ENDFB6 LIBRARY, DUMMY TAPE HEADER 3.006000+3 5.963400+0 -1 0 0 1 325 1451 1 0.000000+0 0.000000+0 0 0 0 6 325 1451 2 1.000000+0 0.000000+0 0 0 10 6 325 1451 3 0.000000+0 0.000000+0 0 0 269 2 325 1451 4 3-LI - 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 5 DIST-JAN90 900123 325 1451 6 ----ENDF/B-VI MATERIAL 325 325 1451 7 -----INCIDENT NEUTRON DATA 325 1451 8 ------ENDF-6 FORMAT 325 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 325 1451 10 DOSIMETRY 325 1451 11 ****************************************************************** 325 1451 12 325 1451 13 ****************************************************************** 325 1451 14 325 1451 15 ENDF/VI EVALUATION 325 1451 16 G. M. Hale and P. G. Young 325 1451 17 325 1451 18 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 325 1451 19 325 1451 20 1. Inclusion of the ENDF/B-VI standard (n,t) cross section 325 1451 21 from the simultaneous standards analysis (ca85) over the 325 1451 22 energy range thermal to 1 MeV. 325 1451 23 2. Replacement of all major cross sections and elastic angular 325 1451 24 distributions at energies between 10^-5 eV and 3 MeV with 325 1451 25 results from the R-matrix analysis performed in conjunction 325 1451 26 with the simultaneous standards analysis. 325 1451 27 3. Revision of the elastic cross sections and angular distri- 325 1451 28 butions at energies between 3 and 20 MeV to match recent 325 1451 29 experimental data, resulting in a general decrease of the 325 1451 30 elastic cross section in this energy range. 325 1451 31 4. Revision of the (n,n')d cross sections to account for 325 1451 32 recent measurements, resulting in a general increase in 325 1451 33 the total (n,n')d cross section that tends to offset the 325 1451 34 decrease in the elastic cross section and maintain about 325 1451 35 the same total cross section as before. 325 1451 36 325 1451 37 325 1451 38 ****************************************************************** 325 1451 39 325 1451 40 325 1451 41 mf=2 --------- resonance parameters ---------------------------- 325 1451 42 325 1451 43 mt=151 effective scattering radius = 2.31175e-13 cm. 325 1451 44 325 1451 45 mf=3 --------- smooth cross sections --------------------------- 325 1451 46 325 1451 47 the 2200 m/s cross sections are as follows: 325 1451 48 mt=1 sigma = 941.6928 barns 325 1451 49 mt=2 sigma = 0.67157 barns 325 1451 50 mt=102 sigma = 0.03850 barns 325 1451 51 mt=105 sigma = 940.9827 barns 325 1451 52 325 1451 53 mt=1 total cross section 325 1451 54 below 3 mev, the values are taken from an r-matrix 325 1451 55 analysis by hale, dodder, and witte (ha84) which takes 325 1451 56 into account data from all reactions possible in 7li 325 1451 57 up to 4 mev neutron energy. total cross section data 325 1451 58 considered in this analysis were those of ha75 and sm77. 325 1451 59 between 3 and 20 mev, the total was taken to be the 325 1451 60 sum of mt=2,4,24,102,103, and 105, which generally 325 1451 61 follows the measurements of sm82, ke79, kn77, go72, 325 1451 62 and fo71. 325 1451 63 mt=2 elastic cross section 325 1451 64 below 3 mev, the values are taken from the r-matrix 325 1451 65 analysis cited for mt=1, which includes the elastic 325 1451 66 measurements of sm82 and la61. above 3 mev, the curve 325 1451 67 is a smooth representation of the data of kn79 and ba63 325 1451 68 up to 7.5 mev, and of that of ho79 between 7.5 and 13 325 1451 69 mev. the curve passes through the average of several 325 1451 70 measurements at 14 mev, and is extrapolated to 20 mev 325 1451 71 using the shape of an optical model calculation. 325 1451 72 mt=4 total inelastic cross section 325 1451 73 sum of mt=51 through mt=81. 325 1451 74 mt=24 (n,2n)alpha cross section 325 1451 75 passes through the point of mather and paine (ma69) at 325 1451 76 14 mev, taking into account the measurements of as63. 325 1451 77 mt=51,52,54-56,58-81 (n,n')d continuum 325 1451 78 represented by continuum-level contributions in 6li, 325 1451 79 binned in .5-mev intervals. The energy-angle spectra 325 1451 80 are determined by a 3-body phase-space calculation, 325 1451 81 assuming isotropic center-of mass distributions. at 325 1451 82 each energy, the sum of the continuum-level 325 1451 83 contributions is normalized to an assumed energy-angle 325 1451 84 integrated continuum cross section which approximates 325 1451 85 the difference of the nonelastic sigma and the 325 1451 86 contribution from the first and second levels in 6li. 325 1451 87 the steep rise of the pseudo-level cross sections from 325 1451 88 their thresholds and the use of fixed bin widths over 325 1451 89 finite angles produces anomolous structure in the 325 1451 90 individual cross sections which is especially apparent 325 1451 91 near the thresholds. some effort has been made to 325 1451 92 smooth out these effects, but they remain to some 325 1451 93 extent. 325 1451 94 mt=53 (n,n1)d discrete level cross section 325 1451 95 has p-wave penetrability energy dependence from threshold 325 1451 96 to 3.2 mev. matched at higher energies to a curve 325 1451 97 through fitted legendre coefficients from experimental 325 1451 98 data of sa82, ho79, sm80, ho68, ba63. 325 1451 99 mt=57 (n,n2)gamma cross section 325 1451 100 is based on the available experimental data, especially 325 1451 101 that of ho79, li80, sm82, ho68. 325 1451 102 gradually to 20 mev. smooth curve drawn through data 325 1451 103 of pr69 and be75. 325 1451 104 mt=102 (n,gamma) cross section 325 1451 105 unchanged from version v, which was based on the thermal 325 1451 106 measurement of jurney (ju73) and the pendlebury 325 1451 107 evaluation (pe64) at higher energies. 325 1451 108 mt=103 (n,p) cross section 325 1451 109 threshold to 9 mev, based on the data of ba65.extended to 325 1451 110 20 mev through the 14 mev data of fr54 and ba53. 325 1451 111 mt=105 (n,t) cross section 325 1451 112 below 3 mev, values are taken from the r-matrix analysis, 325 1451 113 which includes (n,t) measurements from re78, la78, br77, 325 1451 114 ov74, and ba75. between 3 and 5 mev, the values are 325 1451 115 based on ba75, and at higher energies are taken from the 325 1451 116 evaluation of pe64, extended to 20 mev considering the 325 1451 117 data of ke58. 325 1451 118 325 1451 119 325 1451 120 mf = 4-------------------angular distributions------------------- 325 1451 121 325 1451 122 mt=2 elastic cross section 325 1451 123 legendre coefficients determined as follows: 325 1451 124 below 4 mev, coefficients up to l=6 were taken from 325 1451 125 the r-matrix analysis , which included the measurements 325 1451 126 la61 and sm 82. above 4 mev, the coefficients represent 325 1451 127 fits to the measurements of ho68, ho79, kn79, sm82, 325 1451 128 de73, ba63, ab70, and hy68. most emphasis was placed 325 1451 129 on the data of ho79, kn79, sm82. extrapolation of the 325 1451 130 coefficients to 20 mev was aided by optical model 325 1451 131 calculations. 325 1451 132 mt=24 (n,2n) cross section 325 1451 133 lab distributions obtained by integrating over energy the 325 1451 134 4-body phase-space spectra that result from transforming 325 1451 135 isotropic center-of-mass distributions to the laboratory 325 1451 136 system. 325 1451 137 mt=51-81 (n,n')d cross sections 325 1451 138 excitation energy binned data is assumed isotropic in the 325 1451 139 center of mass reference system. mt = 53 and 57 325 1451 140 are real levels. mt = 57 is assumed to be isotropic 325 1451 141 in the two-body reference system. mt = 53 is given as 325 1451 142 anisotropic, based on fits of legendre expansions to 325 1451 143 the experimental data of ab70, ba63, ho68, ho79, me65, 325 1451 144 hy68, wo62, sa82. 325 1451 145 mt=105 (n,t).cross section (to be added) 325 1451 146 legendre coefficients obtained from the r-matrix analysis 325 1451 147 are supplied at energies below 4 mev. the analysis 325 1451 148 takes into account (n,t) angular distribution 325 1451 149 measurements from kn83, co82, dr82, br77, ba75, and 325 1451 150 ov74. 325 1451 151 325 1451 152 325 1451 153 mf = 5---------secondary energy distributions-------------------- 325 1451 154 325 1451 155 mt=24 (n,2n) 325 1451 156 lab distributions obtained by integrating over angle the 4- 325 1451 157 body phase-space spectra that result from transforming 325 1451 158 isotropic center-of-mass distributions to the laboratory 325 1451 159 system. 325 1451 160 325 1451 161 325 1451 162 mf = 12--------gamma-ray multiplicities-------------------------- 325 1451 163 325 1451 164 mt=57 (n,n2) gamma 325 1451 165 energy taken from aj74. multiplicity assumed to be one. 325 1451 166 mt=102 (n,gamma) 325 1451 167 energies and transition arrays for radiative capture taken 325 1451 168 from ju73, as reported in aj74. the lp flag was used to 325 1451 169 describe the mt=102 photons. 325 1451 170 325 1451 171 325 1451 172 mf = 14--------gamma-ray angular distributions------------------- 325 1451 173 325 1451 174 mt=57 (n,n2)gamma 325 1451 175 the gamma is assumed isotropic. 325 1451 176 mt=102 (n,gamma) 325 1451 177 .the two high-energy gammas are assumed isotropic. data on 325 1451 178 the 477 kev gamma indicate isotropy. 325 1451 179 325 1451 180 325 1451 181 mf=33----------cross section covariances------------------------- 325 1451 182 (to be added later) 325 1451 183 325 1451 184 the relative covariances for mt=1,2, and 105 below 4 mev are 325 1451 185 given in file 33. they are based on calculations using the co- 325 1451 186 variances of the r-matrix parameters in first-order error 325 1451 187 propogation. 325 1451 188 mt=1 total 325 1451 189 relative covariances entered as nc-type sub-subsection, 325 1451 190 implying that they are to be constructed from those for 325 1451 191 mt=2 and 105. they are not intended for use at energies 325 1451 192 above 4 mev. 325 1451 193 mt=2,105 elastic and (n,t) 325 1451 194 relative covariances among these two cross sections are 325 1451 195 entered explicitly as ni-type sub-subsections in the 325 1451 196 lb=5 (direct) representation at energies below 4 mev. 325 1451 197 although values for the 3.95-4.05 mev bin are repeated 325 1451 198 in a 4-20 mev bin, the covariances are not intended for 325 1451 199 use at energies above 4 mev. 325 1451 200 325 1451 201 325 1451 202 --------------------- references -------------------------------- 325 1451 203 325 1451 204 ab70 U.Abbondanno, Nuo.Cim. A166,139(1970). 325 1451 205 aj74 f.ajzenberg-selove and t.lauritsen, nucl. phys. a227,55 325 1451 206 (1974). 325 1451 207 ar64 a.h.armstrong, j.gammel, l.rosen, and g.m.frye, nucl. phys. 325 1451 208 52,505 (1964). 325 1451 209 as63 v.j.ashby et al, phys. rev. 129,1771 (1963). 325 1451 210 ba53 m.e.battat and f.l.ribe, phys.rev. 89,80 (1953). 325 1451 211 ba63 r.batchelor and j.h.towle, nucl. phys. 47,385 (1963). 325 1451 212 ba65 r.bass, c.bindhardt, and k.kruger, eandc(e)-57u (1965). 325 1451 213 ba75 c.m.bartle, proc. conf. on nuclear cross sections and 325 1451 214 technology, vol.2,688 (1975), and private communication 325 1451 215 (1976). see also nucl. phys. a330, 1 (1979). 325 1451 216 be75 besotosnyj et al., yk-19, 77 (1975). 325 1451 217 br77 r.e.brown,g.g.ohlsen,r.f.haglund, and n.jarmie, phys. rev. 325 1451 218 16c, 513 (1977). 325 1451 219 ca85 a.d.carlson,w.p.poenitz,g.m.hale, and r.w.peele, nuclear 325 1451 220 data for basic and applied science (santa fe, n.m.), 1429 325 1451 221 (1985). 325 1451 222 co67 j.a.cookson and d.dandy, nucl. phys. a91,273 (1967). 325 1451 223 co82 h.conde,t.andersson,l.nilsson, and c.nordborg, nuclear data 325 1451 224 for science and technology (antwerp, belgium), 447 (1982). 325 1451 225 de73 F.Demanins et al., infn/be-73 (1973). 325 1451 226 dr82 m.drosg,d.m.drake,r.a.hardekopf, and g.m.hale, la-9129-ms 325 1451 227 (1982). 325 1451 228 dr85 m.drosg et al., santa fe conf.1, 145(1985). 325 1451 229 fo71 d.g.foster and d.w.glasgow, phys. rev. c3,576 (1971). 325 1451 230 fr54 g.m.frye, phys. rev. 93,1086 (1954). 325 1451 231 go72 c.a.goulding and p.stoler, eandc(us)-176u,161 (1972). 325 1451 232 ha75 j.a.harvey and n.w.hill, nuclear cross sections and 325 1451 233 technology (washington, d.c.), 244 (1975). 325 1451 234 ha84 g.m.hale, nuclear standard reference data (geel,belgium) 325 1451 235 iaea tecdoc-335, 103 (1984). describes preliminary analysis. 325 1451 236 ho68 j.c.hopkins,d.m.drake, and h.conde, nucl. phys. a107,139 325 1451 237 (1968), and j.c.hopkins, d.m.drake, and h.conde, la-3765 325 1451 238 (1967). 325 1451 239 ho79 h.h.hogue et al., n.s.&e. 69, 22 (1979). 325 1451 240 ju73 e.t.jurney, lasl, private communication (1973). 325 1451 241 ke58 r.d.kern and w.e.kreger, phys. rev. 112, 926 (1958). 325 1451 242 ke79 j.d.kellie,g.p.lamaze, and r.b.schwartz, nuclear cross 325 1451 243 sections for technology (knoxville, tn.), 48 (1979). 325 1451 244 kn77 h.h.knitter,c.budtz-jorgensen,m.mailly, and r.vogt, eur- 325 1451 245 5726e (1977). 325 1451 246 kn79 h.d.knox,r.m.white, and r.o.lane, n.s.&e. 69, 223 (1979). 325 1451 247 kn83 h.h.knitter,c.budtz-jorgensen,d.l.smith, and d.marletta, 325 1451 248 n.s.&e. 83, 229(1983). 325 1451 249 la61 r.o.lane,a.s.langsdorf,j.e.monahan, and a.j.elwyn, ann. 325 1451 250 phys.12, 135 (1961). 325 1451 251 la78 g.p.lamaze,o.a.wasson,r.a.schrack, and a.d.carlson, n.s.&e. 325 1451 252 68, (1978). 