60-Nd-143 JAEA+ EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata DIST-DEC21 20100119 ----JENDL-5 MATERIAL 6028 -----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/MT4,16,17,22,24,28,32,41,102-107,112) 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 5 keV For JENDL-2, resonance energies were adopted from Tellier /1/, and those not measured by Tellier were taken from Rohr et al./2/ and Musgrove et al./3/ after normalization to Tellier's data. Radiation widths were derived from capture areas measured by Rohr et al. below 2 keV and Musgrove et al. above 2.5 keV; for the resonances not measured by Tellier, neutron widths were determined from capture areas by assuming the average radiation widths of 0.077 eV for s-wave resonances and 0.085 eV for p-wave ones. Scattering radius was determined from systematics of measured values. A negative resonance was added at -6 eV so as to reproduce the capture cross section of 325+-10 barns compiled by Mughabghab et al./4/ For JENDL-3, total spin J of some resonances was estimated with a random number method. For JENDL-3.2, these resonance parameters were modified so as to reproduce the capture area data measured at ORNL, by taking account of the correction factor (0.9507) announced by Allen et al./5/ The parameters of a negative resonance and scattering radius were adjuseted to get better agreement with recommended thermal cross sections/4/. In JENDL-4, the data for 55.4 - 446.5 eV were replaced with the ones obtained by Barry et al./6/ Unresolved resonance region : 5.0 keV - 200.0 keV The unresolved resonance paramters (URP) were determined by ASREP code /7/ so as to reproduce the evaluated total and capture cross sections calculated with optical model code OPTMAN /8/ and CCONE /9/. 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 4.0821e+02 Elastic 8.3074e+01 n,gamma 3.2511e+02 1.2850e+02 n,alpha 2.2186e-02 ---------------------------------------------------------- (*) 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 /9/. MT= 16 (n,2n) cross section Calculated with CCONE code /9/. MT= 17 (n,3n) cross section Calculated with CCONE code /9/. MT= 22 (n,na) cross section Calculated with CCONE code /9/. MT= 24 (n,2na) cross section Calculated with CCONE code /9/. MT= 28 (n,np) cross section Calculated with CCONE code /9/. MT= 32 (n,nd) cross section Calculated with CCONE code /9/. MT= 41 (n,2np) cross section Calculated with CCONE code /9/. MT= 51-91 (n,n') cross section Calculated with CCONE code /9/. MT=102 Capture cross section Calculated with CCONE code /9/. MT=103 (n,p) cross section Calculated with CCONE code /9/. MT=104 (n,d) cross section Calculated with CCONE code /9/. MT=105 (n,t) cross section Calculated with CCONE code /9/. MT=106 (n,He3) cross section Calculated with CCONE code /9/. MT=107 (n,a) cross section Calculated with CCONE code /9/. MT=112 (n,pa) cross section Calculated with CCONE code /9/. MF= 4 Angular distributions of emitted neutrons MT= 2 Elastic scattering Calculated with CCONE code /9/. MF= 6 Energy-angle distributions of emitted particles MT= 16 (n,2n) reaction Calculated with CCONE code /9/. MT= 17 (n,3n) reaction Calculated with CCONE code /9/. MT= 22 (n,na) reaction Calculated with CCONE code /9/. MT= 24 (n,2na) reaction Calculated with CCONE code /9/. MT= 28 (n,np) reaction Calculated with CCONE code /9/. MT= 32 (n,nd) reaction Calculated with CCONE code /9/. MT= 41 (n,2np) reaction Calculated with CCONE code /9/. MT= 51-91 (n,n') reaction Calculated with CCONE code /9/. MT=102 Capture reaction Calculated with CCONE code /9/. ***************************************************************** Nuclear Model Calculation with CCONE code /9/ ***************************************************************** 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./10/ (+) proton omp: Koning,A.J. and Delaroche,J.P./11/ deuteron omp: Lohr,J.M. and Haeberli,W./12/ triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./13/ He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./13/ alpha omp: McFadden,L. and Satchler,G.R./14/ (+) (+) omp parameters were modified. 2) Two-component exciton model/15/ * Global parametrization of Koning-Duijvestijn/16/ was used. * Gamma emission channel/17/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Width fluctuation correction/18/ 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/19/. Parameters are shown in Table 2. * Gamma-ray strength function of generalized Lorentzian form /20/,/21/ 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-143 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 7/2 - * 1 0.74205 3/2 - 2 1.22804 13/2 + 3 1.30586 1/2 - 4 1.40708 9/2 - * 5 1.43123 11/2 - 6 1.50600 5/2 + 7 1.55554 5/2 - 8 1.55644 3/2 + 9 1.55880 9/2 + 10 1.60838 1/2 + 11 1.69000 5/2 + 12 1.73921 9/2 - 13 1.77485 1/2 + 14 1.79952 3/2 + 15 1.85150 7/2 - 16 1.85256 3/2 - 17 1.90030 7/2 - 18 1.91081 5/2 - 19 1.92060 5/2 - 20 1.96600 3/2 + 21 1.98822 11/2 - 22 1.99640 5/2 + 23 2.00467 1/2 - 24 2.01130 9/2 + 25 2.01887 15/2 - 26 2.01920 7/2 - 27 2.03560 7/2 - 28 2.06385 9/2 + 29 2.06684 13/2 - 30 2.07400 5/2 + 31 2.07513 11/2 - 32 2.09060 7/2 + 33 2.09439 11/2 - 34 2.10100 5/2 - 35 2.12582 3/2 - 36 2.13443 9/2 - 37 2.13700 3/2 - 38 2.14790 3/2 + 39 2.17358 7/2 + 40 2.18300 9/2 - ------------------- *) 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-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 Nd-141 17.8113 1.0106 -0.4633 0.5388 0.1405 4.2362 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 Pr-140 16.9753 0.0000 -0.5433 0.5678 -0.9137 3.4023 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 Ce-138 16.8661 2.0430 -0.4123 0.5781 1.0263 5.6162 Ce-137 18.4300 1.0252 0.5020 0.5105 0.0280 4.2432 -------------------------------------------------------- Table 3. Gamma-ray strength function for Nd-144 -------------------------------------------------------- * E1: ER = 15.05 (MeV) EG = 5.28 (MeV) SIG = 317.00 (mb) * M1: ER = 7.82 (MeV) EG = 4.00 (MeV) SIG = 0.76 (mb) * E2: ER = 12.02 (MeV) EG = 4.38 (MeV) SIG = 3.40 (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. et al.: "Neutron Cross Sections, Vol. I, Part A", Academic Press (1981). 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) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999) [in Japanese]. 8) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004). 9) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007). 10) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007). 11) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003) [Global potential]. 12) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974). 13) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept. J.H.Williams Lab., Univ. Minnesota (1969). 14) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966). 15) Kalbach,C.: Phys. Rev. C33, 818 (1986). 16) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004). 17) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985). 18) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980). 19) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151 (1994). 20) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990). 21) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).