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Nuclear Data Libraries for Advanced Systems: Fusion Devices FENDL-3 J-Ch Sublet CEA, DEN, Cadarache,13108 Saint Paul Lez Durance, France. Outline. Transport files Status Evolution N-transport Processing Benchmarking Activation files Status Benchmarking Processing Uncertainties.

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  1. Nuclear Data Libraries for Advanced Systems: Fusion DevicesFENDL-3 J-Ch Sublet CEA, DEN, Cadarache,13108 Saint Paul Lez Durance, France

  2. Outline • Transport files • Status • Evolution • N-transport • Processing • Benchmarking • Activation files • Status • Benchmarking • Processing • Uncertainties

  3. The actual FENDL-2.1 n-transport • H H-1 H-2 H-3 • He He-3 He-4 • Li Li-6 Li-7 • Be Be-9 • B B-10 B-11 • C C-12 • N N-14 N-15 • O O-16 • F F-19 • Na Na-23 • Mg Mg-nat • Al Al-27 • Si Si-28 Si-29 Si-30 • P P-31 • S S-nat • Cl Cl-35 Cl-37 • K K-nat • Ca Ca-nat • Ti Ti-46 Ti-47 Ti-48 Ti-49 Ti-50 • V V-nat • Cr Cr-50 Cr-52 Cr-53 Cr-54 • Mn Mn-55 • Fe Fe-54 Fe-56 Fe-57 Fe-58 • Co Co-59 • Ni Ni-58 Ni-60 Ni-61 Ni-62 Ni-64 • Cu Cu-63 Cu-65 • Ga Ga-nat • Zr Zr-nat • Nb Nb-93 • Mo Mo-92 Mo-94 Mo-95 Mo-96 Mo-97 Mo-98 Mo-100 • Sn Sn-nat • Ta Ta-181 • W W-182 W-183 W-184 W-186 • Au Au-197 • Pb Pb-206 Pb-207 Pb-208 • Bi Bi-209 71 isotopes/elements JEFF-3.1 = 313 ENDF/B-VII = 323 TENDL-2008 = 350 for the same Z range

  4. Some questions, remarks on the actual content • Why: Ga, Na ?? • No n-absorbers, Ag, In, Cd, you do not stop 14 MeV neutron by • just adding material in front of it • Missing : Hf, Gd, lanthanides, fusion engineer will love some of • their characteristics, as fission engineer do • No thermal data, the neutron just vanished ?? of course not • they thermalise. • g-production in n-transport file • No g-transport, photonuclear

  5. Suggestions, to start on a better footage • 8 Elements == > isotopes • H in H20, D in D20, C in graphite, Al, … • Update each isotopes to the latest releases, it will improve the • file immediately, assuring a much better start • Serious thoughts need to be given for the source of: • H1  ENDF/B-VII corrected • O16  ENDF/B-VII • H-2  BRC • Mn-55  ENDF/B-VIIa, ORNL • Cd-113  ENDF/B-VIIa, ORNL • Fe’s, why mixed ?? • … • Known deficiencies: Fast Be-9 (JEFF-3.1 is better), Cu, …

  6. What is important • Inelastic, super elastic • Cross section shape • levels • Angular distribution • Emitted particles spectrum – no histogram • (n,n’), (n,Xn), (n,Xg) • (n,g) • The unresolved resonance range - continuum • Probability tables – range • Competition widths, open channels, formalism • Self-shielding effects – not negligible • Neutron slow down, so what has been miss-predicted at high energy is likely to make the prediction at lower energies even worst…

  7. NJOY Smoothing One histogram up to 7.5 Kev !! Not physical Not acceptable

  8. Inelastic scattering distributions histograms Spectra Smoothing effects: Inelastic Fission (n,2n) 1e-5 to 10 keV histo. replaced by sqrt(E) Else it overestimates the flux

  9. Effect of the coarse energy steps above 10 MeV It is significant for above 10 Mev threshold reaction rate !! 4% due to the flux not the cross-section !! Every MeV is too coarse 200 KeV steps are better Shape close to an exponential above 10 MeV

  10. Effects of Interpolation for Energy Distributions Incident energy interpolation in the continuum inelastic INT=2 INT=22 (unit-base or UK law 2) MCNP and TART developers recognized this many year ago  hard coded in the codes

  11. Processing • The unresolved resonance range is an important energy range • Different processing codes = different data handling • CALENDF (in contrast to NJOY or PREPRO) generates URR resonances from the averaged parameters of the MF-2, and then produce the probability tables. • In the URR, one needs to self-shield the open competitive channels, and not only take the smooth MF-3 part • The Monte Carlo code TRIPOLI can handle those tables (switch on and off) and demonstrates their impact • The PURR pt’s for MCNP can be used in the same way, however, the only formalism allowed is SLBW ….

