activities performed within the program of nuclear safety

KIT – University of the State of Baden-Württemberg andNational Large-scale Research Center of the Helmholtz Association

Program Nuclear Safety Research, Institute for Materials Research I: Applied Materials Physics, Institute for Materials Research II: Materials and Biomechanics, Institute for Materials Research III: Materials Processing Technology, Institute for Pulsed Power and Microwave Technology

www.kit.edu

Activities performed within the program of Nuclear Safety Research on Structural and Cladding Materials for Innovative Reactor System able to Transmute Nuclear Waste

C. Fazio, M. Rieth, R. Lindau, J. Aktaa, H-C. Schneider, J. Konys, M. Yurechko, G. Müller, A. Weisenburger

Joint EC-IAEA Topical Meeting on “Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems

5-9 October 2009, F4E, Barcelona, Spain

Program Nuclear Safety Research1 25.09.2009 C. Fazio – IAEA-EC Workshop October 2009

Outline The Karlsruhe Institute of Technology

Program Nuclear Safety Research in the Helmholtz Association (HGF) and related materials activities

Advanced fuel cycle studies including P&T

Fast neutron systems and materials issues

The GETMAT project and related structural materials activities performed at KITODS development Join and Weld TechniquesMaterials Qualification in different heat transfer mediaModeling

Summary and outlook


Karlsruhe Institute of Technology: since October 1, 2009

10km

, 15

min

10km

, 15

min

ForschungszentrumKarlsruhe GmbH

12 Programs27 Institutes

3.800 Employees385 Mio. € Budget

UniversitätKarlsruhe (TH)

11 Faculties118 Institutes

4.000 Employees250 Mio. € Budget

18.500 Students

++++

Natural Sciences and Engineering

Excellent overlap and complement great synergies

C E N T R E


Program Nuclear Safety Research in the HGFProgramme TopicsSafety Research for Nuclear Reactors Light water reactors Innovative reactor concepts

Safety Research for Nuclear Waste Disposal Characterisation and immobilisation of high-level waste Reduction of radiotoxicity (P&T) Long-term safety of nuclear waste disposal

Radiation Research Emergency management Radiation research at ISF

Deckgebirge

Wirtsgestein Strecken-lagerung

Bohrlochlagerung

Biosphäre

Overburden

Host rock-

Biosphere


Radiotoxicity of the spent nuclear fuel

Generic objectives of P/T strategies: reduce the burden on a

geological storage in terms of waste mass minimization,reduction of the heat load and of the source of potential radiotoxicity.

Radiotoxicity of 1 ton Spent FuelSeparation of Pu and MA

1E+02

1E+03

1E+04

1E+05

1E+06

1E+07

1E+08

1E+09

1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07

Time after Discharge [years]

Rad

ioto

xici

ty [S

v / t

HM

]

without99,9%Pu99,9%Pu,MANat-U

It is a generally agreed conclusion that fast neutron spectrum systems are more appropriate for transmutation of TRU


Advanced Fuel Cycles including P&T

Pu+MAMulti-recycling

FuelFabrication

DedicatedTransmuter

Reprocessing

FP, Losses at

reprocessingFP, Losses at

reprocessing

Pu+MAMulti-recycling

FuelFabrication

DedicatedTransmuter

Reprocessing

FP, Losses at

reprocessingFP, Losses at

reprocessing

Repository

Reprocessing

UOX-PWR

PuMulti-

recycling

MOX-PWR

Fuel Fabrication

Reprocessing

Pu

PuMulti-

recycling

MOX-PWR

Fuel Fabrication

Reprocessing

Pu

MA

MAPu

FP, Losses atreprocessing

Repository

Reprocessing

UOX, MOXPWR

Multi-recycling

Dedicated Transmuter

ADS, FR

Fuel Fabrication

Reprocessing

Pu+MA

Pu+MA

FP, Losses at

reprocessing

FP, Losses at

reprocessing

Last Transmuter

U

Repository

Reprocessing

UOX, MOXPWR

Multi-recycling


ADS, FR

Fuel Fabrication

Multi-recycling


ADS, FR

Fuel Fabrication

Reprocessing

Pu+MA

Pu+MA

FP, Losses at

reprocessing

FP, Losses at

reprocessing

Last Transmuter

U

« Double strata scenario»: Pu still a resource. Gen-IV FR deployment delayed

Reduction of Pu+MA stockpile (Pu considered as waste)

In both advanced fuel cycle scenarios the dedicated transmuter can be a fast reactor or and Accelerator Driven Subcritical System (ADS)


Austenitic steel AISI 316L for vessel and component materials.

