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Associazione Euratom-ENEA sulla fusione. Experimental validation of FENDL-3 nuclear data library Paola Batistoni ENEA - Fusion and Nuclear Technology Dept. Frascati - ITALY FENDL-3 1 st Research Co-ordination Meeting IAEA Hq, Vienna, 2-5 December 2008. - PowerPoint PPT Presentation
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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
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)
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
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
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)
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
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
Neutron energyE> 10 MeV
Nb-93(n,2n)
Ni-58(n,2n)
Bulk Shield experiment
Neutron energyE> 3 MeV
Fe-56(n,p)
Al-27(n,a)
Bulk Shield experiment
Neutron energyE> 0.8 MeV
Ni-58(n,p)
In-115(n,n’)
Bulk Shield experiment
Neutron energyE eV
Mn-55(n,)
Au-197(n,)
Bulk Shield experiment
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 :
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
Nb-93(n,2n)E > 10 MeV
Al-27(n,a)E > 5 MeV
Streaming experiment
Ni-58(n,p)E > 1.5 MeV
Au-197(n,g)E 5 eV
Streaming experiment
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 :
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
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
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
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
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
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
• 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
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
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
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%
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)
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
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.
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