9225 0 0 0 9.22340+ 4 2.32030+ 2 1 1 0 29225 1451 1 0.00000+ 0 1.00000+ 0 0 0 69225 1451 2 1.00000+ 0 2.00000+ 7 0 0 10 69225 1451 3 0.00000+ 0 0.00000+ 0 0 0 254 1069225 1451 4 92-U-234 EVAL-September2002 9225 1451 5 DIST-October2002 9225 1451 6 ----B-404-ISTC MATERIAL 9225 9225 1451 7 -----INCIDENT NEUTRON DATA 9225 1451 8 -----ENDF/B-VI FORMAT 9225 1451 9 *****************************************************************9225 1451 10 UNRESOLVED RESONANCE PARAMETERS FOR 1.5-140 KEV REGION, 9225 1451 11 TOTAL, ELASTIC SCATTERING, INELASTIC SCATTERING, FISSION, 9225 1451 12 CAPTURE,(N,2N) AND (N,3N) CROSS SECTIONS AS WELL AS 9225 1451 13 ANGULAR AND ENERGY DISTRIBUTIONS OF SECONDARY AND PROMPT 9225 1451 14 FISSION NEUTRONS WERE EVALUATED BY 9225 1451 15 V.M. MASLOV, Yu.V. PORODZINSKIJ, N.A. TETEREVA, 9225 1451 16 M. BABA, A. HASEGAWA, 9225 1451 17 N.V. KORNILOV, A.B. KAGALENKO /1/. 9225 1451 18 9225 1451 19 9225 1451 20 MF=1 GENERAL INFORMATION 9225 1451 21 MT=451 DESCRIPTIVE DATA AND DICTIONARY 9225 1451 22 MT=452 NUMBER OF NEUTRONS AND DICTIONARY 9225 1451 23 SUM OF MT=455 and 456. 9225 1451 24 MT=455 DELAYED NEUTRONS PER FISSION 9225 1451 25 ARE DEFINED USING SYSTEMATICS BY TUTTLE/2/. SIX GROUP 9225 1451 26 DECAY CONSTANTS WERE ADOPTED FROM BRADY ET AL./3/ 9225 1451 27 9225 1451 28 MT=456 PROMPT NEUTRONS NUMBER 9225 1451 29 ESTIMATED WITH SYSTEMATICS /4/, WHICH WAS NORMALIZED IN 9225 1451 30 THE ENERGY RANGE 2.5-4 MeV TO MEASURED DATA BY MATHER ET 9225 1451 31 AL./5/. NU-BAR OF Cf-252 SPONTANEOUS FISSION WAS ASSUMED 9225 1451 32 TO BE 3.756. FOR INCIDENT NEUTRON ENERGIES HIGHER THAN 9225 1451 33 (N,NF) REACTION THRESHOLD, NU-BAR WAS CALCULATED TAKING 9225 1451 34 INTO ACCOUNT PARTIAL CONTRIBUTIONS OF (N,XNF) REACTIONS 9225 1451 35 /1/. 9225 1451 36 9225 1451 37 MF=2 RESONANCE PARAMETERS 9225 1451 38 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 9225 1451 39 (RESOLVED RESONANCE REGION = 1.0E-5 EV TO 1.5 KEV), 9225 1451 40 (UNRESOLVED RESONANCE REGION = 1.5 KEV TO 140 KEV) 9225 1451 41 9225 1451 42 RESOLVED MLBW RESONANCE PARAMETERS RECOMMENDED IN 9225 1451 43 JENDL-3.2 WERE ADOPTED. THESE ARE RESONANCE PARAMETERS BY 9225 1451 44 JAMES ET AL./6/, MODIFIED ASSUMING AVERAGE RADIATION 9225 1451 45 WIDTH OF 0.026 EV. FISSION WIDTH OF NEGATIVE 2.06-eV RESO-9225 1451 46 NANCE WAS VARIED TO FIT THERMAL FISSION CROSS SECTION 9225 1451 47 VALUE BY WAGEMANS ET AL./7/ 9225 1451 48 9225 1451 49 ENERGY-DEPENDENT UNRESOLVED RESONANCE PARAMETERS COVER 9225 1451 50 ENERGY RANGE FROM 1.5 TO 140 KEV. PARAMETERS WERE 9225 1451 51 OBTAINED TO REPRODUCE SMOOTH TOTAL AND CAPTURE CROSS 9225 1451 52 SECTIONS, CALCULATED WITH STATISTICAL MODEL. CAPTURE 9225 1451 53 CROSS SECTION DATA BY MURADYAN ET AL. /8/ IN THE ENERGY 9225 1451 54 RANGE OF 0.03 - 2 KeV ARE DESCRIBED. 