Alternative materials for pump impeller under investigation within other projectsMaxthal, SiSiC, Noriloy

Fe, Al based coating and surface modification (GESA) as corrosion protection barrier for e.g. cladding

9Cr ferritic/martensitic (FM) steel T91 foreseen for core structure and spallation target (ADS) Steels investigated: T91, HT9, Manet, Optifer, HCM12a + 1.4970(15-15Ti)

Innovative reactor systems: reference materials

T91 or AISI316L for HX

EFIT

ELSY



Ref. SMINS, 2007

JSFR



Cut-away view of a proposed 2400 MWth indirect-cycle GFR

re-fuellingequipment

corecontrol and shutdown rod drives

steel reactor pressure vessel

core barrel

main heat exchanger (indirect cycle)

Decay heat removal heat exchanger

Materials:• RPV: F/M steel (e.g. T91)• IHX: Ni-based alloys• Core: SiC/SiC back-up are

considered and for the low power experimental device steel (e.g. austenitic) has been envisaged)Commonalities with V/HTR


Motivation for Materials selection: F/M vs. austenitic steel

7

9

11

13

15

17

19

50 100 150 200 250 300 350 400Temperature, °C

Mea

n Th

erm

al E

xpan

sion

, 10-6

/K

AISI316L

T91

5

10

15

20

25

30

0 100 200 300 400 500 600Temperature, °C

Ther

mal

Con

duct

ivity

( ),

W/m

K

AISI 316L

T91

Data AISI316L from AAA handbook; Data T91 from RCC-MR

• Lower Swelling: impact on dose

However, experience on austenitic steels for the nuclear use is available

• Better thermal properties: impact on e.g. primary system design

higher thermal conductivity

lower thermal expansion


Motivation for materials selection: ODS vs. F/M

Thermal Creep strength

480 500 520 540 560 580 600 620 640 6600

100

200

300

400

500

600

700

1Cr0.5Mo

9Cr ODS2

Alloy 617Ni-23CrCr12Co10Mo

12-14Cr-ODS

105 h

Cre

ep s

treng

th in

MP

aTemperature in °C

9Cr F/M

A. Möslang, et al, FZK

0 5 10 15 20 25

Eurofer-ODS (FZK)0.5% Y2O3Ttest = Tirr = 250°C

irradiated, 15 dpa unirradiated

Strain [%] Somewhat less irrad. hardening Still work hardening almost no

loss of uniform elongation (Au ~7%)

R. Lindau, E. Materna-Morris, A. Möslang, IMF I

Mechanical strength after neutron irradiation


Summary of innovative systems and reference materials

System SFR LFR & ADS GFR

Coolant Na, few bars390-600°C

Lead alloys, few bars(Pb, LBE)

He, 70 bars480-850°C

Core structure

Cladding: ODS(15Cr-15Ni Ti stabilised austenitic steel)Wrapper: 9Cr MS

Cladding:9Cr MS, ODSWrapper: 9Cr MS

ADS target: 9Cr MS350-550°C100dpa+He+H

SiC-SiCf composite or (backup) ODSLow-power dn: steel structure under consideration

Core Inlet/Outlet Temp. 390-750°C 400-480°C 500-1200°C

Max. Dose up to 200dpa 100dpa 60-90dpa

Out of core struct. and others

prim/sec/steam circ.: 9-12Cr MS390-600°CVessel: steel

HX: T91 or 316LVessel: AISI316L

HX: Ni-based alloyvessel: 9-12Cr MS350-500°C<<1dpa


Objective of the GETMAT Project

Cross-cutting items: ODS alloys development and characterisation Joining and welding Materials compatibility with coolants and

corrosion protection barriers Improvement and extension of 9-12 Cr F/M

steels qualification (PIE of relevant ongoing irradiation experiments)

