| U. Fischer et al | FZK Contributions to FENDL-3 CRP KIT - Die Kooperation von Forschungszentrum Karlsruhe GmbH und Universität Karlsruhe (TH) IAEA CRP

| U. Fischer et al | FZK Contributions to FENDL-3 CRP

KIT - Die Kooperation von

Forschungszentrum Karlsruhe GmbH

und Universität Karlsruhe (TH)

IAEA CRP on “Nuclear Data Libraries for Advanced Systems: Fusion Devices (FENDL-3)”- Contributions by FZKU. Fischer, A. Konobeyev, P. Pereslavtsev, S. P. Simakov

Association FZK-EuratomKarlsruhe Institute of Technology Forschungszentrum Karlsruhe

1st Research Coordination Meeting on “Nuclear Data Libraries for Advanced Systems – Fusion Devices (FENDL)”, IAEA, Vienna, December 2-5, 2008





Overview

• Recent FZK activities related to fusion and high energy nuclear data

• Proposed CRP contributions by FZK

Data evaluationsData files Validation analyses





Nuclear Data Evaluations for Fusion Technology and High Energy Applications

• Objectives:To provide cross-section data evaluations required for fusion neutronics calculations (transport and transmutation/activation)

• Evaluation of neutron induced cross-sections for EFF/JEFF (Joint Evaluated Fusion Fission File) up to 150 MeV neutron energy– Production of complete data files (ENDF6 data format) for processing

with NJOY (use with MC & SN codes)

– Suitable for fusion and other applications

– Complete data files for 182,183,184,186W, 181Ta, 55Mn, 52Cr evaluation

– Largely with co-variance data

– Contributions to FENDL-3 project co-ordinated by IAEA/NDS• Evaluation of d + 6,7Li data up to 50 MeV for IFMIF D-Li source term

simulation using McDeLicious MC code





ENDF/B, JEFF-3.1, EAF2007

Experiment

Evaluated file

RIPL-2RIPL-2

ECISECIS

Transmission coefficients

el, non, tot

Nuclear model parameters Mass table

Nuclear level structure

INPUT files

Collective excitations

Flowchart of the evaluation procedure





0 20 40 60 80 100 120 1400

100

200

300

400

500

600

700

800

900

1000

1100

1200

Haight, 07 ENDF/B-VII EAF-2007 present result

Cr(n,xp)

Cro

ss s

ectio

n [m

b]

Neutron energy [MeV]

0 20 40 60 80 100 120 1400

50

100

150

200

250

Haight, 07 ENDF/B-VII EAF-2007 present result

Cr(n,x)

Cro

ss s

ectio

n [m

b]


Evaluation of n + 52Cr cross-section data up to 150 MeV for the Joint European Fusion Fission File (JEFF)

10-2 10-1 100 101 102

102

103

104

105

Carlton, 2000 Agrawal, 84 Foster, 71 Abfalterer, 01 ENDF/B-VII present result

Total cross section

Cro

ss s

ectio

n [m

barn

]


0 2 4 6 8 10 12 14

101

102

103

Takahashi, 88 Yabuta, 88 Matsuyama, 92 JEFF-3.1 present result

14 MeV 52Cr(n,xn)

ds/

dE

[m

b/M

eV

]

Neutron emission energy [MeV]

0 20 40 60 80 100 120 140 160 180100

101

102

103

Olsson, 87 ENDF/B-VII present work

n + 52Cr Elastic scatteringE

n= 21.6 MeV

d/d

[mb

/sr]

c.m.

