93-Np-237 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,S.Chiba,+ DIST-DEC21 20130704 ----JENDL-5 MATERIAL 9346 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-02 Fission cross section was evaluated with GMA code. 07-05 New calculation was made with CCONE code. 07-07 Re-calculation was made with CCONE code. 07-08 Fission cross section was revised. 07-11 Isomeric ratio of the (n,2n) reaction was given. 07-12 Resonance parameters 08-01 Fission cross section was revised. 08-02 Fission cross section and nu-p were revised. CCONE calculation was made with revised parameters. 08-03 Data were compiled as JENDL//AC-2008/1/. 09-03 Resonance parameters and fission cross section were modified. 09-08 (MF1,MT458) was evaluated. 10-01 Data of prompt gamma rays due to fission were given. 10-02 Covariance data were given. 13-07 (MF32,MT151) was corrected. (MF8,MT102) MATP was removed. 20-01 The resolved resonance parameters were replaced by N.Iwamoto. 21-10 JENDL-5b3 revised by N.Iwamoto (MF2/MT151) URP: derived with ASREP (MF3,MT1,2) recalculated (MF3,6/MT102) replaced (MF8/MT4,16,17,18,37,102) taken from JENDL/AD-2017 or added 21-10 JENDL-5b3 revised by N.Iwamoto (MF3/MT102) replaced into JENDL-4.0 and adjusted (MF2/MT151) URP: derived with ASREP (MF3,MT1,2) recalculated 21-11 revised by O.Iwamoto (MF3/MT19-21,38) deleted 21-11 above 20 MeV, JENDL-4.0/HE merged by O.Iwamoto 21-11 (MF6/MT5) recoil spectrum added by O.Iwamoto 21-12 (MF1/MT452,456) by O.Iwamoto revise nu-p (En > 20 MeV) by a systematics (MF6/MT5) by O.Iwamoto modify multiplicity of ZAP=1 to compensate the revision of nu-p MF= 1 MT=452 Total neutron per fission Sum of MT=455 and 456. MT=455 Delayed neutrons (same as JENDL-3.3) Nu-d was based on the experimental data of Saleh et al./2/, Charlton et al./3/, Piksaikin et al./4/ and Zeinalov et al./5/ Decay constants were detemined from the experimental data of Piksaikin et al./4/. MT=456 Prompt neutrons per fission The following data were fitted by GMA code /6/: Nu-p measured by Veeser/7/, Frehaut et al./8/, Malinovskii et al./9/, and nu-total by Boikov et al./10/, Thierens et al./11/, Mueller et al./12/ and Khokhlov et al./13/ 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/14/. 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 2000/15/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/16/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/17/ 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 500eV) The parameters given in JENDL-3.3 were modified: Parameters of the resonances below 3.9 eV were modified so as to reproduce the experimental data of Esch et al./18/ and Tovesson and Hill/19/, and the thermal capture cross section of 178.3 b and fission of 0.0202 b. Background cross section was given to the capture to get agreement with the average cross section of Esch et al./18/ Thermal cross sections at 0.0253 eV were based on: fission: Wagemans et al./20,21/, Kozharin et al./22/ capture: Kobayashi et al./23/, Katoh et al./24/, Harada et al./25/, Bringer et al./26/, Esch et al./18/, and others. In JENDL-5, the resonance parameters up to 109.1 eV were replaced with the data of Rovira et al./27/. The parameters of negative resonance were modified so as to reproduce the thermal capture cross section/28/. Unresolved resonance parameters (500eV - 30keV) Parameters were determined with ASREP code/29/ so as to reproduce the cross sections in the energy range from 500 eV to 30 keV. They are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ. (*) (barns) (barns) ------------------------------------------------------- Total 1.88148E+02 Elastic 1.42031E+01 Fission 2.05940E-02 6.69249E-01 n,gamma 1.73925E+02 6.86223E+02 ------------------------------------------------------- (*) 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/30/. The model parameters were determined by considering integral experimental data as well as measured cross-section data. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./31/ MT= 2 Elastic scattering cross section Calculated as total - non-elastic scattering cross sections. MT=16 (n,2n) cross section Calculated with CCONE code. The experimental data of Gromova et al./32/, Nishi et al./33/ and Landrum et al./34/ were used to determine the model parameters for the CCONE calculation. MT=18 Fission cross section The following experimental data reported after 1982 were analyzed with the GMA code /6/: Behrens+/35/, Cance+/36/, Alkhazov+/37/, Meadows/38/, Wu+/39/, Garlea+/40/, Goverdovskij+/41/, Goverdovskij+/42/, Zasadny+/43/, Goverdovskij+/44/, Kanda+/45/, Kovalenko+/46/, Alkhazov+/47/, Gul+/48/, Terayama+/49/, Meadows /50/, Desdin+/51/, Merla+/52/, Garlea+/53/, Shcherbakov+/54/, Furman+/55/, Baba+/56/, Tovesson+/57/. The data measured relatively to U235 fission were transformed to cross sections by using the U235 fission cross section of JENDL/AC-2008. The results of GMA were used to determine the parameters in the CCONE calculation. Further modification was made for JENDL-4.0 in the energy range from 500eV to 200keV, by eye-guiding the experimental data of Tovesson and Hill /19/. 