten Salt Reactor

Gas Fast Reactor

Supercritical Water-cooled ReactorVery High Temperature Reactor

(12-20y) R&D (~1 B€) before a 1st prototype or techno demo


Fast Reactors & recycling for Sustainable Nuclear Power

R T

Udep

FP

U Pu MA

R T

Udep

MA

U Pu

FPR T

Udep

FP MA

U Pu

Homogeneous Recycling

Heterogeneous Recycling

U & Pu Recycling

Natural resources conservation Waste minimisation Proliferation resistance (intl standards)

Type of nuclear materials Detection, technical difficulty, cost, time…

Strategies for a flexible management of actinides in Gen IV fast neutron systems.

Implementation depending on international standards and national optimization criteria (economics & waste).


Technical challenges & Leading physical phenomena

60-year lifetime

Fast neutron damage (fuel and core materials) Effect of irradiation on microstructure, phase instability, precipitation Swelling growth, hardening, embrittlement Effect on tensile properties (yield strength, UTS, elongation…) Irradiation creep and creep rupture properties Hydrogen and helium embrittlement

High temperature resistance (SFR > 550°C, V/HTR > 850-950°C) Effect on tensile properties (yield strength, UTS, elongation…) High temperature embrittlement Effect on creep rupture properties Creep fatigue interaction Fracture toughness

Corrosion resistance (primary coolant, power conversion, H2 production) Corrosion and stress-corrosion cracking (IGSCC, IASCC, hydrogen cracking & chemical compatibility…)

Requirements for materials in future nuclear systems (1/2)


Additional requirements

Material availability and cost

Fabricability, joining technology

In service inspection Non destructive examination techniques

Safety approach and licensing Codes and design methods

R&D effort needed to establish or complement mechanical design rules and standards

Decommissioning and waste management

Requirements for materials in future nuclear systems (2/2)


Structural materials for Innovative Reactor Systems

SFR GFR LFR VHTR SCWR MSR Fusion

CoolantT (°C)

Liquid Na few

bars

He, 70 bars480-850

Lead alloys

550-800

He, 70 bars

600-1000

Water280-55024 MPa

Molten salt

500-720

He, 80 b300-480

Pb-17Li 480-700

Core Structures

Wrapper F/M

steels

Cladding AFMA

F/M ODS

Fuel & core

structures

SiCf-SiC composite

Target, Window Cladding

F/M steels ODS

CoreGraphite

Control rodsC/C

SiC/SiC

Cladding & core

structures

Ni basedAlloys &

F/M steels

Core structure

Graphite

Hastelloy

First wallBlanket

F/M steelsODS

SiCf-SiC

Temp. °C 390-700 600-1200 350-480 600-1600 350-620 700-800 500-625

DoseCladding 200 dpa

60/90 dpa

Cladding ~100 dpa

ADS/Target~100 dpa

7/25 dpa~ 100 dpa

+ 10 ppmHe/dpa+ 45 ppmH/dpa

Other components

IHX or turbine

Ni alloys

IHX or turbine

Ni alloys


A new generation of sodium cooledFast Reactors

Reduced investment cost Simplified design, system innovations(Pool/Loop design, ISIR – SC CO2 PCS)

Towards more passive safety features+ Better managt of severe accidents

Integral recycling of actinidesRemote fabrication of TRU fuel

SFR Steering

Committee

U.S.A.U.S.A.JapanJapan

FranceFrance

South KoreaSouth KoreaEuratom Euratom countriescountries

ElectricalPower

GeneratorTurbine

Condenser

Heat Sink

Pump

Pump

Pump

PrimarySodium（Cold）

Cold Plenum

Hot Plenum

PrimarySodium（Hot）

Control Rods

Heat Exchanger

Steam Generator

Core

SecondarySodium

ElectricalPower

GeneratorTurbine

Condenser

Heat Sink

Pump

Pump

Pump

PrimarySodium（Cold）

Cold Plenum

Hot Plenum

PrimarySodium（Hot）

Control Rods

Heat Exchanger

Steam Generator

Core

SecondarySodium

2009: Feasibility – 2015: Performance 2020+ : Demo SFR (FR, US, JP…)

Sodium Fast Reactor (SFR)