9225 1451 55 ENDF/B PROCESSING CODES /9,10/ IGNORE DIRECT INELASTIC 9225 1451 56 SCATTERING CONTRIBUTION. TO COMPENSATE THAT DEFICIENCY WE 9225 1451 57 INCREASED AVERAGE INELASTIC SCATTERING WIDTHS, CAPTURE 9225 1451 58 WIDTHS ABOVE 50 KEV ALSO WAS SLIGHTLY INCREASED TO KEEP 9225 1451 59 CAPTURE CROSS SECTION UNDISTORTED AS COMPARED WITH 9225 1451 60 CALCULATED BY PHYSICALLY CORRECT (PC) CODES. AS A RESULT, 9225 1451 61 TOTAL,ELASTIC SCATTERING AND CAPTURE CROSS SECTIONS, 9225 1451 62 CALCULATED WITH THESE PCC CODES,ARE REPRODUCED WITH 9225 1451 63 CONVENTIONAL ENDF PROCESSING CODES USING AVERAGE RESONANCE9225 1451 64 PARAMETERS GIVEN MF=2 MT=151. 9225 1451 65 9225 1451 66 2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9225 1451 67 2200 M/S(B) RES. INTEG.(B) 9225 1451 68 TOTAL 119.23 9225 1451 69 ELASTIC 19.416 9225 1451 70 FISSION 67.96 6.637 9225 1451 71 CAPTURE 99.75 631.980 9225 1451 72 9225 1451 73 MF=3 NEUTRON CROSS SECTIONS 9225 1451 74 FROM 1.5 KEV UP TO 140 KEV EVALUATED CROSS SECTIONS WERE 9225 1451 75 REPRESENTED WITH THE UNRESOLVED RESONANCE PARAMETERS. 9225 1451 76 9225 1451 77 MT= 1, 2, 4, 51-86, 91 - TOTAL, ELASTIC AND INELASTIC 9225 1451 78 SCATTERING CROSS SECTIONS. 9225 1451 79 TOTAL, ELASTIC AND DIRECT INELASTIC FOR ROTATIONAL GROUND 9225 1451 80 STATE BAND LEVELS MT=51,52,53,54 (COUPLED LEVELS) 9225 1451 81 AS WELL AS OPTICAL TRANSMISSION COEFFICIENTS ARE OBTAINED 9225 1451 82 IN A RIGID ROTATOR MODEL COUPLED CHANNELS CALCULATIONS. 9225 1451 83 DIRECT EXCITATION OF GAMMA- AND BETA-VIBRATIONAL, OCTUPOLE9225 1451 84 AND K=2+ QUADRUPOLE BAND LEVELS,MT=56-65,67-73,75,76,78, 9225 1451 85 80,83,85,86 ARE OBTAINED IN A SOFT ROTATOR MODEL COUPLED 9225 1451 86 CHANNEL CALCULATIONS, FOR NORMALIZATION PURPOSES THESE 9225 1451 87 DIRECT INELASTIC CROSS SECTIONS WERE SUBTRACTED FROM MT=2 9225 1451 88 ELASTIC SCATTERING CROSS SECTION. DIRECT INELASTIC 9225 1451 89 CONTRIBUTIONS WERE ADDED INCOHERENTLY TO HAUSER-FESHBACH 9225 1451 90 CALCULATIONS OF COMPOUND NUCLEUS INELASTIC SCATTERING 9225 1451 91 CROSS SECTIONS. 9225 1451 92 9225 1451 93 THE DEFORMED OPTICAL POTENTIAL ADOPTED WAS THAT FOR 232Th,9225 1451 94 THEN EVALUATED VALUE OF S-WAVE STRENGTH FUBCTION 9225 1451 95 S0= 0.95x10-4(EV)-1/2 WAS FITTED: 9225 1451 96 9225 1451 97 VR=(45.722-0.334xE) MEV; RR =1.2668 FM; AR =.6468 FM; 9225 1451 98 WD=(3.145+0.455xE)MEV; E< 8 MEV RD =1.25 FM; 9225 1451 99 WD= 6.785 MEV; E>= 8 MEV AD =.5246 FM; 9225 1451 100 VSO= 6.2 MEV; RS0=1.12 FM; ASO=.47 FM; 9225 1451 101 B2= .190; B4=.072; 9225 1451 102 9225 1451 103 9225 1451 104 FISSION, CAPTURE AND COMPOUND INELASTIC SCATTERING CROSS 9225 1451 105 SECTIONS WERE CALCULATED WITH HAUSER-FESHBACH-MOLDAUER/11/ 9225 1451 106 APPROACH, AT INCIDENT NEUTRON ENERGIES HIGHER THAN 1.3 MEV 9225 1451 107 (LEVEL OVERLAPPING ENERGY) TEPEL ET AL./12/ THEORY WAS 9225 1451 108 EMPLOYED. 