[deg]

0 2 4 6 8 10 12 14 16 18 20 22 2410-4

10-3

10-2

10-1

100

101

102

103

Oblozinsky, 92 JEFF-3.1=ENDF/B-VII present result

14.6 MeV 52Cr(n,xg)

ds/

dE

d [

mb

/Me

V s

r]


=100o

Photon emission energy [MeV]





Evaluation of n + 181Ta cross-section data up to 150 MeV for the Joint European Fusion Fission File (JEFF)

0 2 4 6 8 10 12 14100

101

102

103

104

Vonach, 80 Takahashi, 89 Takahashi, 92 Hermsdorf, 82 ENDF/B-VI JENDL-3.3 GNASH

14 MeV 181Ta(n,xn)

ds/d

E [

mb/

MeV

]


0 2 4 6 8 10 12 14 16 18 20100

101

102

103

104

Marcinkowski, 93 REX ENDF/B-VI JENDL-3.3 GNASH

20 MeV 181Ta(n,xn)

ds/d

E [

mb/

MeV

]


6 8 10 12 14 16 18 20 22 24 26 28 300

500

1000

1500

2000

2500

Veeser, 77 Frehaut, 80 Takahashi, 89 ENDF/B-VI JENDL-3.3 JEFF-3.0 GNASH

181Ta(n,2n)180Ta

Cro

ss s

ectio

n [m

b]


8 10 12 14 16 18 20 22 24 26 28 300

2

4

6

8

10

12

14

16

18

20

22

Woelfle, 88 Xiangzhong, 92 Kasugai, 92 Han-Lin, 89 Begun, 02 Filatenkov, 01 ENDF/B-VI JENDL-3.3 JEFF-3.0 GNASH evaluation

181Ta(n,p)181Hf

Cro

ss s

ectio

n [m

b]

Neutron energy [MeV]0 20 40 60 80 100 120 140 160 180

100

101

102

103

104

Hansen, 85 Benenson, 73 Cross, 60 Hudson, 62 ENDF/B-VI JENDL-3.3 evaluation

n + 181Ta Elastic scatteringE

n= 15 MeV

d/d

[mb

/sr]

c.m.

[deg]

10-1 100 101 102

2

4

6

8

10 Total cross section

Finlay, 93 Poenitz, 81 Tsubone, 84 Byoun, 73 Poenitz, 83 Foster, 71 ENDF/B-VI JENDL-3.3 Koning Young

Cro

ss s

ectio

n [b

arn]






Evaluation of n + 55Mn cross-section data up to 150 MeV for the Joint European Fusion Fission File (JEFF)

0 20 40 60 80 100 120 140 160 180100

101

102

103

Schweitzer, 78 Becker, 66 Holmqvist, 70 JENDL-HE ECIS evaluation

n + 55Mn Elastic scatteringE

n= 3.4 MeV

d/

d

[mb

/sr]

c.m.

[deg]

10-3 10-2 10-1 100 101 102

100

101

102

Abfalterer, 01 Pineo, 70 Foster, 71 Cierjacks, 69 Garg, 78 JENDL-HE ECIS evaluation

Total cross section

Cro

ss s

ectio

n [b

arn]


0 2 4 6 8 10 12 14100

101

102

103

104

Takahashi, 89 Baba, 88 Hermsdorf, 82 JENDL-HE evaluation

14 MeV 55Mn(n,xn)

ds/d

E [

mb/

MeV

]


0 2 4 6 8 10 12 1410-1

100

101

102

Salnikov, 72 Takahashi, 89 Baba, 88 evaluation

En=14 MeV

d/d

Ed

[m

b/sr

MeV

]

Emitted neutron CM energy [MeV]

10 15 20 25 30 35 400

500

1000

1500

Paulsen, 65 Menlove, 67 Ikeda, 88 Filatenkov, 99 Lu, 89 Greenwood, 87 Borman, 69 Auchampaugh, 77 Soewarsono, 92 Iwamino, 92 EAF-2005 JENDL-HE evaluation

55Mn(n,2n)54Mn

Cro

ss s

ectio

n [m

b]


0 10 20 30 40100

101

102

Qaim, 94 Prasad, 71 Minetti, 67 Allan, 61 Weigold, 60 Mitra, 66 EAF-2005 (EFF-2.4) JENDL-HE evaluation

55Mn(n,p)55Cr

Cro

ss s

ectio

n [m

b]






