JEFF-3.1 General Purpose Neutron File, May 2005. 2310 0 0 0 9.423900+4 2.369984+2 1 1 2 09437 1451 1 0.000000+0 0.000000+0 0 0 0 69437 1451 2 1.000000+0 3.000000+7 0 0 10 319437 1451 3 0.000000+0 0.000000+0 0 0 855 829437 1451 4 94-Pu-239 BRC,CAD,+ EVAL: ROMAIN, MORILLON, DOSSANTOS-UZARRALDE 9437 1451 5 DIST-MAY05 REV1-MAY05 20050504 9437 1451 6 ----JEFF-31 MATERIAL 9437 9437 1451 7 -----INCIDENT NEUTRON DATA 9437 1451 8 ------ENDF-6 FORMAT 9437 1451 9 ***************************** JEFF-3.1 *************************9437 1451 10 ** **9437 1451 11 ** Original data taken from: JEFF-3.0 **9437 1451 12 ** **9437 1451 13 ******************************************************************9437 1451 14 9437 1451 15 05-01 NEA/OECD (Rugama) 8 delayed neutron groups 9437 1451 16 Jefdoc-976(Spriggs,Campbel and Piksaikin,Prg Nucl Eng 41,223(2002)9437 1451 17 9437 1451 18 2003-06 CAD (Dupont) Unresolved Resonance Parameters 9437 1451 19 (MF=2,MT=151,LRU=2) for L=1 and AJ=1.0 resonances: 9437 1451 20 AMUN changed from 1. to 2. AND GN0 divided by 2. 9437 1451 21 9437 1451 22 ***************************** JEFF-3.0 ***********************9437 1451 23 9437 1451 24 NEW evaluation 9437 1451 25 9437 1451 26 This evaluation is built from contributions of several individuals9437 1451 27 in various laboratories. 9437 1451 28 ** BRC : J.P. Delaroche, P. Dossantos-Uzarralde, S. Hilaire, 9437 1451 29 C. Le luel, M. Lopez-Jimenez, P. Morel, B. Morillon, 9437 1451 30 P. Romain. 9437 1451 31 ** CAD : E. Dupont, E. Fort, O. Serot, J-Ch Sublet. 9437 1451 32 ** + : H. Derrien, T. Nakagawa. 9437 1451 33 *************************************************************** 9437 1451 34 MF=1 Descriptive and Nubar Information ************************ 9437 1451 35 *************************************************************** 9437 1451 36 MT=452: Number of neutrons per fission 9437 1451 37 Total Nubar. Sum of MT=455 and 456. 9437 1451 38 9437 1451 39 MT=455: Delayed nubar evaluation (from WPEC/SG 6) 9437 1451 40 See JEF/DOC-920 9437 1451 41 Energy dependent delayed neutron spectrum introduced 9437 1451 42 9437 1451 43 MT=456: Prompt nubar evaluation (E.Fort and B.Morillon) 9437 1451 44 The evaluation below 650 eV is based on experimental 9437 1451 45 data [30]. 9437 1451 46 From 650 eV to 30 MeV, the adopted values are obtained 9437 1451 47 from the Los Alamos model upgraded by G.Vladuca and 9437 1451 48 A.Tudora (multiple fission chances included) [31]. 9437 1451 49 The model parameters are slightly different from those 9437 1451 50 adopted in [31]. 9437 1451 51 9437 1451 52 MT=458: Energy release due to fission (O.Serot and B.Morillon). 9437 1451 53 The kinetic energy of the fragments results from a 9437 1451 54 compilation of recent experimental measurements. 9437 1451 55 The kinetic energy of the prompt fission neutrons is 9437 1451 56 consistent with the nup-value and the average energy 9437 1451 57 deduced from prompt fission spectra. The energy 9437 1451 58 released by the emission of prompt gamma rays is 9437 1451 59 obtained from the systematics proposed by Frehaut. 9437 1451 60 The components of the prompt energy release are 9437 1451 61 consistent with the nup calculations performed with 9437 1451 62 a model similar to Madland's. (JEFDOC xxx) 9437 1451 63 Actually they are all energy dependent but ENDF format 9437 1451 64 does not allow for such representation. 9437 1451 65 9437 1451 66 9437 1451 67 ******************************************************************9437 1451 68 MF=2 9437 1451 69 ******************************************* 9437 1451 70 PU239 RESONANCE DATA 0 keV TO 2.5 keV 9437 1451 71 Principal evaluators: H.Derrien, T.Nakagawa 9437 1451 72 ******************************************* 9437 1451 73 Resonance region evaluation by H.Derrien and T. Nakagawa 9437 1451 74 discussed below. This evaluation extended resonance region to 9437 1451 75 2.5 keV. 9437 1451 76 The present file contains the resonance parameters obtained 9437 1451 77 from a SAMMY fit analysis of high resolution experimental data, 9437 1451 78 performed at ORNL (Oak Ridge Nationnal Laboratory, USA) by 9437 1451 79 H.Derrien and G.De Saussure and at JAERI (Tokai-Mura Research 9437 1451 80 Establishment, Japan) by H.Derrien and T.Nakagawa. 9437 1451 81 The file contains three independant sections: 9437 1451 82 1) the first corresponds to the energy range 0 keV to 1 keV. 9437 1451 83 The corresponding set of resonance parameters contains 398 9437 1451 84 resonances in the energy range 0 keV to 1 keV, 4 ficticious 9437 1451 85 negative energy resonances and 3 ficticious resonances above 9437 1451 86 1 keV; 9437 1451 87 2) the second corresponds to the energy range 1 keV to 2 keV. 9437 1451 88 The corresponding set of resonance parameters contains 435 9437 1451 89 resonances in the energy range 0.980 keV to 2.02 keV, 3 9437 1451 90 ficticious resonances below 0.9 keV and 3 ficticious resonances 9437 1451 91 above 2.02 keV; 9437 1451 92 3) the third corresponds to the energy range 2 keV to 2.5 keV. 9437 1451 93 The corresponding set of resonance parameters contains 218 9437 1451 94 resonances in the energy range 1.98 keV to 2.53 keV, 3 9437 1451 95 ficticious resonances below 1.98 keV and 3 ficticious resonances9437 1451 96 above 2.53 keV. 9437 1451 97 In all sections the ficticious resonance parameters take 9437 1451 98 into account the contribution of all the external truncated 9437 1451 99 resonances in such a way that no total, scattering, fission and 9437 1451 100 capture smooth files are needed in the corresponding energy 9437 1451 101 ranges for the reproduction of the cross sections within the 9437 1451 102 experimental errors. 9437 1451 103 The following experimental data base has been used in the 9437 1451 104 SAMMY fits: 9437 1451 105 - absorption and fission from R. Gwin et al. [1,4]; 9437 1451 106 - fission from R. Gwin et al. [5,7], J. Blons [3], L.W. Weston 9437 1451 107 et al. [8,15]; 9437 1451 108 - transmission from R.R. Spencer et al. [10], J.A. Harvey et al.9437 1451 109 [9]. 9437 1451 110 Prior to the fits the experimental fission and absorption cross 9437 1451 111 sections were normalised,directly or indirectly to the 0.0253 eV 9437 1451 112 values obtained by the ENDF/B-VI standard evaluation group [11]. 9437 1451 113 The transmission data were considered as accurate absolute 9437 1451 114 measurements (R.R.Spencer total cross section at 0.0253 eV is 9437 1451 115 1025.0 b in excellent agreement with the 1027.3 b standard value)9437 1451 116 Details on the analysis are found in [14],[16],[17]. 