9225 1451 109 9225 1451 110 ADOPTED LEVEL SCHEME OF U-234 FROM NUCLEAR DATA SHEETS /13/. 9225 1451 111 9225 1451 112 9225 1451 113 LEVEL SCHEME: 9225 1451 114 -------------------------------------------------------- 9225 1451 115 NO. ENERGY(MEV) SPIN-PARITY K-PARITY* 9225 1451 116 -------------------------------------------------------- 9225 1451 117 9225 1451 118 G.S. .000000+00 0+ 0+ 9225 1451 119 .434980-01 2+ 0+ 9225 1451 120 .143350-00 4+ 0+ 9225 1451 121 .296070-00 6+ 0+ 9225 1451 122 .497040+00 8+ 0+ 9225 1451 123 .741200+00 10+ 0+ 9225 1451 124 .786290+00 1- 0- 9225 1451 125 .809880+00 0+ 0+ 9225 1451 126 .849300+00 3- 0- 9225 1451 127 .851700+00 2+ 0+ 9225 1451 128 .926740+00 2+ 2+ 9225 1451 129 .947850+00 4+ 0+ 9225 1451 130 .962600+00 5- 0- 9225 1451 131 .968600+00 3+ 2+ 9225 1451 132 .989450+00 2- 2- 9225 1451 133 .102370+01 4+ 2+ 9225 1451 134 .102380+01 12+ 0+ 9225 1451 135 .102383+01 3- 2- 9225 1451 136 .104450+01 0+ 0+ 9225 1451 137 .106930+01 4- 2- 9225 1451 138 .108530+01 2+ 0+ 9225 1451 139 .109090+01 5+ 2+ 9225 1451 140 .109590+01 6+ 0+ 9225 1451 141 .112520+01 7- 0- 9225 1451 142 .112670+01 2+ 9225 1451 143 .112760+01 5- 2- 9225 1451 144 .115000+01 4+ 0+ 9225 1451 145 .116520+01 3+ 9225 1451 146 .117210+01 6+ 2+ 9225 1451 147 .117420+01 1+ 9225 1451 148 .119470+01 6- 2- 9225 1451 149 .121460+01 4+ 9225 1451 150 .123720+01 1- 9225 1451 151 .126180+01 7+ 2+ 9225 1451 152 .127440+01 5+ 9225 1451 153 .127750+01 7- 2- 9225 1451 154 .129260+01 8+ 0+ 9225 1451 155 9225 1451 156 *) K-PARITY ARE SHOWN ONLY FOR THE LEVELS, 9225 1451 157 IDENTIFIED WITHIN RIGID AND SOFT ROTATOR MODELS 9225 1451 158 9225 1451 159 9225 1451 160 OVERLAPPING LEVELS ARE ASSUMED ABOVE 1.3 MEV 9225 1451 161 9225 1451 162 9225 1451 163 MT=16,17,37. (N,2N) AND (N,3N) CROSS SECTION FROM 9225 1451 164 STATISTICAL MODEL CALCULATIONS /1/ WITH ACCOUNT OF 9225 1451 165 PRE-EQUILIBRIUM NEUTRON EMISSION (MODIFIED STAPRE CODE/14/ 9225 1451 166 WAS USED). PRE-EQUILIBRIUM NEUTRON EMISSION CONTRIBUTION WAS 9225 1451 167 FIXED ACCORDING TO CONSISTENT DESCRIPTION OF(N,F) AND (N,XN) 9225 1451 168 REACTION DATA FOR 238U AND 232Th TARGET NUCLIDES. 