92-U -232 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al. DIST-MAY10 20100323 ----JENDL-4.0 MATERIAL 9219 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-01 Fission cross section was modified. 06-10 Nu-p was modified. 07-05 New calculation was performed with CCONE code. 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=455 and 456. MT=455 Delayed neutron data Nu-d was determined from Tuttle's systematics/2/. Six group decay constants were taken from the paper of Brady and England/3/. MT=456 Number of prompt neutrons per fission Nu-p measured by Jaffey and Lerner/4/ at the thermal neutron energy was adopted. An energy dependent term was determined from the systematics of Ohsawa /5/. Nu-p of Cf-252 SF = 3.756+-0.031 /6/ was used. 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/7/. 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 2000/8/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/9/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/10/ 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 - 200 eV) For JENDL-3.3, recommendation by Mughabghab /11/ was adopted, and its formula was changed from Reich-Moore to Multilevel Breit-Wigner type. Background cross section was given to reproduce measured fission cross sections/12,13/ at valleys of resonance levels. For the present file, the capture and fission widths of a negative resonance were modified so as to reproduce thermal cross sections. The thermal cross sections to be reproduced: Fission = 76.5 +- 4.1 b Elson et al./14/, Cabell et al./15/, Gryntakis/16/ Capture = 75.4 +- 1.5 b Halperin et al./17/, Cabell et al./15/ Unresolved resonance parameters (200 eV- 40 keV) Parameters were determined with ASREP code/18/ so as to repruduce the following cross sections: Total = sum of partial cross sections Elastic = calculated with CCONE code/19/ Fission = results of GMA analysis Capture = calculated with CCONE code Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 162.72 elastic 10.81 fission 76.52 364 capture 75.39 173 ------------------------------------------------------- (*) 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 total (MT=1), 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 From 200 eV to 40 keV: sum of partial cross sections. Above 40 keV: the cross section was calculated with CCONE code using CC OMP of Soukhovitskii et al./20/ MT=2 Elastic scattering cross section From 200 eV to 40 keV: calculated with CCONE code. Above 40 keV: total - nonelastic scattering cross sections MT=18 Fission cross section Below 200 eV, background cross section was given. Above 200 eV, the following experimental data were analyzed with the GMA code /21/: Authors Energy range Data points Reference Auchampaugh+ 0.15 - 1.93 keV 265 /12/ Farrell 0.15 - 21.3 keV 1173 /13/ Fursov+ 0.135 - 14.7 MeV 77 /22/(*1) (*1) Relative to Pu-239 fission. Data were converted to cross sections using JENDL-3.3 data. In the energy range where experimental data were scarce, cross-section curve was determined by eye-guiding. 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 Prompt neutron spectra Calculated with CCONE code. MT=455 Delayed neutron spectra Results of summation calculation made by Brady and England /3/ were 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./23/ for U-235 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 4 MeV and 4 - 7 MeV, and above 7 MeV, respectively. MT=456 Covariance was obtained by fitting a linear function to the data at thermal energy and 5 MeV assuming errors of 2% and 5%, respectively. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Standard deviations were adopted from the data of Mughabghab /11/. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/24/ and the covariances of model parameters used in the theoretical calculations. Covariances of the total, elastic-scattering and capture cross sections were determined by considering the experimental data (see MF=3). For the following cross sections, covariances were further modified. MT=1,2 Total and elastic scattering cross sections In the resonance region (up to 200 eV), uncertainty of 10 % was added. MT=18 Fission cross section In the resonance region, error of 5% was added to the contributions from uncertainties of resonance parameters. Above the resonance region, cross section was evaluated with GMA code/21/. Standard deviations of 20% were added in the energy region from 200 eV to 20 keV. Above 8 MeV, they were assumed to be 10%. MT=102 Capture cross section In the resonance region, addtional error of 2 % was given. 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/25/ * Global parametrization of Koning-Duijvestijn/26/ was used. * Gamma emission channel/27/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/28/ 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/29/. 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/30/,/31/ 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 /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 = 0.066 beta_6 = 0.0015 * Calculated strength function S0= 1.05e-4 S1= 2.49e-4 R'= 9.65 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of U-232 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04757 2 + * 2 0.15657 4 + * 3 0.32260 6 + * 4 0.54100 8 + * 5 0.56319 1 - 6 0.62897 3 - 7 0.69121 0 + 8 0.73456 2 + 9 0.74690 5 - 10 0.80580 10 + 11 0.83307 4 + 12 0.86679 2 + 13 0.91142 3 + 14 0.91510 7 - 15 0.97068 4 + 16 0.98480 6 + 17 1.01685 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 -------------------------------------------------------- U-233 18.3972 0.7861 2.4694 0.3819 -0.8199 2.9895 U-232 18.3293 1.5757 2.6095 0.3887 -0.1141 3.8805 U-231 18.2614 0.7895 2.6793 0.4123 -1.1756 3.4264 U-230 18.1935 1.5825 2.6739 0.3937 -0.1508 3.9419 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- U-233 5.970 0.800 5.450 0.520 U-232 5.800 1.040 5.000 0.600 U-231 6.000 0.800 5.600 0.520 U-230 3.800 1.040 3.900 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-233 20.2008 0.9172 2.6000 0.3312 -1.4942 2.9172 U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383 U-231 20.0518 0.9211 2.6000 0.3325 -1.4903 2.9211 U-230 19.9772 1.8463 2.6000 0.3332 -0.5651 3.8463 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-233 20.2008 0.9172 0.0200 0.3576 -0.6166 2.7172 U-232 20.1263 1.8383 -0.0200 0.3745 0.1401 3.8383 U-231 20.0518 0.9211 -0.0600 0.3758 -0.7765 2.9211 U-230 19.9772 1.8463 -0.1000 0.3771 0.1493 3.8463 -------------------------------------------------------- Table 7. Gamma-ray strength function for U-233 -------------------------------------------------------- K0 = 1.501 E0 = 4.500 (MeV) * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.00 (mb) ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb) * M1: ER = 6.66 (MeV) EG = 4.00 (MeV) SIG = 2.69 (mb) * E2: ER = 10.24 (MeV) EG = 3.31 (MeV) SIG = 6.53 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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