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 Calculted with CCONE code. The experimental data of Esch et al./58/, Kobayashi et al./59/, Weston and Todd/60/, Linder et al./61/, and Buleeva et al./62/ 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/63/. MF= 6 Energy-angle distributions Calculated with CCONE code. Distributions from fission (MT=18) are not included. MF= 8 Radioactive Decay Decay chain will be given in the decay data file. MT=16 Decay data of Np-236: taken from ENDF (as of 2007) /64/ MF= 9 Multiplicities for Production of Radioactive Nuclides MT=16 Meta-stable state (T-1/2 =22.5H) production was assumed to be 70% from the cross sections measured around 14 MeV. 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./65/ 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 4% and 10 % was assumed below 2 MeV and above 6 MeV, respectively. Between 2 and 6 MeV, 7% was assumed./66/ MT=456 Covariance was obtained by GMA fitting to the experimental data (see MF1,MT456). Obtained standard deviation was multiplied by a factor of 3, because it was too small. MF=32 Covariances of resonance parameters Format of LCOMP=0 was adopted. Standard diviations of resonance parameters were taken from JENDL-3.3 covariance file /66/, which were estimated from errors reported in Refs./67,68,69,70/. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/71/ 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 500 eV), standard deviation (SD) of 8 % was added to the contributions from resonance parameters. Above 500 eV, covariances were obtaibed with CCONE and KALMAN codes, and experimental data. MT=18 Fission cross section SD of 4% was added in the energy region up to 1 eV, and 15% from 1 eV to 500 eV. Above 500 eV, covariances were obtained with GMA code/6/. SD was multiplied by a factor of 2.0. MT=102 Capture cross section Additional SD of 4% was given from 0.1 eV 500 eV. 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/30/ 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/72/ * Global parametrization of Koning-Duijvestijn/73/ was used. * Gamma emission channel/74/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/75/ 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/76/. 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/77/,/78/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,3,5,7 (see Table 2) * optical potential parameters /31/ Volume: V_0 = 49.8581 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 = 1.25 fm a_v = 0.57 fm Surface: W_0 = 17.1839 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 = 0.212764 beta_4 = 0.066 beta_6 = 0.0015 * Calculated strength function S0= 0.99e-4 S1= 2.49e-4 R'= 9.76 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Np-237 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 5/2 + * 1 0.03320 7/2 + * 2 0.05954 5/2 - 3 0.07592 9/2 + * 4 0.10296 7/2 - 5 0.13000 11/2 + * 6 0.15851 9/2 - 7 0.19146 13/2 + * 8 0.22596 11/2 - 9 0.26754 3/2 - 10 0.26990 15/2 + 11 0.28135 1/2 - 12 0.30506 13/2 - 13 0.31680 7/2 + 14 0.32442 7/2 - 15 0.33236 1/2 + 16 0.34850 17/2 + 17 0.35970 5/2 - 18 0.36859 5/2 + 19 0.37093 3/2 + 20 0.39552 15/2 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-238 18.3635 0.0000 2.2742 0.3206 -0.9887 1.4167 Np-237 18.2971 0.7795 2.4371 0.3964 -0.9743 3.1574 Np-236 18.2307 0.0000 2.1332 0.3000 -0.7998 1.1669 Np-235 18.1643 0.7828 2.2924 0.3974 -0.9420 3.1307 Np-234 18.0979 0.0000 2.1332 0.2845 -0.6773 1.0000 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Np-238 6.199 0.460 5.848 0.370 Np-237 6.250 0.950 5.200 0.600 Np-236 5.500 0.600 5.200 0.400 Np-235 6.250 0.950 5.200 0.600 Np-234 6.200 0.460 5.850 0.370 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-238 20.5728 0.0000 2.6000 0.3280 -2.4114 2.0000 Np-237 20.4984 0.9094 2.6000 0.3286 -1.5020 2.9094 Np-236 17.3240 0.0000 2.6000 0.3613 -2.5262 2.0000 Np-235 20.3497 0.9133 2.6000 0.3299 -1.4981 2.9133 Np-234 20.2753 0.0000 2.6000 0.3306 -2.4114 2.0000 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Np-238 20.5728 0.0000 0.2800 0.3661 -1.7021 2.0000 Np-237 20.4984 0.9094 0.2400 0.3674 -0.7921 2.9094 Np-236 17.3240 0.0000 0.2000 0.4067 -1.7431 2.0000 Np-235 20.3497 0.9133 0.1600 0.3699 -0.7870 2.9133 Np-234 20.2753 0.0000 0.1200 0.3712 -1.6997 2.0000 -------------------------------------------------------- Table 7. Gamma-ray strength function for Np-238 -------------------------------------------------------- K0 = 1.300 E0 = 4.500 (MeV) * E1: ER = 10.98 (MeV) EG = 2.17 (MeV) SIG = 309.44 (mb) ER = 14.08 (MeV) EG = 4.66 (MeV) SIG = 540.00 (mb) * M1: ER = 6.62 (MeV) EG = 4.00 (MeV) SIG = 2.08 (mb) * E2: ER = 10.17 (MeV) EG = 3.25 (MeV) SIG = 6.65 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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