JEFF-3.1 General Purpose Neutron File, May 2005. 2310 0 0 0 9.223300+4 2.310380+2 1 1 2 49222 1451 1 0.000000+0 1.000000+0 0 0 0 69222 1451 2 1.000000+0 2.000000+7 0 0 10 319222 1451 3 0.000000+0 0.000000+0 0 0 511 879222 1451 4 92-U -233 SAEI+ Eval-MAR00 T.Mutsunobu, T.Kawano 9222 1451 5 DIST-MAY05 REV1-MAY05 20050504 9222 1451 6 ----JEFF-31 MATERIAL 9222 9222 1451 7 -----INCIDENT NEUTRON DATA 9222 1451 8 ------ENDF-6 FORMAT 9222 1451 9 9222 1451 10 ***************************** JEFF-3.1 *************************9222 1451 11 Original data taken from: JENDL 3.3 + ENDF/B-VI.4 9222 1451 12 9222 1451 13 (MF 12-15 From ENDF/B-VI.4) 9222 1451 14 MT=458(ENERGY RELEASE IN FISSION) evaluation from ENDF/B-VI.4 9222 1451 15 25/05/05 (NEA/OECD)the 4th time group has been corrected: 9222 1451 16 from 1.330420-2 to 1.330420-1 9222 1451 17 05/01 NEA/OECD (Rugama) 8 delayed neutron groups 9222 1451 18 Jefdoc-976(Spriggs,Campbel and Piksaikin,Prg Nucl Eng 41,223(2002)9222 1451 19 **************************** JENDL 3.3 ***************************9222 1451 20 HISTORY 9222 1451 21 82-06 Evaluation for JENDL-2 was made by N. Asano (SAEI), 9222 1451 22 H. Matsunobu (SAEI) and Y.Kikuchi(JAERI). 9222 1451 23 87-03 Re-evaluation for JENDL-3 was made by H.Matsunobu (SAEI) 9222 1451 24 Main part of revision was the cross sections above 10 keV 9222 1451 25 and angular and energy distributions of neutrons. 9222 1451 26 Data were compiled by T. Nakagawa (JAERI). 9222 1451 27 94-04 JENDL-3.2. 9222 1451 28 Resonance parameters reevaluated by H.Derrien (JAERI)/1/ 9222 1451 29 Cross sections reevaluated by H.Matsunobu(SAEI) 9222 1451 30 Fission spectrum reevaluated by T.Ohsawa(Kinki univ.) 9222 1451 31 Compiled by T.Nakagawa (NDC/JAERI) 9222 1451 32 9222 1451 33 ***** Modified parts for JENDL-3.2 ******************** 9222 1451 34 (2,151); New analysis with SAMMY 9222 1451 35 MF=3 ; all data except fission cross section below 6.75 9222 1451 36 MeV and total cross section 9222 1451 37 MF=4 ; for inelastic scattering 9222 1451 38 (5,18) 9222 1451 39 *********************************************************** 9222 1451 40 9222 1451 41 00-02 JENDL-3.3. Evaluation was made by H. Matsunobu (SAEI) 9222 1451 42 and T.Kawano(Kyushu Univ.) and compiled by O.Iwamoto 9222 1451 43 (NDC/JAERI). 9222 1451 44 9222 1451 45 ***** Modified parts for JENDL-3.3 ******************** 9222 1451 46 (1,452), (1,455), (1,456) 9222 1451 47 (2,151) unresolved resonance region 9222 1451 48 (3,2), (3,16), (3,17), (3,18), (3,102) 9222 1451 49 (5,16), (5,17), (5,91), (5,455) 9222 1451 50 *********************************************************** 9222 1451 51 9222 1451 52 02-01 Covariances were taken from JENDL-3.2 covariance file except9222 1451 53 for MF/MT=31/455, 31/456/, 33/18, and 33/102, which were 9222 1451 54 evaluated by Matsunobu./8/ 9222 1451 55 9222 1451 56 MF=2 Resonance Parameters 9222 1451 57 MT=151 9222 1451 58 9222 1451 59 2200-m/s cross sections and calculated res. integrals 9222 1451 60 2200 m/s Res. Integ. 9222 1451 61 total 588.38 b - 9222 1451 62 elastic 11.97 b - 9222 1451 63 fission 531.16 b 773 b 9222 1451 64 capture 45.25 b 138 b 9222 1451 65 9222 1451 66 MF=3 Neutron Cross Sections 9222 1451 67 Smooth part (above 30 keV) 9222 1451 68 9222 1451 69 MT=18 Fission 9222 1451 70 Results of recent simultaneous evaluation of fission cross9222 1451 71 sections /2/ were adopted. 9222 1451 72 9222 1451 73 MF=5 Energy Distributions of Secondary Neutrons 9222 1451 74 MT=16,17,91 9222 1451 75 Calculated with pre-compound and multi-step evaporation 9222 1451 76 theory code EGNASH /3,4/. 9222 1451 77 MT=455 Delayed neutron spectrum. 9222 1451 78 Summation calculation made by Brady and England /5/ was 9222 1451 79 adopted. 9222 1451 80 9222 1451 81 MF=31 Covariances of Average Number of Neutrons per Fission 9222 1451 82 MT=452 9222 1451 83 Constructed from MT=455 and 456. 9222 1451 84 MT=455 9222 1451 85 Based on experimental data. A chi-value was 0.13. 9222 1451 86 MT=456 9222 1451 87 Based on experimental data. A chi-value was 0.96. 9222 1451 88 9222 1451 89 MF=32 Covariances of Resonance Paremeters (ref.6) 9222 1451 90 MT=151 9222 1451 91 Resolved resonance 9222 1451 92 Based on the SAMMY analysis by Derrien /1/. 