98-Cf-249 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,+ DIST-MAY10 20100318 ----JENDL-4.0 MATERIAL 9852 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 05-12 Fission cross section above 70 eV was revised. 07-09 Theoretical calculation was made with CCONE code. Data were compiled as JENDL/AC-2008/1/. 09-08 (MF1,MT458) was evaluated. 09-12 New theoretical calculation was made with CCONE code. 10-01 Data of prompt gamma rays due to fission were given. 10-03 Covariance data were given. MF= 1 General information MT=452 Number of Neutrons per fission Sum of MT's = 455 and 456. MT=455 Delayed neutron data (Same as JENDL-3.3) Semi-empirical formula by Tuttle/2/. Decay constants were evaluated by Brady and England/3/. MT=456 Number of prompt neutrons per fission (same as JENDL-3.3) At the thermal energy, the data of Volodin et al./4/ was adopted. An energy dependent term was based on the semi- empirical formula by Howerton/5/. 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/6/. 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 2000/7/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/8/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/9/ 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 (MLBW: 1.0e-5 - 70.0 eV) Resonance energies, neutron and fission widths were taken from the experimental data of Benjamin et al./10/ Radiative widths were assumed to be 0.04 eV according to Dabbs et al./11/ A negative resonance was added so as to reproduce thermal cross sections. capture = 507+-20 /12,13/ fission = 1671+-28 /14,15,16,17/ Unresolved resonance parameters (70 eV - 30 keV) Parameters (URP) were determined with ASREP code/18/ so as to reproduce the cross sections in this energy region. URP are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 2185.5 elastic 6.25 fission 1672.7 2220 capture 506.5 696 ------------------------------------------------------- (*) 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/19/. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./20/ MT= 2 Elastic scattering cross section Calculated as total - non-elastic scattering cross sections. MT=18 Fission cross section In the energy range from 70 to 300 eV, the fission cross section was determined from the data measured by Dabbs et al./21/ Resonance structure was kept in the present file. Above 300 eV, the following experimental data were analyzed with the GMA code/22/: Authors Energy range Data points Reference Silbert 280 eV - 2.9 MeV 2083 /23/ Fursov+ 0.5 - 7 MeV 13 /24/ (*1) Dabbs+ 280 eV - 17 MeV 956 /21/ Kuprijanov+ 0.13 - 7.4 MeV 43 /25/ (*1) (*1) Relative to Pu-239 fission. Data were converted to cross sections using JENDL-3.3 data. The results of GMA were used to determine the parameters in the CCONE calculation. In the energy region from 10 to 20 MeV, the data of JENDL-3.3 were adopted. MT=19, 20, 21, 38 Multi-chance fission cross sections Calculated with CCONE code, and renormalized to the total fission cross section (MT=18). MF= 4 Angular distributions of secondary neutrons MT=2 Elastic scattering Calculated with CCONE code/19/. MT=18 Fission Isotropic distributions in the laboratory system were assumed. MF= 5 Energy distributions of secondary neutrons MT=18 Prompt neutrons Calculated with CCONE code/19/. MT=455 Delayed neutrons Calculated by Brady and England/3/. MF= 6 Energy-angle distributions Calculated with CCONE code/19/. 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./26/ for Pu-239 thermal fission were adopted. MF=31 Covariances of average number of neutrons per fission MT=452 Number of neutrons per fission Sum of covariances for MT=455 and MT=456. MT=455 Error of 15% was assumed. MT=456 Covariance was obtained by fitting a linear function to the data at 0.0 and 5.0 MeV with an uncertainty of 1% and 5%, respectively. The uncertainty of 1% was based on the data of Volodin et al./4/ MF=32 Covariances of resonance parameters MT=151 Resolved resonance parameterss Format of LCOMP=0 was adopted. Uncertainties of parameters were taken from Mughabghab /27/. For the parameters without any information on uncertainty, the following uncertainties were assumed: Resonance energy 0.1 % Neutron width 10 % Capture width 20 % Fission width 10 % They were further modified by considering experimental data of the fission and capture cross sections at the thermal neutron energy. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/28/ and the covariances of model parameters used in the cross-section calculations. For the fission cross section, covariances obtained with the GMA analysis were adopted. Standard deviations (SD) were multiplied by a factor of 1.5. SD of 6% were assumed in the energy region above 6 MeV. In the resolved resonance region, the following standard deviations were added to the contributions from resonance parameters: Elastic scattering 20 % 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 Estimated with CCONE and KALMAN codes. ***************************************************************** Calculation with CCONE code ***************************************************************** Models and parameters used in the CCONE/19/ 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/29/ * Global parametrization of Koning-Duijvestijn/30/ was used. * Gamma emission channel/31/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/32/ 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/33/. 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/34/,/35/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,2,5,7 (see Table 2) * optical potential parameters /20/ Volume: V_0 = 49.97 MeV lambda_HF = 0.01004 1/MeV C_viso = 15.9 MeV A_v = 12.04 MeV B_v = 81.36 MeV E_a = 385 MeV r_v = 1.2568 fm a_v = 0.633 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 = 1.1803 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 = 0.05 beta_6 = 0.0015 * Calculated strength function S0= 1.13e-4 S1= 3.05e-4 R'= 8.98 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Cf-249 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 9/2 - * 1 0.06249 11/2 - * 2 0.13620 13/2 - * 3 0.14496 5/2 + 4 0.18794 7/2 + 5 0.22000 15/2 - * 6 0.24307 9/2 + 7 0.31500 17/2 - * 8 0.31500 11/2 + 9 0.37952 7/2 + 10 0.38400 13/2 + 11 0.41680 1/2 + 12 0.42500 19/2 - 13 0.43756 9/2 + 14 0.44000 3/2 + 15 0.44298 7/2 + 16 0.46000 5/2 + 17 0.50070 9/2 + ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cf-250 19.3542 1.5179 1.0208 0.3946 -0.1776 3.8329 Cf-249 19.0966 0.7605 0.7689 0.3773 -0.6847 2.7641 Cf-248 19.0305 1.5240 0.7495 0.3921 -0.0631 3.7096 Cf-247 18.9645 0.7635 0.5924 0.3748 -0.6176 2.6898 Cf-246 18.8984 1.5302 0.7686 0.4030 -0.1574 3.8448 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Cf-250 6.200 1.150 5.450 0.600 Cf-249 6.000 1.200 6.000 0.800 Cf-248 6.200 1.040 5.400 0.600 Cf-247 6.700 1.000 5.900 0.800 Cf-246 6.200 1.040 5.400 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cf-250 21.4621 1.7709 2.5000 0.2827 -0.1068 3.2709 Cf-249 20.0514 0.8872 2.5000 0.3486 -1.7837 3.0872 Cf-248 21.3142 1.7780 2.5000 0.3226 -0.6352 3.7780 Cf-247 21.2402 0.8908 2.5000 0.3232 -1.5224 2.8908 Cf-246 21.1662 1.7852 2.5000 0.3238 -0.6280 3.7852 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cf-250 21.4621 1.7709 0.9600 0.3499 0.0610 3.7709 Cf-249 20.0514 0.8872 0.9200 0.3794 -1.0059 3.0872 Cf-248 21.3142 1.7780 0.8800 0.3522 0.0693 3.7780 Cf-247 21.2402 0.8908 0.8400 0.3533 -0.8173 2.8908 Cf-246 21.1662 1.7852 0.8000 0.3544 0.0777 3.7852 -------------------------------------------------------- Table 7. Gamma-ray strength function for Cf-250 -------------------------------------------------------- K0 = 1.500 E0 = 4.500 (MeV) * E1: ER = 11.31 (MeV) EG = 2.67 (MeV) SIG = 255.17 (mb) ER = 14.32 (MeV) EG = 4.20 (MeV) SIG = 510.34 (mb) * M1: ER = 6.51 (MeV) EG = 4.00 (MeV) SIG = 1.84 (mb) * E2: ER = 10.00 (MeV) EG = 3.11 (MeV) SIG = 7.35 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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