Typical content of ENDF data files

MF=1 General description. MT=451

MF=2 Resonance parameters. MT=151

MF=3 MT=1, 2, 4, 5, 16, 22, 28, 51-60, 91, 102,103, 104, 105, 106, 107

MF=4 Angular distributions. MT=2, 51-60 (Legendre polynomials)

MF=6Energy-angular distributions. MT=16,22,28,91, 103-107 (below 20 MeV), MT=5 (above 20 MeV): particles, photons, recoils

MF=12Photon production multiplicities and transmission probabilities. MT=16,22,28,51-60,91,102-107

MF=14 Photon angular distributions. MT=16,22,28,51-60,91,102-107

MF=15 Photon energy spectra. MT=16,22,28,91,102-107





Benchmark analyses

• Objectives:To check and validate the new data evaluations for a reliable use in application calculations (ITER TBM, Demo, IFMIF) and provide feed-back to the evaluators.

- Computational analyses of integral benchmark experiments (14 MeV neutrons) by means of Monte Carlo calculations with the MCNP code

- Comparisons for neutron and photon spectra and other measured responses (e. g. reaction rates)

- Cross-checks with calculations using data evaluations from different sources (JEFF-3.0/3.1, FENDL-2.0/2.1, JENDL-3.3, ENDF/B-VI/VII

- Checks of neutron emission cross-section data





Benchmark analyses on Ta

• Integral benchmark experiments

– Lawrence Livermore National Laboratory (LLNL) pulsed sphere experiments with central 14 MeV neutron source: Time-of-flight spectra for spherical Ta shells (wall thicknesses of 3.4 and 10.2 cm).

– Lewis Research Centre (LRC) experiment on 9 cm thick Ta shell with central Am-Be neutron source: Neutron leakage spectrum measurement by proton recoil scintillation detector.

• Differential experimental data

– Few measurements of secondary neutron energy/angular distributions for incident energies 5 to 20 MeV by IPPE (Obninsk), IRK (Vienna), OSA (Osaka), TOH (Tohoku), TUD (Dresden)

• Available data evaluations

– ENDF/B-V (VI, VII), JENDL-3.3 (FENDL-2.1, JEFF-3.1), FZK 2006 (JEFF-3.2)





Livermore (1986):

Shell #1: outer R = 3.4 cm (1 mfp) Shell #2: outer R = 10.2 cm (3 mfp)14 MeV-neutron source: TiT +d, Ed = 400 keVMethod: Time-of-Flight, L = 10 m n-detector: Θ = 26o, Ethresh > 1 MeV

L = 10 m

Θ = 26o

d-beam

n-detector

Ta-Shell

TiT-Target:14 MeV-Source

R

Shell: outer R = 12.1 cm inner r = 3.1 cm (wall 9 cm, 4 mfp)241Am-Be neutron source, <E> = 4 MeVMethod: Proton-recoil scintil. at L = 2 m n-detector: Θ = 26o, Ethresh > 1 MeV

Lewis Research Center (1974):

Ta shell

Am-Be source

n-detector

L = 2 m

Ta benchmark experiments





0 2 4 6 8 10 12 14 160.0

0.5

1.0

1.5

FZK'06

JEFF-3.2T

JENDL-3.3

C/E

Neutron Energy, MeV

ENDF/B-VI

10-4

10-3

10-2

10-1

FZK'06

JENDL-3.3

LLNL' Experiment

ENDF/B-VI

Ta shell: R = 3.4 cm (1mfp)

Lea

kage

Spe

ctru

m, 1

/sr/

MeV

0 2 4 6 8 10 12 14 160.0

0.5

1.0

1.5

JEFF-3.2T

FZK'06

JENDL-3.3

C/E

Neutron Energy, MeV

ENDF/B-VI

10-4

10-3

10-2

10-1

JEFF-3.2TFZK'06

JENDL-3.3

LLNL' Experiment

ENDF/B-VI

Ta shell: R = 10.2 cm (3mfp)

L

eaka

ge S

pect

rum

, 1/s

r/M

eV

Benchmarking of Ta data

LLNL Ta spheres experiments (14 MeV neutron source)

