JEFF-3.1.2 general purpose neutron file update released Oct.2011 1 0 0 0 9.123200+4 2.300450+2 1 1 2 19134 1451 1 0.000000+0 1.000000+0 0 0 0 69134 1451 2 1.000000+0 2.000000+7 2 0 10 319134 1451 3 0.000000+0 0.000000+0 0 0 216 259134 1451 4 91-Pa-232 ORNL,TIT EVAL-DEC99 R. Q. WRIGHT, N. TAKAGI 9134 1451 5 DIST-JAN09 20090105 9134 1451 6 ----JEF-31 MATERIAL 9134 9134 1451 7 -----INCIDENT NEUTRON DATA 9134 1451 8 ------ENDF-6 FORMAT 9134 1451 9 ***************************** JEFF-3.1.2 ************************9134 1451 10 ** **9134 1451 11 ** Original data taken from: JEFF-3.1.1 **9134 1451 12 ** **9134 1451 13 ******************************************************************9134 1451 14 *************************** JEFF-3.1.1 *************************9134 1451 15 ** **9134 1451 16 ** Original data taken from: JEFF-3.1 **9134 1451 17 ** **9134 1451 18 ******************************************************************9134 1451 19 ***************************** JEFF-3.1 *************************9134 1451 20 ** **9134 1451 21 ** Original data taken from: ENDF/B-VI.8 **9134 1451 22 ** **9134 1451 23 ******************************************************************9134 1451 24 9134 1451 25 **************************************************************** 9134 1451 26 9134 1451 27 ENDF/B-VI MOD 2 Revision, December 1999, R.Q. Wright (ORNL) 9134 1451 28 9134 1451 29 The JENDL-3.2 evaluation was adopted and updated as follows: 9134 1451 30 9134 1451 31 MF=1, 9134 1451 32 MT=456 NUMBER OF PROMPT NEUTRONS 9134 1451 33 Changed from LNU=1, polynomial representation to LNU=2, 9134 1451 34 tabulated representation. MT=452, 455, and 456 are now 9134 1451 35 represented as tabulated functions. 9134 1451 36 9134 1451 37 MF=2 RESONANCE PARAMETERS-- MLBW FORMALISM USED. 9134 1451 38 The resolved resonance parameter evaluation is based on 9134 1451 39 the measurements of Danon [Da1996]. Preliminary results were 9134 1451 40 given in the paper by Moore et al. [Mo1994]. 9134 1451 41 For the levels below 5 eV, the parameters from Table 1 9134 1451 42 of Moore et al. were used. The 2nd column of Table 1 of that 9134 1451 43 reference is actually g*Gamma-n0, not 2g*Gamma-n0 as 9134 1451 44 indicated. 9134 1451 45 For the 5 to 10 eV range, the parameters from Table 1 of 9134 1451 46 Danon are used. In this table, Column 2 should be 9134 1451 47 2g*Gamma-n0, not 2g*Gamma-n as indicated. In the current 9134 1451 48 evaluation the neutron widths for the 6.44 and 8.85 eV 9134 1451 49 resonances were both increased slightly. 