2007+

RussiaRussia

ChinaChina


SFR Primary system

New 9-12%Cr F/M steel vs Advanced AusteniticGood physical and thermal properties, dilatation, low costBetter creep resistance (T91, T92 (Fe-9Cr-xW-V-N…))

Compactness, mass reduction of components DBTT but Improved toughness Weldability (%Cr dependent) Good compatibility with sodium impurities (C, O, N)(Demonstrated in Phénix 2ry system & Steam generator + 150 000 h Irradiation experiments of T91 & ODS (SuperNova))

Compact component and system designs (piping, IHX…)

Potential margin for temperature increase (< 600°C) (especially if using a gas turbine power conversion system)

Allowable departure from the negligible creep regime?

New materials for sodium fast reactors (1/2)


J.L. Séran, A. Alamo, A. Maillard, H. Touron, J.C. Brachet, P. Dubuisson, O. Rabouille J. Nucl. Mater. 212-215 (1994) 588-593.

Great stability of fracture properties 9% Cr Martensitic steels

-100

0

100

200

300

400 450 500 550 600

F17Cr SLF17Cr STM12Cr (HT9) SLM12Cr (HT9) STM9Cr (EM10) SLM9Cr (EM10) ST

DB

TT

(°C

)

IRRADIATION TEMPERATURE (°C)

17%Cr

12%Cr

9%Cr

Dose: 70 - 110 dpa - Phenix irradiation

Unirradiated

Sodium Fast Reactor structural materials: F/M Steels


Advanced fuel cladding

316 Ti 15-15 Ti F/M ODSReduced swelling with neutron fluence

EM10 & 15-15 Ti 100 dpa @ 400-700°CT92/HC & ODS 200 dpa @ 480 – 750/800°C

Weldability & joining techniquesGood compatibility with sodium impurities (C, O, N)

Increased fuel burnup 200 GWd/t & 200 dpa

Increased safety with low sodium content in the core & low sodium void effect Better prevention of severe accidents

New materials for sodium fast reactors (2/2)

Research in progress on hardening of F/M steels with micro/nano structures (dispersion / precipitates)

2nd generation ODSF/M steels with carbo/nitride precipitates

Ferritic-ODS by HIP


Sodium Fast Reactor cladding material

Swelling of austenitic Phénix claddings compare to F/M materials

0

1

2

3

4

5

6

7

8

9

10

60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 dose (dpa)

Average 316 Ti

Ferritic-martensitic (F/M) steels, ODS included

Average 15/15Ti Best lot of 15/15Ti

Embrittlement limit

(%)

Swelling of advanced austenitic steels and ferrito-martensitic steels

used as fuel cladding in Phenix


Safety enhancement of Fast Reactor core

Low reactivity sodium void effect high BU coreLarge diameter fuel pin with thin spacer wireODS cladding for low swelling(Experiments in Phenix (Supernova, Matrix1&2) + in Joyo)

15-15 Ti lot CE

15-15 Ti bas C

12-25 Ti N915-25 bas Ti

15-25 Ti Nb DS5

15-25 Ti Nb DS4

16-25 Ti Nb V TS2

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

MA 957 MA 956

V/V%

T °C

15-15 Ti lot CE

15-15 Ti bas C

12-25 Ti N915-25 bas Ti

15-25 Ti Nb DS5

15-25 Ti Nb DS4

16-25 Ti Nb V TS2

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

-5

0

5

10

15

400 450 500 550

MA 957 MA 956

V/V%

T °C

COEX COCA

MOX fuel fabrica-tion from co-precipita-ted UPu solution from the COEX process To be first tested in Phenix (Copix expt)

Various recycling modes of minor actinides in Fast Reactors: Homogeneous (~2% MA): GACID Heterogeneous in blanket (10-20% MA): Curios, Amboine2-Joyo expts


A novel type of Gas-cooled Fast Reactor: an alternative to the Sodium Fast Reactor, and a sustainable version of the VHTR

Robust heat resisting fuel (<1600°C) 1200 MWe – THe ~ 850 °C - Cogeneration of electricity, H2, synfuel, process heat Safe management of cooling accidents Potential for integral recycling of Actinides

GFR Steering

Committee

JapanJapan


FranceFrance

Euratom Euratom countriescountries

Generation IV Gas Fast Reactor (GFR)

2012 : Feasibility ~2020 : ETDR (EU ?)2020: Performance 2030+: GFR Prototype

GCFR 5-6 EU PCRD

System Arrangement GFR signed Nov. 30 Nov.,2006Project Arrangements “Fuel “ & “Design-Safety-Integration” in 2007 U.S.A.U.S.A.