0 2 4 6 8 100.7

0.8

0.9

1.0

1.1

JEFF-3.2T

FZK'06

JENDL-3.3

C/E

Neutron Energy, MeV

ENDF/B-VI

100

101

102

Ta ShellFZK'06

JENDL-3.3

241Am-Be Source

ENDF/B-VI

Ta shell: R = 12.07cm (3.4mfp)

Neu

tron

spe

ctra

, 1/c

m2 /s/M

eV

LRC Ta shell experiment with Am-Be source (4 MeV mean neutron energy)





0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 153x100

101

102

103

4x103

JEFF-3.2T

FZK'06

Experiment:

JENDL-3.3

(n,n')

- IRK - Osaka - Tohoku

181Ta(n,xn), E = 14.1 MeV

d(E

)/dE

, mb/

MeV

Secondary Neutron Energy, MeV

(n,2n)

ENDF-B6


Ta(n,xn) single and double differential neutron emission distributions

0 1 2 3 4 5 6 7 8101

102

103

JEFF-3.2T

FZK'06

Experiment:IPPE'82(n

,2n)

ENDF/B-VI

JENDL-3.3

181Ta(n,n'), E = 7.94 MeV

d(E

)/dE

, mb/

MeV


(n,n')

10

20

30

40

50

60

0 30 60 90 120 150 1800

5

10

15ENDF/B-VI

JEFF-3.2T

Eout

= 4.0 - 6.0 MeV

Angle, degrees

ENDF/B-VI

Experiment: IPPE'82

JEFF-3.2T

Eout

= 2.0 - 4.0 MeV

181Ta(n,xn), Ein = 7.94 MeV

d/d,

mb/

sr

80

90

100

110

120

130

JEFF-3.2T

ENDF/B-VI Eout

= 0.6 - 2.0 MeV

Ein=14.1 MeV

Ein=7.94 MeV

Ein=7.94 MeV





0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0101

102

103

JEFF-3.2T

39 excited levels in 181Ta below 1.50 MeV =>

FZK'06

Experiment IPPE'82

ENDF-B6JENDL-3.3

181Ta(n,n'), E = 5.19 MeV

d(E

)/dE

, mb/

MeV


0

10

20

30

40

0 30 60 90 120 150 1800

5

10

15 ENDF/B-VI

JEFF-3.2T

Eout

= 3.0 - 4.0 MeV

Angle, degrees

ENDF/B-VI

Experiment: IPPE'82

JEFF-3.2T

Eout

= 2.0 - 3.0 MeV

181Ta(n,n'), Ein = 5.19 MeV

d/d,

mb/

sr

80

100

120

140

160

180

ENDF/B-VI

JEFF-3.2T

Eout

= 0.6 - 2.0 MeV

Ta(n,xn) single and double differential neutron emission distributions


Ein=5.19 MeV





Benchmark analyses on Mn

• Integral neutron transport benchmark experiment: OKTAVIAN pulsed Mn sphere (dia. 61 cm) with central 14 MeV neutron

source: – n-leakage spectrum measured by Time-of-flight technique– g-leakage spectrum measured by pulse height technique (NaI detector)

• Differential experimental data– Two measurements of secondary neutron energy/angular distributions

at incident energy 14 MeV by TOH (Tohoku) and OKTAVIAN

• Available data evaluations– ENDF/B-VII and VII, JENDL-3.3 (adopted in FENDL-2.1 and JEFF-3.1)

and FZK 2006 (JEFF-3.2)

• Integral neutron activation benchmark experiment: JAERI/USDOE experiment with Mn-Cu alloy foil in quasi-mono energetic 14

MeV neutron source: – 3 radioactive isotopes were detected - Mn-54, Mn-56, V-52





- Mn spherical shell, R = 30.5 cm (3.4 mfp for 14 MeV neutrons) - 14 MeV-neutron source: TiT +d, Ed = 240 keV- n-spectrometry: Time-of-Flight, NE218-detector at 11 m/55o, E = 0.076 -14 MeV- -spectrometry: Pulse-height/unfolding, NaI-detector at 5.8 m/55o, E = 0.5-11 MeV

