Experimental validation of FENDL-3 nuclear data library Paola Batistoni
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Experimental validation of FENDL-3 nuclear data
library
Paola BatistoniENEA - Fusion and Nuclear Technology Dept.
Frascati - ITALY
FENDL-3 1st Research Co-ordination MeetingIAEA Hq, Vienna, 2-5 December 2008
Associazione Euratom-ENEA sulla fusione
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Machine operation, protection, maintenance & Safety
• Nuclear heating (e.g.: < 14 KW in TFC)
• Fast neutron fluence (e.g.: <1023 n/m2 Nb3Sn, <5•1021 n/m2
insulators) < 1(3) appm He-production thick (thin) < 1 dpa for re-welding• Integrated dose (e.g. <10 MGy to insulators) • Activation (short and long term, dust and ACP estimate, waste estimate)• Shutdown dose rates (repair, replacement maintenance)• Decay heat (heat removal capabilities after accident) • Shielding properties of inner components (streaming paths)
• Tritium breeding Calculations validated by experiments (nuclear data &
codes) with assessment of uncertainties
Radiation loads in fusion devices (e.g. ITER)
(e.g. ITER: 500 MW fusion power, 1.78 x 1020 n/s)
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Neutronic calculations and analyses shall conform to the ITER Management and Quality Program.
ITER Project Management and Quality Program: Quality Assurance in Neutronic Analyses (ITER_D_23H9A4, May 2006)
• Computer software & data must be verified and validated prior to use
• …………. • The FENDL-2.1 library is the reference nuclear data library for
nuclear analyses for ITER• ……
QA in ITER Neutronics Analyses
FENDL-2.1 has been verified and validated for neutronic calculations using available fusion benchmark experiments performed during the ITER R&D activities
Same procedure for FENDL-3
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Frascati 14-MeV Neutron Generator (FNG)
• ENEA, Frascati Research Centre (Italy)
• Accelerator based T(d, n) Ed =300 keV
• European facility for fusion neutronics
Shielding activation development of detectors
• Operating since 1992
• MAIN PARAMETERS
14-MeV neutron intensity 1011 n/s14-MeV neutron flux 1010 n/(cm2s)D+ beam energy / current 260 keV / 1 mATarget tritium content 370 GBq
Generatore di neutroni FNG
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Benchmark experiments performed at FNG and used for FENDL-2.1 /MC validation :
• Stainless steel experiment
• Bulk shield experiment (inboard shield, stainless steel & water)
• Streaming experiment (shield with streaming channel, stainless steel & water)
• Silicon Carbide (SiC) block
• Tungsten block
• HCPB Breeder blanket (Be / Li2CO3) experiment
• HCLL Breeder blanket (LiPb) experiment in progress
• Shutdown dose rate
Fusion neutronics experiments at FNG
All experiments available in SINBAD Fusion benchmarks database (NEA/OECD)
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Main materials in FNG experiments
ENDF/B-VI.8ENDF/B-VI.8
ENDF/B-VIENDF/B-VI
ENDF/B-VI ENDF/B-VI
Li-6Li-7
JENDL-FFJENDL-FFENDF/B-VI (C-nat)C-12
ENDF/B-VI.8JENDL-FFENDF/B-VIO-16
Pb
JENDL-FFJENDL-FFENDF/B-VIBe-9
ENDF/B-VI.8ENDF/B-VIENDF/B-VICr-50,52,53,54
ENDF/B-VI.8JENDL-FF (W-nat)ENDF/B-VIW-182,3,4,6
JEFF-3.0 (EFF-3.0)
ENDF/B-VI.8ENDF/B-VIENDF/B-VI
ENDF/B-VIENDF/B-VI
Ni-58,60Ni61,62,64
JEFF-3.0 (EFF-3.1)EFF-3.0
ENDF/B-VI
Fe-56
FENDL-2.1FENDL-2.0FENDL-1.0Materials
Si-28 BROND-2 ENDF/B-VI.8 ENDF/B-VI.8
Fe-54,57,58,59
ENDF/B-VI
ENDF/B-VI ENDF/B-VI
ENDF/B-VI ENDF/B-VI.8
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Mock-up of the ITER inboard first wall/shielding blanket/vacuum vessel/toroidal magnet
Measurements of the neutron flux as a function of depth by activation foils
Bulk Shield experiment
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Neutron energyE> 10 MeV
Nb-93(n,2n)
Ni-58(n,2n)
Bulk Shield experiment
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Neutron energyE> 3 MeV
Fe-56(n,p)
Al-27(n,a)
Bulk Shield experiment
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Neutron energyE> 0.8 MeV
Ni-58(n,p)
In-115(n,n’)
Bulk Shield experiment