9222 1451 93 Unresolved resonance 9222 1451 94 The covariances were obtained by using kalman. 9222 1451 95 9222 1451 96 MF=33 Covariances of Cross Sections (ref.6) 9222 1451 97 MT=1 9222 1451 98 Based on experimental data. A chi-value was 0.244. 9222 1451 99 MT=2 9222 1451 100 Constructed from mt=1, 4, 16, 17, 18, and 102. 9222 1451 101 MT=4 9222 1451 102 Based on experimental data. A chi-value was 0.151. 9222 1451 103 MT=16 9222 1451 104 Based on experimental data. 9222 1451 105 MT=17 9222 1451 106 Taken from mt=16 9222 1451 107 MT=18 9222 1451 108 The covariances were obtained by the simultaneous evaluation. 9222 1451 109 /2/ 9222 1451 110 MT=102 9222 1451 111 Based on experimental data on alpha values. A chi-value was 9222 1451 112 0.186. 9222 1451 113 9222 1451 114 MF=34 Covariances of Angular Distributions (ref.6) 9222 1451 115 MT=2 9222 1451 116 The covariances of p1 coefficients were obtained by using 9222 1451 117 kalman. A chi-value was 1.474. 9222 1451 118 9222 1451 119 MF=35 Covariances of Energy Distributions 9222 1451 120 MT=18 9222 1451 121 The covariances were obtained by using kalman./7/ 9222 1451 122 9222 1451 123 9222 1451 124 References 9222 1451 125 1) Derrien H.: J. Nucl. Sci. Technol., 31, 379 (1994) 9222 1451 126 2) Kawano T. et al.: JAERI-Research 2000-004 (2000). 9222 1451 127 3) Yamamuro N.: JAERI-M 90-006 (1990). 9222 1451 128 4) Young P.G. and Arthur E.D.: LA-6947 (1977). 9222 1451 129 5) Brady M.C. and England T.R.: Nucl. Sci. Eng., 103, 129(1989). 9222 1451 130 6) Shibata K. and Hasegawa A.: JNC TJ9400 2000-004 (2000) 9222 1451 131 [in Japanese]. 9222 1451 132 7) Kawano T. et al.: JAERI-Research 99-009 (1999) [in Japanese]. 9222 1451 133 8) Matsunobu H.: Private communication (2001). 9222 1451 134 9222 1451 135 9222 1451 136 ******************** JEDL-3.2 ************************************9222 1451 137 92-U -233 SAEI+ Eval-Mar87 H.Matsunobu,Y.Kikuchi,T.Nakagawa 9222 1451 138 Dist-Sep89 Rev2-Apr94 9222 1451 139 History 9222 1451 140 82-06 Evaluation for JENDL-2 was made by N. Asano (SAEI), 9222 1451 141 H. Matsunobu (SAEI) and Y.Kikuchi(JAERI). 9222 1451 142 87-03 Re-evaluation for JENDL-3 was made by H.Matsunobu (SAEI) 9222 1451 143 Main part of revision was the cross sections above 10 keV 9222 1451 144 and angular and energy distributions of neutrons. 9222 1451 145 Data were compiled by T. Nakagawa (JAERI). 9222 1451 146 94-04 JENDL-3.2. 9222 1451 147 Resonance parameters reevaluated by H.Derrien (JAERI)/1/ 9222 1451 148 Cross sections reevaluated by H.Matsunobu(SAEI) 9222 1451 149 Fission spectrum reevaluated by T.Ohsawa(Kinki univ.) 9222 1451 150 Compiled by T.Nakagawa (NDC/JAERI) 9222 1451 151 9222 1451 152 ***** Modified parts for JENDL-3.2 ******************** 9222 1451 153 (2,151); New analysis with SAMMY 9222 1451 154 MF=3 ; all data except fission cross section below 6.75 9222 1451 155 MeV and total cross section 9222 1451 156 MF=4 ; for inelastic scattering 9222 1451 157 (5,18) 9222 1451 158 *********************************************************** 9222 1451 159 9222 1451 160 9222 1451 161 MF=1 General Information 9222 1451 162 MT=451 Comments and dictionary 9222 1451 163 MT=452 Nu-total 9222 1451 164 Sum of Nu-d and Nu-p 9222 1451 165 MT=455 Nu-d 9222 1451 166 Below 4 MeV 9222 1451 167 Nu-d = 0.0075094 + 4.627E-5*ln(E(MeV)) 9222 1451 168 Between 4 and 20 MeV 9222 1451 169 Based on the data of Masters et al. /2/ and Evans et al. 9222 1451 170 /3/. 9222 1451 171 MT=456 Nu-p 9222 1451 172 Renormalization was made to 3.756 of Cf-252. 9222 1451 173 Below 1 MeV 9222 1451 174 Nu-p = 2.486 + 0.1121*(E-DE), 9222 1451 175 where DE is difference of average fragment kinetic energy9222 1451 176 between incident and thermal neutron energies. It was 9222 1451 177 taken from data of Boldeman et al. /4/. 9222 1451 178 Between 1 and 2.73 MeV 9222 1451 179 Nu-p = 2.436 + 0.1279*E 9222 1451 180 Between 2.73 and 7.47 MeV 9222 1451 181 Nu-p = 2.327 + 0.1678*E 9222 1451 182 Above 7.47 MeV 9222 1451 183 Nu-p = 2.857 + 0.09689*E 9222 1451 184 9222 1451 185 MF=2 Resonance Parameters 9222 1451 186 MT=151 9222 1451 187 a) Resolved resonance region ( 1 eV to 150 eV) 9222 1451 188 Resolved resonance parameters for the Reich-Moore formula 9222 1451 189 were obtained by using SAMMY/5/. Details are given in 9222 1451 190 Appendix. 