L = 11 m

Θ = 55o

d-beam

γ-detectorNaI( pulse height)

Mn shell

TiT-Target:14 MeV-Source

R

n-detectorNE-218(TOF)

L = 5.8 m

Benchmarking of Mn dataOKTAVIAN benchmark experiment





10-3

10-2

10-1

100

101

0.1 1 100.4

0.6

0.8

1.0

1.2

1.4 ENDF/B-VI JENDL-3.3

FZK

C/E

Neutron Energy, MeV

FZK

JENDL-3.3

ENDF/B-VI

Experiment

n-L

eaka

ge

, 1/M

eV Mn shell

Benchmarking of Mn data

OKTAVIAN benchmark experiment: neutron leakage

spectrum

0 2 4 6 8 10 12 14101

102

103

2x103

FZK'07

Experiment:Tohoku'88OKTAVIAN'89

JENDL-3.3

(n,n')

55Mn(n,xn), E = 14.1 MeV

d(E

)/d

E, m

b/M

eV


(n,2n)

ENDF/B-VI.8

Mn(n,xn) neutron emission spectra

at 14 MeV incident neutron energy





10-5

10-4

10-3

10-2

10-1

100

0.5 1 100.0

0.2

0.4

0.6

0.8

1.0

1.2

ENDF/B-VI

JENDL-3.3

FZK

C/E

-ray Energy, MeV

Contributionfrom Source gammas

FZK

JENDL-3.3

ENDF/B-VIExperiment

-Le

aka

ge

, 1

/Me

V

60cm dia. Mn shell

Benchmarking of Mn data

0 2 4 6 8 10 12 14 16 18 20 22 2410-2

10-1

100

101

102

103

104

(n,2n), Q=-(10.23+0.06)MeV

NJS'77 (E = 14.1MeV)

FZK'07

ORNL'76 (E = 15.5MeV)

JENDL-3.3

(n,), Q=+7.27MeV

55Mn(n,x), E = 14-16 MeV

d(E

)/d

E, m

b/M

eV

Photon Energy, MeV

(n,n'), Q = (-0.0 +0.125) MeV

ENDF/B-VI.8

Mn(n,x) neutron emission spectra

at 14 MeV incident neutron energy

OKTAVIAN benchmark experiment: photon leakage

spectrum





ITER Streaming Experiment at FNG

Measurements of neutron/photon flux spectra by TUD (K. Seidel et al.)

• Mock-up of ITER inboard blanket/shield system with streaming channel in the blanket and cavity at the bottom of channel.

• Neutron and photon flux spectra measured at positions A (41.4 cm) and B (87.6 cm) with source on and off axis (A0, B0).

• Additional measurements with detectors shifted off the axis by 7.5 cm, 15.0 and 9.0 cm (A1, A2 and B1).

• Neutron spectra measured in the energy range between about 20 keV and 15 MeV. – A set of gas-filled proportional counters and a

stilbene scintillation spectrometer used in the energy range up to 3 MeV.

– NE-213 scintillation spectrometer for energy range 1 to 15 MeV.

• Photon flux spectra measured with NE-213 spectrometer above 0.2 MeV.

Position A41.4 cm

Position B87.6 cm





ITER Streaming Experiment

Neutron flux spectra at position B0





ITER Streaming ExperimentPhoton flux spectra at position

B0






Photon flux spectra at position B1





ITER Streaming ExperimentC/E comparison for neutron flux

integrals Position A0 Statistical error (fsd) in A0-position: Energy, MeV FENDL-2.0 FENDL-2.1 JEFF-3.1 FENDL-2.0 FENDL-2.1 JEFF-3.1

0.1 - 1.0 1.37 1.39 1.42 0.59% 0.57% 0.63% 1.0 - 5.0 1.37 1.38 1.34 0.81% 0.80% 0.90%

