60-Nd-145 JAEA+ EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata DIST-DEC21 20100119 ----JENDL-5 MATERIAL 6034 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 09-12 The resolved resonance parameters were evaluated by A.Zukeran,K.Shibata. The data above the resolved resonance region were evaluated and compiled by N.Iwamoto. 21-11 revised by O.Iwamoto (MF8/MT4,16,17,22,24,28,32,33,41,102-108) added MF= 1 General information MT=451 Descriptive data and directory MF= 2 Resonance parameters MT=151 Resolved and unresolved resonance parameters Resolved resonance region (MLBW formula) : below 4 keV For JENDL-2, resonance energies were taken from Tellier /1/, and after calibration, data of Rohr et al./2/ and Musgrove et al./3/ were adopted for the levels not measured by Tellier. Neutron widths were adopted from Tellier, and radiation widths were obtained from the capture areas measured by Rohr et al. and Musgrove et al. The average radiation width of 0.087 eV was assumed for the resonances whose capture area was not measured, and to estimate neutron widths from the capture areas for the resonances not measured by Tellier. A negative resonance was added so as to reproduce the thermal capture and total cross sections given by Mughabghab et al./4/ For JENDL-3, total spin j of some resonances was tentative- ly estimated with a random number method. For JENDL-3.2, the capture data measured at ORELA of ORNL were renormalized (factor = 0.9507)/5/. The neutron width and/or the radiation width was revised to reproduce the renormalized capture area for each resonance above 2.592 keV. In JENDL-4, the data for 4.36 - 497.87 eV were replaced with the ones obtained by Barry et al./6/ The parameters for the negative resonance were adjusted so as to reproduce the thermal capture cross section recommended by Mughabghab /7/. Unresolved resonance region : 4.0 keV - 200.0 keV The unresolved resonance paramters (URP) were determined by ASREP code /8/ so as to reproduce the evaluated total and capture cross sections calculated with optical model code OPTMAN /9/ and CCONE /10/. The unresolved parameters should be used only for self-shielding calculation. Thermal cross sections and resonance integrals at 300 K ---------------------------------------------------------- 0.0253 eV res. integ. (*) (barn) (barn) ---------------------------------------------------------- Total 6.9165e+01 Elastic 1.9711e+01 n,gamma 4.9455e+01 2.2269e+02 n,alpha 8.0333e-05 ---------------------------------------------------------- (*) Integrated from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections MT= 1 Total cross section Sum of partial cross sections. MT= 2 Elastic scattering cross section Obtained by subtracting non-elastic scattering cross sections from total cross section. MT= 4 (n,n') cross section Calculated with CCONE code /10/. MT= 16 (n,2n) cross section Calculated with CCONE code /10/. MT= 17 (n,3n) cross section Calculated with CCONE code /10/. MT= 22 (n,na) cross section Calculated with CCONE code /10/. MT= 24 (n,2na) cross section Calculated with CCONE code /10/. MT= 28 (n,np) cross section Calculated with CCONE code /10/. MT= 32 (n,nd) cross section Calculated with CCONE code /10/. MT= 33 (n,nt) cross section Calculated with CCONE code /10/. MT= 41 (n,2np) cross section Calculated with CCONE code /10/. MT= 51-91 (n,n') cross section Calculated with CCONE code /10/. MT=102 Capture cross section Calculated with CCONE code /10/. MT=103 (n,p) cross section Calculated with CCONE code /10/. MT=104 (n,d) cross section Calculated with CCONE code /10/. MT=105 (n,t) cross section Calculated with CCONE code /10/. MT=106 (n,He3) cross section Calculated with CCONE code /10/. MT=107 (n,a) cross section Calculated with CCONE code /10/. MT=108 (n,2a) cross section Calculated with CCONE code /10/. MF= 4 Angular distributions of emitted neutrons MT= 2 Elastic scattering Calculated with CCONE code /10/. MF= 6 Energy-angle distributions of emitted particles MT= 16 (n,2n) reaction Calculated with CCONE code /10/. MT= 17 (n,3n) reaction Calculated with CCONE code /10/. MT= 22 (n,na) reaction Calculated with CCONE code /10/. MT= 24 (n,2na) reaction Calculated with CCONE code /10/. MT= 28 (n,np) reaction Calculated with CCONE code /10/. MT= 32 (n,nd) reaction Calculated with CCONE code /10/. MT= 33 (n,nt) reaction Calculated with CCONE code /10/. MT= 41 (n,2np) reaction Calculated with CCONE code /10/. MT= 51-91 (n,n') reaction Calculated with CCONE code /10/. MT=102 Capture reaction Calculated with CCONE code /10/. ***************************************************************** Nuclear Model Calculation with CCONE code /10/ ***************************************************************** Models and parameters used in the CCONE calculation 1) Optical model * coupled channels calculation coupled levels: 0,4 (see Table 1) * optical model potential neutron omp: Kunieda,S. et al./11/ (+) proton omp: Koning,A.J. and Delaroche,J.P./12/ deuteron omp: Lohr,J.M. and Haeberli,W./13/ triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./14/ He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./14/ alpha omp: McFadden,L. and Satchler,G.R./15/ (+) omp parameters were modified. 