Improved modelling and experimental validation


ODS alloys developed within the GETMAT projectPowder Metallurgy

Casting Process using Hallow Jet Nozzle

• Fe-14Cr 1W 0,3Si 0,3Mn 0,15Ni 0,4Ti + 0.3 Y2O3 Produced by CEA

• T91-ODS (9Cr) produced by FZK

Chemical Composition

• T91-ODS (9Cr) produced by SCK-CEN

C Si Mn P S Cr Ni Mo V Nb Al N

0.08 -012

0.2- 0.5

0.3- 0.6

max0.02

max0.01

8.0- 9.5

max0.4

0.85 -1.05

0.18 -0.2

0.06 -0.1

max0.04

0.03 -0.07

Chemical composition T91 (wt%)


After gas atomisation Powder characterisationby sievingStack of sieveswith different mesh size: 180, 150, 106,75, 45, 20 m

Production of 9 Cr ODS Steel at KIT

Mechanicalalloying under protective gas atmosphere(Ar, N2, H2)

View into milling container

valve

tovacuum pump

ball valve

tubular furnace

Encapsulation of MA powder in HIP canister

Capsule after HIPThermo-mechanical treatment

plate6x350x600

mmR. Lindau, KIT


9Cr ODS martensite Superior processing formability; cold-rolling Uniform structure and isotropic strength, disappearance of strength

anisotropy Good radiation resistance No HT He-embrittlement compared to austenitic SS But, inferior creep strength than higher Cr-ferritic ODS

ODS-Steels for Advanced Fission Reactors

No phase transformation higher operational temperature Superior tensile and creep strength compared to 9Cr ODS Better HT corrosion resistance Good radiation resistance Advantageous reprocessing behaviour But, inferior impact behaviour than lower Cr-ferritic ODS

12 to 20Cr ODS ferritic steel


GETMAT: Investigated Weld and Joining technologies

F/MF/M ODSODS ODSODS

Fusion Welding Solid State Welding• TIG (filler wire)• EB (without filler)

• Diffusion• Explosive• EMP• Friction Stir


EB and TIG of P91: Optimisation of heat treatment

TIGHV - weld

melting zone heat affected zone

base material

I - weld

500 µm

melting zone

heat affected zone

base material

EB

as received 1050C1h/770C2h TIG 770C2h TIG 10501h/7702h EB EB 7702h EB 10501h/7702h

M. Rieth, KIT


from Quadakkers J., Werstoffe und Korrosion, 36, 141-150 (1985)

Corrosion in gas (He), effect of impurities

100 200 300 400 500 600 70010-14

10-12

10-10

10-8

10-6

10-4

10-2

TmaxTmelt

Liquid metal corrosion

CO(Fe3O4)

COS

Oxy

gen

con

cent

ratio

nin

Pb/

Bi [

wt%

]

Temperature [°C]

LBE-Oxidation

Corrosion in HLM, effect of impurities

1

23

1 2 3Carburisation Oxidation

Haynes 230

C. Cabet, CEA A. Weisenburger, G. Mueller, FZK

GETMAT: Compatibility with coolants


GETMAT: Compatibility with coolants

SSRT in SCW condition

Temperature [oC] 550

Pressure [bar] 250

Inlet Conductivity [S.cm-1] < 0.1

Outlet Conductivity S.cm-1] 0.4 – 1.2

Inlet Dissolved O2 [ppb] 100 ppb

Dielectric constant seems to play an important role.

Novotny et al., SMINS 2007, FZK

The issues of Na cooled systems

Evidence of good materials compatibility in Na. See Na cooled systems (e.g. FFTF, Phénix, BOR60, Joyo, etc.)

Innovative features for Na cooled system (e.g. compactness, higher burn- up, etc.) could require demonstration experiments.


