Structural Materials for Advancer Reactor Systems: IAEA ... · Structural Materials for Advancer Reactor Systems: IAEA Support activities ... 29 March 2011 17 . ... RCM held in May

IAEA International Atomic Energy Agency

Structural Materials for Advancer Reactor

Systems: IAEA Support activities

Andrej Zeman

NAPC / Physics section

IAEA TWG Gas Cooled Reactors Vienna, 29 March 2011

IAEA

Outline

PS introduction

On-going activities

Coordinated Research

Recent activities & education

IAEA TWG Gas Cooled Reactors

Vienna., 29 March 2011 2

IAEA

The PS supports the IAEA Member

States regarding utilization of: Accelerators

Research reactors

Cross-cutting material research

Controlled fusion

Nuclear instrumentation

PS implements P&B activities based on

MS demand. Organisation of Int.

conferences, Technical and expert

meetings, CRP, Networks, DBs,

Technical Cooperation, etc. Objective is to promote nuclear science &

technology, specifically applied physics and

material science related to nuclear energy.

Physics section profile



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Application of accelerators

In total more than 15.000 accelerators used

world-wide, multidisciplinary use.

Small and medium size facilities; particle and

X-ray machines (CC scheme),

Research & industrial applications, non-

nuclear (semiconductors, medicine, biology,

geology, archeology, etc.) & nuclear (fusion

and fission reactors)

Applications various probing methods (IBA,

PIXE, PIGE, SAXS, XFR, etc.), recently

development & characterization of novel

materials for hydrogen production, storage and

conversion.

www-naweb.iaea.org/napc/physics/accelerators/database/index.html

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Research reactors utilisation

Support of basic & applied research (neutron physics, material science, industrial applications)

Non-nuclear areas: biology, medicine, semiconductors, hydrogen energy systems (storage & conversion).

Operational safety: monitoring and assessment of core components

Approx. 670 research reactors constructed

around the world, about 240 are still operating

Irradiation programs (radio-isotopes, R&D

structural materials, nuclear and non-nuclear

energy applications)

Training activities and know-how

dissemination (professionals & students)

www-naweb.iaea.org/napc/physics/research_reactors/database/database.html

IAEA

Activities linked with the rationale of IAEA’s program: 1.4.2.1 (D2) Enhancement of utilization and applications of

RRs, promote the continued development of scientific research and technological development using research reactors.

1.4.3.1 (D3) Accelerator techniques for modification and analysis of materials for nuclear, analytical and computational investigative tools, on the engineering front - new material performance testing technologies,

1.4.4.1 (D4) Supporting plasma physics and fusion research, support of advanced devices operating plasmas that are used for materials research and industrial applications.

PS activities – programmatic view



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Outline

PS introduction

Overview & lessons learned

Coordinated research

Activities and education



IAEA

Agency should encourage and assist research and development and practical application of atomic energy for peaceful uses and to foster the exchange of scientific and technical information

Support bilateral and international initiatives and their joint R&D on innovative approaches to nuclear power

Secretariat has to promote the exchange of relevant technical information among interested MS and foster HR trainings

IAEA should identify and explore innovative institutional and infrastructural solutions supporting the future deployment of innovative nuclear energy systems

Coordination and strengthening of research activities among MS (e.g. CRP, WGs, Expert meetings, etc.)

53rd IAEA General Conference, Vienna, 14-18 Sep 2009




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Zeman et al., Int. School of Physics (ITEP), Moscow, 12-18 February 2007

System of barriers (FP): Fuel matrix - Cladding – Reactor (PL) –

Containment – Emergency planning (transition regimes, sever

accidents)

R(P)V – key component (non-replaceable), LWR (Gen II+ from 1980’s)

designed 40y, some operators plan to extend up to 60 (80y).

Structural materials - have to assure all (designed) parameters of

component will remain in defined limits!

Reactor core components – degradation due to ageing and other factors

(radiation, temperature, chemistry, etc.), it’s linked with component

reliability

Degradation: embrittlement, thermal creep, swelling, cracking, etc. to be

carefully considered in design phase (engineering approach)

Need to understand material behaviour in range of design limits and

beyond (transient and accident scenarios)!

“Defense in depth” principles




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Design criteria for innovative GCR (GFR/VHTR) Very good overall efficiency (operating at a higher

gas temperature) and modular construction

Proliferation resistance and better fuel management

(higher burn-up)

Bigger primary components (lower power density)

Implementation of passive safety systems

Limited number of reactor/years vs. operational

experiences.