9134 1451 50 The resulting parameters give a very good fit to the 9134 1451 51 fission cross section for the energy range below 10 eV. 9134 1451 52 9134 1451 53 2200-m/s cross sections and resonance integrals 9134 1451 54 2200 m/s value Res. Int. 9134 1451 55 Total 1762.19 b - 9134 1451 56 Elastic 32.75 b - 9134 1451 57 Fission 1517.34 b 868 b 9134 1451 58 Capture 212.10 b 146 b 9134 1451 59 9134 1451 60 MF=3 NEUTRON CROSS SECTIONS 9134 1451 61 MT=1 TOTAL CROSS SECTION 9134 1451 62 For the energy range 10 eV to 35 keV, the total cross 9134 1451 63 section is calculated as sum of MT = 2, 18 and 102, using 9134 1451 64 the revised values of MT = 2, 18, and 102. above 35 keV 9134 1451 65 The cross section is 133.35 barns at 10 eV. The revised 9134 1451 66 total is about the same as the JENDL-3.2 evaluation; it is 9134 1451 67 slightly higher below 4 eV and slightly lower between 4 eV 9134 1451 68 and 3 keV. 9134 1451 69 9134 1451 70 MT=2 ELASTIC SCATTERING CROSS SECTION 9134 1451 71 The elastic cross section is revised below 100 eV to join 9134 1451 72 smoothly with the resolved resonance range; it is unchanged 9134 1451 73 above 100 eV. The elastic cross section is 17.6 barns at 10 9134 1451 74 eV, and is slightly higher between 10 and 100 eV. 9134 1451 75 9134 1451 76 MT=18 FISSION CROSS SECTION 9134 1451 77 The fission cross section is revised below 35 keV to join 9134 1451 78 smoothly with the resolved resonance range; it is unchanged 9134 1451 79 above 35 keV. The cross section at 10 eV (100.0 barns) is 9134 1451 80 based on the average fission from 2 to 10 eV, as caluclated 9134 1451 81 from the resolved resonance parameters. 9134 1451 82 The average fission cross section = 108.4 barns for the 9134 1451 83 energy range 5.5 to 10 eV. For the energy range 100 ev to 9134 1451 84 30 keV, the revised cross section is very close to the 9134 1451 85 ENDF/B-VI 235U fission cross section. 9134 1451 86 9134 1451 87 MT=102 CAPTURE CROSS SECTION 9134 1451 88 The capture cross section is revised below 35 keV to join 9134 1451 89 smoothly with the resolved resonance range. It is unchanged 9134 1451 90 above 35 keV. 9134 1451 91 The cross section at 10 eV (15.75 barns) is based on the 9134 1451 92 average capture from 2 to 10 eV (resolved resonance range). 9134 1451 93 9134 1451 94 ---------------------------------------------------------------- 9134 1451 95 REFERENCES 9134 1451 96 9134 1451 97 Da1996 Y. Danon, M.S. Moore, P.E. Koehler et al. Nucl.Sci.Eng. 9134 1451 98 124, 482 (1996). 9134 1451 99 Mo1994 M.S. Moore, P.E. Koehler, P.E. Littleton et al., Int. 9134 1451 100 Conference on Nuclear Data for Science and Technology, 9134 1451 101 Gatlinburg, Tennessee, May 9-13, 1994 (American Nuclear 9134 1451 102 Society, 1994) Vol. 2, p. 1075. 9134 1451 103 9134 1451 104 **************************************************************** 9134 1451 105 9134 1451 106 JENDL-3.2 COMMENTS FOLLOW-- 9134 1451 107 9134 1451 108 HISTORY 9134 1451 109 88-08 NEW EVALUATION WAS MADE BY N. TAKAGI (TOKYO INSTITUTE OF 9134 1451 110 TECHNOLOGY, TIT) /1/. 9134 1451 111 94-06 JENDL-3.2. 