9137 0 0 0 9.123300+4 2.310380+2 1 1 0 29137 1451 1 0.000000+0 1.000000+0 0 0 0 69137 1451 2 1.000000+0 2.000000+7 0 0 10 69137 1451 3 0.000000+0 0.000000+0 0 0 189 689137 1451 4 91-Pa-233 EVAL-September 2003 9137 1451 5 DIST-October 2003 9137 1451 6 ----ENDF/B-VI MATERIAL 9137 9137 1451 7 -----INCIDENT NEUTRON DATA 9137 1451 8 ------ENDF-6 FORMAT 9137 1451 9 ----B-404-ISTC MATERIAL 9137 9137 1451 10 -----INCIDENT NEUTRON DATA 9137 1451 11 -----ENDF/B-VI FORMAT 9137 1451 12 *****************************************************************9137 1451 13 UNRESOLVED RESONANCE PARAMETERS FOR 16.5 EV-70 KEV REGION, 9137 1451 14 TOTAL, ELASTIC SCATTERING, INELASTIC SCATTERING, FISSION, 9137 1451 15 CAPTURE,(N,2N) AND (N,3N) CROSS SECTIONS AS WELL AS 9137 1451 16 ANGULAR AND ENERGY DISTRIBUTIONS OF SECONDARY AND PROMPT 9137 1451 17 FISSION NEUTRONS WERE EVALUATED BY 9137 1451 18 V.M. MASLOV,N.A. TETEREVA, 9137 1451 19 M. BABA, A. HASEGAWA, 9137 1451 20 N.V. KORNILOV, A.B. KAGALENKO /1/. 9137 1451 21 9137 1451 22 9137 1451 23 MF=1 GENERAL INFORMATION 9137 1451 24 MT=451 DESCRIPTIVE DATA AND DICTIONARY 9137 1451 25 MT=452 NUMBER OF NEUTRONS 9137 1451 26 SUM OF MT=455 and 456. 9137 1451 27 MT=455 DELAYED NEUTRONS PER FISSION ARE ADOPTED FROM JENL-3.3, 9137 1451 28 /2/,WHICH ARE BASED ON TUTTLE' RECOMENDATIONS /3/. 9137 1451 29 MT=456 PROMPT NEUTRONS NUMBER 9137 1451 30 ESTIMATED WITH SYSTEMATICS /4/. FOR INCIDENT NEUTRON 9137 1451 31 ENERGIES HIGHER THAN (N,NF) REACTION THRESHOLD, NU-BAR 9137 1451 32 WAS CALCULATED TAKING INTO ACCOUNT PARTIAL CONTRIBUTIONS 9137 1451 33 OF (N,XNF) REACTIONS /1,5/. 9137 1451 34 9137 1451 35 MF=2, MT=151 RESONANCE PARAMETERS 9137 1451 36 ADOPTED FROM JENDL-3.3/4/. 9137 1451 37 RESOLVED RESONANCES FOR SLBW FORMULA: FROM 9137 1451 38 1.0E-5 TO 16.5 EVPARAMETERS WERE TAKEN FROM THE 9137 1451 39 RECOMMENDATION BY MUGHABGHAB /6/. 9137 1451 40 9137 1451 41 9137 1451 42 2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9137 1451 43 2200 M/S(B) RES. INTEG.(B) 9137 1451 44 TOTAL 53.05 - 9137 1451 45 ELASTIC 13.02 - 9137 1451 46 FISSION 0. 1.61 9137 1451 47 CAPTURE 40.03 872.89 9137 1451 48 9137 1451 49 MF=3 NEUTRON CROSS SECTIONS 9137 1451 50 FROM 16.5 EV UP TO 78 KEV EVALUATED CROSS SECTIONS WERE 9137 1451 51 REPRESENTED WITH THE UNRESOLVED RESONANCE PARAMETERS. 9137 1451 52 ENERGY-DEPENDENT UNRESOLVED RESONANCE PARAMETERS WERE 9137 1451 53 OBTAINED TO REPRODUCE FISSION CROSS SECTION, CALCULATED 9137 1451 54 WITH STATISTICAL MODEL. 9137 1451 55 9137 1451 56 MT= 1, 2, 4, 51-86, 91 - TOTAL, ELASTIC AND INELASTIC 9137 1451 57 SCATTERING CROSS SECTIONS. 