Overview

Critical issues: Structure-Systems-Components and lessons learned (failure of

graphite components, axial crack discovered in the reactor core

Hinkley Point B).

Reactor vessel and core structures / internals, heat exchangers – long

term stability of properties (time & temperature)

Core qualification – engineering approach!



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Technology roadmap of innovative GCR (VHTR, GFR)

Overview

Development & design of full

scale heavy components (RPV,

IHX, Brayton cycle turbo-

machinery

Number of challenges related

to the high performance structural

materials (primary and

secondary)

New materials for fuel

applications (composite

ceramics clad mixed carbide fuel,

advanced fuel particles)

Components for high temperature

process heat (materials for fuel

cells and reaction vessels

What’s really achievable

today, especially in terms of

design, demonstration and

pilot installation?



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Temperature windows and radiation damage

Overview

IAEA CS Meeting, 3 Sep 2010, INTEC, KAERI, Daejeon, Republic of Korea

fusion SiC

V alloy, ODS steel

F/M steel



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Overview IAEA Experts Meeting, 3 Sep 2010, INTEC, KAERI,

Daejeon, Republic of Korea

Innovative GCR systems:

R&D of reactor, core-structural materials

and fuel, several "technological" issues have to be

solved, specifically

(1) Low deformation Swelling, creep and embrittlement

Chemical compatibility and corrosion

Stability at high temperature and

phase transformation

(2) Good performance - mechanical properties Long term stability, trensient and accident

conditions.

Fuel cladding chemical Interaction

reprocessing.

Price, production and fabricability (joining).



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Overview IAEA CS Meeting, 9-12 March 2010, Vienna, Austriaa



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Overview – R&D support

Challenges related to R&D of structural materials incl. viability

concerns faced in the deployment of new fleet of innovative

GCR systems. Some of the issues are linked with specific conditions (e.g. coolant,

high temperature regime, high burn-ups).

In principle R&D has cross-cutting in nature, i.e. common to the

different GCR an fusion reactors as well.

In any case, the assessment of the materials performance and the

prediction of the materials behavior under the specific conditions are

primary issues for the qualification of potential high temperature

stability and irradiation resistant structural materials

International collaboration in the exploitation of experimental

facilities and expertise, harmonisation of experimental procedures

and the creation of a comprehensive database are further important

issues.



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Accomplishment of addressed issues, further international research efforts

require to qualify new and commercially available materials under the

extreme conditions. Several activities have already been initiated in area of structural materials

related to the design safety areas,

Closer collaboration through new IAEA Coordinated Research Projects is

envisaged, it reflects R&D needs identified by the international platforms and

TWGs.

Areas related to R&D of advanced GCR technology (1/2): 1. Advanced thermo-mechanical, irradiation degradation and embrittlement

assessment of the properties of VHTR & GFR candidate structural materials

(joints/welds & coated systems, taking into account coolant properties, high temp

and extended operation period;

2. R&D efforts for the development and harmonisation of codes-of-practice for

advanced testing (non-standard and miniaturised specimens), environmental

testing and test methods and performance assessment under transient off-normal

and accidental conditions (BDBA).



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Areas related to R&D of advanced GCR technology (2/2): 3. Physically-based modelling and experimental validation to contribute to a better

understanding of the materials performance in the respective conditions and

environments, including damage interactions;

4. Strengthening of cross-cutting interactions among interested parties.

The classes of materials investigated comprise: High temperature range (> 800°C): Ni-based alloys, advanced ODS and

refractory-based systems, ceramics (silicon carbide composites), graphite and

carbon-carbon composites.

Intermediate temperature range (600-800°C): traditional and modified austenitic

steels, ODS F/M steels, iron or Ni-based superalloys, refractory alloys;

Low temperature range (300-600°C): austenitic steels, ferritic/martensitic steels,

and ODS alloys;

The activities include (1) Cross cutting R&D of structural materials, and (2) pre-

normative research, codes and standards. Main focus is to contribute to broader

energy issues, specifically energy systems evaluation and energy security.



IAEA


Current design and technological issues related to high temperature

applications in innovative GCR:

Corrosion behavior in impure helium of Haynes 230, a nickel base

alloy candidate for heat exchangers in VHTR

Primary focus on formation and the subsequent destruction of the

surface oxide layer at 900 °C and 980 °C.

