Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft Association FZK-Euratom Status of...

Forschungszentrum Karlsruhein der Helmholtz-Gemeinschaft

Association FZK-Euratom

Status of Neutronics Tools & Data for IFMIF-EVEDA

U. Fischer, S. SimakovAssociation FZK-Euratom

Forschungszentrum KarlsruheGermany

Design Review Meeting IFMIF-EVEDA Test Facilities, Forschungszentrum Karlsruhe, March 26-27, 2007



U. Fischer, S. Simakov, Status Neutronics Tools & Data IFMIF Test Facility Review Meeting, FZK, March 26-27, 2007

Dedicated computational tools and data required:

– D-Li neutron source term simulation

– Neutron/photon transport calculations (En> 20 MeV)

– Activation and transmutation calculations (En> 20 MeV)

– Shielding calculations

IFMIF Neutronics




D-Li Neutron Source Term




• Enhancement to MCNP with ability to sample the generation of d-Li source neutrons on the basis of tabulated double-differential d + 6,7Li cross-sections - Modelling of deuteron beam configuration, orientation and

profile- 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)) - Updated evaluation (new methodology, P. Pereslavtsev, 2005)- McDeLicious standard code for IFMIF neutronics calculations

• Monte Carlo Code McDeLicious Developed to enable a proper representation of the d-Li neutron source term in Monte Carlo transport calculations for IFMIF

D-Li Neutron Source Term




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]

New d + 6,7Li Data EvaluationP. Pereslavtsev et al,

2005

0 10 20 30 40 5010-3

10-2

10-1

100

101

102

Baba, 03 evaporation preequilibrium stripping 1 level 2 level 3 level 4 level 5 level total

Li(d,xn), E

d=40 MeV

d2/

d/d

E [

mb/

sr M

eV]

Emitted neutron energy [MeV]

Neutron emission spectra in forward direction




McDeLicious validation analyses: Thick Li-target neutron forward yields as function of deuteron incidence energy

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

Deuteron Energy [MeV]

2005




Neutron Cross-Section Data (E> 20 MeV)

- General Purpose Data Files -




• IFMIF project – 1H, 56Fe, 23Na, 39K, 28Si, 12C, 52Cr, 51V, 6,7Li, 9Be (INPE/FZK)- 180,182-184,186W, 181Ta (EFF/JEFF)

• 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 • Few other evaluations

– KAERI (12C, 27Al, 56Fe, 208Pb), BNL (12C, 56Fe,208Pb, 209Bi), IPPE (232-238U, 237,239Np, 236-244Pu)

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




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




Activation and Transmutation - Tools & Data -




• 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)– Uses lumped MT=5 reaction cross-sections with production yield vectors (LAW=0) on MF=6 file– Validated through series of benchmark calculations, tested and qualified for SS-316 & V/V-alloy samples in

IFMIF activation experiments

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

Activation and Transmutation- Tools and Data -




• European Activation File EAF-2005 (UKAEA)– EAF data library extended to 60 MeV (based on TALYS + SAFEPAQ)– EAF data format using non-standard MT-numbers– Can be used with FISPACT-2005 inventory code (EASY system, UKAEA)– 62,637 reactions from 50 data sources, 86 different reaction types– Extended decay data library with 2,192 nuclides

• EAF-2007: deuteron induced reactions– 66,864 reactions from different sources, mostly TALYS (– New calculated data library produced by TALYS for data using global

parameters, included long-lived isomers as targets– New calculations using TALYS with local parameters produced by M.

Avrigeanu for 27Al, 54,56,57,58Fe and 63,65Cu (<50 MeV)

• EAF-2007: proton induced reactions– 67,925 reactions, only based on data from TALYS calculations (so far)

FISPACT-2007 modified to handle d and p as incoming particle




NPI Rez Activation Experiments

MCNPX model

D2O Target (Ø30×26 mm2)

Activation Foils

1.7 mm 47 mm ≈ 700 cm

NeutronDetector

24

6

10

12

0 5 10 15 20 25-5-4-3-2-1012345

Ep = 36.5 MeV, p-Flux [1/cm2]