9137 1451 58 TOTAL, ELASTIC AND DIRECT INELASTIC FOR ROTATIONAL GROUND9137 1451 59 STATE BAND LEVELS MT=51,53,66(COUPLED LEVELS) 9137 1451 60 AS WELL AS OPTICAL TRANSMISSION COEFFICIENTS ARE OBTAINED9137 1451 61 IN A RIGID ROTATOR MODEL COUPLED CHANNELS CALCULATIONS. 9137 1451 62 DIRECT INELASTIC CONTRIBUTIONS WERE ADDED INCOHERENTLY 9137 1451 63 TO HAUSER-FESHBACH CALCULATIONS OF COMPOUND NUCLEUS 9137 1451 64 INELASTIC SCATTERING CROSS SECTIONS. 9137 1451 65 9137 1451 66 THE DEFORMED OPTICAL POTENTIAL ADOPTED WAS THAT FOR 9137 1451 67 232Th, /7/ THEN EVALUATED VALUE OF S-WAVE STRENGTH 9137 1451 68 FUNCTION S0= 0.80x10-4(EV)-1/2 WAS FITTED: 9137 1451 69 9137 1451 70 VR=(45.722-0.334xE) MEV; RR =1.2668 FM; AR =.6468 FM; 9137 1451 71 WD=(3.145+0.455xE) MEV; E< 8 MEV RD =1.25 FM; 9137 1451 72 WD= 6.785 MEV; E>= 8 MEV AD =.5246 FM; 9137 1451 73 VSO= 6.2 MEV; RS0=1.12 FM; ASO=.47 FM; 9137 1451 74 B2= .179; B4=.070; 9137 1451 75 9137 1451 76 9137 1451 77 FISSION, CAPTURE AND COMPOUND INELASTIC SCATTERING CROSS 9137 1451 78 SECTIONS WERE CALCULATED WITH HAUSER-FESHBACH-MOLDAUER /8/ 9137 1451 79 APPROACH, AT INCIDENT NEUTRON ENERGIES HIGHER THAN 0.354 MEV 9137 1451 80 (LEVEL OVERLAPPING ENERGY) TEPEL ET AL./9/ THEORY WAS 9137 1451 81 EMPLOYED. 9137 1451 82 THE LEVEL SCHEME WAS TAKEN FROM NUCLEAR DATA SHEETS/10/. 9137 1451 83 9137 1451 84 NO. ENERGY(MEV) SPIN-PARITY 9137 1451 85 GS 0.0 3/2- 9137 1451 86 1 0.0067 1/2- 9137 1451 87 2 0.0572 7/2- 9137 1451 88 3 0.0706 5/2- 9137 1451 89 4 0.0865 5/2+ 9137 1451 90 5 0.0947 3/2+ 9137 1451 91 6 0.1036 7/2+ 9137 1451 92 7 0.1090 9/2+ 9137 1451 93 8 0.1634 11/2+ 9137 1451 94 9 0.1691 1/2+ 9137 1451 95 10 0.1792 9/2- 9137 1451 96 11 0.2017 3/2+ 9137 1451 97 12 0.2123 5/2+ 9137 1451 98 13 0.2379 9/2+ 9137 1451 99 14 0.2573 5/2- 9137 1451 100 15 0.2796 7/2+ 9137 1451 101 16 0.3004 7/2- 9137 1451 102 17 0.3061 7/2+ 9137 1451 103 18 0.3661 9/2+ 9137 1451 104 CONTINUUM LEVELS WERE ASSUMED ABOVE 0.367 MEV. 9137 1451 105 9137 1451 106 MT=16,17,37. (N,2N) AND (N,3N) CROSS SECTION FROM 9137 1451 107 STATISTICAL MODEL CALCULATIONS /1/ WITH ACCOUNT OF 9137 1451 108 PRE-EQUILIBRIUM NEUTRON EMISSION (MODIFIED STAPRE CODE/11/ 9137 1451 109 WAS USED). PRE-EQUILIBRIUM NEUTRON EMISSION CONTRIBUTION WAS 9137 1451 110 FIXED ACCORDING TO CONSISTENT DESCRIPTION OF(N,F) AND (N,XN) 9137 1451 111 REACTION DATA FOR 238U AND 232Th TARGET NUCLIDES. 9137 1451 112 9137 1451 113 MT=18, 19, 20, 21,38. FISSION CROSS SECTION IS CALCULATED 9137 1451 114 WITHIN STATISTICAL MODEL /11/. 9137 1451 115 MEASURED FISSION DATA /12,13,14,15,16/ANALYSIS WAS 9137 1451 116 ACCOMPLISHED. 