325 1451 253 li80 p.w.lisowski et al., la-8342 (1980). 325 1451 254 ma69 d.s.mather and l.f.paine, awre-o-47/69 (1969). 325 1451 255 me65 f.merchez,n.v.sen,v.regis, and r.bouchez, compt. rend. 260, 325 1451 256 3922 (1965). 325 1451 257 ov74 j.c.overley,r.m.sealock, and d.h.ehlers, nucl. phys. a221, 325 1451 258 573 (1974). 325 1451 259 pe64 e.d.pendlebury, awre-o-60/64 (1964). 325 1451 260 pr69 g.presser et al., nuc.phys. a131, 679(1969). 325 1451 261 re78 c.renner,j.a.harvey,n.w.hill,g.l.morgan, and k.pusk, bull. 325 1451 262 am. phys. soc. 23, 526 (1978). 325 1451 263 sa82 e.t.sadowski,h.knox,d.a.resler, and r.o.lane, bap 27,624(c5) 325 1451 264 (1982). 325 1451 265 sm77 a.b.smith,p.guenther,d.havel, and j.f.whalen, anl/ndm-29 325 1451 266 (1977). 325 1451 267 sm82 a.b.smith,p.t.guenther, and j.f.whalen, nucl. phys. a373, 325 1451 268 305 (1982). 325 1451 269 wo62 c.wong,j.d.anderson, and j.w.mcclure, nucl. phys. 33,680 325 1451 270 (1962). 325 1451 271 325 1451 272 ****************************************************************** 325 1451 273 1 451 275 1 325 1451 274 3 105 63 1 325 1451 275 325 1 0 276 325 0 0 277 3.006000+3 5.963400+0 0 0 0 0 325 3105 278 4.783800+6 4.783800+6 0 0 1 178 325 3105 279 178 5 0 0 0 0 325 3105 280 1.000000-5 4.734203+4 2.530000-2 9.409827+2 9.400000+0 4.879280+1 325 3105 281 1.500000+2 1.219570+1 2.500000+2 9.442800+0 3.500000+2 7.977700+0 325 3105 282 4.500000+2 7.033700+0 5.500000+2 6.361300+0 6.500000+2 5.850200+0 325 3105 283 7.500000+2 5.445400+0 8.500000+2 5.113700+0 9.500000+2 4.837100+0 325 3105 284 1.500000+3 3.847000+0 2.500000+3 2.979100+0 3.500000+3 2.518100+0 325 3105 285 4.500000+3 2.221400+0 5.500000+3 2.011000+0 6.500000+3 1.851600+0 325 3105 286 7.500000+3 1.724300+0 8.500000+3 1.622100+0 9.500000+3 1.535800+0 325 3105 287 1.500000+4 1.230100+0 2.000000+4 1.073700+0 2.400000+4 9.867000-1 325 3105 288 3.000000+4 8.930000-1 3.500000+4 8.359000-1 4.500000+4 7.562000-1 325 3105 289 5.500000+4 7.052000-1 6.499999+4 6.725000-1 7.500000+4 6.532000-1 325 3105 290 8.500000+4 6.450000-1 9.500000+4 6.468000-1 1.000000+5 6.515000-1 325 3105 291 1.100000+5 6.687000-1 1.200000+5 6.976000-1 1.300000+5 7.406000-1 325 3105 292 1.400000+5 8.013000-1 1.500000+5 8.843000-1 1.600000+5 9.968000-1 325 3105 293 1.700000+5 1.148500+0 1.800000+5 1.352200+0 1.900000+5 1.620700+0 325 3105 294 2.000000+5 1.965300+0 2.100000+5 2.376300+0 2.200000+5 2.804700+0 325 3105 295 2.300000+5 3.139300+0 2.350000+5 3.231100+0 2.360000+5 3.241600+0 325 3105 296 2.370000+5 3.249500+0 2.380000+5 3.254600+0 2.390000+5 3.257100+0 325 3105 297 2.400000+5 3.256800+0 2.410000+5 3.253900+0 2.420000+5 3.248500+0 325 3105 298 2.430000+5 3.240300+0 2.440000+5 3.229700+0 2.450000+5 3.216400+0 325 3105 299 2.500000+5 3.114500+0 2.600000+5 2.789600+0 2.700000+5 2.402600+0 325 3105 300 2.800000+5 2.036600+0 3.000000+5 1.474300+0 3.100000+5 1.274600+0 325 3105 301 3.250000+5 1.049800+0 3.350000+5 9.365000-1 3.500000+5 8.056000-1 325 3105 302 3.600000+5 7.376000-1 3.750000+5 6.565000-1 3.850000+5 6.128000-1 325 3105 303 4.000000+5 5.591000-1 4.100000+5 5.294000-1 4.250000+5 4.919000-1 325 3105 304 4.350000+5 4.706000-1 4.500000+5 4.431000-1 4.600000+5 4.273000-1 325 3105 305 4.750000+5 4.065000-1 4.850000+5 3.944000-1 5.000000+5 3.782000-1 325 3105 306 5.200000+5 3.597000-1 5.400000+5 3.441000-1 5.700000+5 3.249000-1 325 3105 307 6.000000+5 3.093000-1 6.500000+5 2.891000-1 7.000000+5 2.740000-1 325 3105 308 7.500000+5 2.623000-1 8.000000+5 2.529000-1 8.500000+5 2.454000-1 325 3105 309 9.000000+5 2.391000-1 9.400000+5 2.349000-1 9.600000+5 2.331000-1 325 3105 310 9.800000+5 2.313000-1 1.000000+6 2.297000-1 1.100000+6 2.230000-1 325 3105 311 1.200000+6 2.183000-1 1.300000+6 2.151700-1 1.400000+6 2.131900-1 325 3105 312 1.500000+6 2.124600-1 1.600000+6 2.127100-1 1.700000+6 2.140400-1 325 3105 313 1.800000+6 2.164500-1 1.900000+6 2.178500-1 2.000000+6 2.163800-1 325 3105 314 2.100000+6 2.130700-1 2.200000+6 2.087900-1 2.300000+6 2.039500-1 325 3105 315 2.400000+6 1.987900-1 2.600000+6 1.879200-1 2.800000+6 1.759800-1 325 3105 316 3.000000+6 1.645200-1 3.100000+6 1.582648-1 3.200000+6 1.513389-1 325 3105 317 3.300000+6 1.442339-1 3.400000+6 1.373084-1 3.500000+6 1.308116-1 325 3105 318 3.600000+6 1.249076-1 3.700000+6 1.196982-1 3.800000+6 1.152424-1 325 3105 319 3.900000+6 1.115712-1 4.000000+6 1.087000-1 4.100000+6 1.063000-1 325 3105 320 4.200000+6 1.040000-1 4.300000+6 1.013000-1 4.400000+6 9.860000-2 325 3105 321 4.500000+6 9.600000-2 4.600000+6 9.300000-2 4.700000+6 9.050000-2 325 3105 322 4.800000+6 8.760000-2 4.900000+6 8.510000-2 5.000000+6 8.300000-2 325 3105 323 5.050000+6 8.230000-2 5.100000+6 8.170000-2 5.150000+6 8.100000-2 325 3105 324 5.200000+6 8.030000-2 5.300000+6 7.900000-2 5.400000+6 7.700000-2 325 3105 325 5.500000+6 7.500000-2 5.600000+6 7.300000-2 5.800000+6 7.050000-2 325 3105 326 6.000000+6 6.800000-2 6.200000+6 6.550000-2 6.400000+6 6.300000-2 325 3105 327 6.600000+6 6.100000-2 6.800000+6 5.900000-2 7.000000+6 5.700000-2 325 3105 328 7.200000+6 5.500000-2 7.400000+6 5.350000-2 7.600000+6 5.200000-2 325 3105 329 8.000000+6 4.900000-2 8.200000+6 4.780000-2 8.400000+6 4.660000-2 325 3105 330 8.500000+6 4.600000-2 8.600000+6 4.540000-2 8.800000+6 4.420000-2 325 3105 331 9.000000+6 4.300000-2 9.500000+6 4.000000-2 1.000000+7 3.800000-2 325 3105 332 1.040000+7 3.640000-2 1.050000+7 3.600000-2 1.080000+7 3.480000-2 325 3105 333 1.100000+7 3.400000-2 1.150000+7 3.300000-2 1.200000+7 3.100000-2 325 3105 334 1.250000+7 2.900000-2 1.300000+7 2.800000-2 1.350000+7 2.700000-2 325 3105 335 1.400000+7 2.600000-2 1.450000+7 2.500000-2 1.500000+7 2.400000-2 325 3105 336 1.550000+7 2.330000-2 1.600000+7 2.250000-2 1.650000+7 2.150000-2 325 3105 337 1.700000+7 2.050000-2 1.750000+7 1.950000-2 1.800000+7 1.850000-2 325 3105 338 1.850000+7 1.760000-2 1.900000+7 1.680000-2 1.950000+7 1.590000-2 325 3105 339 2.000000+7 1.500000-2 0.000000+0 0.000000+0 0.000000+0 0.000000+0 325 3105 340 325 3 0 341 325 0 0 342 0 0 0 343 5.010000+3 9.926921+0 -1 0 0 1 525 1451 1 0.000000+0 0.000000+0 0 0 0 6 525 1451 2 1.000000+0 0.000000+0 0 0 10 6 525 1451 3 0.000000+0 0.000000+0 0 0 301 2 525 1451 4 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 5 DIST-JAN90 900108 525 1451 6 ----ENDF/B-VI MATERIAL 525 525 1451 7 -----INCIDENT NEUTRON DATA 525 1451 8 ------ENDF-6 FORMAT 525 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 525 1451 10 DOSIMETRY 525 1451 11 ****************************************************************** 525 1451 12 525 1451 13 ****************************************************************** 525 1451 14 525 1451 15 ENDF/VI EVALUATION 525 1451 16 G. M. Hale and P. G. Young 525 1451 17 525 1451 18 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 525 1451 19 525 1451 20 1. Inclusion of the ENDF/B-VI standard (n,alpha) and (n,alpha1) 525 1451 21 results from the simultaneous standards analysis (Ca85) over 525 1451 22 the standard energy range thermal to 100 keV. 525 1451 23 2. Replacement of all major cross sections and elastic angular 525 1451 24 distributions from 10-5 eV to 1 MeV with results from the 525 1451 25 R-matrix analysis performed in conjunction with the 525 1451 26 simultaneous standards analysis. 525 1451 27 3. Replaced the total cross section 1-20 MeV with results 525 1451 28 from a covariance analysis of available data. 525 1451 29 4. Revised elastic and inelastic cross sections for low-lying 525 1451 30 levels incorporating new elastic, inelastic, and (n,xgamma) 525 1451 31 experimental data. We attempted to better reconcile the 525 1451 32 inelastic and gamma ray data. 525 1451 33 5. Refit all elastic angular distributions from 1-20 MeV with 525 1451 34 Legendre expansions and incorporated results from new 525 1451 35 measurements. 525 1451 36 6. Fit inelastic neutron angular distributions for first 5 525 1451 37 excited states of B10 with Legendre expansions. 525 1451 38 7. Incorporated new (n,t2alpha) cross section data into MT113 525 1451 39 and adjusted (n,alpha) cross sections above standard region 525 1451 40 for better consistency with data as well as other cross 525 1451 41 sections (esp. total and elastic) determined by data. 525 1451 42 525 1451 43 *****Note that covariance data will be added at a later date. 525 1451 44 525 1451 45 ****************************************************************** 525 1451 46 525 1451 47 mf=2 --------- resonance parameters ---------------------------- 525 1451 48 525 1451 49 mt=151 effective scattering radius = 4.129038-13 cm 525 1451 50 525 1451 51 525 1451 52 mf=3 --------- smooth cross sections --------------------------- 525 1451 53 525 1451 54 the 2200 m/s cross sections are as follows, 525 1451 55 mt=1 sigma = 3842.146 barns 525 1451 56 mt=2 sigma = 2.142435 barns 525 1451 57 mt=102 sigma = 0.5 barns 525 1451 58 mt=103 sigma = 0.000566 barns 525 1451 59 mt=107 sigma = 3839.496 barns 525 1451 60 mt=113 sigma = 0.0069993 barns 525 1451 61 mt=600 sigma = 0.000566 barns 525 1451 62 mt=800 sigma = 241.2677 barns 525 1451 63 mt=801 sigma = 3598.228 barns 525 1451 64 525 1451 65 mt=1 total cross section 525 1451 66 0 to 1 mev, calculated from r-matrix parameters obtained 525 1451 67 from simultaneous standards analysis (ca85) used to 525 1451 68 obtain the endf/b-vi standard cross sections. 525 1451 69 1 to 20 mev, covariance analysis of measurements of di67, 525 1451 70 ts62,fo61,co52,au79, and co54, constrained to match 525 1451 71 r-matrix fit at 1 mev. glucs covariance analysis code 525 1451 72 (he80) was used in the calculations. 525 1451 73 525 1451 74 mt=2 elastic scattering cross section 525 1451 75 0 to 1 mev, calculated from the r-matrix parameters 525 1451 76 described for mt=1. experimental elastic scattering data 525 1451 77 included in the fit are those of as70 and la71. 525 1451 78 1 to 6 mev, smooth curve through measurements of la71, po70, 525 1451 79 sa88, and ho69, constrained to be consistent with total 525 1451 80 and reaction cross section measurements. 525 1451 81 6 to 14 mev, smooth curve through measurements of ho69,co69, 525 1451 82 te62,va70, va65, sa88, and gl82. Note that the data of 525 1451 83 sa88 above 9 MeV were discounted. 525 1451 84 14 to 20 mev, optical model extrapolation from 14 mev data 525 1451 85 525 1451 86 mt=4 inelastic cross section 525 1451 87 thres.to 20 mev, sum of mt=51-85 525 1451 88 525 1451 89 mt=51-61 inelastic cross sections to discrete states 525 1451 90 mt=51 q=-0.717 mev mt=55 q=-4.774 mev mt=59 q=-5.923 mev 525 1451 91 52 -1.740 56 -5.114 60 -6.029 525 1451 92 53 -2.154 57 -5.166 61 -6.133 525 1451 93 54 -3.585 58 -5.183 525 1451 94 thres.to 20 mev, based on (n,nprime) measurements of po70, 525 1451 95 co69,ho69,va70,sa88, and gl82, and the (n,xgam) measure- 525 1451 96 ments of da56,da60,ne70, and di88, using a gamma-ray 525 1451 97 decay scheme from analysis of aj88. hauser-feshbach 525 1451 98 calculations were used to estimate shapes and relative 525 1451 99 magnitudes where experimental data were lacking. 525 1451 100 525 1451 101 mt=62-85 inelastic cross sections to groups of levels in 525 1451 102 0.5-mev wide bands centered about the q-values given 525 1451 103 below (used in lieu of mt=91 and file 5) 525 1451 104 mt=62 q=-6.5 mev mt=70 q=-10.5 mev mt=78 q=-14.5 mev 525 1451 105 63 -7.0 71 -11.0 79 15.0 525 1451 106 64 -7.5 72 -11.5 80 15.5 525 1451 107 65 -8.0 73 -12.0 81 16.0 525 1451 108 66 -8.5 74 -12.5 82 16.5 525 1451 109 67 -9.0 75 -13.0 83 17.0 525 1451 110 68 -9.5 76 -13.5 84 17.5 525 1451 111 69 -10.0 77 -14.0 85 18.0 525 1451 112 thres. to 20 mev, integrated cross section obtained by sub- 525 1451 113 tracting the sum of mt=2,51-61,103,104,107,and 113 from 525 1451 114 mt=1. cross section distributed among the bands with 525 1451 115 an evaporation model using a nuclear temperature given 525 1451 116 by t=0.9728*sqrt(en) in mev,taken from ir67. 525 1451 117 525 1451 118 mt=102 (n,gamma) cross section 525 1451 119 0 to 1 mev, assumed 1/v dependence with thermal value of 525 1451 120 0.5 barn. 525 1451 121 1 to 20 mev, assumed negligable, set equal to zero. 525 1451 122 525 1451 123 mt=103 (n,p) cross section 525 1451 124 thres.to 20 mev, sum of mt=600-605 525 1451 125 525 1451 126 mt=104 (n,d) cross section 525 1451 127 thres. to 20 mev, based on be9(d,n)b11 measurements of si65 525 1451 128 and ba60, and the (n,d) measurement of va65. 525 1451 129 525 1451 130 mt=107 (n,alpha) cross section 525 1451 131 0 to 20 mev, sum of mt=800,801. 525 1451 132 525 1451 133 mt=113 (n,t2alpha) cross section 525 1451 134 0 to 2.3 mev, based on a single-level fit to the resonance 525 1451 135 measured at 2 mev by da61, assuming l=0 incoming neu- 525 1451 136 trons and l=2 outgoing tritons. the thermal measure- 525 1451 137 ment (7+-2 mb) of ka87 was included in the analysis. 525 1451 138 2.3 to 20 mev, smooth curve through measurements of fr56, 525 1451 139 wy58, qa88, following general shape of da61 measurement 525 1451 140 from 4 to 9 mev. we assumed that the experimental data 525 1451 141 of qa88 supercedes reference qa85. 525 1451 142 525 1451 143 mt=600-605 (n,p) cross section to discrete levels 525 1451 144 0 to 20 mev, crudely estimated from the calculations 525 1451 145 of po70 and the (n,xgamma) measurements of ne70. cross 525 1451 146 section for mt=600 assumed similar to mt=113 below 525 1451 147 1 mev. gamma-ray decay scheme for be-10 from aj88. 