Retrieved by E4-util: 2010/02/01,11:45:31 1 0 0 0 9.424000+4 2.379920+2 1 1 59 09440 1451 1 0.000000+0 1.000000+0 0 0 0 69440 1451 2 1.000000+0 2.0000E+07 0 0 10 09440 1451 3 0.000000+0 0.000000+0 0 0 541 739440 1451 4 94-Pu-240 IPPE EVAL-DEC02 A.V.IGNATYUK ET AL. 9440 1451 5 RUSFOND MANUAL DIST-DEC06 9440 1451 6 ----RUSFOND MATERIAL 9440 9440 1451 7 -----INCIDENT NEUTRON DATA 9440 1451 8 -------ENDF-6 FORMAT 9440 1451 9 9440 1451 10 RUSFOND DESCRIPTION 9440 1451 11 (See RUSFOND manual for details) 9440 1451 12 9440 1451 13 DATA from BROND-3 are adopted with the next exclusions 9440 1451 14 1. Upper energy is restricted to 20 MeV 9440 1451 15 2. 8 group DN data are included from JEFF-3.1 in MF=1, MT=455 9440 1451 16 3. DN spectra the same as for U-235 are adopted with 8 group 9440 1451 17 yields from JEFF-3.1(MF=5, MT=455) 9440 1451 18 4. MT=5 in MF=3 and MF=6 removed 9440 1451 19 5. Decay data in MF=8, MT=16, 17, 102 are removed 9440 1451 20 6. Fission product yields from JEFF-3.1 is included in MF=8 9440 1451 21 MT=454 and MT=459 9440 1451 22 7. MF6 was splited by SIXPAK + MF4,5 MT18 from VII.0 9440 1451 23 C.mpiled by M.N.Nikolaev JUN 2006 9440 1451 24 9440 1451 25 BROND-3 Description 9440 1451 26 9440 1451 27 NEUTRON CROSS SECTION EVALUATIONS for Pu-240 9440 1451 28 at NEUTRON ENERGIES up to 150 MeV. 9440 1451 29 9440 1451 30 E.V. Gai, A. V. Ignatyuk, V. P. Lunev, Yu. N. Shubin, 9440 1451 31 M.V.Skripova, N. N. Titarenko 9440 1451 32 9440 1451 33 INTRODUCTION 9440 1451 34 To develop main concepts of the accelerator-driven 9440 1451 35 power systems and the corresponding nuclear waste 9440 1451 36 management, it is necessary to have sufficiently complete 9440 1451 37 and reliable nuclear data on spectra and reaction cross 9440 1451 38 sections for structural materials, fissioning actinides, and 9440 1451 39 the most important fission products in a very broad energy 9440 1451 40 range. In practice, the energy interval from thermal 9440 1451 41 energies up to a few thousands megaelectron-volts should be 9440 1451 42 covered [1]. The status of available nuclear data differs 9440 1451 43 strongly for the energy regions below and above 20 MeV. Huge 9440 1451 44 efforts have been made to create libraries of evaluated neutron 9440 1451 45 data (ENDF/B, JENDL, BROND, etc.) for the low-energy region. 9440 1451 46 In spite of some differences between the evaluations, most data 9440 1451 47 are reasonable enough, and their accuracies satisfy requests 9440 1451 48 of major current applications. For energies above 20 MeV, 9440 1451 49 data are rather scarce and are not yet systematised. 9440 1451 50 A lack of experimental data has to be compensated by 9440 1451 51 the development of reliable calculation methods. The codes 9440 1451 52 based on the intranuclear cascade model combined with the 9440 1451 53 evaporation model have been successfully applied for 9440 1451 54 energies above a few hundred megaelectron-volts [2-4]. At 9440 1451 55 lower energies, however, nuclear structure effects are so 9440 1451 56 prominent that their description requires more detailed 9440 1451 57 consideration of competitive reaction mechanisms. For this 9440 1451 58 reason it was decided that the energy region from 20 to 150 9440 1451 59 MeV requires special consideration, and the evaluated data 9440 1451 60 files for this region should be prepared for the most 9440 1451 61 important structural and fissile materials in the same 9440 1451 62 manner as for the energy region below 20 MeV [1]. In 9440 1451 63 accordance with that, the evaluated data files for about 30 9440 1451 64 of most important structural and shielding materials were 9440 1451 65 extended in the ENDF/B-VI library up to 150 MeV by the Los 9440 1451 66 Alamos group [5]. The similar evaluations for the most 9440 1451 67 important fissile isotopes Th-232, U-238 and Pu-239 have been 9440 1451 68 performed at the Institute of Physics and Power Engineering 9440 1451 69 [6-8]. 9440 1451 70 The main results of experimental data analysis and 9440 1451 71 calculations aimed at the creation of the complete file of 9440 1451 72 neutron data for Pu-240 are briefly discussed below. 9440 1451 73 9440 1451 74 INCIDENT NEUTRON ENERGIES BELOW 20 MeV 9440 1451 75 9440 1451 76 The BROND-2 file for Pu-240 has been prepared by the 9440 1451 77 Minsk group in 1980 [9]. During past years new experimental 9440 1451 78 data on the total neutron cross sections, fission cross 9440 1451 79 sections and fission neutron multiplicities have appeared, 9440 1451 80 and the theoretical models used for the neutron cross 9440 1451 81 section calculations have been improved essentially. Also 9440 1451 82 updated evaluations of the recommended data for plutonium 9440 1451 83 isotopes have been performed at other laboratories, which 9440 1451 84 are included in the last versions of ENDF/B-VI, JEFF-3.0 and 9440 1451 85 JENDL-3.3 libraries. Nowadays a new version of the BROND-3 9440 1451 86 library is under formation in the Russian Nuclear Data 9440 1451 87 Center, which should include the more precise neutron cross 9440 1451 88 sections for the main fuel elements, structural and 9440 1451 89 technological materials of nuclear reactors. 