93-Np-238 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,K.Furutaka,+ DIST-MAY10 20100318 ----JENDL-4.0 MATERIAL 9349 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 07-07 New theoretical calculation was made with CCONE code. 07-10 New theoretical calculation was made with CCONE code. 08-01 Resolved resonance parameters were revised. Data were compiled as JENDL/AC-2008/1/. 09-02 (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's=455 and 456. MT=455 Delayed neutron data Determined from systematics by Tuttle/2/, Benedetti et al./3/ and Waldo et al./4/, and partial fission cross sections calculated with CCONE code/5/. Decay constants were taken from the evaluation of Brady and England/6/. MT=456 Number of prompt neutrons per fission Based on the data of Solonkin et al./7/ (2.3+-0.5 at 0.0253 eV) and Ohsawa's systematics/8/. A constant term is an average of these two. 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 (MLBW: 1.0e-5 - 6.65 eV) Evaluated by Furutaka/13/. Parameters were obtained, starting from the parameters evaluated by Morogovskij/14/, with SAMMY code /15/ to reproduce the fission cross section measured by Danon et al./16/ Their data were normalized to 2130 b at 0.0253 eV. The capture width was fixed to 50 meV. A negative resonance was assumed to reproduce the thermal cross sections: efective capture = 479+-24 /17/ fission = 2201+-34 /18,16,19/ Doppler as well as resolution broadenings were taken into account in the analysis: temperature was assumed to be 300 K. For resolution broadening, parameters of SAMMY's original resolution-broadening function were chosen to approximately reproduce the experimental resolution function described by equation (11) of ref./16/. Un-resolved resonance parameters (6.65 eV - 10 keV) Parameters (URP) were determined with ASREP code /20/ 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 2693.3 elastic 12.26 fission 2201.6 1100 capture 479.5 201 ------------------------------------------------------- (*) In the energy range from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections All the cross-section data above 6.65 eV were calculated with CCONE code/5/. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./21/ MT=18 Fission cross section Calculated with CCONE code. The simulated (n,f) cross section of Britt and Wilhelmy/22/, and the experimental data of Danon et al./16/ 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 Prompt neutrons Calculated with CCONE code. MT=455 Delayed neutrons Calculated by Brady and England /6/. 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 Pu-239 thermal fission were adopted. MF=31 Covariances of average number of neutrons per fission MT=452 Number of neutrons per fission Sum of covariances for MT=455 and MT=456. MT=455 Error of 15% was assumed. MT=456 Covariance was obtained by fitting a linear function to the data at 0.0 and 5.0 MeV with an uncertainty of 22% which was estimated from the experimental data of Solonkin et al./7/ MF=32 Covariances of resonance parameters MT=151 Resolved resonance parameterss Format of LCOMP=1 was adopted. Covariances of parameters were taken from the results of SAMMY analysis/13/. The uncertainty of capture width was assumed to be 30%. 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 cross-section calculations. Covariances of the fission cross section were determined by considering the experimental data (see MF=3). In the resolved resonance region, the following standard deviations were added to the contributions from resonance parameters: Total 0 - 10 % Elastic scattering 20 % Fission 0 - 10 % Capture 0 - 10 % 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/5/ 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 = 0,1,2,4,7 (see Table 2) * optical potential parameters /21/ 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= 0.87e-4 S1= 3.05e-4 R'= 9.37 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Np-238 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 2 + * 1 0.02643 3 + * 2 0.06233 4 + * 3 0.08667 3 + 4 0.10615 5 + * 5 0.12165 4 + 6 0.13604 3 - 7 0.16168 6 + * 8 0.16553 5 + 9 0.17915 4 - 10 0.18288 2 - 11 0.21552 3 - 12 0.21795 0 - 13 0.21870 6 + 14 0.23283 5 - 15 0.24396 1 + 16 0.24640 1 + 17 0.25033 1 + 18 0.25039 2 - 19 0.25885 4 - 20 0.27552 5 + 21 0.27764 2 + 22 0.28580 1 - 23 0.29703 6 - 24 0.29837 3 + 25 0.29923 3 + 26 0.29979 1 - 27 0.30068 1 - 28 0.30074 6 - 29 0.30540 1 - 30 0.31270 5 - 31 0.31506 4 + 32 0.32431 4 - 33 0.32521 1 - 34 0.32860 6 + 35 0.33400 1 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-239 18.4349 0.7762 2.6850 0.3834 -0.8836 3.0329 Np-238 18.3685 0.0000 2.2742 0.3205 -0.9882 1.4160 Np-237 18.3022 0.7795 2.4371 0.3963 -0.9739 3.1569 Np-236 18.2358 0.0000 2.1332 0.2999 -0.7994 1.1664 Np-235 18.1694 0.7828 2.2924 0.3973 -0.9417 3.1303 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Np-239 6.250 0.800 5.250 0.600 Np-238 6.200 0.460 5.850 0.370 Np-237 6.000 0.950 5.570 0.600 Np-236 6.100 0.600 6.080 0.600 Np-235 6.250 0.950 5.630 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-239 20.6471 0.9056 2.6000 0.3273 -1.5058 2.9056 Np-238 20.5728 0.0000 2.6000 0.3280 -2.4114 2.0000 Np-237 21.9626 0.9094 2.6000 0.3162 -1.4591 2.9094 Np-236 22.2477 0.0000 2.6000 0.3139 -2.3586 2.0000 Np-235 20.3497 0.9133 2.6000 0.3299 -1.4981 2.9133 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-239 22.1219 0.9056 0.3200 0.3278 -0.5333 2.6056 Np-238 20.5728 0.0000 0.2800 0.3661 -1.7021 2.0000 Np-237 22.3287 0.9094 0.2400 0.3268 -0.5253 2.6094 Np-236 22.2477 0.0000 0.2000 0.3977 -2.2734 2.7000 Np-235 22.1666 0.9133 0.1600 0.3517 -0.7691 2.9133 -------------------------------------------------------- Table 7. Gamma-ray strength function for Np-239 -------------------------------------------------------- K0 = 1.300 E0 = 4.500 (MeV) * E1: ER = 10.98 (MeV) EG = 2.17 (MeV) SIG = 311.00 (mb) ER = 14.08 (MeV) EG = 4.66 (MeV) SIG = 540.00 (mb) * M1: ER = 6.61 (MeV) EG = 4.00 (MeV) SIG = 2.08 (mb) * E2: ER = 10.15 (MeV) EG = 3.24 (MeV) SIG = 6.65 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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