Material Compatibility with LBE at different Oxygen Concentrations and Temperatures

30µm

100 200 300 400 500 600 70010-14

10-12

10-10

10-8

10-6

10-4

10-2

TmaxTmelt

Liquid metal corrosion

CO(Fe3O4)

COS

Oxy

den

conc

entr

atio

n in

Pb/

Bi [

wt%

]

Temperature [°C]

LBE-Oxidation

test matrix

austenitic steel F/M steel

oxide

200µm 10000 h

- huge oxidation rate- frequent spallation of

oxides due to growth stress

- reduced heat removal capability 1W/mK)(K

43OM

Severe dissolution ofalloying elements (Ni).dissolution rate up to 1 µm/h

G. Müller, A. Weisenburger, KIT


Summary of long term corrosion experimentsCORRIDA, COSTA, IPPE

0 5000 10000 15000 200000

20

40

60

80

100

IPPE - spinel + magnetite

0.22*t0,5

Spinel (IPPE) SpinelMagn Spinel (COSTA) SpinelMagn Spinel (pressurized tube) SpinelMagn Spinel (velocity exp.) SpinelMagn

scal

e th

ickn

ess

[µm

]

Zeit [h]

0.5*t0,5

T91 / 550°C / 10-6wt% O

21.3(log(t+267)-51.6

IPPE - spinel

CORRIDA metal recesion

T91 550°C PbBi 10-6 wt% O2

Further experimental data (T/ wt% O2): 480/10-6; 500°/10-6 Pb; 450/10-8; 300/10-8



Collaboration with Design team: Impact of oxide scale

XT-ADS heat exchanger (straight tubes)

18% increases in heat transfer surfaceStrong secondary system operating pressure variation, (operation at a power of 70 MW: 30 bar at BOL and 17 bar at EOL; operation at a power of 50 MW: 40 bar at BOL and 28 bar at EOL)

EFIT Steam Generator (Helical configuration)

15% increases in heat transfer surface

ELSY Steam Generator (Spiral configuration)

23% increases in heat transfer surfaceSG in AISI316 and no oxide 28% increase in heat transfer surface respect to T91 without oxide

Heat Exchanger / Steam Generator

0 100 200 300 400 500 600

fissile height <mm BFC>

30

03

50

40

04

50

50

05

50

inner

clad tem

pera

ture

<°C

>

Cladding

Oxide layers thicker than 20 to 30 µm have to be avoided

L. Mansani, Ansaldo D. Struwe, W. Pfrang, KIT


GESA aluminized samples: qualification program

600°C

Corrosion resistance:GESA samples tested for 10000 h in flowing LBE up to 600°C, flow rate 1 m/s and oxygen potential equivalent to 10-6 wt%

High Flow rate resistance:GESA samples tested for 2000 h in flowing LBE at 550°C, up to 3 m/s and oxygen potential equivalent to 10-6 wt%

GESA treated Pressurised tube test:At 550°C with an internal pressure corresponding to a hoop stress up to 200 MPa

GESA treated LCF Test in LBE and airAt 550 °C

Proton irradiation/flowing LBE combined effect: LISOR experiment. Conditions: ~ 2.5 dpa max dose; 300-350°C; 1 m/s flow rate.Dai & Gavillet, to be published

Crosssection of GESAtreated clad

No corrosion attack observed,However control of Al content is relevant

No flow velocity effect: no dissolution attack, no severe oxidation, no erosion. 1 m/s 1,8 m/s 3 m/s

0,7 % Strain No change in the oxidation behaviour

112.5 MPa

No reduction of LCF

No corrosive attack

Al O Thin oxide layer



Creep-to-rupture T91/GESA at 550 °C in air and LBECollaboration with PROMETEY

3 10 100 1000 10000

150

200

250

air 550°C PbBi 550°C gesapbbi550 °C

Stre

ss in

MPa

Time to rupture in h

Reduction of creep strength of T91 in LBE

Reduction of surface energy by LBE and penetration at grain boundaries

Direct contact of “fresh” metal surface with LBE

creep rate in LBE higher compared to aircreep rate in LBE/air depends stressAt low stress (normal operation) LBE effect is reduced - threshold stress!! G. Müller, A. Weisenburger, KIT


Creep-Rupture Tests in Stagnant LeadDesign of new CRISLA Facility for Stagnant Lead Containing Alloys

weight pan

inductiveextenso-meter

inductiveextenso-meter

thermocouplethermocouple

pull rodpull rod

capsulewith a specimeninside

capsulewith a specimeninside

CRISLA: Creep-Rupture Tests inStagnant Lead Alloys

Pt/airoxygensensor

bellows

thermo- couple

specimen

gasinlet

gasoutlet

elongationmeasuring

part

capsule including creep specimen

in lead

M. Yurechko, J. Konys KIT


KArlsruhe SOdium LAboratory

Objectives: technical competence with

respect safety, operation, thermal-hydraulics (e.g.

fuel bundle, HEX) component interaction

(e.g. pool with internals). Qualification of system

analysis tools. Maintenance of sodium know-

how education of PhD students.