2) Two-component exciton model/16/ * Global parametrization of Koning-Duijvestijn/17/ was used. * Gamma emission channel/18/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Width fluctuation correction/19/ was applied. * Neutron, proton, deuteron, triton, He3, alpha and gamma decay channel were taken into account. * Transmission coefficients of neutrons were taken from optical model calculation. * The level scheme of the target is shown in Table 1. * Level density formula of constant temperature and Fermi-gas model were used with shell energy correction/20/. Parameters are shown in Table 2. * Gamma-ray strength function of generalized Lorentzian form /21/,/22/ was used for E1 transition. For M1 and E2 transitions the standard Lorentzian form was adopted. The prameters are shown in Table 3. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Level Scheme of Nd-145 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 7/2 - * 1 0.06722 3/2 - 2 0.07250 5/2 - 3 0.65767 11/2 - 4 0.74828 9/2 - * 5 0.78045 3/2 - 6 0.91983 1/2 - 7 0.92072 9/2 - 8 0.93705 5/2 - 9 1.01122 11/2 + 10 1.05141 5/2 - 11 1.08525 3/2 + 12 1.11120 13/2 + 13 1.15026 7/2 - 14 1.16105 5/2 + 15 1.16232 9/2 - 16 1.21370 1/2 - 17 1.24973 5/2 - 18 1.28560 5/2 - 19 1.31680 3/2 - 20 1.32630 1/2 + 21 1.33860 7/2 - 22 1.40090 3/2 - 23 1.40130 15/2 - 24 1.40392 5/2 - 25 1.42760 13/2 - ------------------- *) Coupled levels in CC calculation Table 2. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Nd-146 18.1900 1.9863 1.6792 0.5692 0.1138 6.4542 Nd-145 18.5400 0.9965 1.1101 0.5235 -0.2928 4.6189 Nd-144 17.5000 2.0000 0.3419 0.6111 0.2496 6.6190 Nd-143 17.7000 1.0035 -0.4179 0.5516 0.0353 4.4179 Nd-142 15.0000 2.0140 -1.2557 0.6895 0.7987 6.4278 Pr-145 16.8637 0.9965 1.7883 0.6002 -0.8883 5.5766 Pr-144 15.5000 0.0000 0.9153 0.6715 -1.9662 5.0412 Pr-143 16.6639 1.0035 0.4682 0.6161 -0.5920 5.4208 Pr-142 16.4000 0.0000 -0.4377 0.7390 -2.6336 6.4135 Pr-141 16.4637 1.0106 -1.2280 0.6590 -0.3966 5.5793 Ce-144 17.4894 2.0000 1.0129 0.5822 0.3675 6.2813 Ce-143 19.6000 1.0035 0.4100 0.4774 0.1189 3.9645 Ce-142 18.9500 2.0140 -0.3155 0.5558 0.6875 5.9346 Ce-141 17.9000 1.0106 -1.0773 0.4985 0.5829 3.4550 Ce-140 17.0742 2.0284 -1.9470 0.5674 1.4861 4.9920 Ce-139 15.5000 1.0178 -1.1255 0.5922 0.4151 4.0889 -------------------------------------------------------- Table 3. Gamma-ray strength function for Nd-146 -------------------------------------------------------- * E1: ER = 14.74 (MeV) EG = 5.78 (MeV) SIG = 310.00 (mb) * M1: ER = 7.79 (MeV) EG = 4.00 (MeV) SIG = 1.01 (mb) * E2: ER = 11.96 (MeV) EG = 4.36 (MeV) SIG = 3.37 (mb) -------------------------------------------------------- References 1) Tellier, H.: CEA-N-1459 (1971). 2) Rohr, G., et al.: "Proc. 3rd Conf. on Neutron Cross Sections and Technology, Knoxville 1971", Vol. 2, 743. 3) Musgrove, A.R. de L., et al.: AEEC/E401 (1977). 4) Mughabghab, S.F.: "Neutron Cross Sections, Vol. I, Part B", Academic Press (1984). 5) Allen, B.J. et al.: Nucl. Sci. Eng., 82, 230 (1982). 6) Barry, D.P. et al.: Nucl. Sci. Eng., 153, 8 (2006). 7) Mughabghab, S.F.: "Atlas of Neutron Resonances", Elsevier (2006). 8) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999) [in Japanese]. 9) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004). 10) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007). 11) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007). 12) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003) [Global potential]. 13) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974). 14) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept. J.H.Williams Lab., Univ. Minnesota (1969). 15) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966). 16) Kalbach,C.: Phys. Rev. C33, 818 (1986). 17) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004). 18) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985). 19) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980). 20) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151 (1994). 21) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990). 22) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).