Neutronics and Nuclear Data for IFMIF - Overview -

Forschungszentrum Karlsruhein der Helmholtz-Gemeinschaft

Association FZK-Euratom

Neutronics and Nuclear Data for IFMIF

- Overview -Ulrich Fischer

Association FZK-Euratom Forschungszentrum Karlsruhe

Y. Chena), S. P. Simakovb), P.P.H. Wilsonc), P. Vladimirovb), F. Wasastjernad)

IAEA TC Meeting, Karlsruhe, October 4-6, 2005

a)ASIPP, Hefei, China; b) Forschungszentrum Karlsruhe, Germany; c)University of Wisconsin-Madison, USA; d) VTT

Processes, Finland



U. Fischer, IFMIF Neutronics & Nuclear Data IAEA TC Meeting, Karlsruhe, October 4-6, 2005

Test Cell

PIE Facilities

Access cell

IFMIF Intense Neutron Source

0 20 40m




IFMIF Intense Neutron Source

• Deuteron beams:– 2 x 125 mA– Ed = 40 MeV

• Neutron production:

1.1 1017 s -1

• Test volumes:– high flux: 0.5 L > 20 dpa/fpy(*)

– medium flux: 6 L > 1 dpa/fpy,– low flux: 7.5 L 0.1-1 dpa/fpy

Beam Spot(20x5cm2)

High Flux

Low FluxMedium Flux

LiquidLi Jet

DeuteronBeam

(*)fpy = full power year




IFMIF neutron flux spectrum

0,01 0,1 1 10 10010-1

100

101

102

103

104

105

106

10-1

100

101

102

103

104

105

106

ITER first wall HFTM (McDeLi) HFTM (McDeLicious)

N

eutr

on fl

ux d

ensi

ty [

1010/c

m2 /M

eV/s

]

Neutron energy [MeV]




Prove IFMIF’s suitability as neutron source for fusion-specific simulation irradiations

Provide reliable data for the technical layout of facility

• Computational tools and data required– D-Li neutron source term simulation– Neutron transport (En> 20 MeV) – Activation and transmutation (En> 20 MeV)

Key role of neutronics in establishing IFMIF as neutron source for fusion material testing

IFMIF Neutronics Issues




• D-Li source neutrons for MC transport simulation – D-Li interaction processes– Neutron yield & spectra (energy, angle)– To be integrated to standard MC code (MCNP)

• MCNP McDeLi (P. P. H. Wilson, 1998) – Semi-empirical reaction model with adjustable parameters– Deuteron beam configuration, orientation and profile

• McDeLi McDeLicious (S. P. Simakov, 2001)– Use of evaluated d+ 6,7Li data to sample source neutrons– d + 6,7Li cross-sections for deuteron energies up to 50 MeV (A.

Konobeyev et al, NSE 139 (2001)) Improved new evaluation 2005 (P. Pereslavtsev, FZK)

D-Li Neutron Source Term




Thick Li-target neutron yields

0 10 20 30 40

100

101

102

Deuteron Energy [MeV]

MCNPX

McDeLicious

McDeLi

Total Yield ( = 4)

Lone Johnson Sugimoto Mann Baba 2002

Neut

ron

Yiel

d,

1010

/C

0 10 20 30 4010-1

100

101 MCNPX

McDeLicious

Forward Yield ( = 0o)

McDeLi

- Daruga - Weaver - Goland - Amols - Nelson - Lone - Salmarsh - Johnson - Sugimoto - Bem 2002 - Baba 2002

Yiel

d,

1010

/sr/

C


Total (4π) Neutron Yields Forward (0o) Neutron Yields




10-5

10-4

10-3

10-2

10-1

100

0 10 20 30 40

=0o

Neu

tron

Yie

ld

[1010

n/s

r/M

eV/C

]

Baba et al. 2002 McDeLicious MCNPX

0 10 20 30 40

Ed = 25 MeV

=20o =10o

0 10 20 30 40


0 10 20 30 40 50

=90o

0 10 20 30 40 50

0 10 20 30 4010-6

10-5

10-4

10-3

10-2

10-1

=60o =40o

D-Li thick target neutron yield spectra

Ed= 25 MeV




10-4

10-3

10-2

10-1

100

1010 10 20 30 40 50

=20o

=0o

Neu

tron

Yie

ld

[1010

n/s

r/M

eV/C

]

Baba et al. McDeLicious MCNPX

=10o

0 10 20 30 40 50

Ed = 40 MeV

0 10 20 30 40 50 60

=45o

0 10 20 30 40 5010-5

10-4

10-3

10-2

10-1 =60o

0 10 20 30 40 50


0 10 20 30 40 50 60

=90o

Ed= 40 MeV

D-Li thick target neutron yield spectra




Li(d,xn) Double Differential Cross Sections

0 5 10 15 20 25 30 35

10-1

100

101

102

Bem et al.