  12. W-186, elastic, URR 3.2E+03 - 100.0E+03

  13. W-186, capture, URR 3.2E+03 - 100.0E+03

  14. New BRC 2H(n,n)2H elastic angular distribution, subtle m c.m. Courtesy of P. Romain B. Morillon H. Duarte

  15. 2H elastic angular distribution influence + 2H BRC BRC-08 JEFF-3.1 Courtesy of P. Romain B. Morillon H. Duarte

  16. Benchmarking • Differentiate between “data testing” and “code comparison testing” • LLNL “type” sphere, other.. • Fe, Cr, Ni, Mn, W, Pb, … • Compare neutron spectra and leakage • Monte Carlo and Sn • Test angular distribution AND their interpretation • This is extremely important for shielding (Fusion) applications

  17. Activation files: EAF-2007 (EAF_XS) • 65,565 excitation functions • 816 targets (1H to 257Fm) • 90 fissionable nuclides • 10-5 eV - 60 MeV • Isotopes with T½ > 0.5 days as targets (some below) • Reactions split up to three isomeric states (g, m and n) • (n,n¢p) and (n,d) kept separate for gas production • 86 different reaction types • 225 branching ratio, including 18 (g,m,n) • JEFF-3.1 (13), ENDF/B-VII (7), JENDF-3.3 (2) • ENDF-6 manual requires a branching ratio for all resonant channels, to be accounted for in file 2, and stored in file 9

  18. Reaction types and MT’s in EAF-2007/A ENDF-102 MT: 1 …117 new MT: 152…200 declared unassigned

  19. MT Values • Grid of reactions including all 36 MT numbers defined in ENDF • and some defined in EAF-2007

  20. EAF2ENDF code • Transforms EAF-format into ENDF-6 format • Modified to handle branching ratio • Modified to handle the new MT’s • Groups partial channels in a complete data file for each of the 816 target isotopes • Main transformation: MF3 (EAF) --> MF3/8/9/10 (ENDF-6) • add MF-1 and MF-2 • All cross sections up to 60 MeV • Automatically starts CHECKR, FIZCON, and PSYCHE • Allow the use of all ENDF utility codes, PREPRO and some NJOY modules

  21. EAF-2007 into ENDF-6 format = EAF-2007/A • MF-1 General information, comments • Including the original EAF comments lines • MF-2 Resonance parameters • skeleton ; r = 1.35x(A 1/3) • all resonant channels in PENDF already (293.6 K) • MF-3 Total cross section channels • MF-8 Flag, file pointer, dictionary • either to MF-3/9 or MF-10 • MF-9 Isomeric branching ratio • energy dependant • MF-10 Split threshold reaction channels • MF-3 and MF-10 cannot be populated simultaneously, • total reaction channels are not stored when partials exist

  22. EAF-2007/A • ENDF rule MAT numbering • MAT=100*Z+3*(A-Amin)+isom • MAT put in strict ascending order • In EAF the isomer immediately follows • Format (CHECKR) and physics checking (PSYCHE and FIZCON) have been performed !!! When relevant • INTER run through all 816 isotopes • The modified nuclide production format in file 8 was approved at the November 2001 CSWEG meeting • All off this greatly facilitate the processing, and the activation and transport file consistency if … • EAF-2007/A includes all necessary dosimetry channels

  23. EAF2ENDF – 2008 status • The code has been used to process • EAF-2003 = JEFF-3.1/A (distributed) • EAF-2005 (not openly distributed) • EAF-2005.1 (not openly distributed) • EAF-2007 (not openly distributed) • Feedbacks • Error in some of the interpolation line of the MF-9 • Bug, need to be addressed • Fissile NJOY processing • Format change suggested by Bob • ZAP incorrect for MT=112, (n,pa) • Bug, need to be addressed • Q value for (n,f) !!! Why ?? • Energy release by fission – neutrinos energy (ER) • > 2.108