In-situ gas-phase analysis coupled to post-exposure surface

analyses, it has been confirmed that Cr-rich surface oxide formed on

Haynes 230 at 900 °C is unstable above a critical temperature (Cr-

rich oxide reacts with carbon in solution in the alloy and produce

chromium and CO).

Effect of carbon monoxide partial pressure in the gas phase as well

as the influence of chromium and carbon in the alloy on critical

temperature (understanding of thermodynamics and kinetics aspects

into account).



IAEA

Boutard et al., IAEA-EC Topical Meeting, F4E Barcelona, 5-9 October 2009

First Wall Dose (dpa) Temperature appmHe/dpa

ITER Austenitic steel <3 dpa <300 oC

DEMO EUROFER 50-89 dpa <550oC

Power Plant ODS Ferritic Steels 100-150 dpa <750 oC

Power Plant SiCf/SiC Composites 100-150 dpa upto 1100oC

~12

(0.1-0.3 in

Fast Fission

Reactors)

Synergies between fusion and fission:

Reduced Activation approach

9%Cr F/M Steels, SiC-SiC, W-alloys

Low level waste after 80-100y

Nb and Mo are dominating



Cross-cutting R&D support

IAEA 4th GIF-INPRO Interface Meeting

IAEA HQs, Vienna. 1-3 March, 2010 20

Outline

Outline

PS introduction

On-going activities

Coordinated Research

Activities and education

IAEA

Last decade, main R&D activities driven by fusion community,

ITER and non-ITER countries contributed

Several significant breakthroughs achieved by knowledge form

other fields (e.g. ball-milling for ODS production)

Multi-disciplinary approach, effective application of lessons-

learned (advanced metallurgy, aerospace industry, nano-

science…)

Role of the theoretical modelling should not be over-estimated,

experimental studies are needed and will be crucial in future

material development

Continuous development of semi-mechanistical and multi-scale

models, especially in terms of radiation degradation

mechanisms



Recent R&D activities

IAEA

CRP Outputs



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Coordinated Research Project (on-going)

(1) Better understanding of radiation effects and mechanisms of

material damage and basic physics of accelerator irradiation under

specific conditions,

(2) Improvement of knowledge and data for the present and new

generation of structural materials,

(3) Contribution to developmental of theoretical models for radiation

degradation mechanism,

(4) Fostering of advanced and innovative technologies by support of

round robin testing, collaboration and networking.

IAEA CRP on Accelerator Simulation and Theoretical Modelling

of Radiation Effects (jointly NA-NE)

Deals with several issues related to the proton and ion beam

irradiation in order to achieve very high radiation damage,

project aims to facilitate following issues:



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Coordinated Research Project (on-going)

Project launched 01/2009, last reporting

RCM held in May 2010

Members have presented recent

achievements on experimental testing

of various ODS (MA957, PM2000,

EUROFER, K3, etc.) irradiated at various

temp (up to 550°C), dpa and dose rates

Studies of synergism H/He, combination (validation) of recent

theoretical models.

IAEA CRP on Accelerator Simulation and Theoretical Modelling

of Radiation Effects (jointly NA-NE) - FACTS

Extensive theoretical and experimental studies are being carried out

among participating laboratories form Belgium, China, European

Commission, France, India, Japan, Korea, Kazakhstan, Poland,

Russia, Spain, Slovakia, Ukraine and USA (18 full members).



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Coordinated Research Project (new)

IAEA CRP on Benchmarking of advanced materials pre-selected

for innovative nuclear reactors (jointly NA-NE)

Critical review of structural materials pre-selected for innovative

reactor systems (focus on FR technology), stimulation of further

technological improvements in SM area Response to MS demand in R&D of SM via

coordinated assessment of key parameters

and technological limits.

Performance testing of materials pre-selected

for primary components of new innovative

reactor systems.

Assessment of candidate materials for reactor vessel, internals and fuel

cladding; harmonisation of analysis, consideration of samples miniaturisation,

round robin, etc.).

Methodology for testing of recent ODS-grade steels from ageing other

degradation mechanism point of view (mechanical properties /microstructure).

Project launched recently, RCM will take place 2-6 May 2011, Vienna IAEA TWG Gas Cooled Reactors


IAEA

Coordinated Research Project (new)

IAEA CRP on Examination of advanced fuel and core structural

materials for fast reactors” (jointly NA and NE)

Promotion of information exchange related to R&D on materials for FR

core components, combined irradiation experiments.