MCDeLicious

Li(d,xn), Ed = 16.6 MeV

= 0o

d2 /d

/dE

, m

b/sr

/MeV

Neutron Energy, MeV

0 10 20 30 40 50 6010-1

100

101

102

= 0o

Li(d,xn), Ed = 40 MeV

MCDeLicious

Neutron Energy, MeV

d2

/d/

dE,

mb/

sr/M

eV

Baba et al.

Ed = 17 MeV, Bem et al.

Ed = 40 MeV, Baba et al.




New d + Li Data Evaluation

P. Pereslavtsev, this meeting

0 5 10 15 20 25 30 3510-3

10-2

10-1

100

101

102

Bem, 03 Bem, 03 evaporation prequilibrium stripping 1 level 2 level 3 level 4 level 5 level total

7Li(d,xn), Ed=16.6 MeV

d2

/d/

dE [

mb/

sr M

eV]

Emitted neutron energy [MeV]




Thick Li-target neutron yields using 2005 D-Li data evaluation (P. Pereslavstev et al.)

0 10 20 30 4010-1

100

101

2001

MCNPX

McDeLicious

Forward Yield ( = 0o)

McDeLi

- Daruga 68 - Weaver 72 - Goland 75 - Amols 76 - Nelson 77 - Lone 77 - Salmarsh 77 - Johnson 79 - Sugimoto 95 - Bem 02 - Baba 01/02

Neu

tron

Yie

ld, 10

10/s

r/C


2005




IFMIF Test Cell Calculations

• Major neutronics task in this context:– Provide the data required for the design and optimisation of

the irradiation test modules and the lay-out of the test cell

Neutron/photon transport calculations (McDeLicious) for flux distributions and nuclear responses such as nuclear heating, radiation damage accumulation and gas production.

• IFMIF’ s primary mission is to generate a materials irradiation database for the design, construction, licensing and operation of DEMO




Materials for IFMIF

• Highest priority: structural materials of the reduced activation ferritic-martensitic (RAFM) type (Eurofer, F82H). A variety of Eurofer specimens will be irradiated in the high flux

test module (HFTM) up to the target fluence of 150 dpa.

• Other materials of (possibly) lower priority:– SiC, V/V-alloy, divertor materials (e. g. W)– Breeder materials, neutron multiplier– Ceramic insulators and others




Element Specification [w%]

Element Specification [w%]

C 0.090-0120 W 1.0-1.2 Mn 0.20-0.60 Ti <0.01 P <0.005 Cu <0.005 S <0.005 Nb <0.001 Si <0.05 Al <0.01 Ni <0.005 N 0.015-0.045 Cr 8.50-9.50 B <0.001 Mo <0.005 Co <0.005 V 0.15-0.25 O <0.01 Ta 0.05-0.09 Fe balance

Chemical Composition of RAFM Steel Eurofer




• IFMIF project (INPE Obninsk/FZK)- 1H, 56Fe, 23Na, 39K, 28Si, 12C, 52Cr, 51V (50 MeV)

- 6,7Li, 9Be (150 MeV)

• LANL 150 MeV data files (ENDF/B-VI.6)- 1,2H, 12C, 16O, 14N, 27Al, 28,29,30Si, 31P, 40Ca, 50,52,53,54Cr, 54,56,57,58Fe, 58,60,61,62,64Ni, 63,65Cu, 93Nb, 182,183,184,186 W, 196,198, 199, 200, 201, 202, 204 Hg, 206, 20, 208 Pb, 209Bi

• NRG evaluations– 40,42-44,46,48Ca- 45Sc, 46-50Ti, 54,56-,58Fe, 70,72-74,76Ge, 204,206-208Pb, 209Bi

• JENDL-HE data file– 1H, 12,13C, 14N, 16O, 24-26Mg, 27Al, 28-30Si, 39,41K, 40,42- 46,48Ca, 46-50Ti,51V, 50,52-54Cr, 55Mn, 54,56-58Fe, 59Co, 58,60-62,64Ni, 63,65Cu, 64,66-68,70Zn,90- 92,94,96Zr,

93Nb, 180,182-184,186W, 196,198-202,204Hg

Neutron Cross-Sections E 20 MeV - General purpose data ENDF evaluations -

T. Fukahori, this meeting




Iron Transmission Benchmark Experiment (Shin et al.)