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Neutron energyE eV
Mn-55(n,)
Au-197(n,)
Bulk Shield experiment
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Nuclear heating well predicted within 15% uncertainty in FW, SB and VV 30% in TF coils (underestimation)
Neutron and gamma fluxes (He-production, radiation damage) well predicted within 15% uncertainty at the VV and 30% at the TF coil (underestimation)
Bulk Shield experimentNuclear heating in steel and copper
Conclusions :
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Mock-up of the ITER inboard first wall/shielding blanket /vacuum vessel/toroidal magnet with a streaming channel (3 cm diam)
Measurements of the neutron flux as a function of depth by activation foils
in the channel in the box in the shield behind
Streaming experiments
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Nb-93(n,2n)E > 10 MeV
Al-27(n,a)E > 5 MeV
Streaming experiment
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Ni-58(n,p)E > 1.5 MeV
Au-197(n,g)E 5 eV
Streaming experiment
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Activation measutrements in the cavity Similar results are obtained with the other activation reactions
Streaming experiment
Results & uncertainties similar to those obtained in bulk shield
The presence of penetrations does not deteriorate C/E values
Conclusions :
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Measurements of the neutron flux in four positions at different depths by activation foils
Silicon Carbide block (SiC, advanced low-activation structural material)
Siicon Carbide (SiC) experiments
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0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 10 20 30 40 50 60
Depth (cm)
C/E
Nb-EFF-2.4
Nb-FEN-2
Nb-EFF-3
Nb-FEN-2.1Nb-93(n,2n)
Al-27(n,a)
Ni-58(n,p)
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
0 10 20 30 40 50 60
Depth (cm)
C/E
Al-EFF-2.4
Al-FEN-2
Al-EFF.3
Al-FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
0 10 20 30 40 50 60
Depth (cm)
C/E
Ni-EFF-2.4
Ni-FEN-2
Ni-EFF.3
Ni-FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
0.00 10.00 20.00 30.00 40.00 50.00 60.00
Depth (cm)
C/E
Au-197(n,g)-EFF-2.4
Au-197(n,g)-FEN-2.0
Au-197(n,g)-EFF-3.0
Au-197(n,g)-FEN-2.1Au-197(n,g)
E > 3 MeV
E > 1 MeV
E eV
E > 10 MeV
Conclusions: Significant underestimation of the fast neutron flux
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DENSIMET-176 (93.2%w W, 2.6%w Fe, 4.2%w Ni, 17.70 g/cm3). DENSIMET-180 (95.0%w W, 1.6%w Fe, 3.4%w Ni, 18.075 g/cm3)
Measurements of the neutron flux in four positions at different depths by activation foils
Tungsten block (armour material for plasma facing components)
Tungsten (W) experiments
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0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Nb-93(n,2n) / EFF-2Nb-93(n,2n) / FEN-2Nb-93(n,2n) / JEN-3.2Nb-93(n,2n) / FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Zr-90(n,2n) / EFF-2Zr-90(n,2n) / FEN-2Zr-90(n,2n) / JEN-3.2Zr-90(n,2n) / FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 5 10 15 20 25 30
Depth (cm)
C/E
Ni-58(n,2n) / EFF-2Ni-58(n,2n) / FEN-2Ni-58(n,2n) / JEN-3.2Ni-58(n,2n) / FEN-2.1
E > 10 MeV
Nb-93(n,2n)
Zr-90(n,2n) Ni-58(n,2n)
W experiment
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0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Ni-58(n,p) / EFF-2Ni-58(n,p) / FEN-2Ni-58(n,p) / JEN-3.2Ni-58(n,p) / FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Al-27(n,a) / EFF2Al-27(n,a) / FEN-2Al-27(n,a) / JEN-3.2Al-27(n,a) / FEN-2.1
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Fe-56(n,p) / EFF-2Fe-56(n,p) / FEN-2Fe-56(n,p) / JEN-3.2Fe-56(n,p) / FEN-2.1
E > 3 MeV
Ni-58(n,p)
Fe-56(n,p)Al-27(n,a)
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
0 5 10 15 20 25 30 35 40
Depth (cm)
C/E
Au-197(n,g) / EFF-2Au-197(n,g) / FEN-2Au-197(n,g) / JEN-3.2Au-197(n,g) / FEN-2.1
Au-197(n,g)
E > 1 MeV E eV
E > 3 MeV
W experiment
Conclusion: Significant improvement from FENDL-2.0 to FENDL-2.1
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Dose rate experiments
