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Experimental validation of FENDL-3 nuclear data library Paola Batistoni ENEA - Fusion and Nuclear Technology Dept. Frascati - ITALY FENDL-3 1 st Research.

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Presentation on theme: "Experimental validation of FENDL-3 nuclear data library Paola Batistoni ENEA - Fusion and Nuclear Technology Dept. Frascati - ITALY FENDL-3 1 st Research."— Presentation transcript:

1 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 Associazione Euratom-ENEA sulla fusione

2 Machine operation, protection, maintenance & Safety Nuclear heating (e.g.: < 14 KW in TFC) Fast neutron fluence (e.g.: <10 23 n/m 2 Nb 3 Sn, <5 10 21 n/m 2 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 10 20 n/s)

3 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

4 Frascati 14-MeV Neutron Generator (FNG) ENEA, Frascati Research Centre (Italy) Accelerator based T(d, n)  E d =300 keV European facility for fusion neutronics Shielding activation development of detectors Operating since 1992 MAIN PARAMETERS 14-MeV neutron intensity 10 11 n/s 14-MeV neutron flux 10 10 n/(cm 2 s) D+ beam energy / current 260 keV / 1 mA Target tritium content 370 GBq Generatore di neutroni FNG

5 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 / Li 2 CO 3 ) 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)

6 Main materials in FNG experiments ENDF/B-VI.8 ENDF/B-VI Li-6 Li-7 JENDL-FF ENDF/B-VI (C-nat) C-12 ENDF/B-VI.8JENDL-FFENDF/B-VIO-16 Pb JENDL-FF ENDF/B-VIBe-9 ENDF/B-VI.8ENDF/B-VI Cr-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.8 ENDF/B-VI Ni-58,60 Ni61,62,64 JEFF-3.0 (EFF-3.1) EFF-3.0 ENDF/B-VI Fe-56 FENDL-2.1FENDL-2.0FENDL-1.0Materials Si-28BROND-2ENDF/B-VI.8 Fe-54,57,58,59 ENDF/B-VI ENDF/B-VI.8

7 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

8 Neutron energy E> 10 MeV Nb-93(n,2n) Ni-58(n,2n) Bulk Shield experiment

9 Neutron energy E> 3 MeV Fe-56(n,p) Al-27(n,a) Bulk Shield experiment

10 Neutron energy E> 0.8 MeV Ni-58(n,p) In-115(n,n’) Bulk Shield experiment

11 Neutron energy E  eV Mn-55(n,  ) Au-197(n,  ) Bulk Shield experiment

12 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 experiment Nuclear heating in steel and copper Conclusions :

13 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

14 Nb-93(n,2n) E > 10 MeV Al-27(n,a) E > 5 MeV Streaming experiment

15 Ni-58(n,p) E > 1.5 MeV Au-197(n,g) E  5 eV Streaming experiment

16 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 :

17  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

18 Nb-93(n,2n) Al-27(n,a) Ni-58(n,p) Au-197(n,g) E > 3 MeV E > 1 MeV E  eV E > 10 MeV Conclusions: Significant underestimation of the fast neutron flux

19 DENSIMET-176 (93.2%w W, 2.6%w Fe, 4.2%w Ni, 17.70 g/cm 3 ). DENSIMET-180 (95.0%w W, 1.6%w Fe, 3.4%w Ni, 18.075 g/cm 3 )  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

20 E > 10 MeV Nb-93(n,2n) Zr-90(n,2n) Ni-58(n,2n) W experiment

21 E > 3 MeV Ni-58(n,p) Fe-56(n,p)Al-27(n,a) 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

22 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

23 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

24 Measurements of tritium production in 4 positions P.2 – P.8 by Li 2 CO 3 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

25 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 measurements Analysis performed with EFF-3 Very similar results obtained with FENDL-2.1

26 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 1TUD 3TUD 5TUD 7TUD 1TUD 3TUD 5TUD 7 Be-9 elastic1.9992.0911.7841.6720.048-0.001-0.053-0.130 (n,2n)0.7160.7100.6640.611-0.016-0.191-0.398-0.619 Li-6 (n,t)0.3260.2480.1790.1520.000 Li-7 (n,n't)0.001 0.000 0.9900.9820.9720.960 C-12,O-16 < 0.1 Be9Li6Li7O16C12total TUD 13.5%0.30%0.60%0.30%0.10%3.58% TUD 34.3%0.20%0.40%0.30%0.03%4.33% TUD 54.0%0.20%0.30%0.40%0.03%4.04% TUD 73.5%0.10%0.20%0.40%0.03%3.53%

27 Analysis of activation measurements in Beryllium Conclusions Neutron flux well predicted in Be within  5-10 % uncertainty (shielding properties)

28 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 Li 2 CO 3 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

29 The use of two types of Li 2 CO 3 pellets ( nat. Li and 95% Li-6 enriched ) will allow to separate contribution from Li-6 and L-7 tritium producing reactions.

30 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/ Li 2 CO 3 ) 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

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