D2O -depth/p-beam direction, mm

Hor

izon

tal d

irec

ton,

mm

37 MeV protons

1 10 4010-2

10-1

100

Spectrum used Activation Caclution

D2O(p,xn), E

p = 37 MeV

Rez Experiment

Neutron Energy, MeV

MCNPX/LA-150

= 0o

Spe

ctra

l Yie

ld,

1010

n/s

r/M

eV/

C

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

EAF-2005

IEAF-2001

C/E

W Activation ProductsC/E ratios of –radio-activities induced in W

p-D2O neutron spectrum




Neutron Spectrum Unfolding Technique

- 37 MeV @ 12 μA proton beam- Heavy water target- Dosimetry foils:

Al, Ti, Fe, Co, Ni, Y, Nb, Lu, Au (Ø12 x 0.1-0.3 mm)

- 10-12 hours irradiation- γ-activities measured by HPGe

detector at cooling 1 h to 100 day;- Guess spectrum: measured by

NE213 detector and simulated by MCNPX.

0 5 10 15 20 25 30

0.5

1.0

1.5

2.0

2.5

24Na196Au

196m2Au

195Au

194Au

198Au 92mNb

90mY

88Y

90mY

57Co

58Co59Fe56M

n

60Co

46Sc48Sc

47Ca57Ni57Co

58Co 60Co

Lu173

Lu174

Lu174m

Lu176m

56Mn

54Mn

51Cr

169 Tm,209Bi(n,3-6n)

115 In,93Nb,103 Rh(n,n')

108

109

1010

2x1010

5th-iteration (final)

Act

ivit

y(C

al)/

Act

ivit

y(E

xp)

Neutron Energy, MeV

D2O(p,xn), E

p = 37 MeV @ 12 A

Guess (initial)

Spe

ctra

l Flu

x, n

/cm

2 /s/M

eV

• Use of modified SAND-II code with cross sections from EAF-2005

• Tested at the NPI cyclotron in neutron spectrum extending up to 35 MeV.




0 20 40 60 80 100 120 1400

200

400

600

800

1000

1200

Fe(n,xp)

LANSCE, 05 EAF-2005 IEAF-2001 GNASH JEFF-3.1

C

ross

sec

tion

[mb]

Neutron energy [MeV]

0 20 40 60 80 100 120 1400

20

40

60

80

100

120

140

160

180

200

Fe(n,x)

LANSCE, 05 EAF-2005 IEAF-2001 GNASH JEFF-3.1

Cro

ss s

ect

ion

[m

b]


0 20 40 60 80 100 120 1400

20

40

60

80

100

120

181Ta(n,x)

LANSCE, 05 EAF-2005 IEAF-2001 Konobeyev GNASH

Cro

ss s

ect

ion

[m

b]


0 20 40 60 80 100 120 1400

200

400

600

800

1000

1200

Ta(n,xp)

LANSCE, 05 EAF-2005 IEAF-2001 GNASH

Cro

ss s

ectio

n [m

b]


Gas production cross-sections




Radiation Shielding




• 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

Radiation Shielding




3-D shielding calculations for IFMIF Test Cell

Dose rate distribution in access room

Geometry model




McCad Model Generation




CAD Interface Programme McCAD

• Automatic conversion of CAD geometry data into semi-algebraic geometry representation for the MCNP Monte Carlo code

- CAD data are imported via data files in neutral format (IGES, STEP)

- Graphical User Interface (GUI) with components for modelling, visualization, and data exchange

- McCAD is able to perform the conversion of the CAD data into MCNP geometry including the generation of voids and error analyses and repair (generation of full MCNP input deck)

- McCAD can also generate CAD geometry models from available input decks (e. g. for visualization)

- Successful test applications to JET and ITER (benchmarking)




Test Cell Model Converted by McCAD

2D cuts generated by MCNP geometry plotter

Vertical cut of test cell Horizontal cut at target mid-plane level




Summary• Codes for IFMIF neutronics

– Transport: McDeLicious, MC/TORT, (MCNPX) – Geometry preparation: CAD/McCAD– Activation/transmutation: ALARA, FISPACT Need to be further developed, in particular: McDeLicious,

MC/TORT, McCAD

• Nuclear Data for IFMIF neutronics– Transport calculations: different sources, urgent need for qualified

IFMIF reference data library (IAEA) – Activation/transmutation: IEAF-2001, EAF-2005/07 (n, p, d), to be

further developed, improved and validated– Validation experiments, to be extended/continued– To be newly developed: dosimetry data file (extended IRDF)

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