Project aims to facilitate collaboration and sharing of experience in

characterization and irradiation behaviour of FR core materials under

high neutron fluence

Phase (I) focused on exchange of information via international platform

with aim to review of existing irradiation capabilities with consideration

of urgent needs from individual MS (in/out-of-pile & PIE)

Phase (II) should be initiated afterwards, internationally coordinated

fast neutron irradiation experiments and PIE of samples provided by

CRP members.

To be launched Q3/2011, MS active in fast reactor programs



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Recent scientific events

http://meeting.iaea.org/



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Upcoming activities



Development of new structural

materials for advanced

fission and fusion reactors

In cooperation with

16 – 20 April 2012

Hosted by JRC Ispra (Italy)

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Education & training activities

Open to IAEA & UNESCO

Member States, see

Support of international and regional education and trainings

Cooperation with ICTP and other collaborating centres

(ANSTO, RID, Elletra, etc.)

More info: www.ictp.it



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International cooperation

Support of basic R&D to be addressed – material research is

cross-cutting activities – invitation could be achieved in the

framework of broader research community (Universities, non-

GIF members, etc.).

Closer interaction with other Int. organisations is needed (BA,

ITER, SNETP, EERA, IEA-FA…).

International coordination is essential – effective allocation of

resources and sharing of best practice in R&D process.

Active contribution/participation in ongoing and new CRPs, TM

and other initiatives (and vice versa).

Support of educational programs (ICTP/UNESCO, WNU,

Scientific workshops) in order to motivate young professionals

and early stage researchers.



IAEA

Thank you for your attention

email: [email protected]

IAEA

Main issues – RPV integrity

RPV steel - shift of ductile-to-brittle transition temperature

Yield and hardness ↑ due to defects acting as barriers to motion of

dislocations

Thermal ageing: long-term degradation process of mechanical properties

Chemical composition (alloying elements: Ni, Cr, Mn; impurities: P, Cu, S)

More-complicated for multi-component system - complex issue, many

parameters and variables

Emergency situation - Pressurised-Thermal-Shock (PTS)

transient

Small break LOCA: fast-cooling of RPV (as consequence to

the water safety injection-system), Initial pressure decrease,

followed with a re-pressurisation

Consequence: HP combined with low temperature can cause

the brittle fracture of the RPV (PWR critical scenario - weld

near by core)



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(3.1) Support of nuclear energy

Enhanced safety features – inherent and passive safety systems

Physical barriers and

protection in depth

Physical barriers and protection in

depth



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Main issues – RPV integrity

DESIGN LIMITS: • No less than 40 y of service life or

2.10E5 h of operation at normal power

• Up to 30 planned shutdowns

• Coolant working pressure at the core

outlet of 10-16 MPa ,

• Coolant temperature at steady

operation of 250 to 289°C (inlet) and

269 to 324°C (outlet)

• RPV temperature (considering heating

due to radiation) up to 300°C,

• Maximum neutron flux density at the

level of the core center of around

10E11 n/(s cm2) with respect to

neutrons with energy greater than 0.5

MeV (1 MeV).

PART OF RPV VVER 440 (230) VVER 440 (213) VVER 1000 (320)

(EN > 0,5 MEV)

BASE MATERIAL 2,3 X 1024 2,6 X 1024 6,3 X 1023

WELD MATERIAL 1,6 X 1024 1,8 X 1024 5,7 X 1023

(EN > 1 MEV)



PART OF RPV VVER 440 (230) VVER 440 (213) VVER 1000 (320)

(EN > 0,5 MEV)



(EN > 1 MEV)





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Main issues – embrittlement

NOK OK Serious problem for RPV (non-replaceable)

Normal (ductile) fracture occurs by direct

breaking of atomic bonds along the

crystallographic planes

Brittle fracture spreads through the grains

and grain boundaries because grains are

oriented in different directions, crack

changes direction at the grain boundary

DBTT limits RPV operation!




