50-Sn-112 JAEA EVAL-Dec09 N.Iwamoto,K.Shibata DIST-DEC21 20100119 ----JENDL-5 MATERIAL 5025 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 09-12 The resolved resonance parameters were evaluated by K.Shibata. The data above the resolved resonance region were evaluated and compiled by N.Iwamoto. 21-10 JENDL-5b3 revised by N.Iwamoto (MF2/MT151) revised (MF3,6/MT600-849) added (MF6/MT108,111,112,115) added (MF8/MT4-115) added (MF9/MT102,103) added (MF10/MT28,32,104,105,111,112) added (MF3/MT1,2,4,103-107) recalculated 21-11 above 20 MeV, JENDL/ImPACT-2018 merged by O.Iwamoto 21-11 (MF6/MT5) recoil spectrum added by O.Iwamoto 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 1.5 keV Resonance parameters were based on Mughabghab et al./1/ Neutron orbital angular momentum l of some resonances was estimated with a method of Bollinger and Thomas/2/. Average radiation width was 110 meV/1/. Scattering radius of 6.3 fm was assumed from the systematics of measured values for neighboring nuclides. A negative resonance was added so as to reproduce the thermal capture cross section given by Mughabghab et al./1/. In JENDL-4, the parameters for the negative resonance were adjusted so as to reproduce the thermal capture cross section recommended by Mughabghab./3/. In JENDL-5 the 21, 46-eV p-wave resonances were deleted, and 236.6-eV p-wave resonance was added, based on the results of Kimura et al./4/. Unresolved resonance region : 1.5 keV - 200 keV The unresolved resonance paramters (URP) were determined by ASREP code /5/ so as to reproduce the evaluated total and capture cross sections calculated with optical model code OPTMAN /6/ and CCONE /7/. 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 5.46995E+00 Elastic 4.60926E+00 n,gamma 8.60688E-01 2.98411E+01 n,alpha 1.02936E-06 1.69676E-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 /7/. MT= 16 (n,2n) cross section Calculated with CCONE code /7/. MT= 17 (n,3n) cross section Calculated with CCONE code /7/. MT= 22 (n,na) cross section Calculated with CCONE code /7/. MT= 24 (n,2na) cross section Calculated with CCONE code /7/. MT= 28 (n,np) cross section Calculated with CCONE code /7/. MT= 32 (n,nd) cross section Calculated with CCONE code /7/. MT= 44 (n,n2p) cross section Calculated with CCONE code /7/. MT= 51-91 (n,n') cross section Calculated with CCONE code /7/. MT=102 Capture cross section Calculated with CCONE code /7/. MT=103, 600-649 (n,p) cross section Calculated with CCONE code /7/. MT=104, 650-699 (n,d) cross section Calculated with CCONE code /7/. MT=105, 700-749 (n,t) cross section Calculated with CCONE code /7/. MT=106, 750-799 (n,He3) cross section Calculated with CCONE code /7/. MT=107, 800-849 (n,a) cross section Calculated with CCONE code /7/. MT=108 (n,2a) cross section Calculated with CCONE code /7/. MT=111 (n,2p) cross section Calculated with CCONE code /7/. MT=112 (n,pa) cross section Calculated with CCONE code /7/. MT=115 (n,pd) cross section Calculated with CCONE code /7/. MF= 4 Angular distributions of emitted neutrons MT= 2 Elastic scattering Calculated with CCONE code /7/. MF= 6 Energy-angle distributions of emitted particles MT= 16 (n,2n) reaction Calculated with CCONE code /7/. MT= 17 (n,3n) reaction Calculated with CCONE code /7/. MT= 22 (n,na) reaction Calculated with CCONE code /7/. MT= 24 (n,2na) reaction Calculated with CCONE code /7/. MT= 28 (n,np) reaction Calculated with CCONE code /7/. MT= 32 (n,nd) reaction Calculated with CCONE code /7/. MT= 44 (n,n2p) reaction Calculated with CCONE code /7/. MT= 51-91 (n,n') reaction Calculated with CCONE code /7/. MT=102 Capture reaction Calculated with CCONE code /7/. MT=600-649 (n,p) reaction Calculated with CCONE code /7/. MT=650-699 (n,d) reaction Calculated with CCONE code /7/. MT=700-749 (n,t) reaction Calculated with CCONE code /7/. MT=750-799 (n,He3) reaction Calculated with CCONE code /7/. MT=800-849 (n,a) reaction Calculated with CCONE code /7/. MT=108 (n,2a) reaction Calculated with CCONE code /7/. MT=111 (n,2p) reaction Calculated with CCONE code /7/. MT=112 (n,pa) reaction Calculated with CCONE code /7/. MT=115 (n,pd) reaction Calculated with CCONE code /7/. MF= 8 Information on decay data MT= 4 (n,n') reaction Decay chain is given in the decay data file. MT= 16 (n,2n) reaction Decay chain is given in the decay data file. MT= 17 (n,3n) reaction Decay chain is given in the decay data file. MT= 22 (n,na) reaction Decay chain is given in the decay data file. MT= 24 (n,na) reaction Decay chain is given in the decay data file. MT= 28 (n,np) reaction Decay chain is given in the decay data file. MT= 32 (n,nd) reaction Decay chain is given in the decay data file. MT= 44 (n,n2p) reaction Decay chain is given in the decay data file. MT=102 Capture reaction Decay chain is given in the decay data file. MT=103 (n,p) reaction Decay chain is given in the decay data file. MT=104 (n,d) reaction Decay chain is given in the decay data file. MT=105 (n,t) reaction Decay chain is given in the decay data file. MT=106 (n,He3) reaction Decay chain is given in the decay data file. MT=107 (n,a) reaction Decay chain is given in the decay data file. MT=108 (n,2a) reaction Decay chain is given in the decay data file. MT=111 (n,2p) reaction Decay chain is given in the decay data file. MT=112 (n,pa) reaction Decay chain is given in the decay data file. MT=115 (n,pd) reaction Decay chain is given in the decay data file. MF= 9 Isomeric branching ratios MT=102 Capture reaction Calculated with CCONE code /7/. MT=103 (n,p) reaction Calculated with CCONE code /7/. MF=10 Nuclide production reactions MT= 28 (n,np) reaction Calculated with CCONE code /7/. MT= 32 (n,nd) reaction Calculated with CCONE code /7/. MT=104 (n,d) reaction Calculated with CCONE code /7/. MT=105 (n,t) reaction Calculated with CCONE code /7/. MT=111 (n,2p) reaction Calculated with CCONE code /7/. MT=112 (n,pa) reaction Calculated with CCONE code /7/. ***************************************************************** Nuclear Model Calculation with CCONE code /7/ ***************************************************************** Models and parameters used in the CCONE calculation 1) Optical model * coupled channels calculation coupled levels: 0,1,5 (see Table 1) * optical model potential neutron omp: Kunieda,S. et al./8/ (+) proton omp: Kunieda,S. et al./8/ deuteron omp: Lohr,J.M. and Haeberli,W./9/ triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./10/ He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./10/ alpha omp: Huizenga,J.R. and Igo,G./11/ (+) omp parameters were modified. 2) Two-component exciton model/12/ * Global parametrization of Koning-Duijvestijn/13/ was used. * Gamma emission channel/14/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Width fluctuation correction/15/ 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/16/. Parameters are shown in Table 2. * Gamma-ray strength function of generalized Lorentzian form /17/,/18/ 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 Sn-112 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 1.25685 2 + * 2 2.15109 2 + 3 2.19090 0 + 4 2.24762 4 + 5 2.35453 3 - * 6 2.47620 2 + 7 2.52105 4 + 8 2.54930 6 + 9 2.55660 1 - 10 2.61800 0 + 11 2.72156 2 + 12 2.75619 2 + 13 2.78392 4 + 14 2.86000 3 + 15 2.91340 4 + 16 2.91771 3 + 17 2.92678 6 + 18 2.94596 4 + 19 2.96700 2 + 20 2.98900 0 + 21 3.07887 0 + 22 3.09307 2 + 23 3.11800 0 + 24 3.13700 5 - 25 3.14941 4 + ------------------- *) Coupled levels in CC calculation Table 2. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Sn-113 15.4000 1.1289 0.7517 0.6142 -0.2165 5.1009 Sn-112 14.1265 2.2678 0.0327 0.6962 0.7947 6.7965 Sn-111 13.0000 1.1390 -0.0310 0.7930 -0.8028 6.6686 Sn-110 13.9128 2.2883 -0.9402 0.7805 0.4433 7.8599 In-112 14.0642 0.0000 1.6631 0.6172 -1.2351 3.7389 In-111 13.4200 1.1390 1.2797 0.6828 -0.3713 5.4885 In-110 13.8530 0.0000 0.8218 0.6504 -1.2221 3.9107 In-109 13.2140 1.1494 0.2127 0.7178 -0.2493 5.6123 Cd-111 15.6000 1.1390 2.3788 0.6387 -1.0720 6.0644 Cd-110 13.9128 2.2883 1.7183 0.7231 -0.0110 7.6874 Cd-109 16.0000 1.1494 1.5974 0.6243 -0.7773 5.7804 Cd-108 13.6986 2.3094 0.6850 0.7416 0.2888 7.5998 Cd-107 15.7000 1.1601 0.3491 0.6866 -0.9318 6.4069 -------------------------------------------------------- Table 3. Gamma-ray strength function for Sn-113 -------------------------------------------------------- * E1: ER = 15.82 (MeV) EG = 5.07 (MeV) SIG = 251.83 (mb) * M1: ER = 8.48 (MeV) EG = 4.00 (MeV) SIG = 0.69 (mb) * E2: ER = 13.03 (MeV) EG = 4.75 (MeV) SIG = 2.77 (mb) -------------------------------------------------------- References 1) Mughabghab, S.F. et al.: "Neutron Cross Sections, Vol. I, Part A", Academic Press (1981). 2) Bollinger, L.M., Thomas, G.E.: Phys. Rev., 171,1293(1968). 3) Mughabghab, S.F.: "Atlas of Neutron Resonances", Elsevier (2006). 4) Kimura,A. et al.: Nucl. Data Sheets, 119, 150 (2014). 5) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999) [in Japanese]. 6) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004). 7) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007). 8) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007). 9) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974). 10) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept. J.H.Williams Lab., Univ. Minnesota (1969). 11) Huizenga,J.R. and Igo,G.: Nucl. Phys. 29, 462 (1962). 12) Kalbach,C.: Phys. Rev. C33, 818 (1986). 13) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004). 14) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985). 15) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980). 16) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151 (1994). 17) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990). 18) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).