9222 1451 191 b) Unresolved resonance region ( 0.15 keV to 30 keV) 9222 1451 192 Parameters were deduced with ASREP code /6/ so as to 9222 1451 193 reproduce the evaluated cross sections in this energy 9222 1451 194 region. 9222 1451 195 9222 1451 196 2200-m/s cross sections and calculated res. integrals 9222 1451 197 2200 m/s Res. Integ. 9222 1451 198 total 588.38 b - 9222 1451 199 elastic 11.97 b - 9222 1451 200 fission 531.16 b 774 b 9222 1451 201 capture 45.25 b 138 b 9222 1451 202 9222 1451 203 MF=3 Neutron Cross Sections 9222 1451 204 Smooth part (above 30 keV) 9222 1451 205 9222 1451 206 MT=1 Total 9222 1451 207 Based on the data of Poenitz /7,8/. Between 10 and 48 9222 1451 208 keV, cross-section curve calculated with the statistical- 9222 1451 209 model code CASTHY /9/ and the coupled-channel theory code 9222 1451 210 ECIS /10/ was normalized at 48 keV. 9222 1451 211 MT=2 Elastic 9222 1451 212 Obtained by subtracting non-elastic scattering cross section 9222 1451 213 from the total cross section. 9222 1451 214 MT=4 and 51-64,91 Inelastic scattering 9222 1451 215 Calculated with CASTHY /9/ and ECIS /10/. Coupled levels were9222 1451 216 first three levels. Deformed OMP recommended by Haouat et 9222 1451 217 al. /11/, was slightly modified so as to reproduce the 9222 1451 218 experimental data of Smith et al. /12/, and spherical OMP was9222 1451 219 the same as that used for JENDL-2. In the energy range above9222 1451 220 8.25 MeV, the cross section was approximated by using an 9222 1451 221 extponential-type fuction, because the cross section curve 9222 1451 222 obtained by CASTHY and ECIS showed large fluctuation. 9222 1451 223 9222 1451 224 Deformed OMP 9222 1451 225 V =46.4-0.3*E , Ws=3.5 +0.4*E , Vso=6.2 (MeV) 9222 1451 226 r0=1.26 , rs=1.26 , rso=1.12 (fm) 9222 1451 227 a0=0.63 , b =0.52 , aso=0.47 (fm) 9222 1451 228 Beta-2=0.20, Beta-4=0.074 9222 1451 229 Spherical OMP 9222 1451 230 V =41.8-0.20*E+0.008*E**2, Ws=6.50-0.15*E, Vso=6.0 (MeV) 9222 1451 231 r0=1.31 , rs=1.36 , rso=1.32 (fm) 9222 1451 232 a0=0.57 , b =0.44 , aso=0.50 (fm) 9222 1451 233 (dir. W.S.) 9222 1451 234 9222 1451 235 Level scheme was taken from Ref. /13/. 9222 1451 236 No. Energy(MeV) Spin-Parity 9222 1451 237 g.s. 0.0 5/2 + * 9222 1451 238 1 0.04035 7/2 + * 9222 1451 239 2 0.0922 9/2 + * 9222 1451 240 3 0.1551 11/2 + * 9222 1451 241 4 0.29882 5/2 - 9222 1451 242 5 0.31191 3/2 + 9222 1451 243 6 0.3208 7/2 - 9222 1451 244 7 0.34047 5/2 + 9222 1451 245 8 0.3537 9/2 - 9222 1451 246 9 0.397 11/2 - 9222 1451 247 10 0.39849 1/2 + 9222 1451 248 11 0.41576 3/2 + 9222 1451 249 12 0.5039 7/2 - 9222 1451 250 13 0.5467 5/2 + 9222 1451 251 14 0.5971 7/2 + 9222 1451 252 Above 0.6 MeV, assumed to be overlapped. Levels with asterisk9222 1451 253 were coupled in the ECIS calculation. 9222 1451 254 9222 1451 255 MT=16,17 (n,2n) and (n,3n) 9222 1451 256 Calculated by using the EGNASH-2 code /14/. The (n,2n) cross 9222 1451 257 section was normalized to fission-spectrum-averaged value of 9222 1451 258 0.00408 b measured by Kobayashi et al./15/. The same 9222 1451 259 normalization factor was also applied to the (n,3n) cross 9222 1451 260 section. 9222 1451 261 MT=18 Fission 9222 1451 262 Based on the experimental data of Gwin et al. /16/, Carlson 9222 1451 263 et al. /17/, Manabe et al. /18/, Kanda et al. /19/, Iwasaki 9222 1451 264 et al. /20/, Meadows /21/, Lisowski et al./22/ and the 9222 1451 265 fission cross section of U-235 obtained by the simultaneous 9222 1451 266 evaluation /23/ and measured by Carlson et al./24/ between 9222 1451 267 13.25 and 20 MeV. 9222 1451 268 MT=102 Capture 9222 1451 269 In the energy range from 30 keV to 1 MeV, the alpha values 9222 1451 270 measured by Hopkins and Diven /25/ were multiplied by the 9222 1451 271 fission cross section. In the high energy region, values 9222 1451 272 calculated with CASTHY and ECIS were normalized to 0.0578 b 9222 1451 273 at 1 MeV. 9222 1451 274 MT=251 Mu-bar 9222 1451 275 Calculated with CASTHY and ECIS. 9222 1451 276 9222 1451 277 MF=4 Angular Distributions of Secondary Neutrons 9222 1451 278 MT=2, 51-64 and 91 9222 1451 279 Calculated with CASTHY and ECIS. 9222 1451 280 MT=16,17 and 18 9222 1451 281 Assumed to be isotropic in the Lab system. 