9437 1451 117 9437 1451 118 ---------------------------------------------------------------- 9437 1451 119 COMMENTS ON THE THERMAL AND LOW ENERGY RANGES 9437 1451 120 9437 1451 121 The thermal cross-section values calculated at 293 K by the 9437 1451 122 resonance parameters of the first section are given in the 9437 1451 123 following table at 293 K and in barns. 9437 1451 124 9437 1451 125 SAMMY RESENDD Proposed standard [11] 9437 1451 126 ------- -------- ---------------------- 9437 1451 127 Fission 747.64 747.90 747.99+-1.87 9437 1451 128 Capture 271.10 270.73 271.43+-2.14 9437 1451 129 Scattering 7.97 7.99 7.88+-0.97 9437 1451 130 ------- -------- ---------------------- 9437 1451 131 Total 1026.71 1026.62 1027.30+-5.00 9437 1451 132 9437 1451 133 One should note that the 293 K cross sections calculated at 9437 1451 134 0.0253 eV depend on the way the Doppler broadening calculation 9437 1451 135 is performed. For instance using a Gaussian broadening function 9437 1451 136 will give a fission cross section about 2.5 barns larger than the9437 1451 137 one obtained from the accurate calculation which conserves the 9437 1451 138 1/v shape of the thermal cross section. The values given in the 9437 1451 139 table above were obtained from SAMMY (Leal-Hwang method) [13,18] 9437 1451 140 and from RESENDD with 0.1% for the interpolation accuracy [20]. 9437 1451 141 The following table shows experimental cross sections 9437 1451 142 averaged over the energy ranges 0.02 eV to 0.06eV and 0.02 eV 9437 1451 143 to 0.65 eV, compared to the calculated values: 9437 1451 144 9437 1451 145 References Average Cross Sections (barns) 9437 1451 146 [1-10] 0.02 - 0.06 eV 0.02 - 0.65 eV 9437 1451 147 ---------------- ----------------------- ----------------------- 9437 1451 148 Exp Calc (293K) Exp Calc (293K) 9437 1451 149 Gwin71 fiss 631.41 843.71 9437 1451 150 Gwin76 fiss 631.41 838.39 9437 1451 151 Gwin84 fiss(*) 631.41 631.75(+0.05%) 837.18 838.69(+0.18%) 9437 1451 152 Deruyter70 fiss 631.41 859.43 9437 1451 153 Wagemans80 fiss 631.41 862.56 9437 1451 154 Wagemans88 fiss 631.41 841.80 9437 1451 155 Gwin71 capture 243.84 243.22(-0.25%) 524.75 518.13(-1.26%) 9437 1451 156 Gwin76 absorpt(*) 875.90 874.29(-0.18%) 1359.96 1357.14(-0.21%) 9437 1451 157 Spencer84 tot(*) 883.20 882.86(-0.04%) 1361.69 1367.6 (+0.43%) 9437 1451 158 ----------------- ---------------------- ----------------------- 9437 1451 159 (*)These data had the largest weight in the thermal fit. The 9437 1451 160 values between the parentheses give the percentage deviation 9437 1451 161 between the calculated data and the experimental data. 9437 1451 162 9437 1451 163 The value of 631.4 barns for all the averaged experimental 9437 1451 164 fission cross sections in the energy range 0.02 eV to 0.06 eV 9437 1451 165 corresponds to the renormalisation of the fission experiments to 9437 1451 166 748.0+-1. barns at 0.0253 eV. ORNL data are consistent within 9437 1451 167 0.8% over the energy range 0.02 eV to 0.65 eV (i.e. over the 0.3 9437 1451 168 eV resonance). Deruyter70 and Wagemans80 data are about 9437 1451 169 2.5% larger and were not included in the SAMMY fit. 9437 1451 170 When normalised on the standard value at 0.0253 eV, Gwin 76 9437 1451 171 absorption agrees with the absorption obtained from Spencer total9437 1451 172 cross section within 0.7% over the 0.3 eV resonance. The present 9437 1451 173 evaluation is essentially the result of a consistent SAMMY 9437 1451 174 analysis of all the available ORNL data with a larger weight on 9437 1451 175 Gwin 1984 fission, Gwin 1976 absorption and Spencer transmission 9437 1451 176 data. 9437 1451 177 After renormalisation of the calculated fission cross section9437 1451 178 on the preliminary 1991 Weston and Todd fission data (see next 9437 1451 179 section) a slight adjustment of the negative resonance parameters9437 1451 180 was performed to keep the values calculated at 0.0253 eV in close9437 1451 181 agreement with the standard values. The 1988 data of Wagemans et 9437 1451 182 al.[21] agree within 0.4% with the calculated values over the 9437 1451 183 energy range from 0.02 eV to 0.65 eV after adjustment of the 9437 1451 184 energy scale to the ORNL scale (the difference was 0.27 eV at 9437 1451 185 20 eV between 1988 Wagemans and ORNL SAMMY fit energy scales). 9437 1451 186 9437 1451 187 ---------------------------------------------------------------- 9437 1451 188 COMMENTS ON THE 0 keV TO 1 keV ENERGY RANGE. 9437 1451 189 9437 1451 190 At the end of 1987, an analysis was completed up to 1 keV. 9437 1451 191 In a preliminary step, a correlated fit of Harvey transmission 9437 1451 192 data, Weston 84 fission data, and Blons fission data was 9437 1451 193 performed with possible adjustment of the normalisation 9437 1451 194 coefficients and of the background corrections. This preliminary 9437 1451 195 step has shown that this adjustment was not necessary to achieve 9437 1451 196 consistency between Harvey data and Weston data. The Blons data 9437 1451 197 needed a large readjustment of the background and normalisation. 9437 1451 198 Therefore, the final fit was performed only on the Harvey 9437 1451 199 transmission data, Gwin 84 fission data (below 30 eV), and 9437 1451 200 Weston 84 fission data, with no background and normalisation 9437 1451 201 adjustment. Blons data, which have better resolution than Weston 9437 1451 202 84 data, were used only to obtain more accurate fission widths 9437 1451 203 of some narrow resonances in the high energy range. 9437 1451 204 In 1989, preliminary results of the 1988 Weston fission 9437 1451 205 measurement [15] were included in the SAMMY experimental data 9437 1451 206 base. One expected from this measurement, which was performed by 9437 1451 207 using a 86-m flight path with a resolution comparable to that of 9437 1451 208 Harvey transmission, a confirmation of the excellent quality of 9437 1451 209 the 1984 measurement. A consistent SAMMY fit of Harvey transmis- 9437 1451 210 sion, Weston 84 fission and preliminary Weston 88 fission was re-9437 1451 211 started from the parameter and covariance files obtained in 1987.9437 1451 212 It appeared that large background and normalisation corrections 9437 1451 213 were needed on the new Weston fission data to obtain consistency 9437 1451 214 with Harvey transmission data. These corrections were comparable 9437 1451 215 to those found in Blons data and were not understood by the 9437 1451 216 authors of the experiment. The last SAMMY runs were performed by 9437 1451 217 not allowing background and normalisation variations on Harvey 9437 1451 218 transmission and Weston 84 fission (very small error bars were 9437 1451 219 assigned to the corresponding parameters in the covariance 9437 1451 220 matrix) and by allowing these variations on Weston 88 data. A 9437 1451 221 new set of resonance parameters was obtained, which was improved 9437 1451 222 compared to the previous set due to the very high resolution of 9437 1451 223 the new Weston fission measurement. 