525 1451 148 525 1451 149 mt=800 (n,alpha0) cross section 525 1451 150 0 to 1 mev, calculated from the r-matrix parameters 525 1451 151 described for mt=1. experimental (n,alpha0) data input 525 1451 152 to the fit were those of ma68 and da61. in addition, the 525 1451 153 angular distributions of va72 for the inverse reaction 525 1451 154 were included in the analysis. 525 1451 155 1 to 20 mev, based on da61 measurements, with smooth extra- 525 1451 156 polation from 8 to 20 mev using 14-MeV data of an69. The 525 1451 157 da61 data above approximately 2 mev were renormalized 525 1451 158 by a factor of approximately 1.4. Note that some of the 525 1451 159 structure seen in da61 was expanded to give consistent 525 1451 160 nonelastic, elastic, and total cross sections when 525 1451 161 compared with experimental data. 525 1451 162 525 1451 163 mt=801 (n,alpha1) cross section 525 1451 164 0 to 1 mev, calculated from the r-matrix parameters 525 1451 165 described for mt=1. experimental (n,alpha1) data in- 525 1451 166 cluded in the fit are those of sc76. in addition, the 525 1451 167 absolute differential cross-section measurements of 525 1451 168 se76 were included in the analysis. 525 1451 169 1 to 20 mev, smooth curve through measurements of da61 and 525 1451 170 ne70, with smooth extrapolation from 15 to 20 mev. the 525 1451 171 da61 data above approximately 2 mev were renormalized 525 1451 172 by a factor of approximately 1.4. Note that some of the 525 1451 173 structure seen in da61 was expanded to give consistent 525 1451 174 nonelastic, elastic, and total cross sections when 525 1451 175 compared with experimental data. 525 1451 176 525 1451 177 525 1451 178 mf=4---------- neutron angular distributions ------------------- 525 1451 179 525 1451 180 mt=2 elastic angular distributions 525 1451 181 0 to 1 mev, calculated from the r-matrix parameters 525 1451 182 described for mf=1,mt=1. experimental angular distri- 525 1451 183 butions input to the fit for both the elastic scatter- 525 1451 184 ing cross section and polarization were obtained from 525 1451 185 available measurements. 525 1451 186 1 to 14 mev, smoothed representation of legendre coeffi- 525 1451 187 cients derived from the measurements of la71, ha73, 525 1451 188 po70, ho69, co69, va69, va65, sa88, gl82, constrained 525 1451 189 to match the r-matrix calculations at en=1 mev. 525 1451 190 14 to 20 mev, optical model extrapolation of 14-mev data 525 1451 191 525 1451 192 mt=51 inelastic angular distribution to first level 525 1451 193 thres. to 12 mev, fit Legendre expansions to exp. data of 525 1451 194 Po70, Gl82, and Sa88. 525 1451 195 12 - 20 MeV, assumed similar distribution as 12 MeV. 525 1451 196 525 1451 197 mt=52-55 inelastic angular distribution to first level 525 1451 198 thres. to 12 mev, fit Legendre expansions to exp. data of 525 1451 199 Sa88. 525 1451 200 12 - 20 MeV, assumed similar distribution as 12 MeV. 525 1451 201 525 1451 202 mt=56-85 inelastic angular distributions 525 1451 203 thres. to 20 mev, assumed isotropic in center of mass 525 1451 204 525 1451 205 525 1451 206 mf=12 -------- gamma ray multiplicities ------------------------ 525 1451 207 525 1451 208 mt=102 capture gamma rays 525 1451 209 0 to 20 mev, capture spectra and transition probabilities 525 1451 210 derived from the thermal data of th67, after slight 525 1451 211 changes in the probabilities and renormalization to the 525 1451 212 energy levels of aj75. the lp flag is used to conserve 525 1451 213 energy and to reduce significantly the amount of data 525 1451 214 required in the file. except for the modification due 525 1451 215 to the lp flag, the thermal spectrum is used over the 525 1451 216 entire energy range. 525 1451 217 525 1451 218 mt=801 0.4776-mev photon from the (n,alpha1) reaction 525 1451 219 0 to 20 mev, multiplicity of 1.0 at all energies 525 1451 220 525 1451 221 525 1451 222 mf=13 -------- gamma-ray production cross sections ------------- 525 1451 223 525 1451 224 mt=4 (n,ngamma) cross section 525 1451 225 thres. to 20 mev, obtained from mt=51-60 using b-10 decay 525 1451 226 scheme obtained from aj88. 525 1451 227 525 1451 228 mt=103 (n,pgamma) cross sections 525 1451 229 thres. to 20 mev, obtained from mt=601-605 using be-10 525 1451 230 decay scheme deduced from aj88. 525 1451 231 525 1451 232 525 1451 233 mf=14 -------- gamma ray angular distributions ----------------- 525 1451 234 525 1451 235 mt=4 (n,ngamma) angular distributions 525 1451 236 thres. to 20 mev, assumed isotropic 525 1451 237 525 1451 238 mt=102 (n,gamma) angular distributions 525 1451 239 0 to 20 mev, assumed isotropic. 525 1451 240 525 1451 241 mt=103 (n,pgamma) angular distributions 525 1451 242 thres. to 20 mev, assumed isotropic 525 1451 243 525 1451 244 mt=801 (n,alpha1/gamma) angular distribution 525 1451 245 0 to 20 mev, assumed isotropic 525 1451 246 525 1451 247 525 1451 248 ----------------------- references ----------------------------- 525 1451 249 525 1451 250 aj75 f. ajzenberg-selove, nucl. phys. a248,6 (1975) 525 1451 251 aj88 f. ajzenberg-selove, nucl. phys. a490,1 (1988) 525 1451 252 an69 B. Antolkovic, Nuc.Phys.A139, 10 (1969). 525 1451 253 as70 a. asami and m.c. moxon, j.nucl.energy 24,85 (1970) 525 1451 254 au79 g.auchampaugh et al., nucl.sci.eng.69,30(1979) 525 1451 255 ba60 r.bardes and g.e. owen, phys.rev.120,1369 (1960) 525 1451 256 be56 r.l. becker and h.h. barschall, phys.rev.102,1384 (1956) 525 1451 257 bo51 c.k.bockelman et al., phys.rev. 84,69 (1951) 525 1451 258 bo69 d.bogart and l.l.nichols, nucl.phys.a125,463 (1969) 525 1451 259 ca85 a.carlson et al., nuc.data for basic & applied science, 525 1451 260 santa fe, nm (1985) p.1429. 525 1451 261 co52 j.h.coon et al., phys.rev. 88,562 (1952) 525 1451 262 co54 c.f.cook and t.w. bonner,phys.rev. 94,651 (1954) 525 1451 263 co67 s.a. cox and f.r. pontet, j.nucl.energy 21,271 (1967) 525 1451 264 co69 j.a. cookson and j.g.locke,nucl.phys.a146,417(1970) 525 1451 265 co73 m.s. coates et al., priv. comm. to l.stewart (1973) 525 1451 266 da56 r.b.day,phys.rev.102,767 (1956) 525 1451 267 da60 r.b. day and m.walt,phys.rev.117,1330 (1960) 525 1451 268 da61 e.a. davis et al., nucl.phys.27,448 (1961) 525 1451 269 di67 k.m. diment, aere-r-5224 (1967) 525 1451 270 di88 j.k.dickens, proc.conf. on nuc.data for sci.& tech.,mito, 525 1451 271 japan (1988) p.213. 525 1451 272 fo61 d.m. fossan et al., phys.rev. 123,209 (1961) 525 1451 273 fr56 g.m. frye and j.h. gammel,phys.rev. 103,328 (1956) 525 1451 274 gl82 s.glendinning, nuc.sci.eng.80,256(1982) 525 1451 275 ha73 s.l.hausladen, thesis, ohio univ. coo-1717-5 (1973) 525 1451 276 he80 d.hetrick & c.y.fu, ornl/tm-7341 (1980). 525 1451 277 hy69 m.hyakutake, eandc(j)-13 (1969) p.29 525 1451 278 ho69 j.c. hopkins, priv. comm. lasl (1969) 525 1451 279 ir67 d.c.irving, ornl-tm-1872 (1967) 525 1451 280 ka87 R.Kavanagh & R.Marcley, Phys.Rev.C36, 1194 (1987). 525 1451 281 la71 r.o. lane et al., phys.rev.c4,380 (1971) 525 1451 282 ma68 r.l.macklin and j.h.gibbons,phys.rev.165,1147 (1968) 525 1451 283 mo66 f.p.mooring et al.,nucl.phys.82,16 (1966) 525 1451 284 ne54 n.g.nereson,la-1655 (1954) 525 1451 285 ne70 d.o.nellis et al., phys.rev. c1,847 (1970) 525 1451 286 po70 d.porter et al., awre o 45/70 (1970) 525 1451 287 qa85 S.Qaim et al., Santa Fe Conf. (1985)p.97. 525 1451 288 qa88 S.Qaim et al., Mito Conf. (1988) p.225. 525 1451 289 sa88 E.T. Sadowski, Ph.D thesis, Ohio U., (Nov.,1988). 525 1451 290 sc76 r.a. schrack et al., proc.icinn(erda-conf-760715-p2),1345 525 1451 291 (1976) 525 1451 292 se76 r.m. sealock and j.c. overley, phys.rev.c13,2149 (1976) 525 1451 293 si65 r.h.siemssen et al., nucl.phys.69,209 (1965) 525 1451 294 sp73 r.r. spencer et al., eandc(e)147,al (1973) 525 1451 295 te62 k.tesch, nucl.phys.37,412 (1962) 525 1451 296 th67 g.e. thomas et al., nucl.instr.meth.56,325 (1967) 525 1451 297 ts63 k.tsukada and o.tanaka,j.phys.soc.japan 18,610 (1963) 525 1451 298 va65 v.valkovic et al., phys.rev. 139,331 (1965) 525 1451 299 va70 b.vaucher et al.,helv.phys.acta 43,237 (1970) 525 1451 300 va72 l.van der zwan and k.w.geiger, nucl.phys. a180,615 (1972) 525 1451 301 wi55 h.b. willard et al., phys.rev. 98,669(1955) 525 1451 302 wy58 m.e. wyman et al., phys.rev.112,1264 (1958) 525 1451 303 525 1451 304 525 1451 305 1 451 307 1 525 1451 306 3 107 61 1 525 1451 307 525 1 0 308 525 0 0 309 5.010000+3 9.926921+0 0 0 0 0 525 3107 310 2.790000+6 2.790000+6 0 0 2 174 525 3107 311 120 5 174 2 0 0 525 3107 312 1.000000-5 1.931800+5 2.530000-2 3.839496+3 9.400000+0 1.989352+2 525 3107 313 1.500000+2 4.961720+1 2.500000+2 3.837760+1 3.500000+2 3.240060+1 525 3107 314 4.500000+2 2.854440+1 5.500000+2 2.579550+1 6.500000+2 2.371760+1 525 3107 315 7.500000+2 2.206610+1 8.500000+2 2.071300+1 9.500000+2 1.957360+1 525 3107 316 1.500000+3 1.553510+1 2.500000+3 1.198710+1 3.500000+3 1.010500+1 525 3107 317 4.500000+3 8.892400+0 5.500000+3 8.028200+0 6.500000+3 7.374000+0 525 3107 318 7.500000+3 6.855100+0 8.500000+3 6.432100+0 9.500000+3 6.078500+0 525 3107 319 1.500000+4 4.821100+0 2.000000+4 4.170300+0 2.400000+4 3.807200+0 525 3107 320 3.000000+4 3.410700+0 4.500000+4 2.808100+0 5.500000+4 2.560500+0 525 3107 321 6.500000+4 2.377500+0 7.500000+4 2.235400+0 8.500000+4 2.121200+0 525 3107 322 9.500000+4 2.027500+0 1.000000+5 1.985100+0 1.200000+5 1.844300+0 525 3107 323 1.500000+5 1.674700+0 1.700000+5 1.572900+0 1.800000+5 1.524800+0 525 3107 324 1.900000+5 1.477400+0 2.000000+5 1.431400+0 2.100000+5 1.385600+0 525 3107 325 2.200000+5 1.341100+0 2.300000+5 1.298200+0 2.350000+5 1.277600+0 525 3107 326 2.400000+5 1.256500+0 2.450000+5 1.236500+0 2.500000+5 1.216800+0 525 3107 327 2.600000+5 1.178100+0 2.700000+5 1.140800+0 2.800000+5 1.105300+0 525 3107 328 3.000000+5 1.037340+0 3.200000+5 9.780500-1 3.400000+5 9.279900-1 525 3107 329 3.600000+5 8.892600-1 3.800000+5 8.630200-1 4.000000+5 8.490400-1 525 3107 330 4.200000+5 8.452800-1 4.400000+5 8.464200-1 4.600000+5 8.438100-1 525 3107 331 4.800000+5 8.287000-1 5.000000+5 7.971800-1 5.200000+5 7.520500-1 525 3107 332 5.400000+5 6.998100-1 5.600000+5 6.466400-1 5.800000+5 5.964400-1 525 3107 333 6.000000+5 5.509900-1 6.500000+5 4.592600-1 7.000000+5 3.929700-1 525 3107 334 7.500000+5 3.441300-1 8.000000+5 3.071500-1 8.500000+5 2.784700-1 525 3107 335 9.000000+5 2.558200-1 9.500000+5 2.376900-1 1.000000+6 2.230560-1 525 3107 336 1.050000+6 2.139881-1 1.100000+6 2.107045-1 1.150000+6 2.113211-1 525 3107 337 1.200000+6 2.145886-1 1.250000+6 2.196000-1 1.300000+6 2.273315-1 525 3107 338 1.350000+6 2.388980-1 1.400000+6 2.538785-1 1.450000+6 2.719450-1 525 3107 339 1.500000+6 2.928000-1 1.550000+6 3.248122-1 1.600000+6 3.721273-1 525 3107 340 1.650000+6 4.273298-1 1.700000+6 4.798547-1 1.750000+6 5.165000-1 525 3107 341 1.800000+6 5.262540-1 1.850000+6 5.215069-1 1.900000+6 5.060907-1 525 3107 342 1.950000+6 4.847658-1 2.000000+6 4.596000-1 2.050000+6 4.280959-1 525 3107 343 2.100000+6 3.958590-1 2.150000+6 3.647927-1 2.200000+6 3.384507-1 525 3107 344 2.250000+6 3.194910-1 2.300000+6 3.052108-1 2.350000+6 2.938520-1 525 3107 345 2.400000+6 2.849606-1 2.450000+6 2.810818-1 2.500000+6 2.878578-1 525 3107 346 2.550000+6 3.034110-1 2.600000+6 3.200000-1 2.650000+6 3.357468-1 525 3107 347 2.700000+6 3.518343-1 2.750000+6 3.815892-1 2.800000+6 4.000000-1 525 3107 348 2.850000+6 3.847509-1 2.900000+6 3.650000-1 2.950000+6 3.468235-1 525 3107 349 3.000000+6 3.300000-1 3.050000+6 3.154825-1 3.100000+6 3.032678-1 525 3107 350 3.150000+6 2.912017-1 3.200000+6 2.820000-1 3.300000+6 2.821814-1 525 3107 351 3.400000+6 2.960000-1 3.600000+6 3.300000-1 3.700000+6 3.474155-1 525 3107 352 3.800000+6 3.650000-1 4.000000+6 3.500000-1 4.200000+6 2.745479-1 525 3107 353 4.300000+6 2.606951-1 4.400000+6 2.450000-1 4.600000+6 2.000000-1 525 3107 354 4.800000+6 1.712930-1 5.000000+6 1.478000-1 5.200000+6 1.318452-1 525 3107 355 5.400000+6 1.161278-1 5.500000+6 1.124321-1 5.600000+6 1.121000-1 525 3107 356 5.800000+6 1.147314-1 6.000000+6 1.165057-1 6.200000+6 1.181696-1 525 3107 357 6.400000+6 1.183391-1 6.600000+6 1.171288-1 6.800000+6 1.148436-1 525 3107 358 7.000000+6 1.104512-1 7.200000+6 1.041907-1 7.400000+6 9.738018-2 525 3107 359 7.600000+6 9.071085-2 7.800000+6 8.463679-2 8.000000+6 7.942750-2 525 3107 360 8.200000+6 7.465784-2 8.400000+6 7.052106-2 8.600000+6 6.689338-2 525 3107 361 8.800000+6 6.367160-2 9.000000+6 6.076367-2 9.200000+6 5.812695-2 525 3107 362 9.400000+6 5.577225-2 9.600000+6 5.369136-2 9.800000+6 5.187475-2 525 3107 363 1.000000+7 5.031229-2 1.050000+7 4.658572-2 1.100000+7 4.403554-2 525 3107 364 1.150000+7 4.218219-2 1.200000+7 4.126500-2 1.250000+7 4.130563-2 525 3107 365 1.300000+7 4.198975-2 1.350000+7 4.319183-2 1.400000+7 4.437300-2 525 3107 366 1.450000+7 4.490981-2 1.500000+7 4.459223-2 1.550000+7 4.317406-2 525 3107 367 1.600000+7 4.164776-2 1.650000+7 4.015499-2 1.700000+7 3.870001-2 525 3107 368 1.750000+7 3.722898-2 1.800000+7 3.576560-2 1.850000+7 3.430929-2 525 3107 369 1.900000+7 3.285713-2 1.950000+7 3.141235-2 2.000000+7 2.998125-2 525 3107 370 525 3 0 371 525 0 0 372 0 0 0 373 1.202400+4 2.398500+1 -1 0 0 01225 1451 1 0.000000+0 0.000000+0 0 0 0 61225 1451 2 1.000000+0 0.000000+0 0 0 10 61225 1451 3 0.000000+0 0.000000+0 0 0 31 21225 1451 4 12-MG- 24 HEDL,ORNL EVAL-NOV79 MANN,LARSON 1225 1451 5 DIST-JAN90 900123 1225 1451 6 ----ENDF/B-VI MATERIAL 1225 1225 1451 7 -----INCIDENT NEUTRON DATA 1225 1451 8 ------ENDF-6 FORMAT 1225 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1225 1451 10 DOSIMETRY 1225 1451 11 ******************************************************************1225 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 1225 1451 13 * * * * * * *1225 1451 14 ******* 24MG(N,P) CROSS SECTION FOR ACTIVATION FILE ****** 1225 1451 15 ***** EVALUATION TAKEN FROM TAGESEN, VONACH AND STROHMAIER BY 1225 1451 16 ***** D.C. LARSON ORNL ******** 1225 1451 17 THE EVALUATION FOR THIS CROSS SECTION WAS TAKEN FROM THE RECENT 1225 1451 18 EVALUATION OF THIS CROSS SECTION FOR AN INTERNATIONAL DOSIMETRY 1225 1451 19 FILE. THIS WORK IS DOCUMENTED IN DETAIL IN REF TA79. THE RESULTS1225 1451 20 OF 17 MEASUREMENTS WERE RE-ANALYZED USING THE LATEST BRANCHING 1225 1451 21 RATIOS, ETC., AND RENORMALIZING WHERE NECESSARY TO UPDATED VALUES 1225 1451 22 OF "STANDARD" CROSS SECTIONS. THIS WAS A VERY CAREFUL EVALUATION 1225 1451 23 WELL DOCUMENTED, AND THE USER IS REFERRED TO THE DOCUMENT TA79 1225 1451 24 FOR ALL THE DETAILS. COMPLETE UNCERTAINTY FILES ARE ALSO GIVEN. 