Sump tank

Na/Na-HEXM

M

Air HEX

M

MH

D P

ump

3.5m

3.2m

Coresimulator

Emergencypump

R. Stieglitz, KIT


GETMAT: PIE of relevant irradiation programsExperiment/

ReactorMATRIXPhenix

LEXUR IIBOR60

ASTIRBR2

IBIS & SUMOHFR

STIP 4-7SINQ MEGAPIE

Spectra Fast neutrons Fast neutrons Thermal & fast neutrons

Thermal & fast neutrons

High energy protons and

neutrons

High energy protons and

neutrons

Materials

T91, T92, EUROFER

T91,T92 CoatedODS (9-20Cr)

T91, 316T91 , 316 coated

15-15 Ti, ODS

T91,T91 coated,

welds,ODS

T91, Eur-ODST91 coated,

SS316L,welds

9-12 CrODS (9-20Cr)

T91SS 316L

TestsExaminations

TensileImpact

CT, fractog.SEM, TEM

Pressurisedtubes

Tensile, CTCorrosion

NDTPressurised

tubes,CT, tensile,

Charpy, SSRT, SEM, TEM,

EPMA,ICPMS,NRA

TensileKLST

SEM, TEM

TensileBending

Charpy, SPTTEM

Tensile,Bending,

SPT, SIMS, XPS, XRD

SEM, TEM

IrradiationTemperature 390 – 530°C 450 & 550°C 350, 450°C 300, 500°C 300-700°C 250-375°C (T91

beam window)

DoseRange 30-65 dpa Up to 16 dpa ~ 5 dpa 2 dpa 10 – 20 dpa ~7 dpa (T91 beam

window)

Environment Na Pb, inert gas Pb-Bi, inert gas Pb-BiNa (SUMO) Inert gas Pb-Bi


• Infrastructure for handling of radioactive and toxic materials up to 1015 Bq• Highly pure inert gas system for cells and glove boxes

• Mechanical testing: tensile, fatigue, charpy (from -180 up to +1200 °C)• Registering hardness test• Crush load test• Metallography• Mechanical machining• Densitometry• Investigation of tritium absorption and desorption• Light microscopy• Scanning electron microscopy incl. EDX, WDX• Transmission electron microscopy incl. EDX, EELS, EFTEM, HAADF

Hot Laboratory at KIT

H-C Schneider, KIT


• Fully automated impact pendulum

Mechanical tests in Hot - Laboratory

27 mm

• Universal testing machine

27 - 38 mm

• Indentation device Indenter in furnace

H-C Schneider, KIT


Scanning Electron Microscope New 200kV TEM Tecnai G² F20 X-TWIN

02/2009

10/2008

SEM and TEM in Hot Laboratory

H-C Schneider, KIT


GETMAT: Objectives of mutliscale modellingTo develop models describing the microstructure evolution in FeCr alloys under thermal ageing and irradiation.

To correlate the microstructural changes to changes in the mechanical properties of FeCr alloys.

To address the effect of additional elements, such as C or N, on the one hand, and Mo, Nb, V, W, Ta and Mn, on the other

To perform experiments to provide detailed information on the microstructure evolution and its correlation to macroscopic property changes in model alloys.


Summary and outlook

Innovative fast neutron systems are currently studied with the goal to address sustainability and waste management and to enhance competitiveness

The development and qualification of structure materials with improved properties is relevant for the safe operation of those systems

Currently ODS and high Cr F/M steels are under consideration within the EC project GETMAT

The experimental and theoretical activities on structural materials development and characterisation performed at KIT are embedded in EU projects and in the national program and are linked to safety related design assessment

KIT has a wealth of well re-known experts, laboratories and facilities to address materials issues in a wide spectrum

ODS and high Cr F/M steels are as well considered in the Strategic Research Agenda of the SNE-TP and are under consideration within the EERA initiative

activities performed within the program of nuclear safety

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