9437 1451 224 The calculated average fission cross section in the energy 9437 1451 225 range from 0.1 keV to 1.0 keV was 3.7% smaller than the values 9437 1451 226 obtained by the ENDF/B-VI standard evaluation group due to the 9437 1451 227 fact that Weston 84 data were 3.1% lower than the average 9437 1451 228 standard value. A new measurement was performed by Weston and 9437 1451 229 Todd in 1991 [22] in order to check their 1984 data. A careful 9437 1451 230 normalisation of the data in the thermal energy range showed 9437 1451 231 that the 1984 data should be renormalised by about +3%. To take 9437 1451 232 into account this renormalisation, the 1989 resonance parameters 9437 1451 233 were modified at JAERI [17] in the following way: 9437 1451 234 1) increase of the fission width by 3% and decrease of the 9437 1451 235 capture width by a quantity equal to the variation of the 9437 1451 236 fission width in the narrow resonances(mainly 1+ resonances); 9437 1451 237 that does not modify the total cross section in the correspond- 9437 1451 238 ing resonances; 9437 1451 239 2) adjustment of the neutron width of the 0+ resonances by a 9437 1451 240 refit of the transmission data and of the renormalised Weston 9437 1451 241 and Todd 1984 data in energy ranges where the contribution of 9437 1451 242 the 0+ resonances is dominant, and increase of the other(small) 9437 1451 243 0+ neutron widths by 3%. No severe inconsistency was observed 9437 1451 244 between the transmission data and the new fission data over the 9437 1451 245 dominant 0+ resonances; the differences between the 1989 fits of 9437 1451 246 the transmission and the new fits were consistent within the 9437 1451 247 experimental error bars. 9437 1451 248 The following table shows the fission cross sections calculat-9437 1451 249 ed from the resonance parameters, the experimental values and 9437 1451 250 the results of the ENDF/B-VI standard evaluation group averaged 9437 1451 251 in the same energy intervals. Weston 1991 data are preliminary. 9437 1451 252 Weston 1984 data are normalised on preliminary Weston 1991. 9437 1451 253 9437 1451 254 Energy Cross Sections (barn) 9437 1451 255 (eV) Calc. Weston 1991 Weston 1984 Standard 9437 1451 256 ---------- ------ ----------- ----------- -------- 9437 1451 257 0.010-10. 80.12 79.98 9437 1451 258 9-20 94.74 94.91 9437 1451 259 20-40 17.52 17.76 17.97 9437 1451 260 40-60 50.64 50.90 50.87 9437 1451 261 60-100 54.42 54.38 54.33 9437 1451 262 100-200 18.63 18.59 18.56 18.66 9437 1451 263 200-300 17.85 17.89 17.88 9437 1451 264 300-400 8.31 8.34 8.43 9437 1451 265 400-500 9.59 9.58 9.57 9437 1451 266 ---------- ------ ----------- ----------- -------- 9437 1451 267 200-500 11.92 11.93 11.93 11.96 9437 1451 268 ---------- ------ ----------- ----------- -------- 9437 1451 269 500-600 15.39 15.57 15.86 9437 1451 270 600-700 4.37 4.30 4.46 9437 1451 271 700-800 5.51 5.53 5.63 9437 1451 272 800-900 4.84 4.89 4.98 9437 1451 273 900-1000 8.33 8.38 8.30 9437 1451 274 ---------- ------ ----------- ----------- -------- 9437 1451 275 500-1000 7.69 7.73 7.73 7.79 9437 1451 276 ---------- ------ ----------- ----------- -------- 9437 1451 277 20-1000 13.09 13.11 13.11 9437 1451 278 -------------------------------------------------------- 9437 1451 279 9437 1451 280 Gwin 1971 and 1976 absorption data were not included in the 9437 1451 281 SAMMY fit in the energy range above 1 eV. Accurate absorption 9437 1451 282 cross sections should be calculated from the parameters obtained 9437 1451 283 from the analysis of the transmission and fission data. The 9437 1451 284 following table shows the calculated average values of the 9437 1451 285 capture, absorption and alpha compared to Gwin 1971 and Gwin 9437 1451 286 1976 data. The calculations were performed with RESENDD, 1% 9437 1451 287 accuracy. 9437 1451 288 9437 1451 289 Energy (eV) Cross Sections (barn) 9437 1451 290 calc. values (293 K) Gwin data 9437 1451 291 ------------ --------------------- ------------------ 9437 1451 292 CAPT ABSORP ALPHA ABSORP ALPHA 9437 1451 293 7.3- 16.0 76.61 196.04 0.64 208.00 0.74(*) 9437 1451 294 16.0- 37.5 20.51 44.55 0.85 46.50 0.89(*) 9437 1451 295 37.5- 50.0 48.72 70.00 2.29 83.15 2.96(*) 9437 1451 296 50.0-100.0 33.60 92.13 0.57 92.84 0.63 9437 1451 297 100.0-200.0 15.58 34.29 0.83 33.66 0.87 9437 1451 298 200.0-300.0 15.85 33.68 0.89 34.69 0.94 9437 1451 299 300.0-400.0 9.69 18.01 1.16 18.31 1.16 9437 1451 300 400.0-500.0 3.96 13.56 0.41 13.56 0.44 9437 1451 301 500.0-600.0 10.87 26.30 0.70 26.54 0.72 9437 1451 302 600.0-700.0 6.53 10.90 1.49 11.57 1.54 9437 1451 303 700.0-800.0 4.95 10.47 0.90 10.52 0.97 9437 1451 304 800.0-900.0 3.65 8.50 0.75 9.30 0.82 9437 1451 305 900.0-999.9 5.06 13.51 0.60 13.23 0.70 9437 1451 306 ------------------------------------------------------ 9437 1451 307 (*) Gwin 1971 data 9437 1451 308 9437 1451 309 If one excepts the energy range 37.5-50 eV, the calculated 9437 1451 310 absorption values agree well with Gwin experimental data; they 9437 1451 311 are on average 1.2% lower in the energy range from 50 eV to 9437 1451 312 1000 eV. 9437 1451 313 9437 1451 314 ---------------------------------------------------------------- 9437 1451 315 COMMENTS ON THE 1 keV TO 2 keV ENERGY RANGE 9437 1451 316 9437 1451 317 Preliminary resonance parameters were obtained in 1989 from 9437 1451 318 the analysis of the Harvey thick sample transmission data and of 9437 1451 319 the preliminary results of Weston 88 fission measurement. Due to 9437 1451 320 lack of time, the medium and thin sample transmission data were 9437 1451 321 not included in the SAMMY data base, and the contribution of the 9437 1451 322 truncated external resonances was not carefully investigated. 9437 1451 323 Nevertheless, the results were used in the ENDF/B-VI file, along 9437 1451 324 with a smooth file in order to agree with the average values of 9437 1451 325 a previous ENDF/B-VI evaluation (this preliminary set of 9437 1451 326 parameters was considered as more useful than the statistical 9437 1451 327 parameters in the energy range 1 keV to 2 keV for the 9437 1451 328 calculation of the self-shielding factors). 9437 1451 329 The analysis was restarted in April 1991 at JAERI with an 9437 1451 330 updated version of SAMMY adapted by T. Nakagawa to the FACOM 780.9437 1451 331 The preliminary set of parameters obtained at Oak Ridge in 1989 9437 1451 332 was used as prior information to start the SAMMY calculations. 