9440 1451 90 The following improvements have been made for the Pu-240 9440 1451 91 file: 9440 1451 92 1. The resolved resonance parameters were adopted from 9440 1451 93 the JENDL-3.3 evaluation, which is based on the analysis 9440 1451 94 performed recently by the ORNL group [10]. The capture widths 9440 1451 95 for 3 and 1.056 eV resonances were slightly modified in 9440 1451 96 JENDL-3.3 to reproduce the thermal capture cross section 9440 1451 97 better and the upper boundary 2.7 keV was chosen, because the 9440 1451 98 capture cross section, estimated by Bouland et al. [10], was 9440 1451 99 too small above this energy. The energy dependent parameters 9440 1451 100 for unresolved resonances (2.7 to 42.98 keV) were taken from 9440 1451 101 the JENDL-3.3 file too. They were determined up to d-wave 9440 1451 102 resonances to reproduce the available experimental data on 9440 1451 103 the capture and fission cross sections. 9440 1451 104 2. The precise measurements of the total neutron cross 9440 1451 105 sections for U-238 have been performed [11] during the last 9440 1451 106 years. These data enable to determine reliably the 9440 1451 107 parameters of the generalized optical potential using which 9440 1451 108 one can calculate the total neutron cross sections for the 9440 1451 109 neighboring nuclei. Because of small differences of the 9440 1451 110 total cross sections for the neighboring nuclei, such 9440 1451 111 estimation of the total cross sections could be more 9440 1451 112 reliable than the direct description of earlier experimental 9440 1451 113 data, whose uncertainties are essentially larger. The 9440 1451 114 corresponding optical-model calculations have been adopted 9440 1451 115 as a new BROND-3 evaluation. In accordance with the changes 9440 1451 116 of the total cross sections the integral elastic and 9440 1451 117 inelastic neutron scattering cross sections have been 9440 1451 118 changed too. The angular distributions of elastic neutron 9440 1451 119 scattering, however, have been taken from the previous BROND- 9440 1451 120 2 file, because a rather small differences between the old 9440 1451 121 and new results. 9440 1451 122 3. There are numerous experimental data on the fission 9440 1451 123 cross sections and all evaluations of them are agree quite 9440 1451 124 well. The experimental data obtained during the last decade 9440 1451 125 [12,13] confirms the previous evaluations up to the neutron 9440 1451 126 energy of 15 MeV, and some improvements of recommended cross 9440 1451 127 sections are required for higher energies only. 9440 1451 128 Nevertheless, we performed the complete statistical analysis 9440 1451 129 of all available experimental data on the basis of the new 9440 1451 130 method of unrecognized errors estimations. As a result the 9440 1451 131 uncertainties of evaluated fission cross sections and the 9440 1451 132 corresponding covariance matrix was determined in a 9440 1451 133 consistent manner for the BROND-3 files. An important 9440 1451 134 feature of the covariance matrix obtained is rather large 9440 1451 135 values of remote nondiagonal elements reflecting the 9440 1451 136 correlations of recommended data at broad energy intervals. 9440 1451 137 4. The improvements of radiative capture and inelastic 9440 1451 138 scattering cross sections connected with the theoretical 9440 1451 139 model development are comparatively small and they are 9440 1451 140 related to the neutron energies above 1 MeV. On the whole 9440 1451 141 the differences between the new evaluation and the ENDF/B-VI 9440 1451 142 or JENDL-3.3 are considerably less than for the BROND-2. The 9440 1451 143 uncertainties and the corresponding covariance matrix for 9440 1451 144 radiative capture cross sections are evaluated on the basis 9440 1451 145 of the new approach. To some extent the (n,2n) and (n,3n) 9440 1451 146 cross sections have been corrected in accordance with the 9440 1451 147 fission and absorption cross section re-evaluations. 