0 10 20 30 40 50 6010-14

10-13

10-12

10-11

10-10

Shin et al. 91 MCNP/INPE-FZK MCNP/LANL-150

t = 50 cm

t = 40 cm

Neu

tron

Flu

ence

, n/

cm2 /M

eV/p

roto

n

Neutron Energy, MeV

t = 20 cm

10 20 30 40 501

2

3

4

5

Cierjacks 66 Perey 72 Larson 81 Abfalterer 2001 FZK/INPE LANL-150

To

tal n

eutr

on

cro

ss-s

ecti

on

[b

]

Neutron Energy, MeV

Fe slabs (20, 40 50 cm) irradiated with source neutrons produced by 65 MeV protons on Cu target

Fe(nat) total neutron cross-section




Li target

Irradiation test modules

IFMIF Test Cell

MCNP model (visualised by

CATIA)

CAD model (CATIA)

Accelerator lines




HFTM container

IFMIF High Flux Test Module (HFTM)

HFTM test rig

vertical cut horizontal cut




Distributions of nuclear responses in the HFTM

Displacement damage rate [dpa / fpy]

50

40

30

20

10

50

40

30

20

10

50

40

30

20

10

7 6 5 4 3 2 1 1Z

0

10

20

30

40

50

60

0 1 2 3 4 5 6 7 8 9 10 11 12

1

2

3

4

52-Beams Footprint 20 x 5 cm2

dpa-rate [1/fpy]Y

Y

X

Z

d-Beams

Beam Footprint

7

6

5

4

3

2

1

X

Z

2420

16

12

8

24

20

16

12

8 4

24

20

16

12

8

7 6 5 4 3 2 1 1

2-Beams, Footprint 20 x 5 cm2

Z

0 1 2 3 4 5 6 7 8 9 10 11 12

1

2

3

4

5Nuclear Heating [W/cm3]

Y

Y

X

Z

d-Beams

Beam Footprint

7

6

5

4

3

2

1

X

Z

Nuclear heating [ W/ cm3]

McDeLicious calculations with simplified geometry model

S. P. Simakov, this meeting




-4

-3

-2

-1

0

1

2

3

4

0 10 20 30 40 50

dpa/fpy

Ver

tical

Dir

ectio

n, c

m

Fe displacement damage

-4

-3

-2

-1

0

1

2

3

4

0 1 2 3

W/g

Ver

tical

Dir

ectio

n, c

m

Nuclear heating

d - beam

Li – jet

HF

TM

(rig matrix)

HFTM

Distributions of nuclear responses in HFTM test rigs

McDeLicious calculations with detailed 3D geometry model

Nuclear heating

Displacement damage




IFMIF Medium Flux Test Module (MFTM)

Creep- Fatigue Test Module (CFTM)Tritium Release

Module (TRM)




MFTM: MCNP geometry model (horizontal cut)

HFTMLi target

Be rigs(TRM)

W spectral shifter plate

Creep-fatigue testing

machine

LFTM

C neutron reflector




0 200 400 600 800 1000 1200 1400 16000

5000

10000

15000

20000

25000

30000

IFMIF: TRM with W

IFMIF: TRM upstream shifted

IFMIF: TRM no W

HCPB (PPCS)

BOR-60

HFR

1 fpy 20.000 h = 2.3 y 40.000 h = 4.6 y

He

prod

uctio

n [a

ppm

]

H-3 production [appm]

He vs. H-3 production rates in Be rigs of MFTM




Nuclear Heating in Pulling Rods of CFTM

-10.0 -9.5 -9.0 -8.5 -8.0 -7.5 -7.0 -6.5 -6.0

2.5

3.0

3.5

4.0

4.5

5.0

-2.0 -1.5 -1.0 -0.5 0.0 0.5 1.0 1.5 2.0

2.5

3.0

3.5

4.0

4.5

5.0

6.0 6.5 7.0 7.5 8.0 8.5 9.0 9.5 10.0

2.5

3.0

3.5

4.0

4.5

5.0

-10.0 -9.5 -9.0 -8.5 -8.0 -7.5 -7.0 -6.5 -6.0-5.0

-4.5

-4.0

-3.5

-3.0

-2.5

-2.0 -1.5 -1.0 -0.5 0.0 0.5 1.0 1.5 2.0-5.0

-4.5

-4.0

-3.5

-3.0

-2.5

6.0 6.5 7.0 7.5 8.0 8.5 9.0 9.5 10.0-5.0

-4.5

-4.0

-3.5

-3.0

-2.5

W/g

-10.0 -9.5 -9.0 -8.5 -8.0 -7.5 -7.0 -6.5 -6.0

-4.5

-4.0

-3.5

-3.0

-2.5

-2.0

0

0.125

0.250

0.375

0.500

0.625

0.750

0.875

1.000

X, cm

Y,

cm

P. Vladimirov, this meeting




• Intermediate Energy Activation File IEAF-2001 (FZK/INPE)