9437 1451 333 Also prior to the analysis, the contribution of the external 9437 1451 334 resonances was calculated by using the set of the 0 keV to 1 keV 9437 1451 335 known resonances, shifted in the energy ranges -1 keV to 0 keV, 9437 1451 336 2 keV to 3 keV, and 3 keV to 4 keV; equivalent contribution was 9437 1451 337 obtained by using 3 ficticious resonances below 1 keV and 3 9437 1451 338 ficticious resonances above 2 keV [17]. The analysis was 9437 1451 339 performed on the thick and medium sample transmissions of Harvey 9437 1451 340 (the thin sample data was not useful in the high energy range) 9437 1451 341 and on the 1988 fission data released by Weston at the beginning 9437 1451 342 of 1991 [15]. The definitive SAMMY fits were performed in April 9437 1451 343 1992 after renormalisation of the 1988 data of Weston to the 9437 1451 344 ENDF/B-VI standard values between 1 keV and 2 keV, in agreement 9437 1451 345 with the 1991 new measurements of Weston and Todd. 9437 1451 346 The average cross sections calculated from the resonance 9437 1451 347 parameters are compared to the experimental values in the 9437 1451 348 following table. 9437 1451 349 9437 1451 350 Energy Cross Sections (barn) 9437 1451 351 (keV) Total Fission Capture 9437 1451 352 -------- ---------------- ---------------- --------------- 9437 1451 353 CALC(a) EXP(b) CALC(a) EXP(c) CALC(a) EXP(d) 9437 1451 354 1.0-1.1 24.47 24.95 5.549 5.581 4.728 5.04 9437 1451 355 1.1-1.2 22.82 23.10 5.985 6.017 3.757 2.95 9437 1451 356 1.2-1.3 22.29 22.90 4.601 4.501 4.287 4.00 9437 1451 357 1.3-1.4 22.63 22.85 6.997 6.997 3.012 2.52 9437 1451 358 1.4-1.5 20.42 20.95 4.041 4.059 3.450 3.57 9437 1451 359 1.5-1.6 18.30 18.95 2.564 2.613 3.521 3.89 9437 1451 360 1.6-1.7 21.82 21.90 3.952 3.955 3.833 4.36 9437 1451 361 1.7-1.8 21.26 21.35 3.400 3.425 4.091 4.37 9437 1451 362 1.8-1.9 23.76 23.30 5.178 5.187 3.639 3.14 9437 1451 363 1.9-2.0 18.48 18.90 2.152 2.180 3.205 4.06 9437 1451 364 -------- ---------------- ---------------- --------------- 9437 1451 365 1.0-2.0 21.63 21.92 4.442 4.446 3.752 3.79 9437 1451 366 ------------------------------------------------------------- 9437 1451 367 (a) total,fission and capture cross sections calculated by 9437 1451 368 RESEND from the resonance parameters. 9437 1451 369 (b) experimental total cross sections from Derrien [23]. 9437 1451 370 (c) Weston and Todd 1988 high resolution fission cross sections 9437 1451 371 [15] normalised to ENDF/B-VI standard in the energy range 9437 1451 372 from 1.0 keV to 2.0 keV. 9437 1451 373 (d) Gwin 1971 experimental data normalised to Gwin 1976 data. 9437 1451 374 9437 1451 375 The difference of 1.3% between the average calculated total 9437 1451 376 cross section and the average experimental cross section in the 9437 1451 377 energy range from 1.0 keV and 2.0 keV is mainly due to the method9437 1451 378 of evaluating the total cross section from the effective cross 9437 1451 379 section of Derrien [23]. The accuracy of the Sammy fit of the 9437 1451 380 experimental transmission data is better than 0.5% on the cross 9437 1451 381 section. The calculated fission cross sections are in very good 9437 1451 382 agreement with the experimental data. The capture data [1] are 9437 1451 383 average values obtained from the data available in the EXFOR 9437 1451 384 file and normalised to Gwin 1976 average values; there are large 9437 1451 385 differences between the calculated data and the experimental 9437 1451 386 data averaged over 0.1keV intervals; but on the interval from 9437 1451 387 1.0 keV to 2.0 keV the average values are consistent within 1.0%.9437 1451 388 9437 1451 389 ---------------------------------------------------------------- 9437 1451 390 COMMENTS ON THE 2.0 keV TO 2.5 keV REGION 9437 1451 391 9437 1451 392 This energy range was also analysed at JAERI [17]. No 9437 1451 393 preliminary set of resonance parameters was available prior to 9437 1451 394 the analysis. More than 90% of the resonances, compared to the 9437 1451 395 low energy range, could still be identified in the transmission 9437 1451 396 data between 2 keV and 2.5 keV. Therefore, the correlated SAMMY 9437 1451 397 analysis of Harvey transmissions and Weston fission was still 9437 1451 398 feasible in this energy range. The resonance parameters obtained 9437 1451 399 are consistent and have nearly the same statistical properties 9437 1451 400 as those of the resonances in the 0 to 2 keV energy range. A 9437 1451 401 quite good fit of the transmission and fission data was obtained 9437 1451 402 without background and normalisation adjustment. However, the 9437 1451 403 calculated fission cross sections are, on average, 1.4% lower 9437 1451 404 than the experimental values. This difference, which however is 9437 1451 405 not larger than the systematic errors on the experimental data, 9437 1451 406 could be due to the difficulties of identifying the wide j=0+ 9437 1451 407 resonances in the experimental data, because the effects of the 9437 1451 408 increasing resolution and Doppler widths. Prior to the SAMMY 9437 1451 409 fits, the fission data of Weston and Todd (1988 high resolution 9437 1451 410 data) were normalised to the ENDF/B-VI standard in the energy 9437 1451 411 range from 1 keV to 2 keV. 9437 1451 412 The cross sections, calculated from the resonance parameters 9437 1451 413 and averaged over 0.1 keV intervals, are given in the following 9437 1451 414 table. 9437 1451 415 9437 1451 416 Energy Cross Sections (barn) 9437 1451 417 (keV) TOTAL FISSION CAPTURE 9437 1451 418 --------- ---------------- ---------------- ------- 9437 1451 419 CALC(a) EXP(b) CALC(a) EXP(c) CALC(a) 9437 1451 420 2.0-2.1 17.34 17.30 2.034 2.062 3.223 9437 1451 421 2.1-2.2 20.27 19.80 2.949 2.999 4.051 9437 1451 422 2.2-2.3 19.34 19.10 2.357 2.393 3.324 9437 1451 423 2.3-2.4 21.28 21.20 3.646 3.679 3.640 9437 1451 424 2.4-2.5 20.03 20.60 3.956 4.024 3.128 9437 1451 425 --------- ---------------- ---------------- ----------- 9437 1451 426 2.0-2.5 19.65 19.60 2.989 3.031 3.473 9437 1451 427 ----------------------------------------------------------- 9437 1451 428 (a) total,fission and capture cross sections calculated by 9437 1451 429 RESENDD, 1% accuracy at 300 K, from the resonance 9437 1451 430 parameters. 9437 1451 431 (b) average total cross sections obtained from the average 9437 1451 432 experimental effective cross sections of Derrien [23]. 