9134 1451 112 NU-P, NU-D AND NU-TOTAL WERE MODIFIED. 9134 1451 113 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9134 1451 114 9134 1451 115 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 9134 1451 116 (1,452), (1,455), (1,456) 9134 1451 117 *********************************************************** 9134 1451 118 9134 1451 119 MF=1 GENERAL INFORMATION 9134 1451 120 MT=451 COMMENT AND DICTIONARY 9134 1451 121 MT=452 NUMBER OF NEUTRONS PER FISSION 9134 1451 122 SUM OF NU-P NAD NU-D. 9134 1451 123 MT=455 DELAYED NEUTRONS PER FISSION 9134 1451 124 AVERAGE VALUES OF SYSTEMATICS BY TUTTLE/2/, BENEDETTI ET 9134 1451 125 AL./3/ AND WALDO ET AL./4/ DECAY CONSTANTS WERE ADOPTED 9134 1451 126 FROM THE EVALUATION BY BRADY AND ENGLAND/5/. 9134 1451 127 MT=456 PROMPT NEUTRONS PER FISSION 9134 1451 128 BASED ON SYSTEMATICS BY MANERO AND KONSHIN/6/, AND BY 9134 1451 129 HOWERTON/7/. 9134 1451 130 9134 1451 131 MF=2 RESONANCE PARAMETERS 9134 1451 132 MT=151 RESONANCE PARAMETERS 9134 1451 133 NO RESONANCE PARAMETERS WERE GIVEN. 9134 1451 134 9134 1451 135 2200-M/S CROSS SECTIONS AND RESONANCE INTEGRALS 9134 1451 136 2200 M/S VALUE RES. INT. 9134 1451 137 TOTAL 1176.23 B - 9134 1451 138 ELASTIC 12.23 B - 9134 1451 139 FISSION 700.00 B 313 B 9134 1451 140 CAPTURE 464.00 B 309 B 9134 1451 141 9134 1451 142 MF=3 NEUTRON CROSS SECTIONS 9134 1451 143 MT=1 TOTAL CROSS SECTION 9134 1451 144 BELOW 1 EV, CALCULATED AS SUM OF MT'S = 2, 18 AND 102. 9134 1451 145 ABOVE 1 EV, OPTICAL MODEL CALCULATION WAS MADE WITH 9134 1451 146 CASTHY/8/. THE POTENTIAL PARAMETERS/9/ USED ARE AS 9134 1451 147 FOLLOWS, 9134 1451 148 V = 41.0 - 0.05*EN (MEV) 9134 1451 149 WS= 6.4 - 0.15*SQRT(EN) (MEV) 9134 1451 150 WV= 0 , VSO = 7.0 (MEV) 9134 1451 151 R = RSO = 1.31 , RS = 1.38 (FM) 9134 1451 152 A = ASO = 0.47 , B = 0.47 (FM) 9134 1451 153 9134 1451 154 MT=2 ELASTIC SCATTERING CROSS SECTION 9134 1451 155 BELOW 1 EV, ASSUMED TO BE THE SAME AS SHAPE ELASTIC 9134 1451 156 SCATTERING CROSS SECTION CALCULATED WITH THE OPTICAL 9134 1451 157 MODEL. ABOVE 1 EV, OPTICAL MODEL CALCULATION WAS 9134 1451 158 ADOPTED. 9134 1451 159 9134 1451 160 MT=4, 91 INELASTIC SCATTERING CROSS SECTIONS. 9134 1451 161 OPTICAL AND STATISTICAL MODEL CALCULATION WAS MADE WITH 9134 1451 162 CASTHY/8/. NO EXCITED LEVELS WERE RECOMMENDED IN 9134 1451 163 REF./10/. 9134 1451 164 NO ENERGY(KEV) SPIN-PARITY 9134 1451 165 G.S. 0.0 2 - 9134 1451 166 LEVELS ABOVE 50 KEV WERE ASSUMED TO BE OVERLAPPING. 9134 1451 167 THE LEVEL DENSITY PARAMETERS WERE TAKEN FROM REF./11/. 9134 1451 168 9134 1451 169 MT=16,17,37 (N,2N), (N,3N) AND (N,4N) REACTION CROSS SECTIONS 9134 1451 170 CALCULATED WITH EVAPORATION MODEL. 