9137 1451 117 THE CONTRIBUTION OF EMISSIVE (N,NF) AND (N,2NF) FISSION TO 9137 1451 118 THE TOTAL FISSION CROSS SECTION WAS ESTIMATED USING FISSION 9137 1451 119 BARRIER PARAMETERS OF 231-PA AND 230-PA, WHICH DESCRIBE 9137 1451 120 INDIRECT 230-PA(N,F) AND 229-PA(N,F)CROSS SECTION DATA BY 9137 1451 121 BRITT AND WILHEMY /16/. 9137 1451 122 MT=102 CAPTURE 9137 1451 123 CAPTURE CROSS SECTION IS CALCULATED WITHIN A STATISTICAL MO- 9137 1451 124 DEL. ABOVE NEUTRON ENERGY 5.5 MEV CAPTURE IS ASSUMED TO BE 9137 1451 125 CONSTANT. COMPETITION OF (N,GF) AND (N,GN') REACTIONS IS 9137 1451 126 TAKEN INTO ACCOUNT. 9137 1451 127 9137 1451 128 MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 9137 1451 129 FOR MT=2,53,66 FROM COUPLED CHANNEL CALCULATIONS 9137 1451 130 (RIGID ROTATOR MODEL), WITH ADDED ISOTROPIC COMPOUND 9137 1451 131 CONTRIBUTION. 9137 1451 132 9137 1451 133 MT=16, 17, 18-21, 38, 51-52,54-65, 67-68 AND 91 ARE ISOTROPIC 9137 1451 134 IN THE LAB SYSTEM. 9137 1451 135 9137 1451 136 MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS 9137 1451 137 9137 1451 138 ENERGY DISTRIBUTIONS FOR MT=16,17,91 WERE CALCULATED WITH 9137 1451 139 A HAUSER-FESHBACH STATISTICAL MODEL OF CASCADE NEUTRON 9137 1451 140 EMISSION TAKING INTO ACCOUNT THE HISTORY OF THE DECAY WITH 9137 1451 141 THE ALLOWANCE OF PREEQUILIBRIUM EMISSION OF THE FIRST 9137 1451 142 NEUTRON, SIMULTANEOUSLY WITH (N,F) AND (N,XNF) REACTION 9137 1451 143 CROSS SECTIONS. 9137 1451 144 9137 1451 145 MT=18,19,20,21,38 9137 1451 146 PROMPT FISSION NEUTRON SPECTRA (PFNS)WERE CALCULATED WITH THE9137 1451 147 SEMI-EMPIRICAL MODEL/1,5,17/, PRE-FISSION NEUTRON EMISSION 9137 1451 148 IN(N,XNF) REACTION, EITHER EQUILIBRIUM AND PRE-EQUILIBRIUM 9137 1451 149 MODES ARE INCLUDED. SPECTRA OF PRE-FISSION (N,XNF) NEUTRONS 9137 1451 150 ARE CALCULATED WITH HAUSER-FESHBACH STATISTICAL MODEL. 9137 1451 151 BASICALLY PFNS FROM FISSION FRAGMENTS (FF) WERE CALCULATED AS9137 1451 152 A SUPERPOSITION OF TWO WATT DISTRIBUTIONS FOR LIGHT AND HEAVY9137 1451 153 FF WITH EQUAL CONTRIBUTIONS, BUT DIFFERENT TEMPERATURE 9137 1451 154 PARAMETERS. FF KINETIC ENERGY, ONE MORE MODEL PARAMETER, 9137 1451 155 MIGHT BE LOWER THAN TKE, WHICH REFLECTS IT'S DEPENDENS ON THE9137 1451 156 MOMENT OF NEUTRON EMISSION. THIS EFFECTIVELY REDUCES AVERAGE 9137 1451 157 ENERGY OF PFNS FOR INCIDENT NEUTRON ENERGIES ABOVE EMISSIVE 9137 1451 158 FISSION THRESHOLD. 9137 1451 159 9137 1451 160 REFERENCES 9137 1451 161 1) Maslov V.M., Baba M., Hasegawa A., Kornilov N.V, 9137 1451 162 Kagalenko A.B., Tetereva N.A. INDC(BLR)-20,IAEA, 9137 1451 163 Vienna (2003). 9137 1451 164 2) JENDL-3.3, Japan Evaluated Nuclear Data Library (2003). 9137 1451 165 3) Tuttle R.J., INDC(NDS)-107/G (1979). 