9437 1451 433 (c) 1988 high resolution data of Weston and Todd [15] 9437 1451 434 normalised to ENDF/B-VI standard in the energy range 9437 1451 435 from 1 keV to 2 keV. 9437 1451 436 9437 1451 437 ---------------------------------------------------------------- 9437 1451 438 FISSION AND CAPTURE RESONANCE INTEGRALS 9437 1451 439 9437 1451 440 The fission and capture resonance integrals are compared to 9437 1451 441 JENDL3 data in the following table: 9437 1451 442 9437 1451 443 Energy range (eV) Fission(barn) Capture(barn) 9437 1451 444 ----------------- ----------------- ----------------- 9437 1451 445 JENDL3 present JENDL3 present 9437 1451 446 0.5 - 5.0 85.725 84.879 28.651 28.723 9437 1451 447 5.0 - 10.0 25.081 25.147 19.059 18.950 9437 1451 448 10.0 - 50.0 96.856 99.715 77.181 74.686 9437 1451 449 50.0 - 100.0 40.479 41.552 25.930 25.376 9437 1451 450 100.0 - 301.0 19.677 20.252 17.952 17.729 9437 1451 451 301.0 -1000.0 10.047 10.317 8.348 8.418 9437 1451 452 1000.0 -2000.0 3.484 3.206 2.840 2.634 9437 1451 453 2000.0 -2.E+07 17.783 (17.783) 5.224 (5.224) 9437 1451 454 ----------------- ----------------- ----------------- 9437 1451 455 Total 299.132 302.851 185.185 181.739 9437 1451 456 ---------------------------------------------------------- 9437 1451 457 9437 1451 458 The JENDL3 resonance parameters are those obtained in 1987 in 9437 1451 459 the energy range 0 keV to 1 keV. They are sligthly different from9437 1451 460 those published in 1989. Which explains the small differences 9437 1451 461 observed between JENDL3 and the present results in this energy 9437 1451 462 range. In the energy range 1 keV to 2 keV, JENDL3 is unresolved 9437 1451 463 range. The fission and capture resonance integrals calculated 9437 1451 464 from ENDF/B-V and those found in BNL-325 are the following: 9437 1451 465 9437 1451 466 ENDF/B-V Fission: 302.13 b Capture: 194.10 b 9437 1451 467 BNL-325 Fission: 310+-10 b Capture: 200+-20 b 9437 1451 468 9437 1451 469 The consequence of changing from the old sets of resonance 9437 1451 470 parameters(ENDF/B-V and previous sets) to the new set is that 9437 1451 471 the capture resonance integral will decrease by 6.7% compared 9437 1451 472 with the ENDF/B-V value. 9437 1451 473 9437 1451 474 ---------------------------------------------------------------- 9437 1451 475 UNRESOLVED RESONANCE REGION 9437 1451 476 9437 1451 477 The average resonance prameters are given in the energy range 9437 1451 478 2.5 keV to 30 keV for 70 energy points. They were obtained by 9437 1451 479 using the Cadarache statistical code FISINGA to fit the gross 9437 1451 480 structure of the Saclay experimental total cross sections [26] 9437 1451 481 below 4 keV and of selected experimental fission cross sections 9437 1451 482 normalised to ENDF/B-VI standard evaluation [11]. Above 4 keV no 9437 1451 483 high resolution total cross section data are available; average 9437 1451 484 total cross sections were calculated to be consistent with the 9437 1451 485 stastistical paramaters obtained in the resolved resonance 9437 1451 486 region [14] and with the Optical Model parameters of Lagrange 9437 1451 487 and Madland [24] obtained by fitting the experimental data in 9437 1451 488 the high energy range. A value of 9.46 fm was used for the 9437 1451 489 effective radius. The values obtained for alpha are consistent 9437 1451 490 with the experimental data. 9437 1451 491 The competitive width is not used for the inelastic scattering9437 1451 492 cross section. For each energy point of the unresolved region the9437 1451 493 neutron width corresponds only to the elastic scattering cross 9437 1451 494 section. The inelastic scattering cross section should be found 9437 1451 495 in file 3. 9437 1451 496 The cross sections obtained at ORNL by processing the 9437 1451 497 evaluated file using NJOY-87.1 are given in the following table, 9437 1451 498 'FISS' for the fission values and 'CAPT' for the capture values. 9437 1451 499 9437 1451 500 Energy Cross sections Energy Cross sections 9437 1451 501 (keV) (barn) (keV) (barn) 9437 1451 502 ------ ----------------- ------ ---------------- 9437 1451 503 FISS CAPT FISS CAPT 9437 1451 504 2.500 4.280 2.456 13.750 1.715 0.942 9437 1451 505 2.550 2.725 2.754 14.250 1.492 0.948 9437 1451 506 2.650 3.103 3.425 14.750 1.797 0.854 9437 1451 507 2.750 4.169 2.010 15.250 1.883 0.797 9437 1451 508 2.850 4.126 2.077 15.750 1.697 0.843 9437 1451 509 2.950 3.362 3.710 16.250 1.801 0.782 9437 1451 510 3.050 3.017 1.998 16.750 1.628 0.824 9437 1451 511 3.150 4.896 1.934 17.250 1.498 0.819 9437 1451 512 3.250 3.954 2.277 17.750 1.862 0.701 9437 1451 513 3.350 1.710 2.166 18.250 1.711 0.736 9437 1451 514 3.450 2.198 2.572 18.750 1.632 0.748 9437 1451 515 3.550 2.214 1.885 19.250 1.738 0.694 9437 1451 516 3.650 2.394 2.948 19.750 1.743 0.677 9437 1451 517 3.750 3.067 1.624 20.500 1.672 0.679 9437 1451 518 3.850 3.556 2.122 21.500 1.646 0.661 9437 1451 519 3.950 2.931 2.397 22.500 1.472 0.697 9437 1451 520 4.125 2.114 2.270 23.500 1.632 0.619 9437 1451 521 4.375 2.509 2.129 24.500 1.636 0.597 9437 1451 522 4.625 2.772 1.715 25.500 1.547 0.607 9437 1451 523 4.875 1.980 2.186 26.500 1.628 0.562 9437 1451 524 5.125 2.406 1.916 27.500 1.544 0.572 9437 1451 525 5.375 2.153 1.953 28.500 1.568 0.549 9437 1451 526 5.625 2.294 1.807 29.500 1.609 0.521 9437 1451 527 9437 1451 528 Average values of the fission cross sections compared to the 9437 1451 529 ENDF/B-VI standard evaluation [11] and alpha values compared to 9437 1451 530 some experimental data are given in the following table. 9437 1451 531 9437 1451 532 Energy Cross sections (barn) Alpha 9437 1451 533 (keV) (1) (2) (3) (4) (5) (6) (7) (8) 9437 1451 534 ------ ------------------------- -------------------------- 9437 1451 535 3- 4 2.992 3.000 2.213 2.20 0.740 0.720 0.895 0.820 9437 1451 536 4- 5 2.394 2.383 2.073 2.07 0.866 0.870 0.821 0.837 9437 1451 537 5- 6 2.266 2.301 1.863 1.91 0.822 0.820 0.867 0.834 9437 1451 538 6- 7 2.006 2.008 1.677 1.63 0.836 0.790 0.816 0.793 9437 1451 539 7- 8 2.134 2.054 1.409 1.34 0.660 0.640 0.630 0.605 9437 1451 540 8- 9 2.207 2.216 1.245 1.23 0.564 0.540 0.575 0.530 9437 1451 541 9-10 1.867 1.864 1.136 1.05 0.608 0.550 0.617 0.569 9437 1451 542 9437 1451 543 1-10 2.628 2.622 2.014 2.06 0.767 0.752 0.806 0.768 9437 1451 544 10-20 1.762 1.764 0.876 0.85 0.497 0.480 0.466 0.498 9437 1451 545 20-30 1.597 1.595 0.606 0.58 0.379 0.350 0.373 0.388 9437 1451 546 ------------------------------------------------------------- 9437 1451 547 (1) Fission cross section, present evaluation (0K) 9437 1451 548 (2) Fission cross section, ENDF/B-VI standard [11] 9437 1451 549 (3) Capture cross section, present evaluation (293 K) 9437 1451 550 (4) Capture cross section, Gwin et al. 1976 [4] 9437 1451 551 (5) Alpha value, present evaluation (293 K) 9437 1451 552 (6) Alpha value from Gwin et al. 1976 [4] 9437 1451 553 (7) Alpha value from Sowerby-Konshin evaluation 1971 [25] 9437 1451 554 (8) Average alpha value from experimental data 9437 1451 555 9437 1451 556 The fission and capture resonance integrals obtained at ORNL 9437 1451 557 are compared to ENDF/B-5 data in the following table. 