94-Pu-242 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, Murata, + DIST-MAY10 20100323 ----JENDL-4.0 MATERIAL 9446 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-08 Nu-p was revised. 06-09 Resonance parameters were revised. 06-12 Fission cross section was revised. 07-05 Data were calculated with CCONE code. Data were compiled as JENDL/AC-2008/1/. 09-03 (1,452), (1,455) and (1,456) were revised. 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 neutrons Determined from nu-d of the following three nuclides and partial fission cross sections calculated with CCONE code/2/. Pu-243 = 0.0153 an average of experimental data of Krick and Evans /3, 4/ Pu-242 = 0.011 0.00160 /5/ was multiplied by 0.7. Pu-241 = 0.0064 0.00911 /5/ was multiplied by 0.7. Values for Pu-242 and 241 were multiplied by 0.7 to reproduce the nu-d of 0.0113+-0.0009 at 14.7MeV/6/ Decay constants were evaluated by Brady and England/7/. MT=456 Number of prompt neutrons per fission Least-squares fitting of a straight line to the experimental data of Khokhlov et al./8/ Since their data were total numbers of neutrons per fission, numbers of delayed neutrons (MT=455) were subtracted. nu-p = 2.8779 + 0.13755*E(MeV) 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/9/. 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 2000/10/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/11/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/12/ 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 (below 1 keV) Resonance parameters of JENDL-3.3 were modified: * Upper boundary was decreased from 1.9 keV to 1 keV. * Capture width of 2.67-eV resonances was changed from 22 meV to 26.8 meV. * Fission width of 53.46-eV resonance was increased from 1.825 micro-eV to 36 micro-eV. Thermal capture cross section of 19.98+-0.66 b to be reproduced was determined from Butler et al./13/, Durham and Molson/14/ and Marie et al./15/. Unresolved resonance parameters (1 keV - 100 keV) Parameters were estimated with ASREP code/16/ so as to total, fission and capture cross sections in this energy region. They are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 28.213 elastic 8.326 fission 0.00244 4.36 capture 19.885 1130 ------------------------------------------------------- (*) 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 Calculated with CCONE code and modified below 500 keV by multiplying an energy-dependent factor so as to reproduce average total cross sections obtained from the data of Young et al. /17/ The calculation was made with CC OMP of Soukhovitskii et al./18/ MT= 2 Elastic scattering cross section Calculated as total cross section - sum of partial cross sections. MT=18 Fission cross section The following experimental data were analyzed in the energy range from 1 keV to 20 MeV with the GMA code/19/: Authors Energy range Data points Reference Bulter 141 keV - 1.66 MeV 65 /20/ Fomushkin+ 14.5 MeV 1 /21/ Bergen+ 0.1 - 2.96 MeV 141 /22/(*1) Auchampaugh+ 0.95 keV - 3.99 MeV 3102 /23/ Meadows 0.397 - 9.92 MeV 49 /24/(*2) Behrens 97.2 keV - 20.0 MeV 133 /25/(*2) Kuprijanov+ 0.127 - 7.4 MeV 71 /26/(*2) Cance+ 2.47 MeV 2 /27/ Alkhazov+ 14.7 MeV 1 /28/ Weigmann+ 0.3 - 9.7 MeV 222 /29/ Arlt+ 14.7 MeV 1 /30/ Meadows 14.7 MeV 1 /31/(*2) Iwasaki+ 0.597 - 6.76 MeV 17 /32/(*2) Staples 0.514 - 19.5 MeV 124 /33/(*2) *1) only the data above 100 keV were used. *2) ratio to U-235 fission cross section 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). MT=102 Capture cross section Calculated with CCONE code. The experimental data of Wisshak and Kaeppeler /34,35/ and Hockenbury et al./36/ were used to determine the parameters in the CCONE calculation. 