9222 1451 282 9222 1451 283 MF=5 Energy Distributions of Secondary Neutrons 9222 1451 284 MT=16,17,91 9222 1451 285 Calculated with PEGASUS /26/. 9222 1451 286 MT=18 Fission spectrum 9222 1451 287 Distributions were calculated with the modified Madland-Nix 9222 1451 288 model/27,28/. The compound nucleus formation cross sections 9222 1451 289 for fission fragments (FF) were calculated using Becchetti- 9222 1451 290 Greenlees potential/29/. Up to 4th-chance-fission were 9222 1451 291 considered at high incident neuttron energies. The Ignatyuk 9222 1451 292 formula/30/ were used to generate the level density 9222 1451 293 parameters. 9222 1451 294 Parameters adopted: 9222 1451 295 Total average FF kinetic energy = 172.311-0.0212*E(MeV)9222 1451 296 Average energy release = 188.438 MeV 9222 1451 297 Average mass number of light FF = 95 9222 1451 298 Average mass number of heavy FF = 139 9222 1451 299 Level density of the light FF = 9.999- 10.094 9222 1451 300 Level density of the heavy FF = 11.89 - 12.20 9222 1451 301 Note that the parameters vary with the incident energy 9222 1451 302 within the indicated range. 9222 1451 303 MT=455 Delayed neutrons 9222 1451 304 Recommendation by Saphier et al. /31/ was adopted. 9222 1451 305 9222 1451 306 References 9222 1451 307 1) Derrien H.: to be published in J. Nucl. Sci. Technol. (1994) 9222 1451 308 2) Master C.F. et al.: Nucl. Sci. Eng., 36, 202 (1969). 9222 1451 309 3) Evans A.E. et al.: Nucl. Sci. Eng., 50, 80 (1973). 9222 1451 310 4) Boldeman J.W. et al.: Nucl. Phys., A265, 337 (1976). 9222 1451 311 5) Larson N.: ORNL/TM-9179/R1 (1985). 9222 1451 312 6) Kikuchi Y.: to be published. 9222 1451 313 7) Poenitz W.P. et al.: Nucl. Sci. Eng., 78, 333 (1981). 9222 1451 314 8) Poenitz W.P. et al.: ANL/NDM-80 (1983). 9222 1451 315 9) Igarasi S. and Fukahori T.: JAERI 1321 (1991). 9222 1451 316 10) Raynal J.: ECIS. 9222 1451 317 11) Haouat G., et al.: Nucl. Sci. Eng., 81, 491 (1982). 9222 1451 318 12) Smith A.B. et al.: 1982 Antwerp, p.039 (1982). 9222 1451 319 13) Lederer D.G. and Shirley V.S.: Table of Isotopes, 7th Ed. 9222 1451 320 (1978). 9222 1451 321 14) Yamamuro N.: JAERI-M 90-006 (1990). 9222 1451 322 15) Kobayashi K.: J. Nucl. Sci. Technol., 10, 668 (1973). 9222 1451 323 16) Gwin R. et al.: Nucl. Sci. Eng., 59, 79 (1976). 9222 1451 324 17) Carlson G.W. and Behrens J.W.: Nucl. Sci. Eng., 66, 205 (1978)9222 1451 325 18) Manabe F., et al. : 1987 Annual Meeting of Atomic Energy 9222 1451 326 Society of Japan, Nagoya, p.167 (1987) in Japanese. 9222 1451 327 19) Kanda K., et al.: 1985 Santa-Fe, p.569 (1985). 9222 1451 328 20) Iwasaki T., et al.: private communication (1987). 9222 1451 329 21) Meadows J.W.: Nucl. Sci. Eng., 54, 317 (1974). 9222 1451 330 22) Lisowski P.W.: 1991 Juelich, p.732 (1992). 9222 1451 331 23) Kanda Y. et al.: 1985 Santa Fe, 2, 1567 (1986). 9222 1451 332 24) Carlson A.D. et al.: 1991 Juelich, p.518 (1992). 9222 1451 333 25) Hopkins J.C. and Diven B.C.: Nucl. Sci. Eng., 12, 169 (1962). 9222 1451 334 26) Iijima S, et al.: JAERI-M 87-025, p.337 (1987). 9222 1451 335 27) Madland D.G. and Nix J.R.: Nucl. Sci. Eng., 81, 213 (1982). 9222 1451 336 28) Ohsawa T. and Shibata T.: 1991 Juelich Conf., 965 (1992). 9222 1451 337 29) Becchetti Jr.F.D. and Greenlees G.W.: Phys. Rev., 182, 1190 9222 1451 338 (1969). 9222 1451 339 30) Ignatyuk A.V.: Sov. J. Nucl. Phys., 29, 450 (1979). 9222 1451 340 31) Saphier D., et al.: Nucl. Sci. Eng., 62, 660 (1977). 9222 1451 341 9222 1451 342 9222 1451 343 ***************************************************************** 9222 1451 344 Appendix RESONANCE DATA ,JAERI DECEMBRE 1992 9222 1451 345 ***************************************************************** 9222 1451 346 9222 1451 347 The Reich-Moore R-matrix resonance parameters were obtained 9222 1451 348 from sequencial Bayesian fits of selected experimental total, fis-9222 1451 349 sion and capture cross sections performed with the computing code 9222 1451 350 SAMMY. The selected experimental data were the following: 9222 1451 351 1) total cross sections measured by Pattenden/1/, Moore/2/, 9222 1451 352 Kolar/3/, Harvey/4/; 9222 1451 353 2) fission cross sections measured by Weston/5,6/, Blons/7/, 9222 1451 354 Deruyter/8/, Wagemans/9/. 9222 1451 355 3) capture cross sections measured by Weston/5,6/. 