5.0 - 10.0 1.73 1.79 1.80 1.22% 1.20% 1.26% E > 10.0 0.78 0.78 0.78 0.70% 0.69% 0.75%

Position A1 Statistical error (fsd) in A1-position: Energy, MeV FENDL-2.0 FENDL-2.1 JEFF-3.1 FENDL-2.0 FENDL-2.1 JEFF-3.1

0.1 - 1.0 Inf Inf Inf 0.17% 0.17% 0.19% 1.0 - 5.0 0.89 0.89 0.90 0.21% 0.21% 0.23%

5.0 - 10.0 0.96 0.97 1.00 0.42% 0.42% 0.47% E > 10.0 0.90 0.89 0.91 0.40% 0.41% 0.45%


0.1 - 1.0 Inf Inf Inf 0.16% 0.15% 0.17% 1.0 - 5.0 0.88 0.89 0.90 0.22% 0.21% 0.22%

5.0 - 10.0 0.92 0.93 0.96 0.49% 0.47% 0.51% E > 10.0 0.90 0.89 0.90 0.49% 0.47% 0.50%

Position B0 Statistical error (fsd) in B0-position: Energy, MeV FENDL-2.0 FENDL-2.1 JEFF-3.1 FENDL-2.0 FENDL-2.1 JEFF-3.1

0.1 - 1.0 0.77 0.79 0.80 0.56% 0.61% 0.59% 1.0 - 5.0 0.88 0.90 0.90 0.65% 0.71% 0.69%

5.0 - 10.0 1.03 1.06 1.10 0.76% 0.83% 0.81% E > 10.0 0.70 0.70 0.70 1.01% 1.11% 1.08%

Position B1 Statistical error (fsd) in B1-position: Energy, MeV FENDL-2.0 FENDL-2.1 JEFF-3.1 FENDL-2.0 FENDL-2.1 JEFF-3.1

0.1 - 1.0 Inf Inf Inf 0.51% 0.56% 0.51% 1.0 - 5.0 0.71 0.72 0.73 0.58% 0.63% 0.58%

5.0 - 10.0 0.84 0.86 0.90 0.73% 0.79% 0.72% E > 10.0 0.69 0.68 0.70 0.73% 0.80% 0.74%







0.4 - 1.0 0.82 0.80 0.80 0.73% 0.68% 0.76% E > 1.0 0.82 0.84 0.83 0.62% 0.59% 0.64%

Position A1 Statistical error (fsd) in A1-position:

Energy, MeV FENDL-2.0 FENDL-2.1 JEFF-3.1 FENDL-2.0 FENDL-2.1 JEFF-3.1 0.4 - 1.0 0.99 1.00 1.01 0.31% 0.32% 0.36% E > 1.0 0.99 1.01 1.00 0.24% 0.23% 0.25%

Position A2 Statistical error (fsd) in A2-position:


Position B0 Statistical error (fsd) in B0-position:


Position B1 Statistical error (fsd) in B1-position:


C/E comparison for photon flux integrals





Neutron Leakage Spectra from Pb-shells at 14 MeV incident neutron energy

IPPE & IRD shell (t = 7.5 cm) TUD shell (t = 22.5 cm)

10-2

10-1

100

0.03 0.1 1 10 200.7

0.8

0.9

1.0

1.1

JEFF-3.1T

C/E

Neutron Energy, MeV

ENDF-B6

Pb shell (R = 25.0, r = 2.5 cm)

- TUD Exp. - ENDF-B6 - JEFF-3.1T

Lea

kage

Spe

ctru

m, 1

/Let

harg

y

0.5

1.0

1.5

2.0

10-2

10-1

100

0.4 1 10 200.80.91.01.11.2

C/E

(IR

D)

JEFF-3.1T

C/E

(IP

PE

)

Neutron Energy, MeV

ENDF-B6

ENDF-B6JEFF-3.1T

Pb shell (R = 12.0, r = 4.5 cm)

- IPPE Exp.(ToF) - IRD Exp.(p-Recoil) - ENDF-B6 - JEFF-3.1T

Lea

kage

Spe

ctru

m, 1

/MeV





Secondary neutron Energy Distributions (SED) from Pb(n,xn) reaction at 14 MeV incident neutron energy