9437 1451 558 9437 1451 559 Energy range Fission (barn) Capture (barn) 9437 1451 560 (eV) ENDF/B-5 present ENDF/B-5 present 9437 1451 561 --------------- ----------------- ----------------- 9437 1451 562 0.5 - 5.0 86.02 85.71 32.31 28.65 9437 1451 563 5.0 - 10.0 26.03 25.08 20.14 19.06 9437 1451 564 10.0 - 50.0 100.25 96.87 78.66 77.19 9437 1451 565 50.0 - 100.0 40.32 40.47 27.23 25.93 9437 1451 566 100.0 - 301.0 19.98 19.68 19.52 17.95 9437 1451 567 301.0 -1000.0 10.15 10.05 8.54 8.35 9437 1451 568 --------------- ----------------- ----------------- 9437 1451 569 0.5 -1000.0 282.76 277.85 186.30 177.13 9437 1451 570 -------------------------------------------------------- 9437 1451 571 9437 1451 572 The fission and capture resonance integrals are obtained by 9437 1451 573 adding the ENDF/B-V value above 1 keV to the present evaluation. 9437 1451 574 These and the corresponding values from ENDF/B-V evaluation are: 9437 1451 575 Present - Fission: 297.22 b Capture: 184.93 b 9437 1451 576 ENDF/B-V - Fission: 302.13 b Capture: 194.10 b 9437 1451 577 9437 1451 578 ---------------------------------------------------------------- 9437 1451 579 REFERENCES 9437 1451 580 9437 1451 581 1. R. Gwin et al., Nucl.Sci.Eng. 45, 25 (1971). 9437 1451 582 2. A.J. Deruyter et al., J.Nucl.En. 26, 293 (1972). 9437 1451 583 3. J. Blons, Nucl.Sci.Eng. 51, 130 (1973). 9437 1451 584 4. R. Gwin et al., Nucl.Sci.Eng. 59, 79 (1976). 9437 1451 585 5. R. Gwin et al., Nucl.Sci.Eng. 61, 116 (1976). 9437 1451 586 6. W. Wagemans, Ann.Nucl.En. 7 #9, 495 (1980). 9437 1451 587 7. R. Gwin et al., Nucl.Sci.Eng. 88, 37 (1984). 9437 1451 588 8. L.W. Weston et al., Nucl.Sci.Eng. 88, 567 (1984). 9437 1451 589 9. J.A. Harvey et al., Nuclear Data for Sci. and Technol., Proc. 9437 1451 590 Int. Conf., May 30 - June 3, 1988, Mito, Japan (Saikon 9437 1451 591 Publishing Co., 1988) p.115. 9437 1451 592 10. R.R. Spencer et al., Nucl.Sci.Eng. 96, 318 (1987). 9437 1451 593 11. A. Carlson et al., preliminary results of the ENDF/B-6 9437 1451 594 standard evaluation (Sep.8, 1987); see W. P. Poenitz et al., 9437 1451 595 Argonne National Laboratory report ANL/NDM-139 [ENDF-358] 9437 1451 596 (1997) 9437 1451 597 12. A.J. Deruyter, J.Nucl.En. 26, 293 (1972). 9437 1451 598 13. N.M. Larson et al., Oak Ridge National Laboratory reports 9437 1451 599 ORNL/TM-7485, ORNL/TM-9179, and ORNL/TM-9719/R1 9437 1451 600 14. H. Derrien and G. DeSaussure, Oak Ridge National Laboratory 9437 1451 601 report ORNL-TM-10986 (1988). 9437 1451 602 15. L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992). 9437 1451 603 16. H. Derrien et al., Nucl.Sci.Eng. 106, 434 (1990). 9437 1451 604 17. H. Derrien and T. Nakagawa, to be published. 9437 1451 605 18. L. Leal and R.N. Hwang, Trans.Am.Nucl.Soc. 55, 340 (1987). 9437 1451 606 19. H. Derrien et aL., Nucl.Sci.Eng. 106, 434 (1990). 9437 1451 607 20. T. Nakagawa, RESENDD a JAERI version of RESEND 9437 1451 608 21. C. Wagemans et al., Nuclear Data for Sci. and Technol., Proc. 9437 1451 609 Int. Conf., May 30 - June 3, 1988, Mito, Japan (Saikon 9437 1451 610 Publishing Co., 1988) p.91. 9437 1451 611 22. L.W. Weston et al., Nucl.Sci.Eng. 115, 164 (1993). 9437 1451 612 23. H. Derrien, to be published in J.Nucl.Sci.Technol. 9437 1451 613 24. Ch. Lagrange and D.G. Madland, Phys.Rev. C 33, 1616 (1986). 9437 1451 614 25. M.G. Sowerby et al., At.En.Rev. 10, 453 (1972) 9437 1451 615 26. H. Derrien, thesis, Univ. Paris - Sud, Orsay Serie A No. 1172 9437 1451 616 (1973). 9437 1451 617 27. A. Lendl et al.,Atomnaya Energiya Vol61,N3,pp215-216,(1986) 9437 1451 618 28. E. Fort et al., paper to SGC/WPEC, (2002) 9437 1451 619 29. F.J. Hambsch et al, Jour. of Nuc. Sci. and Tech., ND-2001 9437 1451 620 procedings, to be published, (2002) 9437 1451 621 30. E. Fort et al., NSE99,pp375-389, (1988) 9437 1451 622 31. G.Vladuca, A.Tudora., Ann.Nuc.Energy. 28, 689 (2001). 9437 1451 623 9437 1451 624 ******************************************************************9437 1451 625 ENERGY REGION 0.03 TO 30 MeV ***********************************9437 1451 626 Principal evaluators : J.P. Delaroche, S.Hilaire, B. Morillon 9437 1451 627 P. Romain. 9437 1451 628 ******************************************************************9437 1451 629 The evaluation above 30 keV is based on a detailed theoretical 9437 1451 630 model analysis utilizing the available experimental data and 9437 1451 631 microscopic level densities as guides to phenomenological models.9437 1451 632 Coupled channel optical model calculations were used to provide 9437 1451 633 the total and direct reaction components of elastic and inelastic9437 1451 634 scattering cross sections and angular distributions for 9437 1451 635 collective levels. These are the (1/2)+, (3/2)+, ..., (11/2)+ 9437 1451 636 members of the ground state band, and the (1/2)-, (3/2)- and 9437 1451 637 (5/2)- members of the experimentally identified octupole band. 9437 1451 638 The plain rotation-vibration model is adopted. The coupling 9437 1451 639 strength for interband transitions is closed to that deduced from9437 1451 640 coupled channel analyses of inelastic scattering data for the 9437 1451 641 Kpi=(0)- vibrational band of U238. 9437 1451 642 Coupled channel calculations are performed using the ECIS 9437 1451 643 code [Ra70] which also provides coumpound nucleus cross sections 9437 1451 644 and transmission coefficients used in pre-equilibrium/evaporation9437 1451 645 emission treated in the Exciton and HAUSER-FESHBACH models 9437 1451 646 implemented in the GNASH code [Yo96]. 9437 1451 647 This reaction code has been modified to include width 9437 1451 648 fluctuation factors, relativistic kinematics, and a more 9437 1451 649 realistic treatment of the fission process. Briefly, the 9437 1451 650 simple double-humped fission barrier model is improved by 9437 1451 651 treating explicitly the coupling between class I and class 9437 1451 652 II states and damping of class II states. 9437 1451 653 Emission of light hadrons up to He4 is explicitly treated in the 9437 1451 654 model calculations. Fission decay of associated residual nuclei 9437 1451 655 is also treated. But none of these emission and fission cross 9437 1451 656 sections are explicitely provided in the files. 