– Complete cross-section data library for activation and transmutation analyses up to En 150 MeV (1 Z 84)

– Validated through series of benchmark calculations, tested and qualified for SS-316 & V/V-alloy samples in IFMIF activation experiment

• ALARA activation code (P. Wilson, UW) – Analytical and Laplacian Adaptive Radioactivity Analysis– Capable of handling an arbitrary number of reaction channels

EAF-2005 (En 60 MeV) for FISPACT inventory calculations recently became available (UKAEA Culham)

Activation and Transmutation Analyses

Tools and Data

R. Forrest, this meeting

U. Fischer, this meeting




IEAF-2001 Validation AnalysesActivation experiments in IFMIF-like white neutron field (U. v. Möllendorff et al., FZK; P. Bem et al. NPI Rez)

Hf-

179n

Hf-

180m

Hf-

181

Hf-

182m

Hf-

183

Ta-

182

Ta-

183

Ta-

184

Ta-

185

W-1

81

W-1

85

W-1

87

10-1

100

101

102

103

F82H (d-Li)EF-97

W#14 W#4

W#7

C/E

W Activation Products

Tungsten

Calculated (C) vs. measured (E) -radioactivities

Na

24Sc

46

Sc 4

7Sc

48

V 4

8C

r 48

Cr

51M

n 52

Mn

54M

n 56

Fe 5

2Fe

59

Co

56C

o 57

Co

58C

o 60

Ni 5

7H

f 18

0mT

a 17

7T

a 18

0T

a 18

2T

a 18

3T

a 18

4W

18710-3

10-2

10-1

100

101

FZK (d-Li): V-alloy FZK (d-Li): F82H

C/E

FZK (d-Li): SS-316

NPI (p-D2O): EF-97

Eurofer-97,SS-316, F82H, V alloy (FZK)

S. P. Simakov, this meeting




10-3 10-2 10-1 100 101 102 103 104 105 10610-6

10-4

10-2

100

102

104

106

W

LFTM

UTM

106ySS-316 & W, 1 year in IFMIF

TRM

BP

Hands-on Limit

Recycling Limit

104y100y1y30d1d

HFTM

Front Wall Liner

Con

tact

-D

ose

Rat

e,

Sv/h

Time after shutdown, years

Induced radioactivity in the IFMIF HFTM components

Contact - dose rate after IFMIF full power irradiation

[Sv/h]

IFMIF HTFM components (CAD model)




Elemental Transmutation of Eurofer in HFTM

B C N O Al Si P S Ti V Cr

Mn Fe Co Ni

Cu

Nb

Mo Ta W

-20

-10

0

10

20

30

Burn-up

Generation

Tra

nsm

utat

ion

rate

, %

/fpy

- IFMIF/HFTM - FPR/HCLL - FPR/HCPB

• Assessed by McDeLicious transport and ALARA/IEAF-2001 inventory calculations




He production cross-section of Fe-nat up to 100 MeV

B. Haight, this meeting




• Dedicated approach for coupled 3D Monte Carlo and deterministic (SN) transport calculations for IFMIF shielding analyses

• Monte Carlo technique used to simulate the particle generation and transport in and around the neutron source region

• Discrete ordinates (SN) method used to treat the deep penetration problem in bulk shield.

• Mapping approach for calculating SN angular fluxes from scored data of Monte Carlo particle tracks crossing a specified surface

• Implemented in interface programme linking MCNP/McDeLicious and 3D SN code TORT of DOORS3.2 package

3-D shielding analyses of IFMIF




3-D shielding calculations for IFMIF Test Cell

Dose rate distribution in access room

Geometry model




Conclusions

• Suitable computational tools, data and models available for IFMIF neutronics and activation analyses – McDeLicious Monte Carlo code for Li(d,xn) neutron source

generation and transport simulation.– MC transport simulation calculations (e. g. test cell):

• Various general purpose intermediate/high energy data evaluations

– Activation and transmutation calculations:• ALARA inventory code with IEAF-2001 data• FISPACT code with EAF-2005

– Coupled Monte Carlo – Discrete Ordinates (SN) computational scheme for full 3D shielding calculations

– Detailed 3D geometry model of the entire IFMIF test cell.

IFMIF neutronics and nuclear data: significant progress




• D-Li neutron source term– d + 6.7Li evaluation to be validated

– Further improvements required ?

• General purpose (ENDF) data evaluations E> 20 MeV– Need for full IFMIF nuclear data library– Benchmark experiments/analyses required– Cross-section measurements

• Activation/transmutation/gas production data– Need for validation ( Benchmark experiments)

– Need for cross-section measurements

– Deuteron induced activation cross-sections

Outlook

Neutronics and Nuclear Data for IFMIF - Overview -

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