1225 1451 25 ***** DECAY OF 24NA (FILES 8/103 AND 9/103) ******** 1225 1451 26 THESE FILES WERE PROVIDED BY C. REICH OF INEL FOR THE ENDF/B-V 1225 1451 27 EVALUATION EFFORT OF CSWEG. 1225 1451 28 THE 24MG(N,P) REACTION LEADS TO 24NA, WHICH HAS A HALFLIFE OF 1225 1451 29 15.02 HOURS. IT DECAYS 99.94% BACK TO THE 4.12 MEV LEVEL IN 24MG 1225 1451 30 WHICH THEN EMITS 2.754 MEV AND 1.369 MEV GAMMA RAYS IN THE DECAY 1225 1451 31 TO THE GROUND STATE OF 24MG. 1225 1451 32 ****** REFERENCES ******* 1225 1451 33 TA79 S. TAGESEN, H. VONACH AND B. STROHMAIER, PHYSICS DATA 1225 1451 34 REPORT NR 13-1, KARLSRUHE (1979) 1225 1451 35 1 451 37 01225 1451 36 3 103 17 01225 1451 37 1225 1 0 38 1225 0 0 39 1.202400+4 2.398500+1 0 0 0 01225 3103 40 -4.731000+6-4.731000+6 0 0 1 411225 3103 41 41 1 0 0 0 01225 3103 42 4.930000+6 2.100000-5 5.200000+6 4.700000-5 5.600000+6 2.100000-41225 3103 43 5.800000+6 1.907000-3 6.100000+6 4.359000-3 6.300000+6 8.705000-31225 3103 44 6.500000+6 3.698000-2 6.700000+6 4.364000-2 6.900000+6 3.575000-21225 3103 45 7.100000+6 5.225000-2 7.310000+6 4.658000-2 7.510000+6 8.020000-21225 3103 46 7.700000+6 9.339000-2 7.900000+6 1.063000-1 8.110000+6 1.199000-11225 3103 47 8.500000+6 1.193000-1 9.500000+6 1.234000-1 1.000000+7 1.482000-11225 3103 48 1.050000+7 1.580000-1 1.150000+7 1.820000-1 1.220000+7 1.928000-11225 3103 49 1.240000+7 2.096000-1 1.260000+7 2.002000-1 1.280000+7 2.008000-11225 3103 50 1.300000+7 1.977000-1 1.320000+7 2.147000-1 1.340000+7 2.177000-11225 3103 51 1.360000+7 2.014000-1 1.380000+7 1.891000-1 1.400000+7 1.965000-11225 3103 52 1.420000+7 1.968000-1 1.440000+7 1.907000-1 1.460000+7 1.823000-11225 3103 53 1.480000+7 1.718000-1 1.500000+7 1.782000-1 1.550000+7 1.570000-11225 3103 54 1.600000+7 1.398000-1 1.700000+7 1.223000-1 1.800000+7 1.135000-11225 3103 55 1.900000+7 9.713000-2 2.000000+7 8.091000-2 0.000000+0 0.000000+01225 3103 56 1225 3 0 57 1225 0 0 58 0 0 0 59 1.302700+4 2.675000+1 -1 0 0 01325 1451 1 0.000000+0 0.000000+0 0 0 0 61325 1451 2 1.000000+0 0.000000+0 0 0 10 61325 1451 3 0.000000+0 0.000000+0 0 0 300 51325 1451 4 13-AL- 27 LANL EVAL-DEC73 P.G. YOUNG, D.G. FOSTER, JR. 1325 1451 5 LA-4726 (1973). DIST-JAN90 900123 1325 1451 6 ----ENDF/B-VI MATERIAL 1325 1325 1451 7 -----INCIDENT NEUTRON DATA 1325 1451 8 ------ENDF-6 FORMAT 1325 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1325 1451 10 DOSIMETRY 1325 1451 11 ******************************************************************1325 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 1325 1451 13 * * * * * * *1325 1451 14 1325 1451 15 REVISED-AUG77. VERSION IV DATA, WITH ADDITION OF FILE 33. 1325 1451 16 1325 1451 17 ******************* REVISED 11/73 ****************** 1325 1451 18 THE NEW EXPERIMENTAL DATA ARE DESCRIBED BELOW. BRIEFLY, 1325 1451 19 THE FOLLOWING IMPROVEMENTS WERE MADE. 1325 1451 20 1. THE RADIATIVE CAPTURE GAMMA-RAY SPECTRUM FOR THERMAL 1325 1451 21 NEUTRONS WAS REVISED TO BETTER AGREE WITH RECENT DATA. 1325 1451 22 2. IMPROVED (N,P) AND (N,ALPHA) CROSS SECTIONS WERE INCOR- 1325 1451 23 PORATED. THESE DATA ARE THE SAME AS WERE PROVIDED 1325 1451 24 EARLIER TO THE STANDARDS SUBCOMMITTEE OF CSEWG. 1325 1451 25 3. FILE 13 AND FILE 15 (N,NGAMMA) AND (N,NPGAMMA) DATA WERE 1325 1451 26 REVISED TO INCLUDE THE MEASUREMENTS OF DI73. 1325 1451 27 4. THE (N,NPRIME) CROSS SECTIONS TO HIGH-LYING PSEUDO LEVELS 1325 1451 28 WERE REVISED FOR NEUTRON ENERGIES NEAR 14 MEV TO GIVE 1325 1451 29 BETTER AGREEMENT WITH MEASUREMENTS OF KA72. IN ADDITION, 1325 1451 30 (N,NPRIME) DATA TO LOWER LEVELS WERE REVISED TO BE 1325 1451 31 CONSISTENT WITH CHANGES MADE UNDER ITEM 3. 1325 1451 32 5. ANISOTROPIC (N,NPRIME) ANGULAR DISTRIBUTIONS BASED ON 1325 1451 33 EXPERIMENTAL DATA WERE INCORPORATED. 1325 1451 34 6. TOTAL NEUTRON CROSS SECTION WAS EXTENSIVELY REVISED, ES- 1325 1451 35 PECIALLY BELOW 1 MEV, TO INCLUDE THE MEASUREMENT OF PE72. 1325 1451 36 **************************************************************** 1325 1451 37 1325 1451 38 MF=2 --- RESONANCE PARAMETERS --- 1325 1451 39 MT=151 EFFECTIVE SCATTERING RADIUS = 0.32752E-12 CM 1325 1451 40 MF=3 --- NEUTRON SMOOTH CROSS SECTIONS --- 1325 1451 41 THE 2200 M/S CROSS SECTIONS ARE AS FOLLOWS, 1325 1451 42 MT=1 SIGMA = 1.580 B MT=2 SIGMA = 1.348 B 1325 1451 43 MT=3 SIGMA = 0.232 B MT=102 SIGMA = 0.232 B 1325 1451 44 MT=1 TOTAL CROSS SECTION 1325 1451 45 BELOW 4.6 KEV, 1/V FIT TO ME52 AND HI59 (NORMALIZED TO ME52),1325 1451 46 USING THERMAL CAPTURE XS OF 232 MB (GO71) AND RESULTING 1325 1451 47 (N,N) XS OF 1.580 B. 1325 1451 48 4.6 - 189 KEV, FROM GA65 NORMALIZED TO ME52 VIA HI59 AT LOW 1325 1451 49 ENERGIES AND TO PE72 AT HIGH ENERGIES. 1325 1451 50 209 - 475 KEV, FROM PE72 WITH ENERGY SCALE CORRECTED TO MATCH1325 1451 51 CI68 AT HIGHER ENERGIES. 1325 1451 52 .475 - 11.5 MEV, COMPOSITE OF PE72 AND SC69, NORMALIZED TO 1325 1451 53 THEIR WEIGHTED AVERAGE, WITH INSERTS OF NORMALIZED CI68 1325 1451 54 WHERE NEEDED TO PRESERVE RESOLUTION. 1325 1451 55 11.5 - 20 MEV, COMPOSITE OF PE72,FO71,AL66,AND PE60 NORM- 1325 1451 56 ALIZED TO THEIR WEIGHTED AVERAGE. SLOPE AT 20 MEV INCLUDES 1325 1451 57 GUIDANCE FROM TA52. 1325 1451 58 SMOOTHING BETWEEN 4.6 KEV AND 2 MEV BY APPROXIMATE BREIT- 1325 1451 59 WIGNER FITS WHERE POSSIBLE, CONNECTED BY POLYNOMIAL FITS. 1325 1451 60 2 - 12 MEV SMOOTHED BY RUNNING CUBIC POLYNOMIAL FITS. 1325 1451 61 ABOVE 12 MEV FITTED SECONDARY POLYNOMIAL TO MIDDLES OF 1325 1451 62 RUNNING POLYNOMIAL FITS. 1325 1451 63 MT=2 ELASTIC SCATTERING CROSS SECTION 1325 1451 64 BELOW 5 MEV, SUBTRACTED NONELASTIC FROM TOTAL. 1325 1451 65 5 TO 16 MEV, MAINLY BASED ON ELASTIC DATA OF HO69, ST59, KI701325 1451 66 ,ST65,CO58,CO59,BE58,MI68,BE56, AND KA72 TOGETHER WITH THE 1325 1451 67 EVAL. TOTAL AND THE NONELASTIC MEASUREMENTS LISTED BELOW. 1325 1451 68 16 TO 20 MEV, SMOOTH EXTRAP. TO 1/2 THE TOTAL AT 20 MEV. 1325 1451 69 MT=3 NONELASTIC CROSS SECTION (IMPLICIT - NOT IN FILE) 1325 1451 70 BELOW 9 MEV, BASED ON THE NONELASTIC MEASUREMENTS OF BA58, 1325 1451 71 BE56, TA55, DE65, ON THE SUM OF MT=4,102,103,107, AND ON 1325 1451 72 THE DIFFERENCE BETWEEN THE EVALUATED TOTAL AND THE ELASTIC 1325 1451 73 MEASUREMENTS OF HO69, KI70, BE56, MI68. 1325 1451 74 9 TO 16 MEV, BASED ON NONELASTIC MEASUREMENTS OF BA58, TA55, 1325 1451 75 GR55, PH52, MA57, DE61, FL56, CH67, AND ON THE DIFFERENCE 1325 1451 76 BETWEEN MT=1 AND MT=2 DATA OF ST59, ST65, CO58, CO59, BE58 1325 1451 77 16 TO 20 MEV, DIFFERENCE BETWEEN MT=1 AND MT=2 1325 1451 78 MT=4 INELASTIC CROSS SECTION 1325 1451 79 THRES. TO 5 MEV, SUM OF MT=51-63. 1325 1451 80 5 TO 9 MEV, BASED ON (N,NPRIME) DATA OF TH63, TO67, THE EVAL.1325 1451 81 (N,NPRIME) DATA OF DI71 WHICH INCLUDES CONSIDERATION OF 1325 1451 82 SEVERAL OTHER MEASUREMENTS, AND THE (N,NG) DATA OF DI71, 1325 1451 83 OR71, AND DI73. 1325 1451 84 9 TO 20 MEV, THE DIFFERENCE BETWEEN MT=1 AND THE SUM OF 1325 1451 85 MT=2,16,102,103,104,105,107. 1325 1451 86 MT=16 (N,2N) CROSS SECTION 1325 1451 87 THRES. TO 20 MEV, ESTIMATED USING A NUCLEAR TEMP. CALCULATION1325 1451 88 ASSUMING THAT HIGHLY EXCITED STATES IN AL-27 DECAY 50 PER 1325 1451 89 CENT BY NEUTRON EMISSION. 1325 1451 90 MT=51-63 INELASTIC CROSS SECTION TO DISCRETE STATES 1325 1451 91 MT=51 Q=-0.843 MEV MT=56 Q=-3.001 MEV MT=60 Q=-4.409 MEV 1325 1451 92 52 -1.013 57 -3.678 61 -4.508 1325 1451 93 53 -2.210 58 -3.956 62 -4.580 1325 1451 94 54 -2.732 59 -4.055 63 -4.811 1325 1451 95 55 -2.980 1325 1451 96 THRES. TO 5 MEV, BASED ON THE (N,NPRIME) DATA OF TO62, WI63,1325 1451 97 TS61, AND THE (N,NG) DATA OF CH68, MA65, DI71, OR71, DI73. 1325 1451 98 5 TO 9 MEV, BASED ON AN EVALUATION OF SEVERAL MEASUREMENTS 1325 1451 99 GIVEN IN DI71 (TABLE B1). 1325 1451 100 9 TO 20 MEV, SMOOTH EXTRAPOLATION PASSING THROUGH 14 MEV DATA1325 1451 101 OF ST65, BO65A. 1325 1451 102 MT=64-90 INELASTIC CROSS SECTION TO GROUPS OF DISCRETE STATES 1325 1451 103 IN BANDS CENTERED ABOUT THE Q-VALUES GIVEN BELOW, 1325 1451 104 MT=64 Q=- 5.25 MEV MT=73 Q=- 9.75 MEV MT=82 Q=-14.25 MEV 1325 1451 105 65 - 5.75 74 -10.25 83 -14.75 1325 1451 106 66 - 6.25 75 -10.75 84 -15.25 1325 1451 107 67 - 6.75 76 -11.25 85 -15.75 1325 1451 108 68 - 7.25 77 -11.75 86 -16.25 1325 1451 109 69 - 7.75 78 -12.25 87 -16.75 1325 1451 110 70 - 8.25 79 -12.75 88 -17.375 1325 1451 111 71 - 8.75 80 -13.25 89 -18.125 1325 1451 112 72 - 9.25 81 -13.75 90 -18.875 1325 1451 113 THRES. TO 20 MEV, INTEGRATED CROSS SECTION OVER BANDS 1325 1451 114 OBTAINED BY SUBTRACTING MT=51-63 FROM MT=4. CROSS SECTION 1325 1451 115 DIVIDED AMONG BANDS ACCORDING TO A NUCLEAR TEMPERATURE 1325 1451 116 CALCULATION USING TEMP. BASED ON (N,NPRIME) DATA OF TH63, 1325 1451 117 GR53. THE CROSS SECTION TO THE BANDS WITH MT=64-71 WAS 1325 1451 118 ADJUSTED EXTENSIVELY TO PRODUCE AGREEMENT WITH THE 14-MEV 1325 1451 119 MEASUREMENTS OF SECONDARY NEUTRON SPECTRA BY KA72. 1325 1451 120 ****PLEASE NOTE THAT MUCH OF THE CROSS SECTION TO BANDS ABOVE EX=1325 1451 121 ****9 MEV SUBSEQUENTLY RESULTS IN CHARGED PARTICLE EMISSION. 1325 1451 122 ****SINCE THESE DATA ARE NOT INCLUDED EXPLICITLY AS (N,NX) REAC- 1325 1451 123 ****TIONS, IT IS IMPORTANT THAT USERS INVOLVED IN CERTAIN CALCU- 1325 1451 124 ****LATIONS (E.G., LOCAL HEATING) BE AWARE OF THIS INFORMATION. 1325 1451 125 MT=102 (N,GAMMA) CROSS SECTION 1325 1451 126 BELOW 1 KEV, 1/V FROM THERMAL VALUE OF 232 MB (GO71). 1325 1451 127 1 TO 140 KEV, FROM BL68. 6-KEV RESONANCE HAS WIDTH DEDUCED 1325 1451 128 FROM TOTAL X-SECT. SMALL RESONANCES NEARBY ALSO FROM TOTAL.1325 1451 129 ABOVE 140 KEV, MAINLY FROM HE50, HE53, AND CA62. 1325 1451 130 FROM TOTAL X-SECT. SMALL RESONANCES NEARBY ALSO FROM TOTAL.1325 1451 131 ABOVE 140 KEV, MAINLY FROM HE50, HE53, AND CA62. 1325 1451 132 MT=103 (N,P) CROSS SECTION (SEE YO72 FOR MORE DETAILS) 1325 1451 133 THRES. TO 4 MEV, BASED ON THE MEASUREMENTS OF HE54 AND GR67, 1325 1451 134 EXTRAPOLATED FROM 3 MEV TO THRESHOLD WITH AN L=0 PENETRA- 1325 1451 135 BILITY FUNCTION FOR THE OUTGOING P+MG27 CHANNEL. 1325 1451 136 4 TO 20 MEV, SMOOTH CURVE THROUGH THE MEASUREMENTS OF HE54, 1325 1451 137 GR67, CA62, BA66, MA60. 1325 1451 138 MT=104 (N,D) CROSS SECTION 1325 1451 139 THRES. TO 20 MEV, SMOOTH CURVE THROUGH SINGLE DATUM OF GL61. 1325 1451 140 MT=105 (N,T) CROSS SECTION 1325 1451 141 THRES. TO 20 MEV, SMOOTH CURVE WITH SAME SHAPE AS MT=104, 1325 1451 142 REACHING MAXIMUM OF 15 MB AT 20 MEV. 1325 1451 143 MT=107 (N,ALPHA) CROSS SECTION (SEE YO72 FOR MORE DETAILS). 1325 1451 144 THRES. TO 20 MEV, BASED ON A SMOOTH CURVE THROUGH EXPERI- 1325 1451 145 MENTAL DATA, MAINLY THOSE OF BU63, PA65, AND LI66, WITH 1325 1451 146 RESULTS FROM GR67, SC61, GR58, BA61, MA60, GA62, IM64, 1325 1451 147 KE59 AND SEVERAL 14-MEV POINTS ALSO BEING CONSIDERED. 1325 1451 148 THE CURVE WAS EXTRAPOLATED FROM 6 MEV TO THRESHOLD WITH AN 1325 1451 149 L=0 PENETRABILITY FUNCTION FOR THE OUTGOING ALPHA+NA24 1325 1451 150 CHANNEL. 1325 1451 151 MT=203 TOTAL H PRODUCTION IS THE SUM OF MTS 75 + 78 +80 + 81 1325 1451 152 + 83 + 84 + 85 + 87 + 88 + 90 + 103. 1325 1451 153 MT=207 TOTAL HE-4 PRODUCTION IS THE SUM OF MTS 82 + 86 + 89 1325 1451 154 + 107. 1325 1451 155 MF=4 --- NEUTRON ANGULAR DISTRIBUTIONS --- 1325 1451 156 MT=2 ELASTIC ANGULAR DISTRIBUTIONS 1325 1451 157 FROM SMOOTH CURVES THROUGH PLOTS OF COEFFICIENTS FROM FITS 1325 1451 158 TO ALL AVAILABLE DATA ABOVE .25 MEV, WITHOUT EVALUATION 1325 1451 159 OF MEASUREMENTS. MAINLY FROM LA57, CH66, TO62, TS61, KI70, 1325 1451 160 BE58, WITH OPTICAL-MODEL BRIDGE TO ST62. SOME OF DATA 1325 1451 161 AUGMENTED BEFORE FITTING BY ZERO-DEGREE POINT SLIGHTLY 1325 1451 162 ABOVE WICK LIMIT, AND COEFFICIENTS ADJUSTED EMPIRICALLY 1325 1451 163 AFTERWARDS TO OBEY WICK LIMIT, BUT ONLY BARELY AT HIGHER 1325 1451 164 ENERGIES. 1325 1451 165 MT=16 (N,2N) ANGULAR DISTRIBUTIONS 1325 1451 166 ASSUMED ISOTROPIC IN CM, USING 2-BODY KINEMATICS TO ESTIMATE 1325 1451 167 TRANSFORMATION TO LAB SYSTEM. 1325 1451 168 MT=51-63 INELASTIC NEUTRON ANGULAR DISTRIBUTIONS 1325 1451 169 ANISOTROPIC DISTRIBUTIONS BASED ON MEASUREMENTS BY TS61,TO62,1325 1451 170 WI63, MI68, TA70, GL72, ST65, BO65A, KA72 WERE IN- 1325 1451 171 CORPORATED OVER VARYING ENERGY RANGES FROM THRES.TO 20 MEV.1325 1451 172 IN SEVERAL REGIONS WHERE DATA WERE LACKING, ISOTROPY WAS 1325 1451 173 ASSUMED. 