9225 1451 169 9225 1451 170 MT=18, 19, 20, 21,38. FISSION CROSS SECTION IS CALCULATED 9225 1451 171 WITHIN STATISTICAL MODEL /1/. MEASURED FISSION DATA /15-28/ 9225 1451 172 ANALYSIS WAS ACCOMPLISHED. THE CONTRIBUTION OF EMISSIVE 9225 1451 173 (N,NF) AND (N,2NF) FISSION TO THE TOTAL FISSION CROSS SECTION9225 1451 174 WAS ESTIMATED USING FISSION BARRIER PARAMETERS OF 234-U AND 9225 1451 175 233-U, WHICH FIT 233-U(N,F) AND 232-U(N,F) CROSS SECTION 9225 1451 176 DATA. 9225 1451 177 MT=102 CAPTURE 9225 1451 178 CAPTURE CROSS SECTION IS CALCULATED WITHIN A STATISTICAL MO- 9225 1451 179 DEL. ABOVE NEUTRON ENERGY 5 MEV CAPTURE IS ASSUMED TO BE 9225 1451 180 CONSTANT. COMPETITION OF (N,GF) AND (N,GN') REACTIONS IS 9225 1451 181 TAKEN INTO ACCOUNT. ADOPTED ESTIMATE OF RADIATION CAPTURE 9225 1451 182 CROSS SECTION IS CONSISTENT WITH CAPTURE CROSS SECTION DATA 9225 1451 183 BY MURADYAN ET AL. /8/ IN THE ENERGY RANGE OF 0.03 - 2 KeV. 9225 1451 184 9225 1451 185 MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 9225 1451 186 FOR MT=2,51,52,53 AND 54 FROM COUPLED CHANNEL CALCULATIONS 9225 1451 187 (RIGID ROTATOR MODEL), 9225 1451 188 FOR MT=56-65,67-73,75,76,78,80,83,85,86 FROM COUPLED CHANNEL 9225 1451 189 MODEL (SOFT ROTATOR MODEL) WITH ADDED ISOTROPIC STATISTICAL 9225 1451 190 CONTRIBUTION. 9225 1451 191 9225 1451 192 MT=16, 17, 18-21, 38, 66,74,77,79,81,82,84 AND 91 ARE ISOTROPIC 9225 1451 193 IN THE LAB SYSTEM. 9225 1451 194 9225 1451 195 MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS 9225 1451 196 9225 1451 197 ENERGY DISTRIBUTIONS FOR MT=16,17,91 WERE CALCULATED WITH 9225 1451 198 A HAUSER-FESHBACH STATISTICAL MODEL OF CASCADE NEUTRON 9225 1451 199 EMISSION TAKING INTO ACCOUNT THE HISTORY OF THE DECAY WITH 9225 1451 200 THE ALLOWANCE OF PREEQUILIBRIUM EMISSION OF THE FIRST 9225 1451 201 NEUTRON, SIMULTANEOUSLY WITH (N,F) AND (N,XNF) REACTION CROSS9225 1451 202 SECTIONS. 9225 1451 203 9225 1451 204 MT=18,19,20,21,38 9225 1451 205 PROMPT FISSION NEUTRON SPECTRA (PFNS)WERE CALCULATED WITH THE9225 1451 206 SEMI-EMPIRICAL MODEL/1/, PRE-FISSION NEUTRON EMISSION IN 9225 1451 207 (N,XNF) REACTION, EITHER EQUILIBRIUM AND PRE-EQUILIBRIUM 9225 1451 208 MODES ARE INCLUDED. SPECTRA OF PRE-FISSION (N,XNF) NEUTRONS 9225 1451 209 ARE CALCULATED WITH HAUSER-FESHBACH STATISTICAL MODEL. 9225 1451 210 BASICALLY PFNS FROM FISSION FRAGMENTS (FF) WERE CALCULATED AS9225 1451 211 A SUPERPOSITION OF TWO WATT DISTRIBUTIONS FOR LIGHT AND HEAVY9225 1451 212 FF WITH EQUAL CONTRIBUTIONS, BUT DIFFERENT TEMPERATURE 9225 1451 213 PARAMETERS. FF KINETIC ENERGY, ONE MORE MODEL PARAMETER, 9225 1451 214 MIGHT BE LOWER THAN TKE, WHICH REFLECTS IT'S DEPENDENS ON THE9225 1451 215 MOMENT OF NEUTRON EMISSION. THIS EFFECTIVELY REDUCES AVERAGE 9225 1451 216 ENERGY OF PFNS FOR INCIDENT NEUTRON ENERGIES ABOVE EMISSIVE 9225 1451 217 FISSION THRESHOLD. 9225 1451 218 9225 1451 219 REFERENCES 9225 1451 220 1) Maslov V., Porodzinskij Yu., Baba M.,Hasegawa A., Kornilov 9225 1451 221 N., Kagalenko A., Tetereva N.A. JAERI-Research 01-0XX, 2002. 9225 1451 222 2) Tuttle R.J.: INDC(NDS)-107/G+Special, p.29 (1979). 