Gas Fast Reactor fuel designs

0 25 50 75 100%vol. of actinides compound in the volume dedicated to fuel

High densitycompartmented

platelet Advancedparticles

HTRs

Claddedpellets


Candidate ceramics materials for the GFR fuelCeramics

for Gas Fast Reactor

Usual low toughnessof ceramics

Composite CERMET

TiC (HIP)

10 µm

10 µm

trans-granulaire

inter-granulaire

Mixed CER & MET matrices

Fibre strenthened

Multi-layer materials

Interfaces

Manufacturing and testing monolithic and composite ceramics (C/C, SiC/SiC) Characterization and optimization

Objectives: Increase ceramics ductility and toughness

Investigation and modelling of phenomena

Matrix& concept in Phénix


43.0mm

5.0mm

Wall thickness: 1.0mm

Goal: 3m (length) x 10mm (inner diameter) Goal: 3m (length) x 10mm (inner diameter) x 1mm (wall thickness) x 1mm (wall thickness)

2D SiC/SiC by NITE Process for GFR Fuel Pin or Plate

Fuel Pin

Fuel Plate

Nite ProcessKyoto University


A few specific R&D areas on ceramics for GFR fuels

Possible applicationsas matrix or interphase in SiC/SiCf composite…

Nano-laminate structureTi3SiC2

Ti, Ti, C, C, SiSi

MAX Phases

Special propertiesDamage tolerantLow density, machinableHigh thermal andelectrical conductivities

Methods used to obtain large-scale bulk Ti3SiC2

CVD, Arc melting, HIP, HP, SHSHigh energy ball milling & reactive sintering to obtain bulk Ti3SiC2 with very fine grain

Synthesis of TiC from nano-powderHIP without grain growthStable under irradiation (electrons & ions)


A nuclear system dedicated to the production of high temperature process heat for the industry and hydrogen

600 MWth - THe >1000 °CThermal neutronsBlock or pebble core concept

Passive safety features I-S Cycle or HT Electrolysis for H2

VHTR Steering

Committee

U.S.A.U.S.A.JapanJapan


FranceFrance

South KoreaSouth KoreaSouth AfricaSouth Africa

Euratom Euratom

Generation IV Very High Temperature Reactor (V/HTR)

2009: Feasibility – 2015: Performance~ 2020: PBMR, NGNP & Other Near Term Projects

2007+

ChinaChina


VHTR vs PWR pressure vessel manufacturing techniques

PWR Vessel

VHTR Vessel

Normal/off-normal service temperatures and vessel size dominate materials requirements

Up to <450/550°C at 5-9 MPa Up to 1 x 1019 n/cm2 fluence

Very large vessel sizes require scale-up of ring forging & on-site joining

technologies

Irradiation resistance to be demonstrated for licensing

9Cr1Mo alloy for pressure vessel of gas cooled reactors


VHTR core material: Graphite & Composites

Graphite (PCEA (UCAR), NBG 17 (SGL)…) Characterization: chemical, structural, thermal & mechanical

properties (20-1000°C), corrosion tests (air; water, O2, CO2…)

Irradiation tests (T ~1050°C, 1-6 dpa G)

Optimization for waste minimization (14C)

Technical file for codification of design standards

C/C & SiCf/SiC composites Manufacturing: 2D & 3D woven fibres (C, Hi-Nicalon S), interphases, CVI or pitch densification, anti-oxidation coating (Si, B)…)

Characterization: chemical, structural, thermal & mechanical properties (20-1000°C), corrosion (air,water, O2, CO2), irradiation tests

Technical file for codification of design standards

500

mm

60 mm

100 mm


Three IHX technologies identified: Plate-machined Heat Exchanger (Fig. 1) Plate-Fin Heat Exchanger (Fig. 2) Tubular concept

