96-Cm-248 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa+ DIST-MAY10 20100319 ----JENDL-4.0 MATERIAL 9649 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-01 Fission cross section was evaluated with GMA code. 06-05 Resolved resonance parameters were modified. 07-03 Fission spectra were evaluated. 07-05 New calculation was made with CCONE code. 08-03 Interpolation of (5,18) was changed. Recalculation with CCONE code was performed. Data were compiled as JENDL/AC-2008/1/. 09-08 (MF1,MT458) was evaluated. 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/. MT=456 Number of prompt neutrons per fission (same as JENDL-3.3) At the 0 eV, the experimental data of Zhuravlev et al./3/ was adopted. An energy-dependent term was based on the semi- empirical formula by Howerton/4/. 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/5/. 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 2000/6/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/7/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/8/ 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 - 1500 eV) Resonance parameters for JENDL-3.3 were evaluated on the basis of the data of Benjamin et al./9/, Moore and Keyworth/10/ and Maguire et al./11/. Comment for JENDL-3.3 is as follows: * RESONANCE ENERGIES, NEUTRON AND RADIATIVE WIDTHS WERE TAKEN FROM THE EXPERIMENTAL DATA OF BENJAMIN ET AL./9/. FOR RESONANCES WHOSE RADIATIVE WIDTH WAS UNKNOWN, THE AVERAGE VALUE OF 0.026 EV/9/ WAS ADOPTED. THE FISSION WIDTHS were adopted from Moore and Keyworth /10/ and Maguire et al./11/ THE AVERAGE FISSION WIDTH OF 0.0013 EV/10/ WAS USED FOR ALL RESONANCES OF WHICH FISSION WIDTH HAD NOT BEEN MEASURED. Then the fission widths were roughly adjusted to the fission cross section measured by Maguire et al. R=9.1 FM WAS ASSUMED TO REPRODUCE THE POTENTIAL SCATTERING CROSS SECTION OF 10.4 BARNS ASSUMED BY BENJAMIN ET AL./9/. THE NEUTRON WIDTH OF THE FIRST RESONANCE WAS SLIGHTLY ADJUSTED TO REPRODUCE THE CAPTURE CROSS SECTION OF 2.57 BARNS AT 0.0253 EV. BACKGROUND CROSS SECTIONS WERE GIVEN ONLY FOR THE FISSION AND TOTAL CROSS SECTIONS BY ASSUMING THE FORM OF 1/V. THE THERMAL CROSS SECTIONS TO BE REPRODUCED WERE ESTIMATED FROM AVAILABLE EXPERIMENTAL DATA. For the present file, a negative resonance was added at -30 eV to reproduce the thermal cross sections, and background cross sections were removed. The thermal cross sections to be reproduced: Fission = 0.337 +- 0.032 b Benjamin et al./12/, Zhuravlev et al./13/, Serot et al./14/ Capture = 2.87 +- 0.26 b Druschel et al./15/, Gavrilov et al./16/ Unresolved resonance parameters (1.5 - 200 keV) Parameters were determined with ASREP code/17/ so as to reproduce the cross sections. They are used only for self- shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 10.198 elastic 6.989 fission 0.337 7.84 capture 2.872 267 ------------------------------------------------------- (*) 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/18/. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./19/ MT= 2 Elastic scattering cross section Calculated as total - non-elastic scattering cross sections. MT=18 Fission cross section The following experimental data were analyzed in the energy range from 1.5 keV to 7 MeV with the GMA code/20/: Authors Energy range Data points Reference Moore+ 1.4 keV - 2.8 MeV 413 /10/ Fomushkin+ 0.3 - 5.5 MeV 20 /21/ Maguire+ 1.4 keV - 80 keV 44 /11/ Fomushkin+ 14.1MeV 1 /22/ Fursov+ 510 keV - 6.8 MeV 37 /23/*1) *1) Ratio to Pu-239 fission, converted to cross section by using JENDL-3.3 data. The results of GMA were used to determine the parameters in the CCONE calculation. Above 8 MeV, JENDL-3.3 was 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. MT=18 Fission Isotropic distributions in the laboratory system were assumed. MF= 5 Energy distributions of secondary neutrons MT=18 Fission neutron spectra Below 6 MeV, calculated by Ohsawa /24/ with modified Madland-Nix formula considering multi-mode fission processes (standard-1, standard-2, superlong). Above 7 MeV, calculated with CCONE code. 