95-Am-243 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa,+ DIST-SEP12 20120206 ----JENDL-4.0u1 MATERIAL 9549 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT Update File Distribution Sep.14,2012 JENDL-4.0u1 History 07-05 Theoretical calculation was made with CCONE code. 07-08 Theoretical calculation was made with CCONE code. 08-03 Interpolation of (5,18) was changed. Data were compiled as JENDL/AC-2008/1/ 09-03 (1,452) and (1,455) were revised. 09-08 (MF1,MT458) was evaluated. 09-10 Fission cross section was revised. 10-01 Data of prompt gamma rays due to fission were given. 10-03 Covariance data were given. 12-02 For MF1/MT458, E_nu and E_R were corrected. As a result, the Q-vaues (= E_R) were modified for MF3/MT18,19,20,21,38. Re-compiled by K. Shibata. MF= 1 MT=452 Total neutron per fission Sum of MT=455 and 456. MT=455 Delayed neutrons Determined from nu-d of the following three nuclides and partial fission cross sections calculated with CCONE code/2/. Am-244 = 0.0085 Saleh et al./3/, Charlton et al./4/ Am-243 = 0.006659 *1) Am-242 = 0.0049 Saleh et al./3/ *1) an average of systematics by Tuttle/5/, Benedetti et al./6/ and Waldo et al./7/ Decay constants were taken from Saleh et al. and Brady and England. MT=456 Prompt neutrons per fission The data measured by Khokhlov et al./8/ and Drapchinsky et al./9/ were fitted by a linear function/10/. MT=458 Components of energy release due to fission Total energy and prompt energy were calculated from mass balance using JENDL-4 fission yields data and mass excess evaluation. Mass excess values were from Audi's 2009 evaluation/11/. Delayed energy values were calculated from the energy release for infinite irradiation using JENDL FP Decay Data File 2000 and JENDL-4 yields data. For delayed neutron energy, as the JENDL FP Decay Data File 2000/12/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/13/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/14/ when they were provided. When the fractions were not given by Sher, averaged fractions were used. MF= 2 Resonance parameters MT=151 Resolved resonance parameters (below 250eV) JENDL-3.3 parameters of resonances below 1.744 eV were modified: to reproduce an effective capture cross section measured by Ohta et al./15/, to delete background cross sections of fission given in JENDL-3.3. Unresolved resonance parameters (250eV - 40keV) Parameters were determined with ASREP code/16/ so as to reproduce the cross sections in the energy range from 250 eV to 40 keV. They are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 85.831 elastic 6.490 fission 0.0816 6.31 capture 79.259 2040 ------------------------------------------------------- (*) In the energy range from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections Cross sections above the resolved resonance region except for the elastic scattering (MT=2) and fission cross sections (MT=18, 19, 20, 21, 38) were calculated with CCONE code/2/. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./17/ The OMP was adjusted to the Am-241(n,tot) cross section/18/. MT= 2 Elastic scattering cross section Calculated as total - non-elastic scattering cross sections. MT=18 Fission cross section (Above 250eV) The following experimental data were analyzed with the GMA code /19/: Wisshak+/20/, Fomushkin+/21/, Fursov+/22/, Kanda+/23/, Knitter+/24/, Golovnya+/25/, Kobayashi+/26/, Laptev+/27/, Aiche+/28/, Baba+/29/ MT=19, 20, 21, 38 Multi-chance fission cross sections Calculated with CCONE code, and renormalized to the total fission cross section (MT=18). MT=102 Capture cross section The experimental data of Weston and Todd/30/ and Wisshak and Kaeppeler/31/ were used to determine the parameters in the CCONE calculation. MF= 4 Angular distributions of secondary neutrons MT=2 Elastic scattering Calculated with CCONE code. MT=18 Fission Isotropic distributions in the laboratory system were assumed. MF= 5 Energy distributions of secondary neutrons MT=18 Prompt neutrons Below 6 MeV, calculated by Ohsawa/32/ with modified Madland-Nix formula considering multi-mode fission processes (standard-1, standard-2, superlong). Above 7 MeV, calculated with CCONE code/2/. MT=455 Delayed neutrons Taken from Brady and England /33/. Normalized yields of 6 groups are those measured by Saleh et al./3/ MF= 6 Energy-angle distributions Calculated with CCONE code. Distributions from fission (MT=18) are not included. MF=12 Photon production multiplicities MT=18 Fission Calculated from the total energy released by the prompt gamma-rays due to fission given in MF=1/MT=458 and the average energy of gamma-rays. MF=14 Photon angular distributions MT=18 Fission Isotoropic distributions were assumed. MF=15 Continuous photon energy spectra MT=18 Fission Experimental data measured by Verbinski et al./34/ for Pu-239 thermal fission were adopted. MF=31 Covariances of average number of neutrons per fission MT=452 Number of neutrons per fission Combination of covariances for MT=455 and MT=456. MT=455 Error was assumed as follows: E < 5 MeV 5 % /3/ above 15 MeV 15 % MT=456 Covariances were obtained by fitting a linear function to the experimental data of Khokhlov et al./8/ and Drapchinsky et al./9/ Obtained standard deviation was multiplied by a factor of 3 so that the minimum error was about 1%. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Standard deviations of resonance parameters were taken from Mughabghab /35/ If no uncertainties were given by Mughabghab, 0.1% and 10% were assumed for resonance energies and other parameters, respectively. Additional errors of 5 % were given to the fission cross section, and the following values for the capture cross section: energy range additional errors 1.0e-5 - 13 eV 5% 13 - 50 eV 10% 50 - 250 eV 15% MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/36/ and the covariances of model parameters used in the theoretical calculations. For the following cross sections, covariances were determined by different methods. MT=1, 2 Total and elastic scattering cross sections In the resolved resonance region, uncertainty of 10% was added to the contributions from resonance parameter uncertainties. Above 250 eV, estimated with CCONE and KALMAN codes. MT=18 Fission cross section Above 250 eV, evaluated with GMA code/19/. Standard deviations were multiplied by a factor of 2.0. MT=102 Capture cross section Above 250 eV, covariance matrix was obtained with CCONE and KALMAN codes/36/. MF=34 Covariances for Angular Distributions MT=2 Elastic scattering Covariances were given only to P1 components. MF=35 Covariances for Energy Distributions MT=18 Fission spectra Below 6 MeV, covarinaces of Pu239 fission spectra given in JENDL-3.3 were adopted after multiplying a factor of 9. Above 6 MeV, estimated with CCONE and KALMAN codes. ***************************************************************** Calculation with CCONE code ***************************************************************** Models and parameters used in the CCONE/2/ calculation 1) Coupled channel optical model Levels in the rotational band were included. Optical model potential and coupled levels are shown in Table 1. 2) Two-component exciton model/37/ * Global parametrization of Koning-Duijvestijn/38/ was used. * Gamma emission channel/39/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/40/ was included. * Neutron, gamma and fission decay channel were included. * Transmission coefficients of neutrons were taken from coupled channel calculation in Table 1. * The level scheme of the target is shown in Table 2. * Level density formula of constant temperature and Fermi-gas model were used with shell energy correction and collective enhancement factor. Parameters are shown in Table 3. * Fission channel: Double humped fission barriers were assumed. Fission barrier penetrabilities were calculated with Hill-Wheler formula/41/. Fission barrier parameters were shown in Table 4. Transition state model was used and continuum levels are assumed above the saddles. The level density parameters for inner and outer saddles are shown in Tables 5 and 6, respectively. * Gamma-ray strength function of Kopecky et al/42/,/43/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,3,6,8 (see Table 2) * optical potential parameters /17/ Volume: V_0 = 48 MeV lambda_HF = 0.004 1/MeV C_viso = 15.9 MeV A_v = 12.04 MeV B_v = 81.36 MeV E_a = 385 MeV r_v = 1.255 fm a_v = 0.58 fm Surface: W_0 = 17.2 MeV B_s = 11.19 MeV C_s = 0.01361 1/MeV C_wiso = 23.5 MeV r_s = 1.15 fm a_s = 0.601 fm Spin-orbit: V_so = 5.75 MeV lambda_so = 0.005 1/MeV W_so = -3.1 MeV B_so = 160 MeV r_so = 1.1214 fm a_so = 0.59 fm Coulomb: C_coul = 1.3 r_c = 1.2452 fm a_c = 0.545 fm Deformation: beta_2 = 0.213 beta_4 = 0.08 beta_6 = 0.0015 * Calculated strength function S0= 0.91e-4 S1= 2.65e-4 R'= 9.45 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Am-243 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 5/2 - * 1 0.04220 7/2 - * 2 0.08400 5/2 + 3 0.09640 9/2 - * 4 0.10920 7/2 + 5 0.14350 9/2 + 6 0.16230 11/2 - * 7 0.18930 11/2 + 8 0.23800 13/2 - * 9 0.24400 13/2 + 10 0.26600 3/2 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Am-244 18.7661 0.0000 1.9765 0.2797 -0.6834 1.0000 Am-243 18.6999 0.7698 2.0985 0.3873 -0.8965 3.0385 Am-242 18.6337 0.0000 1.6845 0.2795 -0.6541 0.9592 Am-241 18.1961 0.7730 1.7328 0.3819 -0.7226 2.8365 Am-240 18.5012 0.0000 1.3474 0.2883 -0.6831 1.0000 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Am-244 6.450 0.650 6.000 0.530 Am-243 6.200 0.800 5.400 0.520 Am-242 6.510 0.600 6.050 0.550 Am-241 6.100 0.800 5.500 0.520 Am-240 6.100 0.650 6.000 0.450 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Am-244 21.5810 0.0000 2.6000 0.3194 -2.3947 2.0000 Am-243 21.5049 0.8981 2.6000 0.3201 -1.4966 2.8981 Am-242 20.8697 0.0000 2.6000 0.3254 -2.4113 2.0000 Am-241 21.3526 0.9018 2.6000 0.3213 -1.4929 2.9018 Am-240 21.2764 0.0000 2.6000 0.3219 -2.3947 2.0000 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Am-244 21.5810 0.0000 0.6400 0.3664 -1.8698 2.2000 Am-243 21.5049 0.8981 0.6000 0.3532 -0.8019 2.8981 Am-242 21.4288 0.0000 0.5600 0.3688 -1.8681 2.2000 Am-241 21.3526 0.9018 0.5200 0.3556 -0.7969 2.9018 Am-240 21.2764 0.0000 0.4800 0.3567 -1.6981 2.0000 -------------------------------------------------------- Table 7. Gamma-ray strength function for Am-244 -------------------------------------------------------- * E1: ER = 11.51 (MeV) EG = 2.76 (MeV) SIG = 245.81 (mb) ER = 14.29 (MeV) EG = 4.18 (MeV) SIG = 491.62 (mb) * M1: ER = 6.56 (MeV) EG = 4.00 (MeV) SIG = 1.29 (mb) * E2: ER = 10.08 (MeV) EG = 3.18 (MeV) SIG = 6.92 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 2) O.Iwamoto: J. Nucl. Sci. 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