9222 1451 356 In the low energy range the data of Pattenden,Moore,Harvey, 9222 1451 357 Kolar, Weston, Deruyter and Wagemans were considered.In the energy9222 1451 358 range from 15 eV to 150 eV, only the data from Kolar, Blons and 9222 1451 359 Weston could be analysed owing to the poor resolution of the other9222 1451 360 data. Some of the data were renormalized to the Axton standard 9222 1451 361 /10/ at 0.0253 eV. The fission cross section measurement of Weston9222 1451 362 /5/ available in the energy range above 1 eV was renormalised to 9222 1451 363 the data of Wagemans/9/ over the energy range from 1 eV to 20 eV, 9222 1451 364 resulting in an increase of the cross section by 2.4% compared to 9222 1451 365 the original EXFOR file. The fission cross section measurement of 9222 1451 366 Blons available in the energy range above 15 eV was renormalized 9222 1451 367 to the the data of Weston over the energy range from 15 eV to 150 9222 1451 368 eV, resulting in an increase of the cross section by 2.9% compared9222 1451 369 to the data in the original EXFOR file. The total cross section of9222 1451 370 Kolar needed a background correction of (2.3-0.038E) barn, E in 9222 1451 371 eV, in the energy range from 77 eV to 150 eV, corresponding to 0 9222 1451 372 to 1.5% of the measured transmission. These renormalizations and 9222 1451 373 background corrections were performed after preliminary fits of 9222 1451 374 the data available from the EXFOR file in order to realize a 9222 1451 375 consistency of +-2% among the cross section of the experimental 9222 1451 376 data base. The final SAMMY fits were performed without renormali- 9222 1451 377 zation and background correction parameters 9222 1451 378 The transmission data of Pattenden, Moore, Harvey and Kolar 9222 1451 379 are not available from the EXFOR file and were not requested from 9222 1451 380 the authors. The SAMMY fits were performed on the experimental ef-9222 1451 381 fective total cross sections using the sample thicknesses and the 9222 1451 382 experimental resolution to calculate the theoritical effective 9222 1451 383 cross sections. Enough informations were found in the publications9222 1451 384 by the authors to ensure the accuracy of the calculations. 9222 1451 385 Due to the high resolution of the transmission measurements of9222 1451 386 Kolar(100 m flight path) and of the fission measurements of Blons 9222 1451 387 (50 m flight path and sample cooled down at liquid nitrogen tempe-9222 1451 388 rature) the analysis could be performed up to 150 ev neutron ener-9222 1451 389 gy. The high resolution fission cross section of Cao/11/ were not 9222 1451 390 included in the experimental data base owing to a severe problem 9222 1451 391 of the renormalization of the data. 9222 1451 392 The capture cross sections of Weston were included in the fits9222 1451 393 below 30 eV only. Above 30 eV the statistical accuracy of the data9222 1451 394 was too poor and the try and error method was used in a prelimina-9222 1451 395 ry work to obtain the capture width of some strong capture reson- 9222 1451 396 ances. The capture width of the other resonances was kept at a 9222 1451 397 constant value of 41 meV close to the average value obtained by 9222 1451 398 fitting the energy range below 30 eV. Some resonances not pertain-9222 1451 399 ing to 233U were disclosed in the experimental data and were iden-9222 1451 400 tified as 195Pt resonances. The experimental data of Table 2 were 9222 1451 401 roughly corrected for these resonances. 9222 1451 402 The values of the cross sections obtained by Axton at 0.0253 9222 1451 403 eV were included in all the experimental data available in the 9222 1451 404 thermal range with the small error bars obtained by Axton, in 9222 1451 405 order to ensure the best agreement between the calculated and the 9222 1451 406 evaluated thermal values. The values calculated from the resonance9222 1451 407 parameters are the followings: 9222 1451 408 9222 1451 409 Calculation Axton 9222 1451 410 (RESENDD 0.1% 300 K) Evaluation 9222 1451 411 9222 1451 412 Fission 531.29 b 530.70+-1.34 b 9222 1451 413 Capture 45.27 b 45.62+-0.70 b 9222 1451 414 Scattering 11.99 b 12.19+-0.67 b 9222 1451 415 9222 1451 416 9222 