  24. NJOY-99 processing • In NJOY-99.125 reconr has been modified to handle the new MT’s allowing pendf files generation • NJOY-99.68 has an automatic loop (10/) that will process all the nuclide production sections found in File 8 • Results can be extracted from the output listing or read from the GENDF output • MATXSR knows how to add the multiple production sections together and generate a single production cross section for each product • ACER activation-dosimetry file, ZZZAAA.30y, .29y, .28y • Specialized output routines could be written for EAF • all fissile isotopes format changed to be handled by /10 in EAF-2007

  25. NJOY-99 processing • Note that there may be multiple sections generating the same ZAm, and they have to be added together for total nuclide production ZAm of product Fe54 to Mn52m (n,t)  2.605400+4 2.505210+5 1 1 0 12625 3105 37 2.931600+2 0.000000+0 2 1 2 12625 3105 38 4.39594+16 1.172175-7 2625 3105 39  cross section • An update for groupr, matxsr, etc.. is needed for those modules to be able to handle properly the new MT’s

  26. Processing codes: the way forward • SAFEPAQ-II • It is use to create the library, so it can seamlessly process, plot, analyse …. • NJOY-99.259, NJOY-08 !! (but I remember waiting for NJOY-03 !!) • RECONR √ , BROADR √ the MF-1 dictionary is more reliable (it always has been …. rather unreliable, needed to be jumped over) • GROUPR, MATXSR, … need to be updated (for the new MT’s), but will they ever be ?? who need, can handle processed files in those format ?? • CALENDF-2005, MLBW or RM formalism in the URR • PREPRO-2007 • LINEAR, RECENT, ACTIVATE, GROUPIE √ does not bother about MT’s, pendf style gendf … • ZVview, √

  27. Activation libraries: the way forward to EAF2009 • Elastic scattering and total inelastic channels (not only to the isomer) to be added, it allows to reconstruct and compare the total with experimental data, a must .. • Positive Q, (n,p) and (n,a) branching ratio in MF-9 • Upper energy limit; 20 Mev , 60 Mev, … 150 MeV • MT-5/MF-6 yields • above 60 Mev • or for all, but the recognized ENDF-6 MT’s channels  pure ENDF-6 format • MT-5 and activation yields in a separate file, for each isotopes • Deuteron √, Proton √, Gamma activation files

  28. Photonuclear gamma files: pendf Start around 1 MeV

  29. Activation libraries: the way forward • Uncertainties • In a single for EAF-2007 • MF-33 like • + one comment line • + isomeric MAT • Provide a measure of the accuracies without correlations • simple MF-33, groups variances without correlations between the cross sections and/or the adjacent groups • Variable group structure It provide a complete set of “exploitable” data for activation, transmutation calculation uncertainties related to and uniquely the cross sections, not the reaction rates nor the neutron flux

  30. FNS decay heat experiment Product Pathways T½ Path % E/C ΔE % Mo 91 Mo 92(n,2n)Mo 91 15.49m 95.2 0.75 5% Mo 92(n,2n)Mo 91m 1.0m 4.7 0.75 5% Mo 91m Mo 92(n,2n)Mo 91m 100.0 0.76 5%

  31. FNS decay heat experiment Experimental uncertainty Calculational uncertainty E over C

  32. FNS decay heat experiment Product Pathways T½ Path % E/C ΔE % Zr 95 Mo 98(n,a )Zr 95 64.0d 100.0 1.35 19% Nb 91m Mo 92(n,d )Nb 91m 60.9d 89.5 1.35 19% Mo 92(n,2n)Mo 91m 10.3 1.35 19% Nb 92m Mo 92(n,p )Nb 92m 10.1d 99.9 1.06 5% Nb 95 Mo 95(n,p )Nb 95 34.9d 90.3 0.83 9% Mo 96(n,d )Nb 95 8.7 0.83 9% Mo 99 Mo100(n,2n)Mo 99 2.7d 99.4 1.04 5%

  33. FNS decay heat experiment Experimental uncertainty Calculational uncertainty E over C

  34. Conclusions • Differentiate between “data testing” and “code comparison testing” • Clearly to really test the data to some desired accuracy, the simulation codes must be consistent to about that accuracy based on the results of code comparison testing; are they all, every time ??? • Emitted particles spectra smoothing and unit base interpolation are also important • Angular distribution and emitted spectra can influence the results as potently as the cross sections themselves • Temperatures dependant libraries for Fusion applications

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