96-Cm-242 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa+ DIST-MAY10 20100318 ----JENDL-4.0 MATERIAL 9631 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 05-07 Fission cross section was evaluated with GMA. 06-04 Resonance parameters were modified. 07-03 Fission spectra below 6 MeV were modified. 07-05 New calculation was made with CCONE code. 07-08 New calculation was made with CCONE code. 07-11 Resonance parameters were modified. 07-12 Resonance parameters were modified. 08-03 Interpolation of (5,18) was changed. Data were compiled as JENDL/AC-2008/1/. 09-02 (1,452), (1,456) and resonance parameters were revised. 09-03 (1,452) and (1,455) were revised. 09-08 (MF1,MT458) was evaluated. 09-11 New 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 Determined from nu-d of the following three nuclides and partial fission cross sections calculated with CCONE code/2/. Cm-243 = 0.002114 Cm-242 = 0.001462 Cm-241 = 0.001014 They are averages of systematics by Tuttle/3/, Benedetti et al./4/ and Waldo et al./5/ Decay constants calculated by Brady and England./6/ were adopted. MT=456 Number of prompt neutrons per fission Based on the systematics recommended by Ohsawa/7/. 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/8/. 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 2000/9/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/10/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/11/ 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 - 275 eV) The resolved resonance parameters of JENDL-3.3 were based on the data of Artamonov et al./12/ and Alam et al./13/ They were adjusted to the thermal cross sections. Background cross section was given to the fission. These parameters were re-adjusted to the fission cross section and fission resonance integral/13/. The negative and first positive resonances were also modified to reproduce the thermal cross sections/14,15/, and energy-dependent fission cross sections measured by Alam et al./13/ Background cross section given in JENDL-3.3 was removed. The fission resonance integral between 0.53eV and 50.93keV is 12.5b which is in agreement with 12.9+-0.7b of Alam et al./13/ The thermal cross sections to be reproduced: Fission = < 5 b Hanna et al./14/ Capture = 19.1 +- 1.5 b Bringer et al./15/ Unresolved resonance parameters (275 eV - 100 keV) Parameters (URP) were determined with ASREP code/16/ 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 36,41 elastic 12.61 fission 4.67 19.3 capture 19.13 133 ------------------------------------------------------- (*) 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/. MT= 2 Elastic scattering cross section Calculated as total - non-elastic scattering cross sections. MT=18 Fission cross section (Above 275 eV) The following experimental data were analyzed in the energy range below 1.4 MeV with the GMA code /18/: Authors Energy range Data points Reference Vorotnikov+ 0.13 - 1.39MeV 38 /19/ Alam+ 0.2 - 97.6keV 49 /13/(*1) (*1) Relative to U-235 fission. Data were converted to cross sections using JENDL-3.3 data. Above 1.4 MeV, the data of JENDL-3.3 was adopted. The results of GMA were used to determine the parameters in the CCONE calculation. 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 /20/ 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 neutron spectra Summation calculation by Brady and England /6/ was adopted. 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./21/ 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 of 15% was assumed below 5 MeV and above 5 MeV, respectively. MT=456 Covariance was obtained by fitting a linear function to the at 0.0 and 5.0 MeV with an uncertainty of 5%. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Standard deviations of resonance energies and neutron widths were taken from Artamonov et al./12/, those of capture widths from Mughabghab /22/, and those of fission widths from Alam et al./13/ MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/23/ 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 resonance region (below 275 eV), uncertainty of 10 % was added. MT=18 Fission cross section Above the resonance region, cross section was evaluated with GMA code/18/. Standard deviation obatianed was adopted after modifications. MT=102 Capture cross section In the resonance region up to 275 eV, addtional error of 5 % was given. Above 275 eV, covariance matrix was obtained with CCONE and KALMAN codes/23/. 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/24/ * Global parametrization of Koning-Duijvestijn/25/ was used. * Gamma emission channel/26/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/27/ 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/28/. 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/29/,/30/ 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 (see Table 2) * optical potential parameters /17/ 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 = 0.2 beta_4 = 0.06 beta_6 = 0.0015 * Calculated strength function S0= 0.91e-4 S1= 2.95e-4 R'= 9.15 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Cm-242 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04213 2 + * 2 0.13700 4 + * 3 0.28800 6 + * ------------------- *) 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-243 18.3259 0.7698 1.3577 0.3635 -0.5169 2.5698 Cm-242 18.6337 1.5428 1.3581 0.3517 0.3362 3.2428 Cm-241 18.5675 0.7730 1.0938 0.3935 -0.8080 2.9546 Cm-240 18.5012 1.5492 1.2421 0.3627 0.2677 3.3492 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Cm-243 6.150 0.600 5.800 0.400 Cm-242 6.200 1.040 4.900 0.600 Cm-241 6.300 0.800 5.000 0.520 Cm-240 6.000 1.040 5.000 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cm-243 20.5699 0.8981 2.6000 0.3281 -1.5248 2.8981 Cm-242 20.4971 1.7999 2.6000 0.3288 -0.6230 3.7999 Cm-241 20.4242 0.9018 2.6000 0.3294 -1.5211 2.9018 Cm-240 20.3513 1.8074 2.6000 0.3300 -0.6156 3.8074 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Cm-243 20.5699 0.8981 0.6600 0.3619 -0.8115 2.8981 Cm-242 20.4971 1.7999 0.6200 0.3631 0.0909 3.7999 Cm-241 20.4242 0.9018 0.5800 0.3643 -0.8066 2.9018 Cm-240 20.3513 1.8074 0.5400 0.3656 0.0996 3.8074 -------------------------------------------------------- Table 7. Gamma-ray strength function for Cm-243 -------------------------------------------------------- * E1: ER = 11.46 (MeV) EG = 2.74 (MeV) SIG = 323.90 (mb) ER = 14.36 (MeV) EG = 4.22 (MeV) SIG = 420.74 (mb) * M1: ER = 6.57 (MeV) EG = 4.00 (MeV) SIG = 1.43 (mb) * E2: ER = 10.10 (MeV) EG = 3.19 (MeV) SIG = 7.07 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 2) O.Iwamoto: J. Nucl. Sci. 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