94-Pu-240 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,+ DIST-MAY10 20100325 ----JENDL-4.0 MATERIAL 9440 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-09 Numbers of neutrons per fission were revised. 07-05 Theoretical calculation was made with CCONE code. 07-11 Fission cross section was revised with simultaneous evaluation. 07-11 Parameters of the -3 eV resonance was modified. 07-12 Fission cross section was revised with a new result of simultaneous evaluation. 08-01 Fission cross section was revised. 08-02 Fission cross section and nu-p were revised. CCONE calculation was made with revised parameters. Data were compiled as JENDL/AC-2008/1/. 09-03 (1,452) and (1,455) were revised. 09-08 (MF1,MT458) was evaluated. 09-10 nu-p and fission cross section were revised. 10-01 Data of prompt gamma rays due to fission were given. 10-02 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 three nuclides and partial fission cross sections calculated with CCONE code/2/. Pu-241 = 0.00911 measured by Benedetti et al./3/ Pu-240 = 0.0065 evaluation for Pu-239 Pu-239 = 0.00406 measured by Benedetti et al./3/ Decay constants were taken from evaluation by Brady and England/4/. MT=456 Number of prompt neutrons per fission Experimental data of Frehaut et al./5/, Vorob'jova et al./6/ and Khokhlov et al./7/ were reproduced with a straight line. The nu-p of Cf-252 spontaneous fission of 3.756/8/ was applied. Nu-p below 5 MeV was slightly increased. 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 2.7 keV) Reich-Moore type resonance parameters given by Bouland et al. /13/ were adopted. Capture widths of -3 and 1.056-eV levels were slightly modified so as to reproduce the thermal capture cross section better. Bouland et al. analyzed the resonance parameters up to 5.7 keV. However, the upper boundary was set to 2.7 keV, because the capture cross section was too small above this energy. Small background to the capture cross section was given in the energy range from 1 to 2.7 keV to the capture cross sections by comparing with the data of Weston and Todd/14/. The fission width of the negative resonance assumed at -3 eV was modified to reproduce the fission cross section of 0.03 b at 0.0253 eV/15/. Unresolved resonance parameters (2.7 - 90 keV) Parameters were determined with ASREP code/16/ to reproduce the total and capture cross sections calculated with CCONE code, and average fission cross section. 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 292.01 elastic 2.667 fission 0.036 8.10 capture 289.31 8500 ------------------------------------------------------- (*) 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), fission (MT=18, 19, 20, 21, 38) and capture cross sections were calculated with CCONE code/2/. The model parameters were determined by considering integral data as well as measured cross-section data. The fission cross section of JENDL-3.3 was considered for determination of fission related parameters. 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 Below 100 keV, JENDL-3.3 was adopted. JENDL-3.3 adopted the data of JENDL-3.2 which were based on the data of Weston and Todd/18/. Above, 100 keV, experimental data measured after 1960 were analyzed by simultaneous fitting of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241 fission cross sections and its ratios by the SOK code/19/. Covariance matrix reported in Iwasaki et al./20/ was also considered in the analysis. -------------------------------------------------------------- Cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 21821.003 2.47E+6 M.Cance+ /21/ 21821.002 2.47E+6 M.Cance+ /21/ 30548.002 1.48E+7 N.A.Khan+ /22/ 40636.005 1.46E+7 M.I.Kazarinova+ /23/ -------------------------------------------------------------- Ratio to U-235(n,f) cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 21764.004 1.96E+5 - 9.75E+6 C.Budtz-Jorgensen+/24/ 21764.002 1.51E+5 - 2.97E+5 C.Budtz-Jorgensen+/24/ 41444.002 5.88E+5 - 2.97E+7 A.V.Fomichev+ /25/ 13801.003 5.14E+5 - 2.95E+7 P.Staples+ /26/ 22211.002 6.69E+5 - 6.57E+6 T.Iwasaki+ /20/ 13576.002 5.20E+4 - 3.28E+5 J.W.Behrens /27/ 12714.002 3.35E+5 - 9.60E+6 J.W.Meadows /28/ 40509.002 1.27E+5 - 7.40E+6 V.M.Kuprijanov+ /29/ 20766.005 5.21E+4 - 7.31E+4 K.Wisshak+ /30/ 20766.003 5.09E+4 - 7.25E+4 K.Wisshak+ /30/ 20766.002 5.62E+4 - 2.13E+5 K.Wisshak+ /30/ 10597.002 3.39E+5 - 2.89E+7 J.W.Behrens+ /31/ -------------------------------------------------------------- Ratio to Pu-239(n,f) cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 12766.003 5.29E+4 - 2.10E+7 L.W.Weston+ /32/ -------------------------------------------------------------- In the energy region from 500 keV to 10 MeV, obtained cross section was increased slightly. 