Key issues to be addressed:

Materials development- Haynes 230- Inconel 617- Ni-ODS

Intermediate Heat Exchanger design- Compactness- High thermomechanical resistance- High thermal efficiency (95%)- Low pressure drop, no leakage

Properties required at 850°C - 950°C - Tensile, long term creep (Fig. 3),

fatigue,creep-fatigue- Corrosion resistance- Fabrication and joining

techniques

SERRATED FINS

Before and after BRAZING

0.8 à 2.5 mm

3.53 à 9.63 mm

MODULE

PLATES/FINS Assembly

VHTR Intermediate Heat Exchanger

Creep strength: Haynes 230 and Inconel 617

FIG. 1

FIG. 2

FIG. 3


Generation IV Systems related R&D Needs

VHTR SFR GFR SCWR LFR MSR

MetalsMechanics, Corrosion

F/M steels ODS

Ni-alloys

F/M steels ODS

Austenitic

Steels

F/M steels ODS

Ni-Alloys

Clad & structures

Ni-alloys

Radiolysis

Hastelloys

Ceramics & Composites

Graphite

C/C, SiCf/SiC

SiC, TiC

Ceramics

Component mock-ups

IHX & HX,

RPV (9 Cr)

Control rods

HX, SG IHX & HXRPV & DHR

IHX & HXPump

1ry system Technology

He test benches & Loops MW

ISIR He test benches & Loops MW

Heat transfer

SCW Loops

Water chemistry

Corrosion

Purity control

MSalt Techno

Loops

Mechanical Design Rules

HT Design Codification

ASMERCCM-R



Structural materials & Components


Typical Tokamak Configuration T-Breeding Blanket:

Dual Coolant Lithium Lead

Design and technology of T- breeding blanket

He, 80 bars Pb-17Li, ~bar

300, 480 0C 480-700 0C

Fusion Power ReactorDual-Coolant T-Blanket

Martensitic Steels (550 0C)

ODS Ferritic steels (700 0C)SiCf-SiC th. & elect. insulator

Dual-Coolant T-Blanket

F W: T max= 625 0C

Channel: Tmax= 500 0C

Insert: Tmax~1000 0 C


Materials science and new materials are key for optimizing 2nd & 3rd

generation LWRs as well as to meet 4th nuclear systems’ objectives :> 2040: Fast reactors with a closed fuel cycle (SFR, GFR, LFR)~2025-30: High temperature reactors (V/HTR) for process heat (H2…)More prospective nuclear systems (SCWR, MSR)

Incremental progress and breakthroughs are sought on a wide span of structural materials for fuel claddings, core structures, reactor cooling systems & components (RPV, IHX, SG…), power conversion systems (electricity, H2…):

Metals: Austenitic steels, 9-12Cr F/M steels, ODS, Ni-alloys…Ceramics & composites: Graphite, C/C, SiCf/SiC & (TiC, ZrC, Ti3SiC2…) Fabrication, characterization, manufacturing, ND examinationMechanical design codification: ASME, RCCM-R + extensions / harmonization needed for fast neutrons, high temperature, lifetime…

Synergies between materials for 4th generation nuclear systems as well as with materials for Fusion (1st wall & blanket)

Ni-alloys, 9-12%Cr F/M steels, ODS, Ceramics (SiC…)

Innovative Reactor Systems & Requirements for Structural Materials

Summary (1/2)


Increased role of Materials science (analytical research and modelling) for a more predictive R&D towards aimed materials properties

Metals Ceramics Fuels

International cooperation to increase and share R&D work and achieve breakthroughs for 21st century nuclear power systemsFederate national programs into a consistent international roadmap

Enhancing R&D and technology demonstrations (Gen IV, EU FP7…)Databases of materials propertiesMulti-scale modelling of materials & fuelsSynergies between Fission and Fusion materials

Progressing towards harmonized international standardsMechanical design rules and standards, CodificationSafety, non-proliferation, physical protection…

Innovative Reactor Systems & Requirements for Structural Materials

Summary (2/2)

Documents

1 Nuclear Energy Division Materials for Generation IV Nuclear Reactors Cargese, Sept. 24 – Oct. 6, 2007 Outlook on Gen IV Nuclear Systems and related Materials