1325 1451 174 MT=64-90 INELASTIC NEUTRON ANGULAR DISTRIBUTIONS 1325 1451 175 ASSUMED ISOTROPIC IN THE CENTER-OF-MASS SYSTEM. 1325 1451 176 MF=5 --- NEUTRON ENERGY DISTRIBUTIONS --- 1325 1451 177 MT= 16 (N,2N) ENERGY DISTRIBUTIONS, BASED ON STATISTICAL THEORY1325 1451 178 MF=12 --- GAMMA RAY MULTIPLICITIES --- 1325 1451 179 MT=102 (N,G) CAPTURE MULTIPLICITIES 1325 1451 180 THE AL-28 DECAY SCHEME FOR EXCITATION BY RADIATIVE CAPTURE 1325 1451 181 WITH THERMAL NEUTRONS FROM HA69 WAS ADOPTED AFTER SOME 1325 1451 182 MODIFICATION TO ENHANCE AGREEMENT WITH EXPERIMENT. THE 1325 1451 183 RESULTS HAVE BEEN COMPARED WITH THE LESS DETAILED DATA OF 1325 1451 184 BA67, JU71, AND MA69, AND ARE REASONABLY CONSISTENT. SOME 1325 1451 185 OF THE LINES IN RA70 BELOW 1 MEV APPEAR TO BE SPURIOUS 1325 1451 186 AND HAVE BEEN DROPPED. SPECTRUM FOR THERMAL NEUTRONS IS 1325 1451 187 ASSUMED TO APPLY AT ALL ENERGIES. 1325 1451 188 MF=13 --- GAMMA RAY SMOOTH CROSS SECTIONS --- 1325 1451 189 MT=4 (N,NG) CROSS SECTIONS 1325 1451 190 THRES. TO 5 MEV, BASED ON THE (N,NPRIME) DATA OF TO62, WI63, 1325 1451 191 TS61, AND THE (N,NG) DATA OF CH68,MA65,DI71,OR71, AND DI73.1325 1451 192 5 TO 20 MEV, MAINLY FROM (N,NG) DATA OF DI73, OR71, DI71, 1325 1451 193 DR73, PE64, BE66, CL69, EN67, BO65, MA68, CA60, PR60, 1325 1451 194 SUPPLEMENTED BY STATISTICAL THEORY CALCULATIONS. 1325 1451 195 MT=28 (N,NPG) CROSS SECTIONS 1325 1451 196 THRES. TO 20 MEV, BASED ON THE MEASUREMENTS OF DI73, OR71, 1325 1451 197 EN67, PR60, CL69, BE66, BO65, AND ON STATISTICAL THEORY 1325 1451 198 MT=103 (N,PG) CROSS SECTIONS 1325 1451 199 THRES. TO 9 MEV, USED DATA OF DI71. 1325 1451 200 9 TO 20 MEV, SMOOTH EXTRAPOLATION. 1325 1451 201 MF=14 --- GAMMA RAY ANGULAR DISTRIBUTIONS --- 1325 1451 202 MT=4 (N,NG) ANGULAR DISTRIBUTIONS, ASSUMED ISOTROPIC. 1325 1451 203 MT=28 (N,NPG) ANGULAR DISTRIBUTIONS ASSUMED ISOTROPIC. 1325 1451 204 MT=103 (N,PG) ANGULAR DISTRIBUTIONS, ASSUMED ISOTROPIC 1325 1451 205 MF=15 --- GAMMA RAY ENERGY DISTRIBUTIONS --- 1325 1451 206 MT=4 (N,NG) ENERGY DISTRIBUTIONS 1325 1451 207 STATISTICAL THEORY CAL. ADJUSTED TO FIT MEAS. OF DI73 1325 1451 208 MT=28 (N,NPG) ENERGY DISTRIBUTIONS 1325 1451 209 STATISTICAL THEORY CAL. ADJUSTED TO FIT MEAS. OF DI73 1325 1451 210 MF=33 --- CORRELATED ERRORS --- 1325 1451 211 BASED ON REVISED VERSION OF TABLE 6 IN LA-4726. ERRORS ARE 1325 1451 212 GIVEN ONLY FOR MT = 1, 2, 4, 16, 102, 103, AND 107. 1325 1451 213 MT=2 IS TAKEN FROM EXPERIMENT BETWEEN 9 AND 17 MEV, AND 1325 1451 214 DERIVED EVERYWHERE ELSE. MT=4 IS DERIVED ABOVE 9 MEV. 1325 1451 215 THE DERIVATION FORMULAS ARE GIVEN BY NC-TYPE SUB-SUBSECTIONS 1325 1451 216 WITH LTY=0 IN MT=2 AND MT=4. 1325 1451 217 1325 1451 218 --------------------------- REFERENCES --------------------------1325 1451 219 BA58 W.P.BALL ET AL.(LRL), PHYS.REV.110,1392 (1958). 1325 1451 220 BA61 B.P.BAYHURST+R.J.PRESTWOOD (LAS),PHYS.REV.121,1438 (1961). 1325 1451 221 BA66 R.BASS, EANDC(E)-66U,P64, AND PRIVATE COMMUNICATION (1966).1325 1451 222 BA67 G.A.BARTHOLOMEW ET AL., NUCL. DATA A3, 367 (1967) 1325 1451 223 BE55 J.R.BEYSTER ET AL.(LAS), PHYS.REV.98,1216 (1955). 1325 1451 224 BE56 J.R.BEYSTER ET AL.(LAS), PHYS.REV.104,1319 (1956). 1325 1451 225 BE58 S.BERKO ET AL., NUCL. PHYS. 6, 210 (1958) 1325 1451 226 BE66 V.M.BEZOTOSNYI ET AL.(RUS),SOV.J.NUCL.PHYS.3,632 (1966). 1325 1451 227 BL68 R.C.BLOCK, PRIVATE COMMUNICATION TO R.J.HOWERTON (1968) 1325 1451 228 BO59 N.A.BOSTROM ET AL.(TNC), WADC-TN-59-107 (1959). 1325 1451 229 BO65 V.N.BOCHKAREV+V.V.NEFEDOV (LEB),SOV.J.NUCL.PHYS.1,574(1965)1325 1451 230 BO65A G.C.BONAZZOLA ET AL.(TUR), PHYS.REV.140,835 (1965). 1325 1451 231 BU63 J.P.BUTLER AND D.C.SANTRY, CAN.J.PHYS.41, 372 (1963.) 1325 1451 232 CA60 R.L.CALDWELL ET AL.(SOC),NUCL.SCI.ENG.8,173 (1960). 1325 1451 233 CA62 G.CALVI ET AL., NUCL. PHYS. 39, 621 (1962) 1325 1451 234 CA67 A.D.CARLSON AND H.H.BARSCHALL, PHYS. REV. 158, 1142 (1967) 1325 1451 235 CH66 J.P.CHIEN AND A.B.SMITH, NUCL. SCI. ENG. 26, 500 (1966) 1325 1451 236 CH67 A.CHATTERJEE+A.M.GHOSE(BOS), PHYS.REV.161,1181 (1967). 1325 1451 237 CH68 K.C.CHUNG ET AL.(KTY),NUCL.PHYS.68,476 (1968). 1325 1451 238 CI68 S.CIERJACKS ET AL., KFK-1000 (1968) 1325 1451 239 CL69 G.CLAYEUX+G.GRENIER (FR),CEA-R-3807,(1969). 1325 1451 240 CO58 J.H.COON ET AL.(LAS), PHYS.REV. 111,250 (1958). 1325 1451 241 CO59 J.H.COON (LAS),PRIVATE COMMUNICATION TO R.J.HOWERTON (1959)1325 1451 242 DE61 YU.G.DEGTYAREV+V.G.NADTOCHII (RUS), SOV.J.AT.ENERGY 11,10431325 1451 243 (1961). 1325 1451 244 DE65 YU.G.DEGTYAREV (RUS), J.NUCL.ENERGY 20,818 (1965). 1325 1451 245 DI71 J.K.DICKENS (ORL), ORNL-TM-3284, 1971. 1325 1451 246 DI73 J.K.DICKENS ET AL., ORNL-TM-4232 (1973). 1325 1451 247 DR70 D.M.DRAKE ET AL. (LAS), NUCL.SCI.ENG.40,294 (1970). 1325 1451 248 EN67 F.C.ENGESSER+W.E.THOMPSON (NRD),J.NUCL.ENERGY 21,487 (1967)1325 1451 249 FE67 J.M.FERGUSON+J.C.ALBERGOTTI (NRD),NUCL.PHYS.98,65 (1967). 1325 1451 250 FL56 N.N.FLEROV+V.M.TALYZIN (RUS),ATOMNAYA ENERGIYA 1,155 (1956)1325 1451 251 FO71 D.G.FOSTER JR. AND D.W.GLASGOW, PHYS. REV. C 3, 576 (1971) 1325 1451 252 GA62 F.GABBARD+B.D.KERN(KTY), PHYS.REV.128,1276 (1962). 1325 1451 253 GA65 J.B.GARG ET AL., PRIV. COMM. TO R.J.HOWERTON (1965) 1325 1451 254 GL61 R.N.GLOVER+E.WIEGOLD (CBR),NUCL.PHYS.24,630 (1961). 1325 1451 255 GL72 D.W.GLASGOW, PRIVATE COMMUNICATION (1972). 1325 1451 256 GO71 D.T.GOLDMAN, UNPUBLISHED EVALUATION (1971) 1325 1451 257 GR53 E.R.GRAVES+L.ROSEN(LAS),PHYS.REV.89,343 (1953). 1325 1451 258 GR55 E.R.GRAVES+R.W.DAVIS(LAS), PHYS.REV.97, 1205(1955). 1325 1451 259 GR58 J.A.GRUNDL ET AL.(LAS),PHYS.REV.109,425 (1958). 1325 1451 260 GR67 J.A.GRUNDL (LAS), NUCL.SCI.ENG.30,69 (1967). 1325 1451 261 HA69 R.HARDELL ET AL., NUCL.PHYS.A126,392 (1969). 1325 1451 262 HE50 R.L.HENKEL AND H.H.BARSCHALL, PHYS. REV. 80, 145 (1950) 1325 1451 263 HE53 R.L.HENKEL, PRIV. COMM. TO SIGMA CENTER (1953) 1325 1451 264 HI59 C.T.HIBDON, PHYS. REV. 114, 179 (1959) 1325 1451 265 HO69 B.HOLMQVIST+T.WIEDLING (AE), AE-366 (1969). 1325 1451 266 IM64 W.L.IMHOF (LOK),PRIVATE COMMUNICATION TO BNL SIGMA CENTER 1325 1451 267 (1964). 1325 1451 268 JO64 G.D.JOANOU+C.A.STEVENS (GA), GA-5884 (1964). 1325 1451 269 JU71 E.T.JURNEY, PRIVATE COMMUNICATION (1971) 1325 1451 270 KA72 J.L.KAMMERDIENER,THESIS, UNIV.OF CAL.AT DAVIS (1972). 1325 1451 271 KE59 B.D.KERN ET AL.(NRD), NUCL.PHYS.10,226 (1959). 1325 1451 272 KI70 W.E.KINNEY+F.G.PEREY (ORL),ORNL-4516,(1970). 1325 1451 273 LA57 A.LANGSDORF ET AL., PHYS. REV. 107, 1077 (1957) 1325 1451 274 LI66 H.LISKIEN AND A.PAULSEN, NUCLEONICS 8,315 (1966). 1325 1451 275 MA57 M.H.MACGREGOR ET AL.(LRL), PHYS.REV.108,726(1957). 1325 1451 276 MA60 G.S.MANI ET AL.(HAR),NUCL.PHYS.19,535 (1960). 1325 1451 277 MA65 S.C.MATHUR ET AL.(TNC),NUCL.PHYS.73,561 (1965). 1325 1451 278 MA68 G.N.MASLOV ET AL.(RUS),SOV.J.AT.ENERGY 24,704 (1968). 1325 1451 279 MA69 R.E.MAERKER AND F.J.MUCKENTHALER, ORNL-4382 (1969). 1325 1451 280 ME52 A.W.MERRISON AND E.R.WIBLIN, PROC.ROY.SOC.(L) 215,278(1952)1325 1451 281 MI68 A.MITTLER ET AL.(KTY), BULL.AM.PHYS.SOC.13,1420 (1968). 1325 1451 282 OR71 V.J.ORPHAN AND C.G.HOOT, GULF-RT-A10743, (1971). 1325 1451 283 PA65 A.PAULSEN+H.LISKIEN (GEL),J.NUCL.ENERGY 19,907 (1965). 1325 1451 284 PE64 J.L.PERKIN (ALD),NUCL.PHYS.60,561 (1964). 1325 1451 285 PE72 F.G.PEREY ET AL., ORNL-4823 (1972). 1325 1451 286 PH52 D.D.PHILLIPS ET AL.(LAS), PHYS.REV.88,600 (1952). 1325 1451 287 PR60 J.T.PRUDHOMME ET AL.(TNC),AFSWC-TR-60-30 (1960). 1325 1451 288 RA70 N.C.RASMUSSEN ET AL., GA-10248 (1970) 1325 1451 289 SC61 H.W.SCHMITT+J.HALPERIN(ORL),PHYS.REV.121,827 (1961). 1325 1451 290 SC70 R.B.SCHWARTZ, PRIVATE COMMUNICATION (1970) 1325 1451 291 ST59 C.ST.PIERRE ET AL.(MON), PHYS.REV.115,999 (1959). 1325 1451 292 ST62 T.P.STUART ET AL., PHYS. REV. 125, 276 (1962) 1325 1451 293 ST65 P.H.STELSON+R.L.ROBINSON (ORL), NUCL.PHYS.68,97 (1965). 1325 1451 294 TA55 H.L.TAYLOR ET AL.(RIC), PHYS.REV.100,174 (1955). 1325 1451 295 TA70 S.TANAKA ET AL., NUCL.DATA FOR REACTORS (IAEA,1970),P317. 1325 1451 296 TH63 D.B.THOMPSON (LAS),PHYS.REV.129,1649 (1963). 1325 1451 297 TO62 J.H.TOWLE AND W.B.GILBOY, NUCL. PHYS. 39, 300 (1962) 1325 1451 298 TO67 J.H.TOWLE+R.O.OWENS(ALD),NUCL.PHYS.A100,257 (1967). 1325 1451 299 TS61 K.TSUKADA ET AL., PHYSICS OF FAST AND INTERMEDIATE 1325 1451 300 REACTORS, VIENNA (1961). 1325 1451 301 WI63 D.WINTERHALTER (ZAG), NUCL.PHYS.43,339(1963). 1325 1451 302 YO72 P.G.YOUNG, LASL MEMO T-2-99 TO THE NORM. AND STANDARDS 1325 1451 303 SUBCOMMITTEE OF CSEWG, (1972). 1325 1451 304 1 451 309 01325 1451 305 3 103 41 01325 1451 306 3 107 31 01325 1451 307 33 103 18 01325 1451 308 33 107 15 01325 1451 309 1325 1 0 310 1325 0 0 311 1.302700+4 2.675000+1 0 0 0 01325 3103 312 -1.827800+6-1.827800+6 0 0 3 1141325 3103 313 3 2 23 4 114 21325 3103 314 1.896100+6 0.000000+0 2.000000+6 9.09000-15 2.050000+6 2.18040-121325 3103 315 2.100000+6 5.23000-10 2.150000+6 9.748700-9 2.200000+6 6.600000-81325 3103 316 2.250000+6 3.302100-7 2.300000+6 1.150000-6 2.350000+6 3.320200-61325 3103 317 2.400000+6 8.000000-6 2.450000+6 1.708900-5 2.500000+6 3.280000-51325 3103 318 2.550000+6 5.835200-5 2.600000+6 9.690000-5 2.650000+6 1.526200-41325 3103 319 2.700000+6 2.290000-4 2.750000+6 3.299800-4 2.800000+6 4.590000-41325 3103 320 2.850000+6 6.208200-4 2.900000+6 8.180000-4 2.950000+6 1.053800-31325 3103 321 3.000000+6 1.330000-3 3.010000+6 1.390700-3 3.100000+6 2.010000-31325 3103 322 3.200000+6 2.460000-3 3.300000+6 2.930000-3 3.330000+6 3.178200-31325 3103 323 3.360000+6 3.625400-3 3.370000+6 3.860300-3 3.380000+6 4.178000-31325 3103 324 3.400000+6 5.100000-3 3.420000+6 6.336000-3 3.450000+6 8.900000-31325 3103 325 3.470000+6 9.864000-3 3.490000+6 1.015600-2 3.500000+6 1.010000-21325 3103 326 3.520000+6 9.780000-3 3.530000+6 9.280000-3 3.580000+6 5.340000-31325 3103 327 3.590000+6 4.830000-3 3.600000+6 4.600000-3 3.610000+6 4.978000-31325 3103 328 3.620000+6 5.652000-3 3.650000+6 8.700000-3 3.660000+6 9.062000-31325 3103 329 3.680000+6 9.212000-3 3.700000+6 9.100000-3 3.800000+6 7.700000-31325 3103 330 3.880000+6 7.076000-3 3.900000+6 7.000000-3 3.990000+6 6.942300-31325 3103 331 4.000000+6 6.966700-3 4.100000+6 7.500000-3 4.150000+6 7.952100-31325 3103 332 4.200000+6 8.600000-3 4.250000+6 9.543700-3 4.300000+6 1.070000-21325 3103 333 4.400000+6 1.350000-2 4.500000+6 1.700000-2 4.550000+6 1.900000-21325 3103 334 4.600000+6 2.155600-2 4.650000+6 2.174200-2 4.700000+6 2.000000-21325 3103 335 4.750000+6 1.750000-2 4.800000+6 1.650000-2 4.850000+6 1.677500-21325 3103 336 4.900000+6 1.800000-2 5.000000+6 2.330000-2 5.100000+6 2.634000-21325 3103 337 5.200000+6 2.833600-2 5.400000+6 3.316400-2 5.600000+6 3.770400-21325 3103 338 5.800000+6 4.207600-2 6.000000+6 4.640000-2 6.200000+6 5.061600-21325 3103 339 6.400000+6 5.486400-2 6.600000+6 5.928800-2 6.800000+6 6.367200-21325 3103 340 7.000000+6 6.780000-2 7.200000+6 7.155200-2 7.400000+6 7.520800-21325 3103 341 7.600000+6 7.880000-2 7.800000+6 8.228000-2 8.000000+6 8.560000-21325 3103 342 8.200000+6 8.897600-2 8.400000+6 9.190400-2 8.600000+6 9.414400-21325 3103 343 8.800000+6 9.605600-2 9.000000+6 9.800000-2 9.200000+6 1.004800-11325 3103 344 9.400000+6 1.023200-1 9.600000+6 1.034000-1 9.800000+6 1.039000-11325 3103 345 1.000000+7 1.040000-1 1.050000+7 1.025200-1 1.100000+7 1.000000-11325 3103 346 1.150000+7 9.601500-2 1.200000+7 9.220000-2 1.250000+7 8.823700-21325 3103 347 1.300000+7 8.500000-2 1.350000+7 8.157400-2 1.400000+7 7.770000-21325 3103 348 1.450000+7 7.397400-2 1.500000+7 7.000000-2 1.550000+7 6.640400-21325 3103 349 1.600000+7 6.270000-2 1.650000+7 5.880700-2 1.700000+7 5.520000-21325 3103 350 1.750000+7 5.168100-2 1.800000+7 4.780000-2 1.850000+7 4.400000-21325 3103 351 1.900000+7 4.020000-2 1.950000+7 3.638500-2 2.000000+7 3.220000-21325 3103 352 1325 3 0 353 1.302700+4 2.675000+1 0 0 0 01325 3107 354 -3.131600+6-3.131600+6 0 0 3 841325 3107 355 3 2 39 4 84 21325 3107 356 3.248700+6 0.000000+0 3.600000+6 3.16000-22 3.650000+6 1.01400-201325 3107 357 3.700000+6 3.25360-19 3.800000+6 3.35000-16 3.850000+6 3.56170-151325 3107 358 3.900000+6 2.88380-14 3.950000+6 1.77820-13 4.000000+6 8.35000-131325 3107 359 4.050000+6 3.67430-12 4.100000+6 1.42220-11 4.150000+6 4.84190-111325 3107 360 4.200000+6 1.45000-10 4.250000+6 4.09650-10 4.300000+6 1.075800-91325 3107 361 4.350000+6 2.626300-9 4.400000+6 5.960000-9 4.450000+6 1.293100-81325 3107 362 4.500000+6 2.678600-8 4.550000+6 5.296900-8 4.600000+6 1.000000-71325 3107 363 4.650000+6 1.832000-7 4.700000+6 3.252000-7 4.750000+6 5.593000-71325 3107 364 4.800000+6 9.320000-7 4.850000+6 1.516200-6 4.900000+6 2.410700-61325 3107 365 4.950000+6 3.746200-6 5.000000+6 5.690000-6 5.100000+6 1.246400-51325 3107 366 5.200000+6 2.550000-5 5.300000+6 4.924200-5 5.400000+6 9.020000-51325 3107 367 5.500000+6 1.580300-4 5.600000+6 2.650000-4 5.700000+6 4.270000-41325 3107 368 5.800000+6 6.670000-4 5.900000+6 1.010000-3 6.000000+6 1.490000-31325 3107 369 6.100000+6 2.140000-3 6.200000+6 2.990000-3 6.300000+6 4.000000-31325 3107 370 6.400000+6 5.140000-3 6.500000+6 6.540000-3 6.600000+6 8.200000-31325 3107 371 6.700000+6 9.870000-3 6.800000+6 1.180000-2 6.900000+6 1.370000-21325 3107 372 7.000000+6 1.590000-2 7.200000+6 2.010000-2 7.400000+6 2.500000-21325 3107 373 7.600000+6 3.000000-2 7.800000+6 3.560000-2 8.000000+6 4.130000-21325 3107 374 8.200000+6 4.710000-2 8.400000+6 5.330000-2 8.600000+6 5.920000-21325 3107 375 