92-U -234 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al. DIST-MAY10 20100302 ----JENDL-4.0 MATERIAL 9225 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 05-12 Fission cross section was evaluated with GMA code. 06-06 Resonance parameters were modified. 07-05 Calculated with CCONE code. 07-11 Resonance parameters were revised. Data were compiled as JENDL/AC-2008/1/. 09-03 Negative resonance, (1,452) and (1,455) were revised. 09-08 (MF1,MT458) was evaluated. 10-01 Data of prompt gamma rays due to fission were given. 10-02 Covariance data were given. MF=1 General Information MT=451 Descriptive data and dictionary MT=452 Number of neutrons per fission Sum of MT=455 and 456. MT=455 Delayed neutrons per fission Determined from nu-d of the following three nuclides and partial fission cross sections calculated with CCONE code/2/. U -235 = 0.01053 U -234 = 0.0067 U -233 = 0.004998 The data for U-235 and U233is average of systematics by Tuttle/3/, Benedetti et al./4/ and Waldo et al./5/ For U-234, determined from experimental data of Borzakov et al./6/, Conant et al./7/ and Keepin et al./8/. Six group decay constants were adopted from Brady and England/9/. MT=456 Prompt neutrons per fission (same as JENDL-3.3) Based on the experimental data by Mather et al./10/ Nu-p of Cf-252 spontaneous fission was assumed to be 3.756. 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 (MLBW: 1.0-5 eV - 1.5 keV) Dridi/15/ analyzed the capture cross section measured at n_TOF with SAMMY code, and obtained the neutron and capture widths of resonances up to 1.5 keV. Their parameters were adopted in the present file. Since cross sections calculated with MLBW formula were almost the same as those with RM formula, MLBW formula was adopted. Fission widths were calculated from the neutron and capture widths of Dridi et al. and fission areas calculated from the resonance parameters of James et al./16/ For new resonances measured by Dridi, fission width was not given. The fission width of 5.17-eV resonance was determined so as to reproduce the fission cross section measured by Heyse et al./17/ A negative resonance was assumed at -0.97 eV by Dridi/15/. Its parameters were adjusted to the thermal cross sections. Scattering radius of 9.6 fm was adopted, which was in good agreement with CCONE calculation. The thermal cross sections to be reproduced: Fission = 0.064 +- 0.014 b Wagemans et al./18/ Capture = 100.2 +- 1.0 b Pomerance/19/, Lounsbury et al./20/, Cabell et al./21/ Bringer et al./22/ Unresolved resonance parameters (1.5 keV - 80 keV) Parameters were determined with ASREP code/23/ so as to reproduce cross sections in this energy region. The parameters are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 118.18 elastic 17.82 fission 0.0670 5.20 capture 100.29 610 ------------------------------------------------------- (*) 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./24/ MT=2 Elastic scattering cross section Calculated as total - non-elstic scattering cross sections MT=18 Fission cross section (Above 1.5 keV) Above 1.5 keV, the following experimental data were analyzed with the GMA code/25/: Authors Energy range Data points Reference James+ 1.40keV - 8.86MeV 2845 /16/ Behrens+ 0.105 - 18.90 MeV 147 /26/(*1) Meadows 0.600 - 9.91 MeV 56 /27/(*1) Goverdovskiy+ 16.0 MeV 1 /28/(*1) Kanda+ 0.490 - 6.97 MeV 30 /29/(*1) Goverdovskiy+ 4.91 - 10.4 MeV 33 /30/(*1) Goverdovskiy+ 0.210 - 0.997 MeV 22 /31/(*1) Meadows 14.7 MeV 1 /32/(*1) Fursov+ 0.130 - 7.40 MeV 71 /33/(*1) (*1) Relative to U-235 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. 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 fission neutrons Calculated with CCONE code. MT=455 Delayed neutrons Taken from Brady and England /9/. 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 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. MT=456 Covariance of obtained by fitting a stlight line to data points with errors of 2% at 1 MeV and 5% at 10 MeV. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Errors of neutron and capture widths were adopted from Ref./35/ Those of fission widths were assumed to be the same as relative error of fission widths given by James et al./16/ Further errors of 10% were added to the total, elastic scattering, fission and capture cross sections in the resonance region. 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=18 Fission cross section Above 1.5 keV, evaluated with GMA code/25/. Variances obtained by GMA were multiplied by a factor of 1.5. MT=102 Capture cross section Above 1.5 keV, 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 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 = 0,1,2,3,4 (see Table 2) * optical potential parameters /24/ 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= 0.94e-4 S1= 2.42e-4 R'= 9.58 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of U-234 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04350 2 + * 2 0.14335 4 + * 3 0.29607 6 + * 4 0.49704 8 + * 5 0.74120 10 + 6 0.78629 1 - 7 0.80988 0 + 8 0.84930 3 - 9 0.85170 2 + 10 0.92674 2 + 11 0.94785 4 + 12 0.96260 5 - 13 0.96860 3 + 14 0.98945 2 - 15 1.02370 4 + 16 1.02380 12 + 17 1.02383 3 - 18 1.04453 0 + 19 1.06930 4 - 20 1.08530 2 + 21 1.09090 5 + 22 1.09590 6 + 23 1.12527 7 - 24 1.12668 2 + 25 1.12760 5 - 26 1.15000 3 + 27 1.16520 3 + 28 1.17210 6 + 29 1.17420 1 + 30 1.19473 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 -------------------------------------------------------- U-235 18.5328 0.7828 2.6265 0.3874 -0.9246 3.1046 U-234 18.4650 1.5689 2.5578 0.3902 -0.1511 3.9087 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 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- U-235 5.790 0.600 5.600 0.320 U-234 6.220 1.040 5.000 0.600 U-233 5.970 0.800 5.400 0.520 U-232 5.800 1.040 5.100 0.600 U-231 6.000 0.800 5.600 0.520 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-235 20.3497 0.9133 2.6000 0.3445 -1.7123 3.1133 U-234 20.2753 1.8304 2.6000 0.3306 -0.5810 3.8304 U-233 18.0364 0.9172 2.6000 0.3977 -2.2218 3.5172 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 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-235 20.8948 0.9133 0.1000 0.3865 -1.0315 3.2133 U-234 20.2753 1.8304 0.0600 0.4009 -0.2030 4.2304 U-233 16.2328 0.9172 0.0200 0.5015 -1.6781 3.9172 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 -------------------------------------------------------- Table 7. Gamma-ray strength function for U-235 -------------------------------------------------------- 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 = 6.64 (MeV) EG = 4.00 (MeV) SIG = 2.68 (mb) * E2: ER = 10.21 (MeV) EG = 3.29 (MeV) SIG = 6.52 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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