9137 1451 166 4) Malinovskij V.V., Yadernye constanty, V.2, 25 (1987). 9137 1451 167 5) Maslov V.M., Porodzinskij Yu.V., Baba M., Hasegawa A., 9137 1451 168 Kornilov N.V., Kagalenko A.B. European Phisical Journal A 9137 1451 169 in print) (2003). 9137 1451 170 6) Mughabghab S.F. Neutron Cross Sections, V. 1, Part B, (1984).9137 1451 171 7) Maslov V.M.,Porodzinskij Yu.V.,Baba M.,Hasegawa A., 9137 1451 172 Kornilov N.V., Kagalenko A.B., Tetereva N.A. INDC(BLR)-16, 9137 1451 173 IAEA, Vienna (2003). 9137 1451 174 8) Moldauer P.A., Phys. Rev., C11, 426 (1975). 9137 1451 175 9) Tepel J.W., Hoffman H.M., Weidenmuller H.A. Phys. Lett. 49, 9137 1451 176 1 (1974). 9137 1451 177 10) Ellis, Y.A. Nucl. 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Eng. 72, 222 (1979).9137 1451 188 17) Maslov V.M., Porodzinskij Yu.V., Baba M., Hasegawa A., 9137 1451 189 Kornilov N.V., Kagalenko A.B., Proc. of the 10th Internatio- 9137 1451 190 nal Seminar on Interaction of Neutrons with Nuclei, 9137 1451 191 May 22-25, 2002,Dubna, E3-2003-10, Russia, 222 (2003). 9137 1451 192 9137 1451 193 1 451 261 09137 1451 194 1 452 10 09137 1451 195 1 455 12 09137 1451 196 1 456 10 09137 1451 197 2 151 769 09137 1451 198 3 1 29 09137 1451 199 3 2 25 09137 1451 200 3 4 28 09137 1451 201 3 16 9 09137 1451 202 3 17 7 09137 1451 203 3 18 25 09137 1451 204 3 19 25 09137 1451 205 3 20 10 09137 1451 206 3 21 7 09137 1451 207 3 37 4 09137 1451 208 3 38 6 09137 1451 209 3 51 22 09137 1451 210 3 52 19 09137 1451 211 3 53 24 09137 1451 212 3 54 18 09137 1451 213 3 55 17 09137 1451 214 3 56 16 09137 1451 215 3 57 16 09137 1451 216 3 58 15 09137 1451 217 3 59 15 09137 1451 218 3 60 15 09137 1451 219 3 61 14 09137 1451 220 3 62 14 09137 1451 221 3 63 13 09137 1451 222 3 64 13 09137 1451 223 3 65 12 09137 1451 224 3 66 17 09137 1451 225 3 67 11 09137 1451 226 3 68 11 09137 1451 227 3 91 16 09137 1451 228 3 102 19 09137 1451 229 4 2 197 09137 1451 230 4 16 2 09137 1451 231 4 17 2 09137 1451 232 4 37 2 09137 1451 233 4 51 2 09137 1451 234 4 52 2 09137 1451 235 4 53 197 09137 1451 236 4 54 2 09137 1451 237 4 55 2 09137 1451 238 4 56 2 09137 1451 239 4 57 2 09137 1451 240 4 58 2 09137 1451 241 4 59 2 09137 1451 242 4 60 2 09137 1451 243 4 61 2 09137 1451 244 4 62 2 09137 1451 245 4 63 2 09137 1451 246 4 64 2 09137 1451 247 4 65 2 09137 1451 248 4 66 170 09137 1451 249 4 67 2 09137 1451 250 4 68 2 09137 1451 251 4 91 2 09137 1451 252 5 16 1181 09137 1451 253 5 17 673 09137 1451 254 5 18 2226 09137 1451 255 5 19 2226 09137 1451 256 5 20 1785 09137 1451 257 5 21 1008 09137 1451 258 5 37 69 09137 1451 259 5 38 564 09137 1451 260 5 91 1173 09137 1451 261