Findings: satisfactory agreement for energy differential cross section

for natPb(n,xn) and 208Pb(n,xn) or 208Pb(n,n’) reactions at 14MeV

Neutron emission for natural LeadNeutron emission and inelastic for Lead-208

0 1 2 3 4 5 6 7 8 9 10 11 12 132x100

101

102

103

3x103

3-2+3-

1/2-

(n,2

n)

5/2-

3/2-

3/2+

7/2-

5/2-

208Pb(n,2n)

208Pb(n,n')

JEFF-3.1T

207Pb levels208Pb levels

208Pb(n,xn) & 208Pb(n,n'), E = 14.1 MeV

d(E

)/dE

, mb/

MeV

Neutron Energy, MeV

ENDF-B6

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14101

102

103

3x103

OKTAVIANJEFF-3.1T

natPb(n,xn), E = 14 MeV

d/d

E, m

b/M

eV

Neutron Energy, MeV

ENDF-B6IPPE (208Pb)





Gamma leakage spectra from Pb spherical shell & γ-spectra from Pb(n,xγ) reaction at 14 MeV incident neutron energy

γ - Leakage Spectrum from Pb-Shell

γ - Spectrum from Pb(n,xγ) reaction

0 2 4 6 8 10 12 14 16 18 20 22 2410-4

10-3

10-2

10-1

100

101

102

103

104

natPb(n,n'n) natPb(n,)

JEFF-3.1T

ENDF-B6

natPb(n,x), E = 14 MeV

d/d

E,

mb/

MeV

Gamma Energy, MeV

Exp: Lubljana'79

0 2 4 6 8 10 120.00.20.40.60.81.01.21.4

C/E

Gamma Energy, MeV

10-5

10-4

10-3

10-2

10-1

JEFF-3.1T

ENDF-B6

Pb shell (R = 20, r = 10cm)

-ray

Lea

kage

Spe

ctru

m, 1

/MeV

OKTAVIAN'89





Evaluation and validation of high energy cross-section data for the FENDL-3 nuclear data library

• Proposed research activities by FZK

– Full evaluations of neutron and proton induced high energy cross-section data (up to 150 MeV) including co-variance information

– Preparation of data files in ENDF-6 format– Validation analyses based on Monte Carlo calculations

of available integral experiments– Preparation of special radiation damage cross-sections

for structural materials.





Proposed research activities by FZK

• Validation analyses– Monte Carlo calculations with the MCNP code will be

performed for the FENDL-3 data evaluations using suitable integral benchmark experiments.

– Measured and calculated responses will be compared and possible discrepancies will be analysed to identify deficient cross-section data.

– This information will be used to improve the FENDL-3 data evaluations.






• Neutron cross-section data– Consistent evaluation of high energy neutron cross-sections (up to

150 MeV) for selected nuclides such as Cr-50, -52, -53, -54, Mn-55, Ta-181, W-182, -183, -184, -186 and others including co-variance data information.

– General purpose 150 MeV data files will be prepared in standard ENDF-6 data and made available for FENDL-3.

• Proton activation cross-section data– Current version of the Proton Activation Data File PADF-2007

(comprising 418,575 excitation functions for 2355 target nuclei from Mg to Ra for proton energies from 0 up to 150 MeV) will be extended to include more nuclides, it will be updated to take into account new experiments, and will be complemented by co-variance data






• Radiation damage data: – Neutron and proton induced displacement cross-

section data will be evaluated for structural materials such as Cr, Fe, Ni and others using experimental information about defect generations in irradiated materials, results of molecular dynamics simulations, and results of binary collision approximation model calculations.

– These data will be prepared in a special data file for FENDL-3.

Documents

| U. Fischer et al | FZK Contributions to FENDL-3 CRP KIT - Die Kooperation von Forschungszentrum Karlsruhe GmbH und Universität Karlsruhe (TH) IAEA CRP