8.800000+6 6.490000-2 9.000000+6 7.020000-2 9.200000+6 7.510000-21325 3107 376 9.400000+6 7.970000-2 9.600000+6 8.380000-2 9.800000+6 8.770000-21325 3107 377 1.000000+7 9.120000-2 1.050000+7 9.961200-2 1.100000+7 1.075000-11325 3107 378 1.150000+7 1.149300-1 1.200000+7 1.212000-1 1.250000+7 1.260400-11325 3107 379 1.300000+7 1.288000-1 1.350000+7 1.279400-1 1.400000+7 1.247000-11325 3107 380 1.450000+7 1.193600-1 1.500000+7 1.129000-1 1.550000+7 1.052100-11325 3107 381 1.600000+7 9.700000-2 1.650000+7 8.840600-2 1.700000+7 7.940000-21325 3107 382 1.750000+7 7.118700-2 1.800000+7 6.370000-2 1.850000+7 5.653700-21325 3107 383 1.900000+7 4.980000-2 1.950000+7 4.343700-2 2.000000+7 3.800000-21325 3107 384 1325 3 0 385 1325 0 0 386 1.302700+4 2.675000+1 0 0 0 1132533103 387 0.000000+0 0.000000+0 0 103 0 3132533103 388 0.000000+0 0.000000+0 0 1 46 23132533103 389 1.000000-5 0.000000+0 1.899000+6 6.250000-2 2.000000+6 4.000000-2132533103 390 2.200000+6 2.900000-2 2.400000+6 2.250000-2 2.600000+6 1.700000-2132533103 391 2.800000+6 1.650000-2 3.000000+6 1.400000-2 3.300000+6 1.200000-2132533103 392 3.700000+6 1.000000-2 4.100000+6 8.000000-3 4.500000+6 6.500000-3132533103 393 5.000000+6 5.000000-3 5.500000+6 4.500000-3 6.000000+6 3.500000-3132533103 394 7.000000+6 6.500000-3 8.000000+6 8.000000-3 9.000000+6 6.500000-3132533103 395 1.100000+7 5.000000-3 1.300000+7 4.500000-3 1.500000+7 6.000000-3132533103 396 1.700000+7 1.000000-2 2.000000+7 0.000000+0 0.000000+0 0.000000+0132533103 397 0.000000+0 0.000000+0 0 2 16 8132533103 398 1.000000-5 0.000000+0 1.899000+6 3.000000-1 2.400000+6 2.000000-1132533103 399 2.800000+6 1.500000-1 3.300000+6 1.000000-1 4.500000+6 8.000000-2132533103 400 5.500000+6 0.000000+0 2.000000+7 0.000000+0 0.000000+0 0.000000+0132533103 401 0.000000+0 0.000000+0 0 1 12 6132533103 402 1.000000-5 0.000000+0 5.500000+6 5.000000-3 8.000000+6 1.450000-2132533103 403 1.300000+7 6.500000-3 1.700000+7 1.000000-2 2.000000+7 0.000000+0132533103 404 132533 0 405 1.302700+4 2.675000+1 0 0 0 1132533107 406 0.000000+0 0.000000+0 0 107 0 3132533107 407 0.000000+0 0.000000+0 0 1 30 15132533107 408 1.000000-5 0.000000+0 3.248000+6 9.000000-2 3.500000+6 4.000000-2132533107 409 4.000000+6 2.000000-2 4.500000+6 8.000000-3 5.000000+6 6.500000-3132533107 410 5.500000+6 3.500000-3 6.000000+6 2.500000-3 7.000000+6 1.600000-3132533107 411 8.000000+6 9.000000-4 9.000000+6 9.000000-4 1.000000+7 6.000000-4132533107 412 1.200000+7 9.000000-4 1.500000+7 1.200000-3 2.000000+7 0.000000+0132533107 413 0.000000+0 0.000000+0 0 2 16 8132533107 414 1.000000-5 0.000000+0 3.248000+6 4.000000-1 4.000000+6 2.000000-1132533107 415 5.000000+6 8.000000-2 6.000000+6 6.000000-2 9.000000+6 5.000000-2132533107 416 1.200000+7 0.000000+0 2.000000+7 0.000000+0 0.000000+0 0.000000+0132533107 417 0.000000+0 0.000000+0 0 1 8 4132533107 418 1.000000-5 0.000000+0 1.200000+7 1.600000-3 1.500000+7 1.300000-3132533107 419 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533107 420 132533 0 421 1325 0 0 422 0 0 0 423 2.104500+4 4.456790+1 1 0 0 02125 1451 1 0.000000+0 0.000000+0 0 0 0 62125 1451 2 1.000000+0 0.000000+0 0 0 10 62125 1451 3 0.000000+0 0.000000+0 0 0 80 42125 1451 4 21-SC- 45 BNL EVAL-JUL79 MAGURNO AND MUGHABGHAB 2125 1451 5 DIST-JAN90 900123 2125 1451 6 ----ENDF/B-VI MATERIAL 2125 2125 1451 7 -----INCIDENT NEUTRON DATA 2125 1451 8 ------ENDF-6 FORMAT 2125 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2125 1451 10 DOSIMETRY 2125 1451 11 ******************************************************************2125 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2125 1451 13 * * * * * * *2125 1451 14 P.C. TO NORMALIZATION AND STANDARDS SUBCOMMITTEE APRIL 1979 2125 1451 15 SC-45(N,GAMMA). EVALUATED BY B.A.MAGURNO AND S.F.MUGHABGHAB 2125 1451 16 THE POSITIVE ENERGY RESONANCE PARAMETERS LISTED BELOW 2125 1451 17 ARE THOSE EVALUATED FOR BNL-325 4-TH ED.(1)THE PARAMETER 2125 1451 18 SOURCES ARE REF.2,3,4. 2125 1451 19 THE BOUND LEVEL PARAMETERS WERE CHOSEN TO REPRODUCE A 2125 1451 20 CAPTURE CROSS SECTION OF 27.2B (26.4B CONTRIBUTED FROM BOUND 2125 1451 21 LEVEL,0.8B CONTRIBUTED FROM POSITIVE LEVELS).THE BOUND LEVEL 2125 1451 22 SPIN(J=3) CHOICE WAS BASED ON KOPECKY (5) POLARIZATION 2125 1451 23 EXPERIMENT WHICH SHOWED THAT A J=4 LEVEL CONTRIBUTION WOULD BE 2125 1451 24 LESS THAN 8PC. THIS IS IN AGREEMENT WITH RESULTS PREDICTED BY 2125 1451 25 MARSHAK(6),LIOU(2),BOLOTIN(7)AND DELANG(8). 2125 1451 26 IN THE FITTING PROCEDURE CARE WAS TAKEN TO REPRODUCE 2125 1451 27 THE ELASTIC SCATTERING CROSS SECTION RECOMMENDED IN REF1(22.1B)2125 1451 28 AND THE 2 KEV"WINDOW" TOTAL CROSS SECTION IN REF 2(0.71B). 2125 1451 29 KENNY (4) TABULATES THE AVERAGE CAPTURE CROSS 2125 1451 30 SECTION FROM 2 TO 10 KEV IN ONE KEV BINS AND FROM 10 KEV TO 90 2125 1451 31 KEV IN 10 KEV BINS.THE SLOPE OF THE AVERAGE CROSS SECTIONS FROM2125 1451 32 40 TO 90 KEV WAS USED TO EXTROPOLATE THE CROSS SECTION TO ONE 2125 1451 33 MEV. 2125 1451 34 THE CAPTURE CROSS SECTION FROM 1 MEV TO 20 MEV IS A 2125 1451 35 LINE FIT TO PERKIN DATA REFERENCE 12 AND A WEIGHTED AVERAGE OF 2125 1451 36 OF REF9,10,11,12 AT 14.6 MEV. 2125 1451 37 OTHER DATA CONSIDERED IN THE INITIAL EVALUATION BUT FOR2125 1451 38 THE MOST PART IGNORED IN THE FINAL ANAYSIS ARE IN REF 13,14. 2125 1451 39 2125 1451 40 -REFERENCES- 2125 1451 41 2125 1451 42 1.BNL-325 4TH ED. S.F.MUGHABGHAB ET.AL. TO BE PUBLISHED 2125 1451 43 2.LIOU ET.AL. NUC.SCI.AND ENG.67,326,1978 2125 1451 44 3.CHO ET.AL 2ND INT.CONF. NUC. DATA FOR REACT.CONF-700606,619,'702125 1451 45 4.KENNY ET.AL. AUS.J.OF PHYSICS 30,605,197 ) 2125 1451 46 5.KOPECKY ET.AL. TO BE PUBLISHED 2125 1451 47 6.MARSHAK H. BULL.AM.PHYS.SOC.22,993,1977 2125 1451 48 7.BOLOTIN H.H. PHYS.REV.168,1317,1968 2125 1451 49 8.DELANG ET.AL. CONF ON NUC STUDY WITH NEUTRON 2125 1451 50 9.WAGNER M. PRIV.COMM.NOV.1976 2125 1451 51 10.BUDNAR ET.AL. INDC(YUG)-5,1977 2125 1451 52 11.CSIKAI ET.AL. NUC.PHYS./A,95,229,1967 2125 1451 53 12.PERKIN J.L. J.NUC.EN.17,349,1963 2125 1451 54 13.ROMANOV ET.AL. YAD.FIZ.1,229,1965 2125 1451 55 14.BOOTH ET.AL. PHYS.REV.112,226,1958 2125 1451 56 ***** 2125 1451 57 SC-45(N,2N) THIS REACTION WAS EVALUATED FOR USE IN THE SPEC- 2125 1451 58 IAL PURPOSE FILE IN ENDF/B V FOR USE WITH THE INTERNATIONAL 2125 1451 59 REACTOR DOSIMETRY FILE. 2125 1451 60 THE SC(N,2N) WAS CALCULATED WITH THRESH AND ADJUSTED TO FIT 2125 1451 61 EXPERIMENTAL DATA. THE ISOMERIC RATIO ADOPTED WAS 0.71 AT 2125 1451 62 14.8 MEV. 2125 1451 63 ***** 2125 1451 64 REFERENCES 2125 1451 65 MUKHERJEE ET AL PROC. PHYS. COC. 77,508,61 2125 1451 66 PERKIN PRIVATE COMMUNICATION 1961 2125 1451 67 PRESTWOOD & BAYHURST PHYS. REV. 121,1438,61 2125 1451 68 RAYBURN PHYS. REV. 122,168,61 2125 1451 69 ARNOLD & RAYBURN BAPS 9, 352,64 2125 1451 70 TIWARI & KONDAIAH, NUC. & SOLID STATE SYM. BOMBAY FEB 66 2125 1451 71 VALLIS AWRE-0-76/77 2125 1451 72 NETHAWAY NUC. PHYS A1 90,635,72 2125 1451 73 BAYHURST ET AL PHYS. REV. C 12,451,75 2125 1451 74 EAPEN & SALAITA J. INORG. NUC. CHEM. 37,1121,75 2125 1451 75 KAO & ALFORD NUC. PHYS. A 237,11,75 2125 1451 76 PRASAD & SARKAR NUC. PHYS. A 94,476,67 2125 1451 77 CSIKAI & PETO ACT. PHYS. ACAD. SCI. HUNGARY 23,87,67 2125 1451 78 CSIKAI & PETO NUC. PHYS. A 122,234,68 2125 1451 79 WEIGEL ET AL RADIOCHEM. ACTA 22,11,75 2125 1451 80 CEVOLANI & PETRALIA NUOVO CIM. 26,1328,62 2125 1451 81 KHURANA & HANS NUC. PHYS. 28,560,61 2125 1451 82 BRAMLITT & FINK PHYS. REV. 131,2649,63 2125 1451 83 THRESH PEARLSTEIN J. NUC. ENERGY 27,81,73 2125 1451 84 1 451 88 02125 1451 85 2 151 197 02125 1451 86 3 102 11 02125 1451 87 33 102 11 02125 1451 88 2125 1 0 89 2125 0 0 90 2.104500+4 4.456790+1 0 0 1 02125 2151 91 2.104500+4 1.000000+0 0 0 1 02125 2151 92 1.000000-5 9.694379+4 1 2 0 02125 2151 93 3.500000+0 4.640000-1 0 0 2 02125 2151 94 4.456790+1 0.000000+0 0 0 474 792125 2151 95 -5.680000+2 3.000000+0 1.199090+2 1.190000+2 9.090000-1 0.000000+02125 2151 96 3.295000+3 3.000000+0 7.842400+1 7.771400+1 7.100000-1 0.000000+02125 2151 97 4.326000+3 4.000000+0 3.135200+2 3.128900+2 6.300000-1 0.000000+02125 2151 98 6.692000+3 3.000000+0 1.378800+2 1.371400+2 7.400000-1 0.000000+02125 2151 99 8.000000+3 4.000000+0 1.427900+2 1.422200+2 5.700000-1 0.000000+02125 2151 100 9.070000+3 4.000000+0 3.640600+1 3.555600+1 8.500000-1 0.000000+02125 2151 101 9.080000+3 3.000000+0 2.993900+2 2.982900+2 1.100000+0 0.000000+02125 2151 102 1.066200+4 3.000000+0 1.080000+1 1.000000+1 8.000000-1 0.000000+02125 2151 103 1.074000+4 4.000000+0 6.520000+0 6.000000+0 5.200000-1 0.000000+02125 2151 104 1.158000+4 4.000000+0 2.905400+2 2.897800+2 7.600000-1 0.000000+02125 2151 105 1.450000+4 3.000000+0 2.056000+1 2.000000+1 5.600000-1 0.000000+02125 2151 106 1.482000+4 4.000000+0 2.623800+1 2.577800+1 4.600000-1 0.000000+02125 2151 107 1.562300+4 4.000000+0 2.860000+1 2.800000+1 6.000000-1 0.000000+02125 2151 108 1.858000+4 3.000000+0 3.281100+1 3.200000+1 8.106100-1 0.000000+02125 2151 109 1.908000+4 4.000000+0 6.069000+1 6.000000+1 6.900000-1 0.000000+02125 2151 110 2.064400+4 4.000000+0 6.749700+1 6.666700+1 8.300000-1 0.000000+02125 2151 111 2.093000+4 3.000000+0 7.206000+2 7.200000+2 6.000000-1 0.000000+02125 2151 112 2.401000+4 3.000000+0 6.108700+1 6.000000+1 1.087000+0 0.000000+02125 2151 113 2.431500+4 4.000000+0 6.046200+1 6.000000+1 4.620000-1 0.000000+02125 2151 114 2.692500+4 4.000000+0 9.094400+1 9.004400+1 9.000000-1 0.000000+02125 2151 115 2.790000+4 3.000000+0 1.108700+2 1.100600+2 8.110000-1 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2.285900+1 2.181800+1 1.040800+0 0.000000+02125 2151 268 8.610500+4 3.000000+0 1.189500+1 1.142900+1 4.660000-1 0.000000+02125 2151 269 8.667000+4 4.000000+0 5.106700+1 1.066700+1 4.040000+1 0.000000+02125 2151 270 9.024000+4 5.000000+0 2.967900+1 2.909100+1 5.880000-1 0.000000+02125 2151 271 9.057500+4 5.000000+0 8.512700+0 7.272700+0 1.240000+0 0.000000+02125 2151 272 9.095000+4 4.000000+0 4.521400+1 4.444400+1 7.700000-1 0.000000+02125 2151 273 9.160000+4 3.000000+0 5.790300+1 5.714300+1 7.600000-1 0.000000+02125 2151 274 9.225000+4 2.000000+0 1.108000+0 6.720000-2 1.040800+0 0.000000+02125 2151 275 9.272000+4 5.000000+0 2.939100+1 2.909100+1 3.000000-1 0.000000+02125 2151 276 9.339000+4 4.000000+0 2.727100+1 2.666700+1 6.040000-1 0.000000+02125 2151 277 9.378000+4 3.000000+0 3.561600+1 3.428600+1 1.330000+0 0.000000+02125 2151 278 9.402000+4 5.000000+0 2.321800+1 2.181800+1 1.400000+0 0.000000+02125 2151 279 9.445000+4 2.000000+0 1.957000+1 1.920000+1 3.700000-1 0.000000+02125 2151 280 9.525000+4 4.000000+0 2.685400+1 2.666700+1 1.870000-1 0.000000+02125 2151 281 9.644000+4 5.000000+0 1.124600+1 1.090900+1 3.370000-1 0.000000+02125 2151 282 9.690000+4 3.000000+0 4.625400+1 4.571400+1 5.400000-1 0.000000+02125 2151 283 9.826000+4 4.000000+0 2.725300+1 2.666700+1 5.860000-1 0.000000+02125 2151 284 9.858500+4 5.000000+0 2.985500+1 2.909100+1 7.640000-1 0.000000+02125 2151 285 9.908000+4 2.000000+0 6.483000+1 6.400000+1 8.300000-1 0.000000+02125 2151 286 9.920000+4 3.000000+0 5.780500+1 5.714300+1 6.620000-1 0.000000+02125 2151 287 2125 2 0 288 2125 0 0 289 2.104500+4 4.456790+1 0 0 0 02125 3102 290 8.760000+6 8.760000+6 0 0 2 242125 3102 291 3 2 24 5 0 02125 3102 292 1.000000-5 0.000000+0 9.694379+4 0.000000+0 9.694379+4 3.000000-22125 3102 293 1.500000+5 2.343000-2 3.000000+5 1.200000-2 3.600000+5 1.020000-22125 3102 294 4.000000+5 9.600000-3 4.800000+5 8.600000-3 5.200000+5 8.390000-32125 3102 295 6.000000+5 8.100000-3 7.000000+5 8.000000-3 8.000000+5 7.700000-32125 3102 296 9.500000+5 7.330000-3 1.100000+6 6.500000-3 1.300000+6 5.000000-32125 3102 297 1.500000+6 4.000000-3 1.700000+6 2.800000-3 1.900000+6 2.300000-32125 3102 298 2.050000+6 2.020000-3 2.500000+6 1.600000-3 3.500000+6 1.080000-32125 3102 299 4.500000+6 9.300000-4 1.460000+7 7.520000-4 2.000000+7 7.200000-42125 3102 300 2125 3 0 301 2125 0 0 302 2.104500+4 4.456790+1 0 0 0 1212533102 303 0.000000+0 0.000000+0 0 102 0 2212533102 304 0.000000+0 0.000000+0 0 1 8 4212533102 305 1.000000-5 1.500000-5 2.530000-2 1.000000-4 1.000000+6 1.400000-3212533102 306 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0212533102 307 0.000000+0 0.000000+0 0 1 28 14212533102 308 1.000000-5 3.000000-5 8.000000+3 1.400000-3 2.000000+4 3.400000-3212533102 309 5.000000+4 5.000000-3 1.000000+5 1.200000-3 1.500000+5 7.000000-4212533102 310 3.000000+5 2.100000-3 5.000000+5 1.800000-3 6.000000+5 3.800000-3212533102 311 8.000000+5 6.100000-3 1.000000+6 6.500000-3 6.000000+6 1.000000-2212533102 312 1.800000+7 4.000000-2 2.000000+7 0.000000+0 0.000000+0 0.000000+0212533102 313 212533 0 314 2125 0 0 315 0 0 0 316 2.204600+4 4.555780+1 -1 0 0 02225 1451 1 0.000000+0 0.000000+0 0 0 0 62225 1451 2 1.000000+0 0.000000+0 0 0 10 62225 1451 3 0.000000+0 0.000000+0 0 0 30 32225 1451 4 22-TI- 46 BRC,ANL EVAL-JAN77 C.PHILIS,O.BERSILLON,D.SMITH,+ 2225 1451 5 DIST-JAN90 900124 2225 1451 6 ----ENDF/B-VI MATERIAL 2225 2225 1451 7 -----INCIDENT NEUTRON DATA 2225 1451 8 ------ENDF-6 FORMAT 2225 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2225 1451 10 DOSIMETRY 2225 1451 11 ******************************************************************2225 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2225 1451 13 * * * * * * *2225 1451 14 ** 2225 1451 15 DOCUMENTATION ANL/NDM-27 2225 1451 16 ** 2225 1451 17 COMMENTS-EVALUATION IN ENERGY RANGES 3.50E+06 TO 1.0E+07 AND 2225 1451 18 