9134 1451 171 9134 1451 172 MT=18 FISSION CROSS SECTION 9134 1451 173 MEASURED THERMAL CROSS SECTION OF 700 BARNS WAS TAKEN FROM9134 1451 174 REF./12/, AND 1/V FORM WAS ASSUMED BELOW 1 EV. FOR 9134 1451 175 ENERGIES ABOVE 1 EV, THE SHAPE WAS ASSUMED TO BE THE SAME 9134 1451 176 AS U-233 FISSION CROSS SECTION AND NORMALIZED TO THE 9134 1451 177 SYSTEMATICS BY BEHRENS AND HOWERTON/13/. 9134 1451 178 9134 1451 179 MT=102 CAPTURE CROSS SECTION 9134 1451 180 MEASURED THERMAL CROSS SECTION OF 464 BARNS WAS TAKEN FROM9134 1451 181 REF./12/, AND 1/V FORM WAS ASSUMED BELOW 1 EV. THE CROSS 9134 1451 182 SECTION SHAPE NEAR 1 EV WAS ADJUSTED SO AS TO REPRODUCE 9134 1451 183 THE RESONANCE INTEGRAL/12/. ABOVE 1 EV, CALCULATED WITH 9134 1451 184 CASTHY. THE GAMMA-RAY STRENGTH FUNCTION WAS ESTIMATED FROM9134 1451 185 GAMMA-GAMMA = 0.040 EV AND LEVEL SPACING = 0.417 EV. 9134 1451 186 9134 1451 187 MT=251 MU-L 9134 1451 188 CALCULATED WITH CASTHY. 9134 1451 189 9134 1451 190 MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 9134 1451 191 MT=2,91 CALCULATED WITH OPTICAL MODEL. 9134 1451 192 MT=16,17,18,37 ISOTROPIC IN THE LAB SYSTEM. 9134 1451 193 9134 1451 194 MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS 9134 1451 195 MT=16,17,37,91 EVAPORATION SPECTRA 9134 1451 196 OBTAINED FROM LEVEL DENSITY PARAMETERS. 9134 1451 197 9134 1451 198 MT=18 MAXWELLIAN FISSION SPECTRUM. 9134 1451 199 TEMPERATURE WAS ESTIMATED FROM Z**2/A DEPENDENCE/14/. 9134 1451 200 9134 1451 201 REFERENCES 9134 1451 202 1) TAKAGI N. ET AL.: J. NUCL. SCI. TECHNOL., 27, 853 (1990). 9134 1451 203 2) TUTTLE R.J.: INDC(NDS)-107/G+SPECIAL, P.29 (1979), 9134 1451 204 3) BENEDETTI G. ET AL.: NUCL. SCI. ENG., 80, 379 (1982). 9134 1451 205 4) WALDO R. ET AL.: PHYS. REV., C23, 1113 (1981). 9134 1451 206 5) BRADY M.C. AND ENGLAND T.R.: NUCL. SCI. ENG., 103, 129(1989). 9134 1451 207 6) MANERO F. AND KONSHIN V.A.: AT. ENERGY REV.,10, 637 (1972). 9134 1451 208 7) HOWERTON R.J.: NUCL. SCI. ENG., 62, 438 (1977). 9134 1451 209 8) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991). 9134 1451 210 9) OHSAWA T., OHTA M.: J. NUCL. SCI. TECHNOL., 18, 408 (1981). 9134 1451 211 10) SCHMORAX M.R.: NUCL. DATA SHEETS, 36, 367 (1982). 9134 1451 212 11) GILBERT A., CAMERON A.G.W.: CAN. J. PHYS., 43, 1446 (1965). 9134 1451 213 12) MUGHABGHAB S.F.: "NEUTRON CROSS SECTIONS, VOL.1, NEUTRON 9134 1451 214 RESONANCE PARAMETERS AND THERMAL CROSS SECTIONS , PART B, 9134 1451 215 Z=61-100", ACADEMIC PRESS (1984). 9134 1451 216 13) BEHRENS J.W., HOWERTON R.J: NUCL. SCI. ENG., 65, 464, (1978). 9134 1451 217 14) SMITH A.B. ET AL.: ANL/NDM-50 (1979). 