9437 1451 657 Above 16.5 MeV the sigma (n,3n), (n,4n) and (n,5n) include 9437 1451 658 components from Light Charged Particles (LPCs). For instance 9437 1451 659 sigma (n,3n) given in MT=17 is the sum of the actual (n,3n) 9437 1451 660 cross section and cross sections associated with LPCs : 9437 1451 661 9437 1451 662 Effective sigma(n,3n)= True sigma(n,3n) + 9437 1451 663 sigma (n,LPC) * sigma(n,3n) / [sigma(n,3n)+sigma(n,4n)+ 9437 1451 664 sigma(n,5n)] 9437 1451 665 Below is provided a table of such relationships between true 9437 1451 666 and effective sigma(n,xn) cross sections. 9437 1451 667 9437 1451 668 ***************************************************************** 9437 1451 669 *Neutron* Sigma* Sigma* Sigma* Sigma* Sigma* Sigma* Sigma* 9437 1451 670 * Energy* (n,3n)* (n,3n)* (n,4n)* (n,4n)* (n,5n)* (n,5n)*(n,LCP)* 9437 1451 671 * * | * (+LCP)* | * (+LCP)* | * (+LCP)* | * 9437 1451 672 * (MeV) * (b) * (b) * (b) * (b) * (b) * (b) * (b) * 9437 1451 673 ***************************************************************** 9437 1451 674 *.1650+2*.1095+0*.1190+0*.0000+0*.0000+0*.0000+0*.0000+0*.9476-2* 9437 1451 675 *.1700+2*.1430+0*.1538+0*.0000+0*.0000+0*.0000+0*.0000+0*.1077-1* 9437 1451 676 *.1750+2*.1809+0*.1931+0*.0000+0*.0000+0*.0000+0*.0000+0*.1215-1* 9437 1451 677 *.1800+2*.2219+0*.2355+0*.0000+0*.0000+0*.0000+0*.0000+0*.1358-1* 9437 1451 678 *.1850+2*.2633+0*.2784+0*.0000+0*.0000+0*.0000+0*.0000+0*.1512-1* 9437 1451 679 *.1900+2*.2995+0*.3162+0*.0000+0*.0000+0*.0000+0*.0000+0*.1669-1* 9437 1451 680 *.1950+2*.3222+0*.3405+0*.1969-4*.2081-4*.0000+0*.0000+0*.1829-1* 9437 1451 681 *.2000+2*.3363+0*.3562+0*.2584-3*.2737-3*.0000+0*.0000+0*.1996-1* 9437 1451 682 *.2050+2*.3368+0*.3585+0*.1271-2*.1353-2*.0000+0*.0000+0*.2175-1* 9437 1451 683 *.2100+2*.3191+0*.3423+0*.3907-2*.4191-2*.0000+0*.0000+0*.2345-1* 9437 1451 684 *.2150+2*.2989+0*.3235+0*.8409-2*.9100-2*.0000+0*.0000+0*.2526-1* 9437 1451 685 *.2200+2*.2791+0*.3047+0*.1515-1*.1654-1*.0000+0*.0000+0*.2695-1* 9437 1451 686 *.2250+2*.2629+0*.2894+0*.2350-1*.2586-1*.0000+0*.0000+0*.2882-1* 9437 1451 687 *.2300+2*.2291+0*.2557+0*.3475-1*.3878-1*.0000+0*.0000+0*.3058-1* 9437 1451 688 *.2350+2*.2019+0*.2280+0*.4841-1*.5467-1*.0000+0*.0000+0*.3234-1* 9437 1451 689 *.2400+2*.1763+0*.2013+0*.6387-1*.7291-1*.0000+0*.0000+0*.3400-1* 9437 1451 690 *.2450+2*.1560+0*.1798+0*.7793-1*.8984-1*.0000+0*.0000+0*.3575-1* 9437 1451 691 *.2500+2*.1367+0*.1586+0*.9692-1*.1124+0*.0000+0*.0000+0*.3735-1* 9437 1451 692 *.2550+2*.1241+0*.1443+0*.1158+0*.1346+0*.0000+0*.0000+0*.3896-1* 9437 1451 693 *.2600+2*.1098+0*.1280+0*.1342+0*.1564+0*.0000+0*.0000+0*.4034-1* 9437 1451 694 *.2650+2*.1004+0*.1170+0*.1518+0*.1769+0*.0000+0*.0000+0*.4170-1* 9437 1451 695 *.2700+2*.9288-1*.1084+0*.1643+0*.1917+0*.4294-7*.5011-7*.4293-1* 9437 1451 696 *.2750+2*.8309-1*.9739-1*.1733+0*.2031+0*.1862-5*.2182-5*.4412-1* 9437 1451 697 *.2800+2*.7845-1*.9216-1*.1818+0*.2136+0*.1668-4*.1959-4*.4548-1* 9437 1451 698 *.2850+2*.7155-1*.8510-1*.1802+0*.2143+0*.8049-4*.9574-4*.4771-1* 9437 1451 699 *.2900+2*.6966-1*.8362-1*.1784+0*.2142+0*.2391-3*.2870-3*.4976-1* 9437 1451 700 *.2950+2*.6734-1*.8218-1*.1678+0*.2048+0*.6368-3*.7772-3*.5197-1* 9437 1451 701 *.3000+2*.6565-1*.8130-1*.1596+0*.1976+0*.1279-2*.1584-2*.5400-1* 9437 1451 702 ***************************************************************** 9437 1451 703 9437 1451 704 The (n,5n) cross section is provided in the above table, but not 9437 1451 705 inserted in the file. 9437 1451 706 MF=3 Smooth Cross Sections ------------------------------------- 9437 1451 707 MT=1 Neutron Total Cross Section. 0.03 to 30 MeV, analysis 9437 1451 708 based on coupled-channel optical calculations and the 9437 1451 709 exp. data of [Po81,Sh78,Po83,Sc74,Fo71,Sm73,Na73,Pe60, 9437 1451 710 Ca73,Li90]. Calculated as the sum of MT=1 and MT=3. 9437 1451 711 MT=2 0.030 to 30 MeV, based on coupled channel and 9437 1451 712 statistical model calculations. 9437 1451 713 MT=3 0.030 to 30 MeV, 9437 1451 714 The sum of partial cross sections is calculated using 9437 1451 715 GNASH, in which the neutron transmission coefficients 9437 1451 716 we use are from ECIS calculations. Compound elastic 9437 1451 717 component is not included in the above sum. 9437 1451 718 MT=4 0.030 to 30 MeV, based on sum of MT=51-91. 9437 1451 719 MT=16 (n,2n) cross section 0.030 to 30 MeV, 9437 1451 720 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 721 MT=17 (n,3n) cross section 0.030 to 30 MeV, 9437 1451 722 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 723 For more information see comments and table above. 9437 1451 724 MT=18 fission cross section 0.030 to 30 MeV, 9437 1451 725 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 726 This file includes components stemming from fission 9437 1451 727 of residuals associated with charged particle emission. 9437 1451 728 MT=37 (n,4n) cross section 0.030 to 30 MeV, 9437 1451 729 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 730 For more information see comments and table above. 9437 1451 731 MT=51-55 Thres. to 30 MeV, coupled-channel optical model 9437 1451 732 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 733 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 734 (3/2)- and (5/2)- members of the octupole band) 9437 1451 735 using the ECIS code. Compound nucleus contributions, 9437 1451 736 obtained from GNASH calculations, are also included. 9437 1451 737 MT=56-63 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 738 calculations using the GNASH code. 9437 1451 739 MOLDAUER width fluctuation factors are turned off beyond9437 1451 740 4 MeV incident energy. 9437 1451 741 MT=64 Thres. to 30 MeV, coupled-channel optical model 9437 1451 742 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 743 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 744 (3/2)- and (5/2)- members of the octupole band) 9437 1451 745 using the ECIS code. Compound nucleus contributions, 9437 1451 746 obtained from GNASH calculations, are also included. 9437 1451 747 MT=65 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 748 calculations using the GNASH code. 9437 1451 749 MOLDAUER width fluctuation factors are turned off beyond9437 1451 750 4 MeV incident energy. 9437 1451 751 MT=66-67 Thres. to 30 MeV, coupled-channel optical model 9437 1451 752 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 753 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 754 (3/2)- and (5/2)- members of the octupole band) 9437 1451 755 using the ECIS code. Compound nucleus contributions, 9437 1451 756 obtained from GNASH calculations, are also included. 9437 1451 757 MT=68-77 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 758 calculations using the GNASH code. 9437 1451 759 MOLDAUER width fluctuation factors are turned off beyond9437 1451 760 4 MeV incident energy. 