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 spectra Calculated with CCONE code. MT=455 Delayed neutron spectra (Same as JENDL-3.3) Results of summation calculation made by Brady and England/7/ 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./37/ 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 10% was assumed below 5 MeV and above 5 MeV, respectively by comparing with experimental data/4, 6/ MT=456 Covariance was obtained by fitting a linear function to the experimental data of Khokhlov et al./8/(see MF1,MT456). Variances were multiplied by a factor of 2. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Standard deviations of resonance energy, neutron and capture widths were taken from Mughabghab /38/ Those of fission width were based on the data of fission area reported by Weigmann et al./29/ and Auchampaugh et al./23/. If no information was available, uncertainties were assumed. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/39/ 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 1 keV), uncertainty of 8 % was added. Above 1 keV, covariance matrix was obtained with CCONE and KALMAN codes/39/. MT=18 Fission cross section In the resonance region from 10 to 1000 eV, addtional error of 50% was given. Above the resonance region, cross section was evaluated with GMA code/19/. Standard deviation obatianed was multiplied by a factor of 2.0. MT=102 Capture cross section In the resonance region from 10 to 1000 eV, addtional error of 10% was given. Above 1 keV, covariance matrix was obtained with CCONE and KALMAN codes/39/. 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/40/ * Global parametrization of Koning-Duijvestijn/41/ was used. * Gamma emission channel/42/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/43/ 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/44/. 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/45/,/46/ 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 /18/ 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 = 0.23892 beta_4 = 0.04807 beta_6 = -0.02376 * Calculated strength function S0= 0.98e-4 S1= 2.93e-4 R'= 9.32 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Pu-242 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04454 2 + * 2 0.14730 4 + * 3 0.30640 6 + * 4 0.51810 8 + * 5 0.77860 10 + 6 0.78045 1 - 7 0.83230 3 - 8 0.86500 3 + 9 0.92700 5 - 10 0.95600 0 + 11 0.99250 2 + 12 1.01950 3 - 13 1.03920 1 + 14 1.06400 4 - 15 1.08440 12 + 16 1.09210 6 + 17 1.10200 2 + 18 1.12200 5 - 19 1.15100 2 - 20 1.15450 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 -------------------------------------------------------- Pu-243 18.6999 0.7698 2.4578 0.3280 -0.3352 2.3315 Pu-242 18.6337 1.5428 2.4520 0.3701 0.0291 3.6198 Pu-241 18.5675 0.7730 2.1853 0.3473 -0.4715 2.5167 Pu-240 18.5012 1.5492 2.1440 0.3871 -0.0917 3.7899 Pu-239 18.4349 0.7762 1.8503 0.3560 -0.5001 2.5655 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Pu-243 5.750 0.680 5.520 0.520 Pu-242 6.100 1.000 4.850 0.600 Pu-241 5.950 0.580 5.480 0.520 Pu-240 6.250 1.040 4.920 0.600 Pu-239 6.050 0.700 5.700 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-243 20.5699 0.8981 2.6000 0.3633 -2.0616 3.3981 Pu-242 20.4971 1.7999 2.6000 0.3503 -0.9450 4.0999 Pu-241 20.4242 0.9018 2.6000 0.3647 -2.0579 3.4018 Pu-240 20.3513 1.8074 2.6000 0.3300 -0.6156 3.8074 Pu-239 20.2784 0.9056 2.6000 0.3523 -1.8394 3.2056 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-243 20.9439 0.8981 0.5400 0.3740 -0.9758 3.0981 Pu-242 20.4971 1.7999 0.5000 0.3933 -0.2466 4.1999 Pu-241 20.4242 0.9018 0.4600 0.3804 -0.9744 3.1018 Pu-240 20.5363 1.8074 0.4200 0.3796 -0.0661 4.0074 Pu-239 20.2784 0.9056 0.3800 0.3901 -1.0534 3.2056 -------------------------------------------------------- Table 7. Gamma-ray strength function for Pu-243 -------------------------------------------------------- K0 = 2.100 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 = 6.57 (MeV) EG = 4.00 (MeV) SIG = 3.39 (mb) * E2: ER = 10.10 (MeV) EG = 3.19 (MeV) SIG = 6.78 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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