9440 1451 148 5. The re-evaluation of secondary fission neutrons has 9440 1451 149 been performed. The new evaluation is based on the optimal 9440 1451 150 fitting of the available experimental data including the 9440 1451 151 recent measurements [14]. The fission neutron spectra are 9440 1451 152 calculated on the basis of the Madland-Nix model [15]. The 9440 1451 153 agreement of the new evaluation with the ENDF/B-VI and JENDL- 9440 1451 154 3.3 is rather good for the energy region corresponding to 9440 1451 155 the first plateau of fission cross section. For higher 9440 1451 156 energies the decrease of the average neutron energies due to 9440 1451 157 the multi-chance fission was taken into account. This effect 9440 1451 158 was neglected in the previous evaluations. 9440 1451 159 6. The gamma-ray spectra and production cross sections 9440 1451 160 were re-evaluated on the basis of the consistent statistical 9440 1451 161 model. The estimated integral gamma-ray yields are rather 9440 1451 162 similar to the ENDF/B-VI evaluations, but there are some 9440 1451 163 regular differences between the evaluated gamma-ray spectra, 9440 1451 164 especially, at high energies. It should be noted that in the 9440 1451 165 JENDL-3.3 files there are no data on the gamma-ray yields, 9440 1451 166 and the JEFF-3.0 file contains the evaluations of gamma-ray 9440 1451 167 yields accompanying the inelastic scattering and (n,xn) 9440 1451 168 reactions only, but there are no data on gamma-rays from 9440 1451 169 fission, which dominate in the integral yields of gamma- 9440 1451 170 rays. 9440 1451 171 9440 1451 172 INCIDENT NEUTRON ENERGIES ABOVE 20 MeV 9440 1451 173 9440 1451 174 Evaluations above 20 MeV are based on nuclear model 9440 1451 175 calculations, whose parameters are to be adjusted on the 9440 1451 176 available experimental data. The coupled-channel optical 9440 1451 177 model realized in the ECIS-87 code [16] was used to 9440 1451 178 calculate the neutron and proton transmission coefficients 9440 1451 179 needed for the statistical model calculations of all 9440 1451 180 competitive reaction cross sections. Besides that the 9440 1451 181 coupled-channel model was also used for the evaluations of 9440 1451 182 the neutron total cross sections (MF=3, MT=1), the integral 9440 1451 183 elastic and inelastic scattering cross sections (MF=3, MT=2, 9440 1451 184 51, 52, 53), and the corresponding neutron angular 9440 1451 185 distributions (MF=4, MT=2, 51, 52, 53). 9440 1451 186 The GNASH code [17] was used to calculate the integral 9440 1451 187 and double-differential cross sections for secondary 9440 1451 188 neutrons, charged particles, and gamma rays and to prepare 9440 1451 189 data in ENDF/B-VI format. The level density description for 9440 1451 190 all reaction channels was obtained with the Gilbert-Cameron 9440 1451 191 approach fitted to experimental data on the density of low- 9440 1451 192 lying discrete levels and neutron resonances. The 9440 1451 193 phenomenological systematics of the level density parameters 9440 1451 194 developed in the framework of the RIPL project [18] were 9440 1451 195 applied to the fission channels and to nuclei for which the 9440 1451 196 corresponding experimental data does not exist. 9440 1451 197 9440 1451 198 Total and Scattering Cross Sections 9440 1451 199 9440 1451 200 Evaluations of neutron total cross sections are usually 9440 1451 201 based on coupled-channel optical model calculations with 9440 1451 202 potential parameters fitted to experimental data. The 9440 1451 203 analysis of such data was performed in many laboratories, 9440 1451 204 and the deformed optical model parameters obtained were used 9440 1451 205 for the neutron cross-section evaluations of heavy nuclei 9440 1451 206 and actinides [19-20]. These sets of parameters give 9440 1451 207 approximately the same total neutron cross sections. 9440 1451 208 However, the calculated neutron absorption cross sections 9440 1451 209 differ significantly for various sets. Discrepancies of the 9440 1451 210 absorption cross section evaluations are essential at 9440 1451 211 neutron energies above 10 MeV, and their effects appear in 9440 1451 212 the evaluated cross sections of partial (n,xn), fission and 9440 1451 213 other reactions. 9440 1451 214 The precise experimental data on the neutron total 9440 1451 215 cross sections for Th and U in the energy region 5 - 500 MeV 9440 1451 216 [11] allow us to estimate more accurately the optimal 9440 1451 217 potential parameters for all actinides. These parameters are 9440 1451 218 given in Table 1. 9440 1451 219 9440 1451 220 Table 1. Deformed optical potential parameters for 238U 9440 1451 221 determined from the neutron total cross section analysis *) 9440 1451 222 ----------------------------------------------------------------- 9440 1451 223 Potential depth (MeV) Energy Geometry 9440 1451 224 region (MeV) parameters (fm) 9440 1451 225 ----------------------------------------------------------------- 9440 1451 226 Vr = 49.70+-16.0tau 060 rr = 1.185, ar = 0.80 9440 1451 233 -.230E +0.00024E**2+Dcoul) 9440 1451 234 9440 1451 235 Wd = 3.660+-8.0tau+.750E 0