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Main issues – swelling

Issue for fuel cladding and core components

Garner, IAEA Satellite meeting on Cross-cutting issues of structural materials for fusion and fission applications, ICFRM-13, Sapporo (Japan), 10-11 September 2009

4.5% swelling

~ 250% CW

1.7% swelling

1150 °C/5 min (WQ)

1.2% swelling

750 °C/1hr (AC) Void swelling in Fe irradiated in the BR-10

fast reactor at 400°C to 25.8 dpa at 4 x 10-7

dpa/sec

Variations in neutron flux-spectra can affect

property changes via transmutation rates

and dpa rates.

While recognized as important the impact of

these effects has often been strongly

underestimated.

Traditionally, predictive swelling equations

for steels have ignored these effects

Long-accepted formula (AISI304): % swelling = A(T) (dpa)2

Now-accepted version: % swelling = A (dpa rate)-0.731 (dpa)2



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Direct impact on mechanical properties

Effects like: Direct Matrix Damage (dpa),

Precipitation, Segregation, Phase

transformation, etc.

Phase stability and role of alloying elements.

Thermal and mechanical treatment (CR) can

accelerate or reduce such processes (e.g.

impact on distribution and size of grains).

Effect of Flux (dose rate), energy spectra (En

> 0.5 MeV) and temperature.

Higher doses (> 10 dpa) – Transmutation

LWR: RPV ~ 0.1 dpa, RVI (up 10 dpa)

PHASE PHASE

TRANSFORMATION TRANSFORMATION

(High DPA)(High DPA)

DIRECT

MATRIX

DAMAGE

PRECIPITATIONPRECIPITATION

SEGREG

ATION

SEGREG

ATION

(GB)

(GB)

PHASE PHASE

TRANSFORMATION TRANSFORMATION

(High DPA)(High DPA)

DIRECT

MATRIX

DAMAGE

PRECIPITATIONPRECIPITATION

SEGREG

ATION

SEGREG

ATION

(GB)

(GB)

Defects loopDefects loop

NanovoidNanovoid

Cu Cu precipitprecipit..

P segregationP segregation

Defects loopDefects loop

NanovoidNanovoid

Cu Cu precipitprecipit..

P segregationP segregation

Main issues – radiation damage

100200

300400

500

175

200

225

250

275

300

0.0

5.0x1023

1.0x1024

1.5x1024

2.0x1024

2.5x1024

ML

T (

ps)

Flu

en

ce (m

-2)

Depth (nm)

151.0

169.6

188.3

206.9

225.5

244.1

262.8

281.4

300.0

100

200

300

400

500

400425450475500525550

160

165

170

175

180

1 (p

s)

t (°C)

Dep

th (

nm

)

156.0

159.0

162.0

165.0

168.0

171.0

174.0

177.0

180.0

V. Slugen, A.Zeman, J.Lipka, L.Debarberis, NDT&E Int. 37 (2004) 651-661

(1 dpa = all atoms in lattice displaced!)

Luckily, lattice behaves

differently, recovery

mechanisms, however only

under certain conditions!



http://en.wikipedia.org/wiki/File:Backpacking_Tent.jpg

http://en.wikipedia.org/wiki/File:CsCl_crystal.png

IAEA

Main issues – prediction models

ΔT = ΔM + ΔP

Total = Matrix + Precipitation

Mechanical (macroscopic) properties

- consequence of micro-structural

changes

Prediction models - designed for EOL

Safety margins vs. lifetime extensions

Study of microstructuctural

mechanisms - more precise models

Benefits for future innovative reactors

(fission and fusion)


ΔT

ΔM

ΔP

Fluence or doseT

0, T

41J,

Y, H

V e

tc. s

hif

t

ΔT

ΔM

ΔP

Fluence or doseT

0, T

41J,

Y, H

V e

tc. s

hif

t

ΔT

ΔM

ΔP

Fluence or doseT

0, T

41J,

Y, H

V e

tc. s

hif

t

FLUENCE [n/cm2], E

n > 1 MeV

1e+17 1e+18 1e+19 1e+20

T

T [

°C]

50

100

150

200

0.35%0.30%

0.25%0.20%

0.15%0.10%

0.08%

UPPER LIMIT

LOW

ER LIM

IT

%Cu =

0.0

8

%P =

0.0

08



IAEA

International cooperation

Fukushima accident Ref: http://www.gereports.com/setting-the-

record-straight-on-mark-i-containment-history/



Documents

Structural Materials for Advancer Reactor Systems: IAEA ... · Structural Materials for Advancer Reactor Systems: IAEA Support activities ... 29 March 2011 17 . ... RCM held in May