9134 1451 218 9134 1451 219 ************************ C O N T E N T S *********************** 9134 1451 220 1 451 245 19134 1451 221 1 452 5 19134 1451 222 1 455 7 19134 1451 223 1 456 5 19134 1451 224 2 151 18 19134 1451 225 3 1 22 19134 1451 226 3 2 20 19134 1451 227 3 4 15 19134 1451 228 3 16 9 19134 1451 229 3 17 6 19134 1451 230 3 18 19 19134 1451 231 3 37 4 19134 1451 232 3 91 15 19134 1451 233 3 102 20 19134 1451 234 4 2 240 19134 1451 235 4 16 10 19134 1451 236 4 17 10 19134 1451 237 4 18 10 19134 1451 238 4 37 10 19134 1451 239 4 91 10 19134 1451 240 5 16 17 19134 1451 241 5 17 22 19134 1451 242 5 18 7 19134 1451 243 5 37 25 19134 1451 244 5 91 10 19134 1451 245 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1452 1 0.000000+0 0.000000+0 0 0 1 59134 1452 2 5 2 9134 1452 3 1.000000-5 2.285800+0 2.530000-2 2.285800+0 5.000000+6 2.845800+09134 1452 4 6.000000+6 2.952800+0 2.000000+7 4.520800+0 9134 1452 5 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1455 1 0.000000+0 0.000000+0 0 0 6 09134 1455 2 1.290000-2 3.470000-2 1.150000-1 2.856000-1 6.706000-1 2.311100+09134 1455 3 0.000000+0 0.000000+0 0 0 1 59134 1455 4 5 2 9134 1455 5 1.000000-5 1.576000-2 2.530000-2 1.576000-2 5.000000+6 1.576000-29134 1455 6 6.000000+6 1.083000-2 2.000000+7 1.083000-2 9134 1455 7 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1456 1 0.000000+0 0.000000+0 0 0 1 59134 1456 2 5 2 9134 1456 3 1.000000-5 2.270040+0 2.530000-2 2.270040+0 5.000000+6 2.830040+09134 1456 4 6.000000+6 2.941970+0 2.000000+7 4.509970+0 9134 1456 5 0.000000+0 0.000000+0 0 0 0 09134 1 099999 0.000000+0 0.000000+0 0 0 0 09134 0 0 0 9.123200+4 2.300450+2 0 0 1 09134 2151 1 9.123200+4 1.000000+0 0 1 1 09134 2151 2 1.000000-5 1.000000+1 1 2 0 09134 2151 3 2.000000+0 9.890000-1 0 0 1 09134 2151 4 2.300450+2 0.000000+0 0 0 78 139134 2151 5 -5.000000+0 2.500000+0 4.906800-1 5.068000-2 4.000000-2 4.000000-19134 2151 6 3.300000-1 1.500000+0 2.310962-1 9.620000-5 4.000000-2 1.910000-19134 2151 7 6.900000-1 2.500000+0 3.400415-1 4.150000-5 4.000000-2 3.000000-19134 2151 8 1.400000+0 1.500000+0 3.400560-1 5.600000-5 4.000000-2 3.000000-19134 2151 9 2.730000+0 1.500000+0 2.176600-1 6.600000-4 4.000000-2 1.770000-19134 2151 10 3.050000+0 2.500000+0 2.101570-1 1.570000-4 4.000000-2 1.700000-19134 2151 11 4.150000+0 2.500000+0 3.753200-1 3.200000-4 4.000000-2 3.350000-19134 2151 12 6.440000+0 1.500000+0 3.086300-1 6.300000-4 4.000000-2 2.680000-19134 2151 13 7.450000+0 2.500000+0 5.261960-1 1.960000-4 4.000000-2 4.860000-19134 2151 14 8.410000+0 2.500000+0 5.192780-1 2.780000-4 4.000000-2 4.790000-19134 2151 15 8.850000+0 2.500000+0 4.537650-1 7.650000-4 4.000000-2 4.130000-19134 2151 16 1.100000+1 1.500000+0 3.613270-1 1.327000-3 4.000000-2 3.200000-19134 2151 17 1.185000+1 2.500000+0 4.408600-1 8.600000-4 4.000000-2 4.000000-19134 2151 18 0.000000+0 0.000000+0 0 0 0 09134 2 099999 0.000000+0 0.000000+0 0 0 0 09134 0 0 0 9.123200+4 2.300450+2 0 0 0 09134 3 1 1 0.000000+0 0.000000+0 0 0 2 569134 3 1 2 6 2 56 5 9134 3 1 3 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.590600-1 0.000000+09134 3 1 4 1.000000+0 0.000000+0 