gamma spectrometer (NE 213 liquid scintillator)
tissue-equivalent dose rate meter(NE 105 plastic scintillator)
Mock-up of the ITER vacuum vessel
Measurement of the shutdown dose rate after long irradiation with 14 MeV neutrons
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• Numerical tools developed combining transport & activation codes (MCNP,FISPACT)
• Validation of different approaches of combination: D1S, R2S
Conclusions:
Shut down dose rate at the vacuum vessel level can be predicted within 25% uncertainty up to 4 months decay time
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Measurements of tritium production in 4 positions P.2 – P.8 by Li2CO3 pellets (nat. Li)(12 pellet/position) and of the neutron flux in the central Be layer by activation foils
Mock-up of the EU breeder blanket Helium Cooled Pebble Bed (HCPB)
HCPB Breeder blanket experiments
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Total uncertainty on C/E comparison ~9% (2 ) C/E sligthly underestimated by 5-15%, generally within total uncertainty
Conclusions:
C/E Comparison for T measurementsAnalysis performed with EFF-3Very similar results obtained with FENDL-2.1
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Cross section sensitivity/uncertainty analysis (FZK, NEA-DB)
Sensitivities (%/%) of calculated T- production vs cross sections
Total uncertainties (2) on total T- production vs cross sections
cross
section
Li-6(n,t) Li-7(n,n't)
TUD 1 TUD 3 TUD 5 TUD 7 TUD 1 TUD 3 TUD 5 TUD 7
Be-9elastic 1.999 2.091 1.784 1.672 0.048 -0.001 -0.053 -0.130
(n,2n) 0.716 0.710 0.664 0.611 -0.016 -0.191 -0.398 -0.619
Li-6 (n,t) 0.326 0.248 0.179 0.152 0.000 0.000 0.000 0.000
Li-7 (n,n't) 0.001 0.001 0.000 0.000 0.990 0.982 0.972 0.960C-12,O-
16 < 0.1
Be9 Li6 Li7 O16 C12 total
TUD 1 3.5% 0.30% 0.60% 0.30% 0.10% 3.58%
TUD 3 4.3% 0.20% 0.40% 0.30% 0.03% 4.33%
TUD 5 4.0% 0.20% 0.30% 0.40% 0.03% 4.04%
TUD 7 3.5% 0.10% 0.20% 0.40% 0.03% 3.53%
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0.80
0.90
1.00
1.10
1.20
1.30
0 5 10 15 20 25
Depth (cm)
C/E
Au(n,g) IRDF02 Nb(n,2n) IRDF02
Al(n,a) IRDF02 Ni(n,p) IRDF02
case c) EFF2.4/Be-9 from FENDL2.0
0.80
0.90
1.00
1.10
1.20
1.30
1.40
0 5 10 15 20 25
Depth (cm)
C/E
Au(n,g) IRDF02 Nb(n,2n) IRDF02
Al(n,a) IRDF02 Ni(n,p) IRDF02
case d) JEFF-3.1
0.80
0.90
1.00
1.10
1.20
1.30
0 5 10 15 20 25
Depth (cm)
C/E
Au(n,g) Nb(n,2n)
Al(n,a) Ni(n,p)
case e) FENDL-2.1
Analysis of activation measurements in Beryllium
Conclusions
Neutron flux well predicted in Be within 5-10% uncertainty (shielding properties)
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HCLL Breeder blanket experiments (2008-2009)
Mock-up of the EU breeder blanket Helium Cooled Pebble Bed (HCPB)
Measurements of tritium production in 15 positions by• Li2CO3 pellets (nat. Li & Li-6
enr.)• TLDs (TLD-600,TLD-700)and of the neutron flux in LiPb by activation foils
(started on 26/11, in progress)
LiPb (15.7 at% nat-Li) 630 kg
Eurofer 11 plates, 9 mm thick
Polyethylene
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0
20
40
60
80
100
120
140
160
180
0 10 20 30Depth (cm)
T-S
peci
fic A
ctiv
ity (
Bq/g
) . Natural Li
Li-6 enriched 95%
The use of two types of Li2CO3 pellets (nat. Li and 95% Li-6 enriched) will allow to separate contribution from Li-6 and L-7 tritium producing reactions.
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Conclusions
• Shielding, activation and dose rate calculations for ITER validated for many materials / components using FENDL- 2.1 reference library
• Neutron fluxes predicted in stainless steel/water shield assemblies within 30% uncertainty at 1 m depth. Underestimation of fast neutron flux observed
• Significant underestimation of the fast neutron flux is found in SiC with FENDL-2.1 (20% at 50 cm)
• Significant improvement from FENDL-2.0 to FENDL-2.1 in tungsten
• Very good prediction of neutron flux in HCPB blanket mock-up (Be/ Li2CO3) Tritium production underestimated by 5-15%, generally within total uncertainty ( 9%, 2)•HCLL blanket experiment in progress
•Only main structural in vessel materials investigated so far