9225 1451 223 3) Brady M.C. and England T.R.: Nucl. Sci. 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Proc. 6th All-Union Conf. on Neutron 9225 1451 256 Physics, Kiev, 2-6 Oct. 1983,2, 134 (1983). 9225 1451 257 ******************************************* 9225 1451 258 1 451 364 09225 1451 259 1 452 8 09225 1451 260 1 455 10 09225 1451 261 1 456 8 09225 1451 262 2 151 489 09225 1451 263 3 1 33 09225 1451 264 3 2 31 09225 1451 265 3 4 32 09225 1451 266 3 16 9 09225 1451 267 3 17 7 09225 1451 268 3 18 31 09225 1451 269 3 19 31 09225 1451 270 3 20 10 09225 1451 271 3 21 8 09225 1451 272 3 38 5 09225 1451 273 3 51 32 09225 1451 274 3 52 29 09225 1451 275 3 53 28 09225 1451 276 3 54 27 09225 1451 277 3 55 19 09225 1451 278 3 56 25 09225 1451 279 3 57 25 09225 1451 280 3 58 24 09225 1451 281 3 59 24 09225 1451 282 3 60 23 09225 1451 283 3 61 23 09225 1451 284 3 62 23 09225 1451 285 3 63 22 09225 1451 286 3 64 22 09225 1451 287 3 65 21 09225 1451 288 3 66 8 09225 1451 289 3 67 21 09225 1451 290 3 68 20 09225 1451 291 3 69 20 09225 1451 292 3 70 20 09225 1451 293 3 71 19 09225 1451 294 3 72 19 09225 1451 295 3 73 19 09225 1451 296 3 74 12 09225 1451 297 3 75 18 09225 1451 298 3 76 18 09225 1451 299 3 77 11 09225 1451 300 3 78 17 09225 1451 301 3 79 11 09225 1451 302 3 80 16 09225 1451 303 3 81 10 09225 1451 304 3 82 9 09225 1451 305 3 83 15 09225 1451 306 3 84 9 09225 1451 307 3 85 14 09225 1451 308 3 86 14 09225 1451 309 3 91 14 09225 1451 310 3 102 25 09225 1451 311 4 2 109 09225 1451 312 4 16 2 09225 1451 313 4 17 2 09225 1451 314 4 18 2 09225 1451 315 4 19 2 09225 1451 316 4 20 2 09225 1451 317 4 21 2 09225 1451 318 4 38 2 09225 1451 319 4 51 109 09225 1451 320 4 52 103 09225 1451 321 4 53 97 09225 1451 322 4 54 94 09225 1451 323 4 55 2 09225 1451 324 4 56 91 09225 1451 325 4 57 88 09225 1451 326 4 58 88 09225 1451 327 4 59 88 09225 1451 328 4 60 88 09225 1451 329 4 61 88 09225 1451 330 4 62 88 09225 1451 331 4 63 88 09225 1451 332 4 64 88 09225 1451 333 4 65 85 09225 1451 334 4 66 2 09225 1451 335 4 67 85 09225 1451 336 4 68 86 09225 1451 337 4 69 85 09225 1451 338 4 70 86 09225 1451 339 4 71 85 09225 1451 340 4 72 85 09225 1451 341 4 73 85 09225 1451 342 4 74 2 09225 1451 343 4 75 85 09225 1451 344 4 76 86 09225 1451 345 4 77 2 09225 1451 346 4 78 85 09225 1451 347 4 79 2 09225 1451 348 4 80 85 09225 1451 349 4 81 2 09225 1451 350 4 82 2 09225 1451 351 4 83 81 09225 1451 352 4 84 2 09225 1451 353 4 85 81 09225 1451 354 4 86 81 09225 1451 355 4 91 2 09225 1451 356 5 16 1072 09225 1451 357 5 17 583 09225 1451 358 5 18 2337 09225 1451 359 5 19 2337 09225 1451 360 5 20 1785 09225 1451 361 5 21 1119 09225 1451 362 5 38 453 09225 1451 363 5 91 1028 09225 1451 364 9225 1 0 365