9437 1451 761 MT=91 Thres. to 30 MeV, 9437 1451 762 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 763 MT=102 0.030-30 MeV, 9437 1451 764 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 765 9437 1451 766 MF=4 Neutron Angular Distributions ----------------------------- 9437 1451 767 Tabulated sigma(theta) values 9437 1451 768 9437 1451 769 MT=2 Elastic scattering angular distribution based on ECIS 9437 1451 770 coupled-channel calculations and GNASH calculations. 9437 1451 771 MT=16,17,37 Isotropic distributions. 9437 1451 772 MT=18 Isotropic distribution. 9437 1451 773 MT=51-55 Thres. to 30 MeV, coupled-channel optical model 9437 1451 774 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 775 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 776 (3/2)- and (5/2)- members of the octupole band) 9437 1451 777 using the ECIS code. Compound nucleus contributions, 9437 1451 778 obtained from GNASH calculations, are also included. 9437 1451 779 MT=56-63 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 780 calculations using the GNASH code. 9437 1451 781 MOLDAUER width fluctuation factors are turned off beyond9437 1451 782 4 MeV incident energy. 9437 1451 783 MT=64 Thres. to 30 MeV, coupled-channel optical model 9437 1451 784 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 785 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 786 (3/2)- and (5/2)- members of the octupole band) 9437 1451 787 using the ECIS code. Compound nucleus contributions, 9437 1451 788 obtained from GNASH calculations, are also included. 9437 1451 789 MT=65 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 790 calculations using the GNASH code. 9437 1451 791 MOLDAUER width fluctuation factors are turned off beyond9437 1451 792 4 MeV incident energy. 9437 1451 793 MT=66-67 Thres. to 30 MeV, coupled-channel optical model 9437 1451 794 calculations [(3/2)+ to (11/2)+] members of the 9437 1451 795 Kpi=(1/2)+ ground state rotational band, and (1/2)-, 9437 1451 796 (3/2)- and (5/2)- members of the octupole band) 9437 1451 797 using the ECIS code. Compound nucleus contributions, 9437 1451 798 obtained from GNASH calculations, are also included. 9437 1451 799 MT=68-77 Thres. to 30 MeV, Compound nucleus reaction theory 9437 1451 800 calculations using the GNASH code. 9437 1451 801 MOLDAUER width fluctuation factors are turned off beyond9437 1451 802 4 MeV incident energy. 9437 1451 803 MT=91 Isotropic distribution. 9437 1451 804 9437 1451 805 MF=5 Neutron Energy Distributions ------------------------------ 9437 1451 806 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 807 Updated Kalbach-Mann systematics used for specifying 9437 1451 808 neutron distributions [Ka87]. Only neutrons given. 9437 1451 809 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 810 Updated Kalbach-Mann systematics used for specifying 9437 1451 811 neutron distributions [Ka87]. Only neutrons given. 9437 1451 812 MT=18 Neutron energy distributions from fission based on the 9437 1451 813 Los Alamos model, with multiple chances (first, second, 9437 1451 814 third, fourth and fifth chance), and upgraded by 9437 1451 815 G.Vladuca and A.Tudora [Vl01]. 9437 1451 816 A linear relation between the average prompt gamma ray 9437 1451 817 energy and the average prompt neutron multiplicity and a9437 1451 818 dependence of the average fission fragments kinetic 9437 1451 819 energy on the incident neutron energy are used. 9437 1451 820 The model parameters are slightly different from those 9437 1451 821 adopted in [Vl01]. 9437 1451 822 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 823 Updated Kalbach-Mann systematics used for specifying 9437 1451 824 neutron distributions [Ka87]. Only neutrons given. 9437 1451 825 MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 826 Updated Kalbach-Mann systematics used for specifying 9437 1451 827 neutron distributions [Ka87]. Only neutrons given. 9437 1451 828 MT=455 Tal England [En89]. 9437 1451 829 9437 1451 830 9437 1451 831 MF=12,13,14,15 Photon-Production Data -----N.Y.I.--------------- 9437 1451 832 9437 1451 833 ---------------------------------------------------------------- 9437 1451 834 REFERENCES 9437 1451 835 9437 1451 836 [Ar84] E. Arthur et al., Nuc.Sci.Eng. 88, 56 (1984). 9437 1451 837 [Ca73] J. Cabe et al., report CEA-R-4524 (1973). 9437 1451 838 [En89] T.R. England et al, Los Alamos reports LA 11151-MS 9437 1451 839 (1988) and LA-11534-T (1989); M.C. Brady and T.R. England, 9437 1451 840 Nucl.Sci.Eng. 103, 129 (1989). 9437 1451 841 [Fo71] D. Foster and D. Glasgow, Phys.Rev. C3, 576 (1971). 9437 1451 842 [Ka87] C. Kalbach, Phys.Rev. C 37, 2350 (1988). 9437 1451 843 [Li90] P. 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Chadwick, 9437 1451 855 in Workshop on Nuclear Reaction Data 9437 1451 856 and Nuclear Reactors, Trieste, Italy (1996). 9437 1451 857 9437 1451 858 ************************ C O N T E N T S *********************** 9437 1451 859 1 451 941 09437 1451 860 1 452 792 09437 1451 861 1 455 10 09437 1451 862 1 456 792 09437 1451 863 1 458 5 09437 1451 864 2 151 1445 09437 1451 865 3 1 108 09437 1451 866 3 2 97 09437 1451 867 3 3 108 09437 1451 868 3 4 106 09437 1451 869 3 16 48 09437 1451 870 3 17 24 09437 1451 871 3 18 97 09437 1451 872 3 37 11 09437 1451 873 3 51 97 09437 1451 874 3 52 95 09437 1451 875 3 53 94 09437 1451 876 3 54 91 09437 1451 877 3 55 90 09437 1451 878 3 56 87 09437 1451 879 3 57 86 09437 1451 880 3 58 85 09437 1451 881 3 59 85 09437 1451 882 3 60 84 09437 1451 883 3 61 83 09437 1451 884 3 62 82 09437 1451 885 3 63 81 09437 1451 886 3 64 81 09437 1451 887 3 65 80 09437 1451 888 3 66 80 09437 1451 889 3 67 80 09437 1451 890 3 68 79 09437 1451 891 3 69 79 09437 1451 892 3 70 78 09437 1451 893 3 71 78 09437 1451 894 3 72 78 09437 1451 895 3 73 77 09437 1451 896 3 74 77 09437 1451 897 3 75 76 09437 1451 898 3 76 75 09437 1451 899 3 77 74 09437 1451 900 3 91 74 09437 1451 901 3 102 97 09437 1451 902 4 2 9478 09437 1451 903 4 16 10 09437 1451 904 4 17 10 09437 1451 905 4 18 10 09437 1451 906 4 37 10 09437 1451 907 4 51 9247 09437 1451 908 4 52 9016 09437 1451 909 4 53 8950 09437 1451 910 4 54 8653 09437 1451 911 4 55 8554 09437 1451 912 4 56 10 09437 1451 913 4 57 10 09437 1451 914 4 58 10 09437 1451 915 4 59 10 09437 1451 916 4 60 10 09437 1451 917 4 61 10 09437 1451 918 4 62 10 09437 1451 919 4 63 10 09437 1451 920 4 64 7630 09437 1451 921 4 65 10 09437 1451 922 4 66 7564 09437 1451 923 4 67 7531 09437 1451 924 4 68 10 09437 1451 925 4 69 10 09437 1451 926 4 70 10 09437 1451 927 4 71 10 09437 1451 928 4 72 10 09437 1451 929 4 73 10 09437 1451 930 4 74 10 09437 1451 931 4 75 10 09437 1451 932 4 76 10 09437 1451 933 4 77 10 09437 1451 934 4 91 10 09437 1451 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