1.25E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA.INTERPOLATION AND2225 1451 19 EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCULATION2225 1451 20 PERFORMED AT BRUYERES LE CHATEL USING A CODE DEVELOPED BY JARRE.2225 1451 21 ** 2225 1451 22 REFERENCES 2225 1451 23 AL61 ALLAN D. NP24,27461 2225 1451 24 BO65 BORMAN M. EANDC-57,17,65 2225 1451 25 CR62 CROSS W. AND PAI H. EANDC(CAN)-16,2,62 2225 1451 26 CH71 GHORAI S. JNE 25,319,1971 2225 1451 27 KO64 KOEHLER D.AND ALFORD W.JNE18,81,64 2225 1451 28 LE69 LEVKOVSKII V. YF 10,44,69 2225 1451 29 LI65 LISKIEN H.AND PAULSEN A. NP63,393,65 2225 1451 30 LU71 LUKIC Y.AND CARROLL E. NSE 43,233,71 2225 1451 31 PA66 PAI H. CJP 44 2337,66 2225 1451 32 PO59 POULARIKAS A. FINK R. PR115,989,59 2225 1451 33 SM75 SMITH D.L.AND MEADOWS J. ANL/NDM10,75 2225 1451 34 1 451 37 02225 1451 35 3 103 11 02225 1451 36 33 103 7 02225 1451 37 2225 1 0 38 2225 0 0 39 2.204600+4 4.555780+1 0 0 0 02225 3103 40 -1.584900+6-1.584900+6 0 0 1 242225 3103 41 24 2 0 0 0 02225 3103 42 1.619400+6 0.000000+0 2.400000+6 4.000000-7 2.750000+6 1.200000-52225 3103 43 3.000000+6 1.000000-3 3.250000+6 3.300000-3 3.500000+6 8.000000-32225 3103 44 4.000000+6 2.600000-2 4.500000+6 4.800000-2 5.000000+6 7.300000-22225 3103 45 5.500000+6 9.900000-2 6.000000+6 1.240000-1 6.500000+6 1.500000-12225 3103 46 7.000000+6 1.730000-1 8.000000+6 2.080000-1 9.000000+6 2.320000-12225 3103 47 1.000000+7 2.520000-1 1.100000+7 2.640000-1 1.200000+7 2.700000-12225 3103 48 1.300000+7 2.680000-1 1.400000+7 2.600000-1 1.500000+7 2.450000-12225 3103 49 1.700000+7 2.100000-1 1.900000+7 1.720000-1 2.000000+7 1.520000-12225 3103 50 2225 3 0 51 2225 0 0 52 2.204600+4 4.555780+1 0 0 0 1222533103 53 0.000000+0 0.000000+0 0 103 0 2222533103 54 0.000000+0 0.000000+0 0 1 6 3222533103 55 1.000000-5 0.000000+0 1.619400+6 1.000000-2 2.000000+7 0.000000+0222533103 56 0.000000+0 0.000000+0 0 1 8 4222533103 57 1.000000-5 0.000000+0 1.619400+6 6.400000-3 1.000000+7 1.000000-2222533103 58 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533103 59 222533 0 60 2225 0 0 61 0 0 0 62 2.204700+4 4.654800+1 -1 0 0 02228 1451 1 0.000000+0 0.000000+0 0 0 0 62228 1451 2 1.000000+0 0.000000+0 0 0 10 62228 1451 3 0.000000+0 0.000000+0 0 0 38 52228 1451 4 22-TI- 47 BRC,ANL EVAL-JAN77 C.PHILIS,O.BERSILLON,D.SMITH+ 2228 1451 5 DIST-JAN90 900124 2228 1451 6 ----ENDF/B-VI MATERIAL 2228 2228 1451 7 -----INCIDENT NEUTRON DATA 2228 1451 8 ------ENDF-6 FORMAT 2228 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2228 1451 10 DOSIMETRY 2228 1451 11 ******************************************************************2228 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2228 1451 13 * * * * * * *2228 1451 14 ** 2228 1451 15 DOCUMENTATION ANL/NDM-27 (N,P) ANL/NDM-28(N,NP) 2228 1451 16 ** 2228 1451 17 COMMENTS- (N,P) EVAL IN ENERGY RANGES 1.10E+06 TO 1.0E+07 AND 2228 1451 18 1.35E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA.INTERPOLATION AND2228 1451 19 EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCULATION2228 1451 20 PERFORMED AT BRUYERES LE CHATEL USING CODE DEVELOPED BY JARRE. 2228 1451 21 PSEUDO-THRESHOLD=0.5 MEV PER P.C. MACFARLANE TO MAGURNO 6/20/83 2228 1451 22 ** 2228 1451 23 (N,NP) AND (N,D) CROSS SECTIONS EVAL BY C.Y.FU ORNL 2228 1451 24 (N,D)CROSS SECTIONS WERE TAKEN TO HAVE A SHAPE SIMILAR TO THAT 2228 1451 25 OF (N,NP),NORMALIZED TO 8+-3 MB AT 14.8 MEV BASED ON THE 2228 1451 26 MEASUREMENT OF HAIGHT AND GRIMES (UCRL-80235) FOR TI-46(9+-4MB) 2228 1451 27 AND TI-48(7+-3 MB).(N,NP)AND (N,D) ADDED FOR DOSIMETRY PURPOSES.2228 1451 28 REFERENCES 2228 1451 29 AL61 ALLAN D. NP24,274,61 2228 1451 30 AR78 ARTHUR E. PC 1978 2228 1451 31 AR75 ARMITAGE P.C. NNDC 2228 1451 32 CR62 CROSS W.AND PAI H. EANDC(CAN)-16,2,62 2228 1451 33 FU79 C.Y.FU PRIVATE COMM 1979 2228 1451 34 GH71 GHORAI S.AND CARROLL E. NSE 43,233,71 2228 1451 35 GO62 GONZALES ET.AL. PR126,271,62 2228 1451 36 HI HILLMAN M. NP37,78,62 2228 1451 37 LE69 LEVKOVSKII V. Y.F.10,44,69 2228 1451 38 PA66 PAI H.CJP 44 2337,66 2228 1451 39 PO59 POULARIKIS A. AND FINK R. PR115,989,59 2228 1451 40 SM75 SMITH D.L.AND MEADOWS J. ANL/NDM10,75 2228 1451 41 TI68 TIKKU ET.AL. SYMP ON NUCL CROSS SECT AND TECH WASH.1968 2228 1451 42 1 451 47 02228 1451 43 3 28 8 02228 1451 44 3 103 15 02228 1451 45 33 28 4 02228 1451 46 33 103 7 02228 1451 47 2228 1 0 48 2228 0 0 49 2.204700+4 4.654800+1 0 0 0 02228 3 28 50 -1.046000+7-1.046000+7 0 0 2 152228 3 28 51 2 2 15 4 0 02228 3 28 52 1.068000+7 0.000000+0 1.100000+7 1.500000-3 1.160000+7 2.800000-32228 3 28 53 1.200000+7 4.380000-3 1.220000+7 5.500000-3 1.260000+7 9.200000-32228 3 28 54 1.320000+7 2.020000-2 1.400000+7 4.720000-2 1.460000+7 7.300000-22228 3 28 55 1.520000+7 9.800000-2 1.600000+7 1.230000-1 1.700000+7 1.480000-12228 3 28 56 1.800000+7 1.580000-1 1.900000+7 1.540000-1 2.000000+7 1.410000-12228 3 28 57 2228 3 0 58 2.204700+4 4.654800+1 0 0 0 02228 3103 59 3.187100+5 3.187100+5 0 0 1 342228 3103 60 34 2 0 0 0 02228 3103 61 1.000000-5 0.000000+0 5.000000+5 0.000000+0 9.000000+5 3.600000-92228 3103 62 1.000000+6 1.900000-7 1.150000+6 1.250000-3 1.225000+6 2.800000-32228 3103 63 1.300000+6 3.800000-3 1.425000+6 3.100000-3 1.440000+6 3.300000-32228 3103 64 1.500000+6 3.900000-3 1.600000+6 4.500000-3 2.250000+6 2.300000-22228 3103 65 2.350000+6 3.050000-2 2.450000+6 3.170000-2 2.600000+6 3.030000-22228 3103 66 2.750000+6 3.060000-2 3.250000+6 5.000000-2 3.750000+6 6.400000-22228 3103 67 4.250000+6 7.400000-2 4.750000+6 8.000000-2 5.000000+6 8.000000-22228 3103 68 6.000000+6 1.000000-1 6.500000+6 1.060000-1 7.000000+6 1.040000-12228 3103 69 8.000000+6 1.190000-1 9.000000+6 1.290000-1 1.000000+7 1.370000-12228 3103 70 1.100000+7 1.400000-1 1.200000+7 1.360000-1 1.300000+7 1.300000-12228 3103 71 1.400000+7 1.200000-1 1.600000+7 9.800000-2 1.800000+7 7.850000-22228 3103 72 2.000000+7 7.000000-2 0.000000+0 0.000000+0 0.000000+0 0.000000+02228 3103 73 2228 3 0 74 2228 0 0 75 2.204700+4 4.654800+1 0 0 0 1222833 28 76 0.000000+0 0.000000+0 0 28 0 1222833 28 77 0.000000+0 0.000000+0 0 1 6 3222833 28 78 1.000000-5 0.000000+0 1.068000+7 9.000000-2 2.000000+7 0.000000+0222833 28 79 222833 0 80 2.204700+4 4.654800+1 0 0 0 1222833103 81 0.000000+0 0.000000+0 0 103 0 2222833103 82 0.000000+0 0.000000+0 0 1 6 3222833103 83 1.000000-5 0.000000+0 8.300000+5 6.400000-3 2.000000+7 0.000000+0222833103 84 0.000000+0 0.000000+0 0 1 8 4222833103 85 1.000000-5 0.000000+0 8.300000+5 6.400000-3 1.000000+7 1.000000-2222833103 86 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222833103 87 222833 0 88 2228 0 0 89 0 0 0 90 2.204800+4 4.753600+1 -1 0 0 02231 1451 1 0.000000+0 0.000000+0 0 0 0 62231 1451 2 1.000000+0 0.000000+0 0 0 10 62231 1451 3 0.000000+0 0.000000+0 0 0 39 52231 1451 4 22-TI- 48 BRC,ANL+ EVAL-JAN77 C.PHILIS,O.BERSILLON,D.SMITH+ 2231 1451 5 DIST-JAN90 900124 2231 1451 6 ----ENDF/B-VI MATERIAL 2231 2231 1451 7 -----INCIDENT NEUTRON DATA 2231 1451 8 ------ENDF-6 FORMAT 2231 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2231 1451 10 DOSIMETRY 2231 1451 11 ******************************************************************2231 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2231 1451 13 * * * * * * *2231 1451 14 ** 2231 1451 15 DOCUMENTATION ANL/NDM-27 (N,P) ANL/NDM-28 (N,NP) 2231 1451 16 ** 2231 1451 17 (N,P) EVAL IN ENERGY RANGES 4.70E+06 TO 1.0E+07 AND 2231 1451 18 1.25E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA. INTERPOLATION 2231 1451 19 AND EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCS 2231 1451 20 PERFORMED AT BRUYERES LE CHATEL USING CODE DEVELOPED BY J.JARRE 2231 1451 21 ** 2231 1451 22 (N,NP) AND (N,D) CROSS SECTIONS EVAL BY C.Y.FU ORNL 2231 1451 23 (N,D) CROSS SECTIONS WERE BASED ON THE MEASUREMENT OF HAIGHT 2231 1451 24 AND GRIMES(UCRL-80235) AT 14.8 MEV. THE SHAPE OF THE CROSS- 2231 1451 25 SECTION CURVE WAS ASSUMED TO BE THAT OF (N,NP).(N,D) AND (N,NP) 2231 1451 26 ADDED TOGETHER FOR DOSIMETRY PURPOSES. 2231 1451 27 REFERENCES 2231 1451 28 AL61 ALLAN D. NP24,274,61 2231 1451 29 BO65 BORMANN M.ET.AL.EANDC(E)57,17,65 2231 1451 30 CR62 CROSS W.AND PAI H. EANDC(CAN)-16 1962 2231 1451 31 FU79 C.Y.FU PRIVATE COMM 1979 2231 1451 32 GA62 GABBARD F.AND KERN B. PR128,1276,62 2231 1451 33 GH71 GHORAI S.ET.AL.JNE25,319,71 2231 1451 34 HI62 HILLMAN M. NP37,78,62 2231 1451 35 LE69 LEVKOVSKII YF10,44,69 2231 1451 36 LU71 LUKIC Y.AND CARROLL E. NSE43,233,71 2231 1451 37 MA75 MANNHART W.AND VONACH H. ZP,A272,279,75 2231 1451 38 PA66 PAI H. CJP44,2337,66 2231 1451 39 PO59 POULARIKAS A.AND FINK R. PR115,989,59 2231 1451 40 SM75 SMITH D.L.AND MEADOWS J. ANL/NDM-10,75 2231 1451 41 TI68 TIKKU L.ET.AL. SYMP.ON NUC CROSS SEC TECH WASH.1968 2231 1451 42 VO68 VONACH H.ET.AL. SYMP ON NUC CROSS SEC TECH WASH.1968 2231 1451 43 1 451 48 02231 1451 44 3 28 8 02231 1451 45 3 103 12 02231 1451 46 33 28 4 02231 1451 47 33 103 7 02231 1451 48 2231 1 0 49 2231 0 0 50 2.204800+4 4.753600+1 0 0 0 02231 3 28 51 -1.144600+7-1.144600+7 0 0 2 132231 3 28 52 2 2 13 4 0 02231 3 28 53 1.168600+7 0.000000+0 1.300000+7 3.000000-3 1.350000+7 4.000000-32231 3 28 54 1.360000+7 4.400000-3 1.400000+7 6.000000-3 1.460000+7 1.030000-22231 3 28 55 1.510000+7 1.500000-2 1.600000+7 2.480000-2 1.660000+7 3.200000-22231 3 28 56 1.760000+7 4.420000-2 1.800000+7 4.970000-2 1.860000+7 4.900000-22231 3 28 57 2.000000+7 4.690000-2 0.000000+0 0.000000+0 0.000000+0 0.000000+02231 3 28 58 2231 3 0 59 2.204800+4 4.753600+1 0 0 0 02231 3103 60 -3.208000+6-3.208000+6 0 0 1 262231 3103 61 26 2 0 0 0 02231 3103 62 3.275000+6 0.000000+0 4.000000+6 8.000000-8 4.200000+6 5.300000-72231 3103 63 4.400000+6 1.600000-5 4.700000+6 2.400000-5 4.900000+6 7.300000-52231 3103 64 5.100000+6 1.000000-4 5.400000+6 4.500000-4 5.700000+6 5.800000-42231 3103 65 6.000000+6 1.900000-3 6.500000+6 3.900000-3 7.000000+6 6.600000-32231 3103 66 7.500000+6 9.800000-3 8.000000+6 1.310000-2 9.000000+6 2.040000-22231 3103 67 1.000000+7 3.000000-2 1.100000+7 4.140000-2 1.200000+7 5.300000-22231 3103 68 1.300000+7 6.100000-2 1.350000+7 6.400000-2 1.400000+7 6.400000-22231 3103 69 1.500000+7 6.200000-2 1.600000+7 5.700000-2 1.800000+7 4.600000-22231 3103 70 1.900000+7 4.100000-2 2.000000+7 3.800000-2 0.000000+0 0.000000+02231 3103 71 2231 3 0 72 2231 0 0 73 2.204800+4 4.753600+1 0 0 0 1223133 28 74 0.000000+0 0.000000+0 0 28 0 1223133 28 75 0.000000+0 0.000000+0 0 1 6 3223133 28 76 1.000000-5 0.000000+0 1.168600+7 9.000000-2 2.000000+7 0.000000+0223133 28 77 223133 0 78 2.204800+4 4.753600+1 0 0 0 1223133103 79 0.000000+0 0.000000+0 0 103 0 2223133103 80 0.000000+0 0.000000+0 0 1 6 3223133103 81 1.000000-5 0.000000+0 3.275000+6 6.400000-3 2.000000+7 0.000000+0223133103 82 0.000000+0 0.000000+0 0 1 8 4223133103 83 1.000000-5 0.000000+0 3.275000+6 6.400000-3 1.000000+7 1.000000-2223133103 84 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133103 85 223133 0 86 2231 0 0 87 0 0 0 88 2.505500+4 5.446610+1 -1 0 0 12525 1451 1 0.000000+0 0.000000+0 0 0 0 62525 1451 2 1.000000+0 0.000000+0 0 0 10 62525 1451 3 0.000000+0 0.000000+0 0 0 213 32525 1451 4 25-MN- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA 2525 1451 5 DIST-APR90 900420 2525 1451 6 ----ENDF/B-VI MATERIAL 2525 2525 1451 7 -----INCIDENT NEUTRON DATA 2525 1451 8 ------ENDF-6 FORMAT 2525 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2525 1451 10 DOSIMETRY 2525 1451 11 ******************************************************************2525 1451 12 2525 1451 13 MF=1 GENERAL INFORMATION 2525 1451 14 MT=451 DESCRIPTIVE DATA AND DICTIONARY 2525 1451 15 2525 1451 16 MF=2 RESONANCE PARAMETERS 2525 1451 17 MT=151 RESOLVED RESONANCE PARAMETERS FOR MLBW FORMULA 2525 1451 18 THE PARAMETERS OF THE LOWEST FOUR RESONANCES WERE TAKEN 2525 1451 19 FROM THE WORK OF MACKLIN /1/. OTHERS WERE TAKEN FROM THE 2525 1451 20 COMPILATION OF MUGHABGHAB ET AL./2/ EXCEPT THAT THE 2525 1451 21 PARAMETERS OF TWO NEGATIVE RESONANCES WERE ADJUSTED SO AS 2525 1451 22 TO FIT TO EXPERIMENTAL THERMAL CROSS SECTIONS. 2525 1451 23 RESONANCE REGION : 1.0E-5 EV TO 100 KEV. 2525 1451 24 SCATTERING RADIUS: 5.15 FM 2525 1451 25 CALCULATED 2200-M/S CROSS SECTIONS AND RES. INTEGRALS 2525 1451 26 2200-M/S RES. INTEG. 2525 1451 27 ELASTIC 2.167 B - 2525 1451 28 CAPTURE 13.413 B 11.73 B 2525 1451 29 TOTAL 15.579 B - 2525 1451 30 2525 1451 31 MF=3 NEUTRON CROSS SECTIONS 2525 1451 32 MT=1 TOTAL 2525 1451 33 BELOW 100 KEV : NO BACKGROUND 2525 1451 34 ABOVE 100 KEV : BASED ON THE EXPERIMENTAL DATA /3,4,5/. 2525 1451 35 MT=2 ELASTIC SCATTERING 2525 1451 36 (TOTAL) - (NONELASTIC CROSS SECTION) 2525 1451 37 MT=3 NON ELASTIC 2525 1451 38 SUM OF MT=4, 16, 22, 28, 102, 103, 104, 105, 106 AND 107 2525 1451 39 MT=4,51-79,91 INELASTIC SCATTERING 2525 1451 40 STATISTICAL-MODEL CALCULATIONS WERE PERFORMED USING THE 2525 1451 41 TNG CODE /6/. THE PRECOMPOUND PROCESS WAS CONSIDERED 2525 1451 42 ABOVE 5 MEV. THE CALCULATED CROSS SECTION OF MT=51 2525 1451 43 WAS MULTIPLIED BY A FACTOR OF 1.2. 2525 1451 44 FOR THE LEVELS OF MT=51,52,57,61,64,65,67,70, 2525 1451 45 THE DIRECT PROCESS COMPONENTS WERE TAKEN INTO ACCOUNT 2525 1451 46 BY THE DWBA CALCULATIONS. 2525 1451 47 THE OPTICAL POTENTIAL PARAMETERS USED ARE AS FOLLOWS/7/ 2525 1451 48 (IN THE UNITS OF MEV AND FM): 2525 1451 49 V = 49.747 - 0.4295*E - 0.0003*E**2 R0 = 1.287 A0 = 0.56 2525 1451 50 WS = 11.2 - 0.09*E RS = 1.345 AS = 0.47 2525 1451 51 VSO= 6.2 RSO= 1.120 ASO = 0.47 2525 1451 52 THE LEVEL SCHEME WAS TAKEN FROM REF./8/. 