1451 417 9222 1451 418 Tables 1 and 2 show the average cross sections calculated 9222 1451 419 from the resonance parameters compared with the average experimen-9222 1451 420 tal data and with average JENDL-3, ENDF/B-VI and JEF-2 data. 9222 1451 421 9222 1451 422 9222 1451 423 Table 1 Fission Cross Sections 9222 1451 424 9222 1451 425 ------------------------------------------------------------- 9222 1451 426 Energy Wagem Deruy Westo Blons Calcu JENDL ENDF6 JEF2 9222 1451 427 Range(eV) 9222 1451 428 ------------------------------------------------------------- 9222 1451 429 0.021-0.031 525.6 526.7 526.8 526.5 528.8 523.9 520.6 9222 1451 430 0.031-0.082 362.9 363.5 363.9 361.7 363.9 361.9 359.7 9222 1451 431 0.082-1.000 151.3 150.9 150.6 150.1 149.4 149.0 148.6 9222 1451 432 1.000-2.100 388.8 387.7 391.7 387.9 383.0 378.9 382.1 9222 1451 433 2.100-2.750 204.4 204.6 207.5 204.6 205.9 198.1 198.8 9222 1451 434 2.750-3.000 50.1 53.4 51.9 50.2 52.9 50.6 50.8 9222 1451 435 3.000-15.00 106.2 105.6 104.9 104.3 103.6 101.2 101.5 9222 1451 436 9222 1451 437 0.021-15.00 134.2 133.5 133.5 132.5 131.7 129.1 129.5 9222 1451 438 9222 1451 439 15.00-30.00 95.03 94.60 95.51 96.49 91.80 92.69 9222 1451 440 30.00-50.00 40.13 40.30 40.27 40.19 38.85 39.16 9222 1451 441 50.00-75.00 40.66 40.49 40.53 40.79 35.80 39.90 9222 1451 442 75.00-100.0 35.57 36.70 36.03 36.58 33.36 32.74 9222 1451 443 100.0-125.0 36.84 36.89 36.97 31.78 29.94 28.94 9222 1451 444 125.0-150.0 21.29 20.29 20.78 16.30 22.10 26.25 9222 1451 445 9222 1451 446 15.00-150.0 41.39 41.48 41.45 39.91 38.49 40.05 9222 1451 447 -------------------------------------------------------------- 9222 1451 448 9222 1451 449 9222 1451 450 Table 2 Capture Cross Sections 9222 1451 451 9222 1451 452 ------------------------------------------------------------- 9222 1451 453 Energy Weston Calcul JENDL3 ENDF6 JEF2.2 9222 1451 454 Range(eV) 9222 1451 455 ------------------------------------------------------------- 9222 1451 456 0.021-0.031 45.17 44.90 44.82 45.40 45.54 9222 1451 457 0.031-0.082 32.51 32.57 32.45 32.58 32.68 9222 1451 458 0.082-1.000 14.06 14.44 13.99 13.24 13.13 9222 1451 459 1.000-2.100 66.83 66.54 70.54 67.46 67.31 9222 1451 460 2.100-2.750 111.83 110.67 106.25 112.04 110.80 9222 1451 461 2.750-3.000 7.50 8.25 8.85 7.53 5.74 9222 1451 462 3.000-15.00 17.43 17.61 19.51 17.66 17.02 9222 1451 463 9222 1451 464 0.021-15.00 24.85 24.97 26.57 24.22 24.43 9222 1451 465 9222 1451 466 15.00-30.00 13.25 13.97 11.92 13.27 12.67 9222 1451 467 30.00-50.00 5.21 5.81 4.85 5.47 5.00 9222 1451 468 50.00-75.00 4.91 5.38 4.42 3.80 5.25 9222 1451 469 75.00-100.0 8.71 9.07 5.39 4.30 5.33 9222 1451 470 100.0-125.0 5.37 6.01 3.55 3.88 4.63 9222 1451 471 125.0-150.0 3.38 3.78 2.12 3.54 4.12 9222 1451 472 9222 1451 473 15.00-150.0 6.39 6.90 4.91 5.16 5.73 9222 1451 474 -------------------------------------------------------------- 9222 1451 475 The experimental capture data of Weston was increased by 9222 1451 476 a background correction according to the evaluation of Reynolds 9222 1451 477 et al./12/ and renormalized to the original value of Weston in 9222 1451 478 the energy range 1.0 eV to 2.75 eV. 9222 1451 479 9222 1451 480 9222 1451 481 RESONANCE INTEGRAL FISSION 9222 1451 482 9222 1451 483 Energy range this work JENDL-3 ENDF/B-6 JEF-2 9222 1451 484 ------------------------------------------------------ 9222 1451 485 0.5 eV-150 eV 710.34 710.53 691.08 697.18 9222 1451 486 150 eV-20 MeV 64.25 63.34 65.29 9222 1451 487 0.5 eV-20 MeV 774.79 754.43 762.47 9222 1451 488 ------------------------------------------------------ 9222 1451 489 9222 1451 490 9222 1451 491 RESONANCE INTEGRAL CAPTURE 9222 1451 492 9222 1451 493 Energy range this work JENDL-3 ENDF/B-6 JEF-2 9222 1451 494 ------------------------------------------------------ 9222 1451 495 0.5 eV-150 eV 131.92 131.77 128.79 127.51 9222 1451 496 150 eV-20 MeV 6.65 7.58 7.24 9222 1451 497 0.5 eV-20 MeV 138.42 136.37 134.75 9222 1451 498 ------------------------------------------------------ 9222 1451 499 9222 1451 500 References of Appendix 9222 1451 501 1) PATTENDEN et al.: Nucl. Sci. Eng., 17, 404 (1963) 9222 1451 502 2) MOORE et al.: Phys. Rev., 118, 714 (1960) 9222 1451 503 3) KOLAR et al.: 1970 Helsinki, Vol.I, 387 (1970) 9222 1451 504 4) HARVEY et al.: 1979 Knoxville, p.690 (1979) 9222 1451 505 5) WESTON et al.: Nucl. Sci. Eng., 34, 1 (1968) 9222 1451 506 6) WESTON et al.: Nucl. Sci. Eng., 42, 143 (1970) 9222 1451 507 7) BLONS et al.: Nucl. Sci. Eng., 51, 130 (1973) 9222 1451 508 8) DERUYTER et al.: Nucl. Sci. Eng., 54, 423 (1974) 9222 1451 509 9) WAGEMANS et al.: 1988 Mito, 91 (1988) 9222 1451 510 10) AXTON et al.: BCMN Report, GE/PH/01/86 (1986) 9222 1451 511 11) CAO et al.: 1970 Helsinki. Vol.I, 419 (1970) 9222 1451 512 12) REYNOLDS et al.: KAPL-M-7323 (1973) 9222 1451 513 9222 1451 514 ******************************************************************9222 1451 515 1 451 602 49222 1451 516 1 452 22 49222 1451 517 1 455 9 49222 1451 518 1 456 20 49222 1451 519 1 458 5 39222 1451 520 2 151 361 49222 1451 521 3 1 275 19222 1451 522 3 2 275 49222 1451 523 3 4 68 39222 1451 524 3 16 55 49222 1451 525 3 17 50 49222 1451 526 3 18 50 49222 1451 527 3 51 59 39222 1451 528 3 52 52 39222 1451 529 3 53 50 39222 1451 530 3 54 35 39222 1451 531 3 55 32 39222 1451 532 3 56 32 39222 1451 533 3 57 32 39222 1451 534 3 58 32 39222 1451 535 3 59 32 39222 1451 536 3 60 31 39222 1451 537 3 61 31 39222 1451 538 3 62 30 39222 1451 539 3 63 30 39222 1451 540 3 64 30 39222 1451 541 3 65 30 39222 1451 542 3 66 28 39222 1451 543 3 67 27 39222 1451 544 3 68 27 39222 1451 545 3 69 26 39222 1451 546 3 91 40 39222 1451 547 3 102 122 49222 1451 548 4 2 314 19222 1451 549 4 16 10 19222 1451 550 4 17 10 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2.572850+0 9.000000+5 2.578040+09222 1452 12 9.500000+5 2.583270+0 1.000000+6 2.588540+0 1.100000+6 2.599160+09222 1452 13 1.200000+6 2.609920+0 1.300000+6 2.620810+0 1.400000+6 2.631840+09222 1452 14 1.500000+6 2.643000+0 1.600000+6 2.654310+0 1.700000+6 2.665750+09222 1452 15 1.800000+6 2.677310+0 1.900000+6 2.689020+0 2.000000+6 2.700860+09222 1452 16 2.100000+6 2.712840+0 2.200000+6 2.724940+0 2.300000+6 2.737190+09222 1452 17 2.400000+6 2.749580+0 2.500000+6 2.762090+0 2.600000+6 2.774740+09222 1452 18 2.700000+6 2.787520+0 2.800000+6 2.800450+0 2.833100+6 2.804750+09222 1452 19 4.000000+6 2.998200+0 6.700000+6 3.442040+0 1.106750+7 4.165400+09222 1452 20 1.410000+7 4.667440+0 1.490000+7 4.799880+0 1.500000+7 4.816430+09222 1452 21 2.000000+7 5.644190+0 9222 1452 22 9222 1 099999 9.223300+4 2.310380+2 0 2 0 09222 1455 1 0.000000+0 0.000000+0 0 0 8 09222 1455 2 1.246670-2 2.829170-2 4.252440-2 1.330420-1 2.924672-1 6.664877-19222 1455 3 1.634781+0 3.554600+0 9222 1455 4 0.000000+0 0.000000+0 0 0 1 99222 1455 5 9 2 0 0 0 09222 1455 6 1.000000-5 6.730000-3 2.530000-2 6.730000-3 5.000000+4 7.358500-39222 1455 7 4.000000+6 7.974620-3 6.700000+6 4.617440-3 1.106750+7 4.617440-39222 1455 8 1.410000+7 4.390000-3 1.490000+7 4.330000-3 2.000000+7 3.947500-39222 1455 9 9222 1 099999 9.223300+4 2.310380+2 0 2 0 09222 1456 1 0.000000+0 0.000000+0 0 0 1 499222 1456 2 49 2 0 0 0 09222 1456 3 1.000000-5 2.480980+0 2.530000-2 2.480980+0 1.000000+1 2.480980+09222 1456 4 5.000000+2 2.486950+0 1.000000+3 2.486950+0 1.000000+3 2.482900+09222 1456 5 1.000000+4 2.482900+0 1.000000+4 2.486680+0 1.000000+5 2.486680+09222 1456 6 1.000000+5 2.491630+0 1.500000+5 2.496310+0 2.000000+5 2.501020+09222 1456 7 2.500000+5 2.505770+0 3.000000+5 2.510550+0 3.500000+5 2.515360+09222 1456 8 4.000000+5 2.520210+0 4.500000+5 2.525090+0 5.000000+5 2.530010+09222 1456 9 5.500000+5 2.534960+0 6.000000+5 2.539940+0 6.500000+5 2.544960+09222 1456 10 7.000000+5 2.550010+0 7.500000+5 2.555100+0 8.000000+5 2.560220+09222 1456 11 8.500000+5 2.565370+0 9.000000+5 2.570550+0 9.500000+5 2.575770+09222 1456 12 1.000000+6 2.581030+0 1.100000+6 2.591640+0 1.200000+6 2.602380+09222 1456 13 1.300000+6 2.613260+0 1.400000+6 2.624270+0 1.500000+6 2.635420+09222 1456 14 1.600000+6 2.646710+0 1.700000+6 2.658130+0 1.800000+6 2.669680+09222 1456 15 1.900000+6 2.681370+0 2.000000+6 2.693200+0 2.100000+6 2.705160+09222 1456 16 2.200000+6 2.717250+0 2.300000+6 2.729480+0 2.400000+6 2.741850+09222 1456 17 2.500000+6 2.754350+0 2.600000+6 2.766980+0 2.700000+6 2.779750+09222 1456 18 2.800000+6 2.792660+0 2.833100+6 2.796960+0 1.500000+7 4.812110+09222 1456 19 2.000000+7 5.640240+0 9222 1456 20 0.000000+0 0.000000+0 0 0 0 09222 1 099999 9.223300+4 