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./25/ 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 4% and 5%, respevtively. The uncertainty at 0 eV was estimated from the experimental data of Zhuravlev et al./3/ MF=32 Covariances of resonance parameters MT=151 Resolved resonance parameterss Format of LCOMP=0 was adopted. Uncertainties of parameters were taken from Mughabghab /26/. 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 % MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/27/ 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 2. SD of 10% were assumed in the energy region above 8 MeV. In the resolved resonance region, the following standard deviations were added to the contributions from resonance parameters: Total 10 - 20 % Elastic scattering 10 - 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 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/18/ 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/28/ * Global parametrization of Koning-Duijvestijn/29/ was used. * Gamma emission channel/30/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/31/ 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/32/. 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/33/,/34/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,2,3,4 (see Table 2) * optical potential parameters /19/ 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 = 0.066 beta_6 = 0.0015 * Calculated strength function S0= 1.21e-4 S1= 3.23e-4 R'= 9.08 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Cm-248 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04340 2 + * 2 0.14360 4 + * 3 0.29810 6 + * 4 0.50500 8 + * 5 0.76070 10 + 6 1.04900 2 + 7 1.04900 1 - 8 1.06130 12 + 9 1.08400 0 + 10 1.09400 3 - 11 1.12600 2 + 12 1.14300 4 + 13 1.17200 5 - 14 1.22200 4 + 15 1.23500 3 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cm-249 19.0966 0.7605 2.3053 0.3737 -0.8387 2.9359 Cm-248 18.5357 1.5240 2.0504 0.3482 0.2818 3.2555 Cm-247 18.3955 0.7635 1.7794 0.3734 -0.6804 2.7533 Cm-246 18.8984 1.5302 1.7310 0.3608 0.1621 3.4286 Cm-245 18.8322 0.7667 1.4601 0.3623 -0.5771 2.6382 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Cm-249 5.500 0.750 5.100 0.430 Cm-248 6.100 1.040 4.950 0.600 Cm-247 5.400 0.800 5.650 0.650 Cm-246 6.300 1.040 5.100 0.600 Cm-245 6.050 0.500 5.700 0.420 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cm-249 21.0062 0.8872 2.6000 0.3388 -1.7502 3.0872 Cm-248 20.9336 1.5000 2.6000 0.3323 -1.0300 3.6000 Cm-247 20.8609 0.8908 2.6000 0.3256 -1.5320 2.8908 Cm-246 20.7882 1.6500 2.6000 0.3263 -0.7728 3.6500 Cm-245 20.7155 0.8944 2.6000 0.3342 -1.6357 2.9944 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cm-249 19.0966 0.8872 0.9000 0.3432 -0.5123 2.4872 Cm-248 20.9336 1.7780 0.8600 0.3706 -0.1043 3.9780 Cm-247 20.8609 0.8908 0.8200 0.3573 -0.8210 2.8908 Cm-246 20.7882 1.7852 0.7800 0.3658 -0.0107 3.8852 Cm-245 20.7155 0.8944 0.7400 0.3596 -0.8163 2.8944 -------------------------------------------------------- Table 7. Gamma-ray strength function for Cm-249 -------------------------------------------------------- * E1: ER = 11.38 (MeV) EG = 2.71 (MeV) SIG = 332.90 (mb) ER = 14.28 (MeV) EG = 4.18 (MeV) SIG = 431.66 (mb) * M1: ER = 6.52 (MeV) EG = 4.00 (MeV) SIG = 1.49 (mb) * E2: ER = 10.01 (MeV) EG = 3.12 (MeV) SIG = 7.06 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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