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, and multiplied by a factor to reproduce well experimental data in the energy range from 170 keV to 4 MeV. The experimental data of Weston and Todd/33/ and Wisshak and Kaeppeler/34,35/ 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/2/. MT=18 Fission Isotropic distributions were assumed in the laboratory system. MF= 5 Energy distributions of secondary neutrons MT=18 Fission spectra Calculated with CCONE code/2/. MT=455 Delayed neutron spectra (Same as JENDL-3.3) Results of summation calculation made by Brady and England/4/ were adopted. MF= 6 Energy-angle distributions Calculated with CCONE code/2/. 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./36/ 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 Number of delayed neutrons per fission Uncertainty of 5% /3/ was assumed below 5 MeV and 15% above 5 MeV. MT=456 Number of prompt neutrons per fission Covariance matrix was obtained by fitting to the experimental data of nu-p (See MF1/MT456). MF=32 Covariances of resonance parameters Only standard deviations of resonance parameters were given on the basis of SAMMY fitting results /13/. No correlation matrix was given. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/37/ 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 In the resolved resonance region, uncertainty of 3% was added to the contributions from resonance parameter uncertainties. Above 2.7 keV, covariances for CCONE calculation were adopted. MT=18 Fission cross section In the resolved resonance region, uncertainty of 2% was added to the contributions from resonance parameter uncertainties. Between 2.7 to 90 keV, covariance matrix was estimated from Ref./18/ Above 90 keV, covariance matrix was obtained by simultaneous evaluation among the fission cross sections of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241(See MF=3, MT=18, and /1/). Since the variances are very small, they were adopted by multiplying a factor of 2. MT=102 Capture cross section Obtained with CCONE and KALMAN codes/37/. Uncertainty of 5% is given in the resolved resonance region. 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/38/ * Global parametrization of Koning-Duijvestijn/39/ was used. * Gamma emission channel/40/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/41/ 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/42/. 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/43/,/44/ 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 /17/ Volume: V_0 = 49.8075 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 = 17.171 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 = 0.233263 beta_4 = 0.06237 beta_6 = -0.02167 * Calculated strength function S0= 1.03e-4 S1= 2.66e-4 R'= 9.42 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Pu-240 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 0 + * 1 0.04282 2 + * 2 0.14169 4 + * 3 0.29432 6 + * 4 0.49752 8 + * 5 0.59734 1 - 6 0.64885 3 - 7 0.74233 5 - 8 0.74780 10 + 9 0.86071 0 + 10 0.90032 2 + 11 0.93806 1 - 12 0.95885 2 - 13 0.99220 4 + 14 1.00193 3 - 15 1.03053 3 + 16 1.03752 4 - 17 1.04180 12 + 18 1.05570 9 - 19 1.07622 4 + 20 1.08945 0 + 21 1.11553 5 - 22 1.13095 2 + 23 1.13697 2 + 24 1.16153 6 - 25 1.17750 3 + 26 1.18050 2 + 27 1.19900 3 + 28 1.22300 2 + 29 1.23246 4 + 30 1.24080 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 -------------------------------------------------------- Pu-241 18.5557 0.7730 2.1853 0.3475 -0.4724 2.5178 Pu-240 18.4895 1.5492 2.1440 0.3873 -0.0924 3.7908 Pu-239 18.4232 0.7762 1.8503 0.3562 -0.5010 2.5665 Pu-238 18.3569 1.5557 1.9652 0.3806 0.0280 3.6616 Pu-237 18.2906 0.7795 1.8799 0.3588 -0.5099 2.5875 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Pu-241 5.949 0.580 5.478 0.520 Pu-240 6.150 1.040 4.900 0.600 Pu-239 6.050 0.700 5.700 0.600 Pu-238 6.000 1.040 4.800 0.600 Pu-237 5.800 0.800 5.800 0.520 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-241 20.4242 0.9018 2.6000 0.3647 -2.0579 3.4018 Pu-240 20.3513 1.8074 2.6000 0.3517 -0.9375 4.1074 Pu-239 20.2784 0.9056 2.6000 0.3523 -1.8394 3.2056 Pu-238 20.2054 1.8150 2.6000 0.3313 -0.6081 3.8150 Pu-237 20.1324 0.9094 2.6000 0.3320 -1.5137 2.9094 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- 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 Pu-238 20.2054 1.8150 0.3400 0.3694 0.1086 3.8150 Pu-237 20.1324 0.9094 0.3000 0.3706 -0.7963 2.9094 -------------------------------------------------------- Table 7. Gamma-ray strength function for Pu-241 -------------------------------------------------------- K0 = 2.000 E0 = 4.500 (MeV) * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.96 (mb) ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb) * M1: ER = 6.59 (MeV) EG = 4.00 (MeV) SIG = 3.29 (mb) * E2: ER = 10.12 (MeV) EG = 3.22 (MeV) SIG = 6.78 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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