2.310380+2 0 0 0 09222 1458 1 0.000000+0 0.000000+0 0 0 18 99222 1458 2 1.682200+8 5.000000+5 4.900000+6 1.000000+5 3.100000+4 4.650000+39222 1458 3 7.720000+6 5.200000+5 5.010000+6 6.000000+4 5.160000+6 6.000000+49222 1458 4 6.930000+6 9.000000+4 1.910400+8 2.284000+5 1.979700+8 2.100000+59222 1458 5 9222 1 099999 9222 0 0 0 9.223300+4 2.310380+2 0 0 1 09222 2151 1 9.223300+4 1.000000+0 0 1 2 09222 2151 2 1.000000-5 1.500000+2 1 3 0 09222 2151 3 2.500000+0 9.620000-1 0 0 1 19222 2151 4 2.310380+2 0.000000+0 0 0 1140 1909222 2151 5 -1.701260+2 3.000000+0 2.505000-1 4.000000-2-1.535000-3 2.720000+09222 2151 6 -1.909390+1 2.000000+0 4.207000-2 4.000000-2 0.000000+0-2.385000-19222 2151 7 -7.970770+0 3.000000+0 5.417000-3 4.000000-2-5.528000-1 0.000000+09222 2151 8 -3.850400+0 3.000000+0 9.742000-4 3.000000-2 1.485000+0 9.105000-39222 2151 9 -2.810000+0 3.000000+0 9.796000-4 9.419000-2 3.811000-3 1.357000+09222 2151 10 1.698170-1 3.000000+0 1.491000-7 6.502000-2 5.588000-2-1.351000-39222 2151 11 4.401140-1 2.000000+0 1.526000-6 9.145000-4 2.318000-1 1.107000-19222 2151 12 1.427570+0 2.000000+0 2.064000-4 4.251000-2 2.133000-2-6.152000-19222 2151 13 1.767460+0 3.000000+0 2.473000-4 4.018000-2-1.829000-1 1.188000-29222 2151 14 2.304780+0 3.000000+0 1.548000-4 3.843000-2 5.444000-2 5.918000-59222 2151 15 3.357970+0 2.000000+0 2.190000-4 3.500000-2 8.190000-1 4.211000-49222 2151 16 3.627080+0 3.000000+0 5.927000-5 3.884000-2 9.213000-2 7.547000-59222 2151 17 4.486160+0 2.000000+0 3.581000-4 3.500000-2-5.513000-5-1.006000+09222 2151 18 5.800280+0 3.000000+0 9.221000-5 3.705000-2 3.212000-1-1.453000-49222 2151 19 6.628340+0 2.000000+0 9.819000-5 4.100000-2 9.103000-6 2.932000-19222 2151 20 6.816230+0 3.000000+0 8.078000-4 3.854000-2-1.329000-1-6.432000-39222 2151 21 7.492010+0 3.000000+0 5.594000-5 2.800000-2 1.261000-1 1.658000-19222 2151 22 8.188730+0 2.000000+0 2.762000-4 2.500000-2 1.668000+0 9.307000-49222 2151 23 8.623480+0 2.000000+0 3.646000-5 5.560000-2-5.782000-3-4.216000-19222 2151 24 9.154360+0 3.000000+0 8.581000-5 3.684000-2 1.638000-1 1.023000-19222 2151 25 1.038130+1 3.000000+0 1.350000-3 4.684000-2-2.554000-1 5.909000-49222 2151 26 1.138530+1 3.000000+0 1.934000-4 4.662000-2-1.292000-2-3.395000-19222 2151 27 1.184180+1 2.000000+0 1.792000-4 4.100000-2-4.844000-1-2.694000-29222 2151 28 1.279370+1 3.000000+0 1.343000-3 3.965000-2 1.212000-1-1.818000-19222 2151 29 1.344730+1 3.000000+0 5.385000-5 5.881000-2 7.963000-2 4.429000-29222 2151 30 1.364280+1 2.000000+0 4.755000-4 3.890000-2-2.966000-1 8.274000-29222 2151 31 1.533060+1 3.000000+0 7.488000-4 5.082000-2-2.825000-4-1.604000-19222 2151 32 1.552340+1 3.000000+0 4.011000-5 3.800000-2 1.011000-2-2.587000-39222 2151 33 1.620440+1 3.000000+0 5.785000-4 3.500000-2 2.544000-1 1.186000-19222 2151 34 1.657750+1 3.000000+0 6.606000-4 4.053000-2-5.831000-2 1.423000-19222 2151 35 1.704720+1 2.000000+0 2.852000-5 4.100000-2 2.167000+0-7.448000-29222 2151 36 1.807410+1 2.000000+0 3.715000-4 4.162000-2 3.057000-1 5.663000-49222 2151 37 1.853370+1 3.000000+0 2.946000-4 4.500000-2-1.809000-1 8.481000-29222 2151 38 1.894660+1 3.000000+0 1.579000-3 4.324000-2-2.347000-1-2.980000-29222 2151 39 2.061500+1 3.000000+0 7.288000-4 4.600000-2 1.284000-1-1.869000-19222 2151 40 2.188760+1 3.000000+0 1.024000-3 5.888000-2-1.001000-3 1.823000-19222 2151 41 2.225310+1 2.000000+0 5.320000-3 3.548000-2 1.920000-1 5.056000-19222 2151 42 2.235910+1 2.000000+0 5.388000-4 4.100000-2 8.589000-1 2.340000-29222 2151 43 2.251620+1 3.000000+0 9.686000-5 1.727000-2 1.330000-1-3.041000-39222 2151 44 2.376290+1 3.000000+0 4.125000-4 3.000000-2 3.750000-1-2.909000-29222 2151 45 2.422440+1 2.000000+0 4.427000-4 2.500000-2-8.873000-2-6.386000-19222 2151 46 2.528300+1 3.000000+0 6.377000-4 3.868000-2-2.555000-1-5.054000-39222 2151 47 2.578570+1 2.000000+0 1.541000-3 2.500000-2 2.654000-2 1.231000+09222 2151 48 2.660890+1 3.000000+0 3.098000-4 5.579000-2-1.010000-2 2.227000-19222 2151 49 2.776370+1 3.000000+0 4.103000-4 3.200000-2 7.243000-1 5.728000-19222 2151 50 2.825500+1 2.000000+0 1.660000-4 3.200000-2-1.964000-1 1.955000-19222 2151 51 2.846250+1 3.000000+0 4.102000-5 7.798000-2 7.941000-3-7.087000-29222 2151 52 2.908300+1 3.000000+0 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