1.000000+1 0.000000+0 1.000000+1 1.333500+29134 3 1 5 3.162280+1 7.912700+1 1.000000+2 4.912960+1 3.162280+2 3.413684+19134 3 1 6 1.000000+3 2.544520+1 3.162280+3 2.038895+1 1.000000+4 1.708060+19134 3 1 7 3.000000+4 1.476440+1 3.250000+4 1.445023+1 3.500000+4 1.415950+19134 3 1 8 4.000000+4 1.390960+1 5.000000+4 1.349490+1 5.021730+4 1.348670+19134 3 1 9 7.000000+4 1.285810+1 1.000000+5 1.214320+1 1.100000+5 1.194180+19134 3 1 10 1.200000+5 1.175310+1 1.500000+5 1.124790+1 1.800000+5 1.081060+19134 3 1 11 2.000000+5 1.054780+1 2.500000+5 9.968470+0 3.000000+5 9.476870+09134 3 1 12 4.000000+5 8.686870+0 5.000000+5 8.091510+0 6.000000+5 7.644420+09134 3 1 13 7.000000+5 7.315260+0 8.000000+5 7.081530+0 9.000000+5 6.925250+09134 3 1 14 1.000000+6 6.831530+0 2.000000+6 7.310760+0 3.000000+6 7.819010+09134 3 1 15 4.000000+6 7.814340+0 5.000000+6 7.464290+0 5.585000+6 6.968820+09134 3 1 16 6.000000+6 6.993430+0 7.000000+6 6.544650+0 8.000000+6 6.138000+09134 3 1 17 9.000000+6 5.817650+0 1.000000+7 5.601610+0 1.100000+7 5.491100+09134 3 1 18 1.200000+7 5.480780+0 1.242870+7 5.501695+0 1.300000+7 5.545290+09134 3 1 19 1.400000+7 5.660120+0 1.500000+7 5.811850+0 1.600000+7 5.980510+09134 3 1 20 1.700000+7 6.141030+0 1.800000+7 6.265730+0 1.824700+7 6.285382+09134 3 1 21 1.900000+7 6.346220+0 2.000000+7 6.405530+0 9134 3 1 22 0.000000+0 0.000000+0 0 0 0 09134 3 099999 9.123200+4 2.300450+2 0 0 0 09134 3 2 1 0.000000+0 0.000000+0 0 0 2 499134 3 2 2 4 2 49 5 9134 3 2 3 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.000000+1 0.000000+09134 3 2 4 1.000000+1 1.760000+1 3.162280+1 1.588000+1 1.000000+2 1.431960+19134 3 2 5 1.000000+3 1.379520+1 1.000000+4 1.254460+1 3.000000+4 1.162840+19134 3 2 6 3.250000+4 1.153840+1 3.500000+4 1.145510+1 4.000000+4 1.131820+19134 3 2 7 5.000000+4 1.106610+1 5.021730+4 1.106040+1 7.000000+4 1.057820+19134 3 2 8 1.000000+5 9.954650+0 1.100000+5 9.768500+0 1.200000+5 9.589910+09134 3 2 9 1.500000+5 9.092570+0 1.800000+5 8.643270+0 2.000000+5 8.366490+09134 3 2 10 2.500000+5 7.743780+0 3.000000+5 7.222580+0 4.000000+5 6.398080+09134 3 2 11 5.000000+5 5.788480+0 6.000000+5 5.328500+0 7.000000+5 4.983510+09134 3 2 12 8.000000+5 4.728490+0 9.000000+5 4.542310+0 1.000000+6 4.412280+09134 3 2 13 2.000000+6 4.519400+0 3.000000+6 5.028580+0 4.000000+6 5.076990+09134 3 2 14 5.000000+6 4.767700+0 6.000000+6 4.308940+0 7.000000+6 3.856530+09134 3 2 15 8.000000+6 3.482500+0 9.000000+6 3.202950+0 1.000000+7 3.024650+09134 3 2 16 1.100000+7 2.946740+0 1.200000+7 2.958650+0 1.300000+7 3.040290+09134 3 2 17 1.400000+7 3.166930+0 1.500000+7 3.317030+0 1.600000+7 3.476580+09134 3 2 18 1.700000+7 3.637010+0 1.800000+7 3.788450+0 1.900000+7 3.917850+09134 3 2 19 2.000000+7 4.019540+0 9134 3 2 20 0.000000+0 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