2525 1451 53 NO. ENERGY(MEV) SPIN-PARITY 2525 1451 54 G.S. 0.0 5/2 - 2525 1451 55 1. 0.126 7/2 - 2525 1451 56 2. 0.984 9/2 - 2525 1451 57 3. 1.290 1/2 - 2525 1451 58 4. 1.292 11/2 - 2525 1451 59 5. 1.293 1/2 - 2525 1451 60 6. 1.528 3/2 - 2525 1451 61 7. 1.884 7/2 - 2525 1451 62 8. 2.015 7/2 - 2525 1451 63 9. 2.198 7/2 - 2525 1451 64 10. 2.215 5/2 - 2525 1451 65 11. 2.252 3/2 - 2525 1451 66 12. 2.267 5/2 - 2525 1451 67 13. 2.312 13/2 - 2525 1451 68 14. 2.366 5/2 - 2525 1451 69 15. 2.398 9/2 + 2525 1451 70 16. 2.427 1/2 + 2525 1451 71 17. 2.563 3/2 - 2525 1451 72 18. 2.727 7/2 - 2525 1451 73 19. 2.753 5/2 - 2525 1451 74 20. 2.822 9/2 - 2525 1451 75 21. 2.824 5/2 - 2525 1451 76 22. 2.873 1/2 - 2525 1451 77 23. 2.954 3/2 - 2525 1451 78 24. 2.976 3/2 - 2525 1451 79 25. 2.992 7/2 - 2525 1451 80 26. 3.006 3/2 - 2525 1451 81 27. 3.036 11/2 - 2525 1451 82 28. 3.038 1/2 - 2525 1451 83 29. 3.040 3/2 + 2525 1451 84 2525 1451 85 LEVELS ABOVE 3.046 MEV WERE ASSUMED TO BE OVERLAPPING. 2525 1451 86 2525 1451 87 MT=16,22,28,103,107 (N,2N),(N,N'A),(N,N'P),(N,P) AND (N,A) 2525 1451 88 CROSS SECTIONS 2525 1451 89 CALCULTED WITH TNG. 2525 1451 90 GLOBAL OPTICAL-POTENTIAL PARAMETERS WERE EMPLOYED 2525 1451 91 FOR PROTONS AND ALPHA-PARTICLES /9,10/. 2525 1451 92 2525 1451 93 MT=102 RADIATIVE CAPTURE CROSS SECTION 2525 1451 94 BELOW 100 KEV : RESONANCE PARAMETERS GIVEN (NO BACKGROUND)2525 1451 95 ABOVE 100 KEV : BASED ON THE EXPERIMENTAL DATA /11/-/15/. 2525 1451 96 (TAKEN FROM JENDL-2 /20/) 2525 1451 97 2525 1451 98 MT=104 (N,D) CROSS SECTION 2525 1451 99 THE EXCITATION FUNCTION OF THE (N,P) CROSS SECTION 2525 1451 100 CALCULATED WITH TNG WAS USED FOR THE (N,D) REACTION BY 2525 1451 101 SHIFTING THE THRESHOLD ENERGY. THE CROSS SECTIONS WERE 2525 1451 102 NORMALIZED TO THE EXPERIMENTAL DATUM AT 14.1 MEV /16/. 2525 1451 103 2525 1451 104 MT=105 (N,T) CROSS SECTION 2525 1451 105 THE EXCITATION FUNCTION OF THE (N,P) CROSS SECTION 2525 1451 106 CALCULATED WITH TNG WAS USED FOR THE (N,T) REACTION BY 2525 1451 107 SHIFTING THE THRESHOLD ENERGY. THE CROSS SECTIONS WERE 2525 1451 108 NORMALIZED TO THE EXPERIMENTAL DATUM AT 14.7 MEV /17/. 2525 1451 109 2525 1451 110 MT=106 (N,HE-3) CROSS SECTION 2525 1451 111 BASED ON THE EXPERMENTAL DATA /18,19/. 2525 1451 112 2525 1451 113 MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 2525 1451 114 MT=2 ELASTIC 2525 1451 115 OPTICAL AND STATISTICAL-MODEL CALCULATIONS 2525 1451 116 2525 1451 117 MF=6 ENERGY-ANGLE DISTRIBUTIONS 2525 1451 118 MT=16 (N,2N) 2525 1451 119 ENERGY DISTRIBUTIONS OF NEUTRONS, RECOILS AND GAMMA-RAYS 2525 1451 120 WERE CALCULATED BY TNG. ANGULAR DISTRIBUTIONS ARE GIVEN 2525 1451 121 ONLY FOR NEUTRONS. 2525 1451 122 2525 1451 123 MT=22 (N,N'A) 2525 1451 124 ENERGY DISTRIBUTIONS OF NEUTRONS, ALPHAS, RECOILS AND 2525 1451 125 GAMMA-RAYS WERE CALCULATED BY TNG. ANGULAR DISTRIBUTIONS 2525 1451 126 ARE GIVEN ONLY FOR NEUTRONS. 2525 1451 127 2525 1451 128 MT=28 (N,N'P) 2525 1451 129 ENERGY DISTRIBUTIONS OF NEUTRONS, PROTONS, RECOILS AND 2525 1451 130 GAMMA-RAYS WERE CALCULATED BY TNG. ANGULAR DISTRIBUTIONS 2525 1451 131 ARE GIVEN ONLY FOR NEUTRONS. 2525 1451 132 2525 1451 133 MT=51 TO 79 INELASTIC SCATTERING TO DISCRETE LEVELS 2525 1451 134 OPTICAL AND STATISTICAL-MODEL CALCULATIONS 2525 1451 135 THE COMPONENTS OF THE DIRECT-INTERACTION PROCESS WERE 2525 1451 136 ADDED TO THE LEVELS OF MT=51,52,57,61,64,65,67,70 BY THE 2525 1451 137 DWBA CALCULATIONS. 2525 1451 138 2525 1451 139 MT=91 INELASTIC SCATTERING TO CONTINUUM LEVELS 2525 1451 140 ENERGY DISTRIBUTIONS OF NEUTRONS, RECOILS AND GAMMA-RAYS 2525 1451 141 WERE CALCULATED BY TNG. ANGULAR DISTRIBUTIONS ARE GIVEN 2525 1451 142 ONLY FOR NEUTRONS. 2525 1451 143 2525 1451 144 MT=103 (N,P) 2525 1451 145 ENERGY DISTRIBUTIONS OF PROTONS, RECOILS AND GAMMA-RAYS 2525 1451 146 WERE CALCULATED BY TNG. 2525 1451 147 2525 1451 148 MT=107 (N,A) 2525 1451 149 ENERGY DISTRIBUTIONS OF ALPHAS, RECOILS AND GAMMA-RAYS 2525 1451 150 WERE CALCULATED BY TNG. 2525 1451 151 2525 1451 152 MF=12 PHOTON PRODUCTION MULTIPLICITIES 2525 1451 153 MT=51 TO 79 2525 1451 154 BRANCHING RATIOS FOR THE LEVELS ARE GIVEN. 2525 1451 155 MT=102 CAPTURE 2525 1451 156 CALCULATED WITH TNG. 2525 1451 157 2525 1451 158 MF=14 PHOTON ANGULAR DISTRIBUTIONS 2525 1451 159 MT=51 TO 79 2525 1451 160 ASSUMED TO BE ISOTROPIC. 2525 1451 161 MT=102 CAPTURE 2525 1451 162 ASSUMED TO BE ISOTROPIC. 2525 1451 163 2525 1451 164 MF=15 PHOTON ENERGY DISTRIBUTIONS 2525 1451 165 MT=102 CAPTURE 2525 1451 166 CALCULATED WITH TNG. 2525 1451 167 2525 1451 168 MF=33 COVARIANCE DATA 2525 1451 169 MT=1,2,3,4,16,22,28,51-79,91,102,103,104,105,106,107 2525 1451 170 ESTIMATED FROM EXPERIMENTAL DATA AND CALCULATION 2525 1451 171 2525 1451 172 REFERNCES 2525 1451 173 1) MACKLIN, R.L.: NUCL. SCI. ENG., 89, 362 (1985). 2525 1451 174 2) MUGABGHAB, S.F., DIVADEENAM, M. AND HOLDEN, N.E.: "NEUTRON 2525 1451 175 CROSS SECTIONS", VOL. 1, PART A, ACADEMIC PRESS (1981). 2525 1451 176 3) CIERJACKS, S., FORTI, P., KOPSCH, D., KROPP, L., NEBE, J. 2525 1451 177 AND UNSELD, H.: "HIGH RESOLUTION TOTAL CROSS SECTIONS 2525 1451 178 FOR NA, CL, K, V, MN AND CO BETWEEN 0.5 AND 30 MEV", 2525 1451 179 KFK-1000 (1968). 2525 1451 180 4) PINEO, W.F.E., DIVADEENAM, M., BILPUCH, E.G., SETH, K.K. 2525 1451 181 AND NEWSON, H.W.: ANN. PHYS., 84, 165 (1974). 2525 1451 182 5) GARG, J.B., RAINWATER, J. AND HAVENS, JR., W.W.: 2525 1451 183 NUCL. SCI. ENG., 65, 76 (1978). 2525 1451 184 6) FU, C.Y.: "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2525 1451 185 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2525 1451 186 MOMENTUM", ORNL/TM-7042 (1980). 2525 1451 187 7) FU, C.Y.: PRIVATE COMMUNICATION (1985). 2525 1451 188 8) ZHOU ENCHEN, HUO JUNDE, ZHOU CHUNMEI, LU XIANE AND 2525 1451 189 WANG LIZHENG: NUCL. DATA SHEETS, 44, 463 (1985). 2525 1451 190 9) PEREY, F.G.: PHYS. REV., 131, 745 (1963). 2525 1451 191 10) HUIZENGA, J.R. AND IGO, G.J.: NUCL. PHYS., 29, 462 (1962). 2525 1451 192 11) GARG, J.B., MACKLIN, R.L. AND HALPERIN, J.: PHYS. REV., 2525 1451 193 C18, 2079 (1978). 2525 1451 194 12) DOVBENKO, A.G., KOLESOV, V.E., KOROLEVA, V.P., TOLSTIKOV, 2525 1451 195 V.A.: ATOM. ENERG., 26, 67 (1969). 2525 1451 196 13) MENLOVE, H.O., COOP, K.L., GRENCH, H.A. AND SHER, R.: 2525 1451 197 PHYS. REV., 163, 1299 (1967). 2525 1451 198 14) SCHWERER, O., WINKLER-ROHATSCH, M., WARHANEK, H. AND 2525 1451 199 WINKLER, G.: NUCL. PHYS., A264, 105 (1976). 2525 1451 200 15) BUDNAR, M., CVELBAR, F., HODGSON, E., HUDOKLIN, A., 2525 1451 201 IVKOVIC, V., LIKAR, A., MIHAILOVIC, M.V., MARTINCIC, R., 2525 1451 202 NAJZER, M., PERDAN, A., POTOKAR, M. AND RAMSAK, V.: 2525 1451 203 "PROMPT GAMMA-RAY SPECTRA AND INTEGARTED CROSS SECTIONS 2525 1451 204 FOR THE RADIATIVE CAPTURE OF 14 MEV NEUTRONS FOR 28 2525 1451 205 NATURAL TARGETS IN THE MASS REGION FROM 12 TO 208", 2525 1451 206 INDC(YUG)-6 (1979). 2525 1451 207 16) COLLI, L., IORI, I., MICHELETTI, S. AND PIGNANELLI, M.: 2525 1451 208 NUOVO. CIM., 21, 966 (1962). 2525 1451 209 17) SUDAR, S. AND CSIKAI, J.: NUCL. PHYS., A319, 157 (1979). 2525 1451 210 18) DIKSIC, M., STROHAL, P. AND SLAUS, I.: J. INORG. NUCL. 2525 1451 211 CHEM., 36, 477 (1974). 2525 1451 212 19) WU, C.H., WOELFLE, R. AND QAIM, S.M.: NUCL. PHYS., A329, 2525 1451 213 63 (1979). 2525 1451 214 20) NAKAGAWA, T.: "SUMMARY OF JENDL-2 GENERAL PURPOSE 2525 1451 215 FILE", JAERI-M 84-103 (1984). 2525 1451 216 2525 1451 217 1 451 220 12525 1451 218 3 16 7 12525 1451 219 33 16 12 12525 1451 220 2525 1 0 221 2525 0 0 222 2.505500+4 5.446610+1 0 0 0 02525 3 16 223 -1.022700+7-1.022700+7 0 0 1 122525 3 16 224 12 2 0 0 0 02525 3 16 225 1.041480+7 0.000000+0 1.100000+7 6.629000-2 1.200000+7 3.244100-12525 3 16 226 1.300000+7 5.688000-1 1.400000+7 7.190000-1 1.450000+7 7.659500-12525 3 16 227 1.500000+7 7.977500-1 1.600000+7 8.293500-1 1.700000+7 8.349000-12525 3 16 228 1.800000+7 8.258000-1 1.900000+7 8.091000-1 2.000000+7 7.900500-12525 3 16 229 2525 3 0 230 2525 0 0 231 2.505500+4 5.446600+1 0 0 0 1252533 16 232 0.000000+0 0.000000+0 0 16 0 3252533 16 233 0.000000+0 0.000000+0 0 1 6 3252533 16 234 1.000000-5 0.000000+0 1.041500+7 2.500000-3 2.000000+7 0.000000+0252533 16 235 0.000000+0 0.000000+0 0 1 16 8252533 16 236 1.000000-5 0.000000+0 1.041500+7 8.100000-2 1.200000+7 8.100000-2252533 16 237 1.300000+7 7.290000-3 1.400000+7 4.050000-3 1.500000+7 3.240000-3252533 16 238 1.700000+7 3.240000-3 2.000000+7 0.000000+0 0.000000+0 0.000000+0252533 16 239 0.000000+0 0.000000+0 0 8 16 8252533 16 240 1.000000-5 0.000000+0 1.041500+7 3.954900-5 1.200000+7 9.471800-4252533 16 241 1.300000+7 2.620600-4 1.400000+7 2.326300-4 1.500000+7 2.291100-4252533 16 242 1.700000+7 2.509400-4 2.000000+7 0.000000+0 0.000000+0 0.000000+0252533 16 243 252533 0 244 2525 0 0 245 0 0 0 246 2.605400+4 5.347600+1 -1 0 0 12625 1451 1 0.000000+0 0.000000+0 0 0 0 62625 1451 2 1.000000+0 0.000000+0 0 0 10 62625 1451 3 0.000000+0 0.000000+0 0 0 178 32625 1451 4 26-FE- 54 ORNL EVAL-NOV89 D.M. HETRICK,C.Y. FU,N.M. LARSON 2625 1451 5 ORNL/TM- DIST-APR90 900326 2625 1451 6 ----ENDF/B-VI MATERIAL 2625 2625 1451 7 -----INCIDENT NEUTRON DATA 2625 1451 8 ------ENDF-6 FORMAT 2625 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2625 1451 10 DOSIMETRY 2625 1451 11 ******************************************************************2625 1451 12 THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE 2625 1451 13 DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1) 2625 1451 14 AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES 2625 1451 15 ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE 2625 1451 16 COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2625 1451 17 REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2625 1451 18 CONTENTS OF THIS EVALUATION WILL BE PUBLISHED AT A LATER DATE. 2625 1451 19 RESONANCE PARAMETERS AND THERMAL CROSS SECTIONS ARE FROM 2625 1451 20 REF. (5) AND ADJUSTED THROUGH SAMMY BY N. LARSON USING J.A. 2625 1451 21 HARVEY'S 54-FE DATA (J.A. HARVEY, PRIVATE COMMUNICATION, 1989). 2625 1451 22 BETWEEN 110 AND 700 KEV, CAPTURE CROSS SECTION BACKGROUND IN 2625 1451 23 FILE 3 WAS BASED ON REF. (6). ABOVE 700 KEV, 54-FE DATA FROM 2625 1451 24 J.A. HARVEY (PRIVATE COMMUNICATION, 1989) USED. 2625 1451 25 2625 1451 26 ----- DESCRIPTION OF FILES 2625 1451 27 (MF-MT) 2625 1451 28 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2625 1451 29 2-151 RESONANCE PARAMETERS; SEE ABOVE. 2625 1451 30 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2625 1451 31 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2625 1451 32 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2625 1451 33 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2625 1451 34 3-1 TOTAL CROSS SECTION; SEE ABOVE. 2625 1451 35 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2625 1451 36 SUBTRACTING THE NONELASTIC FROM THE TOTAL 2625 1451 37 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 2625 1451 38 3-28, 3-102, 3-103, 3-104, AND 3-107. 2625 1451 39 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, .. 2625 1451 40 .., 3-57, AND 3-91 2625 1451 41 3-16 (N,2N) CROSS SECTIONS WERE CALCULATED BY THE TNG 2625 1451 42 CODE (2,3,4). 2625 1451 43 3-22 (N,NA) CROSS SECTIONS WERE CALCULATED BY THE TNG 2625 1451 44 CODE (2,3,4). 2625 1451 45 3-28 (N,NP) CROSS SECTIONS WERE CALCULATED BY THE TNG 2625 1451 46 CODE (2,3,4). 2625 1451 47 3-51 TO 3-57 INELASTIC SCATTERING EXCITING LEVELS; FOR EN>6 MEV,2625 1451 48 RESULTS ARE FROM TNG (2,3,4) WHICH INCLUDES DIRECT 2625 1451 49 INTERACTION CROSS SECTIONS FROM DWUCK (1) CALCULATIONS. 2625 1451 50 FOR EN<6 MEV, TNG CALCULATIONS WERE ADJUSTED TO FIT THE 2625 1451 51 DATA OF GUENTHER ET AL. (7). 2625 1451 52 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM ARE TNG 2625 1451 53 CALCULATED, BUT ADJUSTED AS FOR 3-51 TO 3-57 FOR EN<6 MEV.2625 1451 54 3-102 (N,G) CAPTURE CROSS SECTION; SEE ABOVE. TNG CALCULATIONS 2625 1451 55 NORMALIZED TO THE DATA WERE ADDED FOR ENERGIES FROM 1.0 2625 1451 56 TO 1.5 MEV. 2625 1451 57 3-103 (N,P) CROSS SECTIONS WERE TAKEN FROM THE GLUCS (8) 2625 1451 58 CALCULATION IN WHICH THIS REACTION WAS STUDIED 2625 1451 59 SIMULTANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2625 1451 60 SECTIONS (9). 2625 1451 61 3-104 (N,D) CROSS SECTIONS WERE TAKEN FROM (N,P) SHAPE WHICH 2625 1451 62 WAS NORMALIZED TO DATA OF GRIMES ET AL. (10) AT 14.8 MEV. 2625 1451 63 3-107 (N,A) CROSS SECTIONS WERE TAKEN FROM ENDF/B-V FOR EN < 9 2625 1451 64 MEV. ABOVE 9 MEV, ENDF/B-V CROSS SECTIONS WERE ADJUSTED 2625 1451 65 TO BE SMALLER TO BETTER FIT MEASURED DATA. 2625 1451 66 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR 2625 1451 67 ELASTIC SCATTERING ARE TAKEN FROM ENDF/B-V. 2625 1451 68 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2625 1451 69 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2625 1451 70 PARAMETERS IN 2/151. 2625 1451 71 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2625 1451 72 NEUTRON AND 53FE RESIDUAL, AND ENERGY DEPENDENT YIELD 2625 1451 73 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2625 1451 74 RAY; TNG CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR2625 1451 75 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2625 1451 76 THE OUTGOING NEUTRON). 2625 1451 77 6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2625 1451 78 NEUTRON, ALPHA, AND 50CR RESIDUAL, AND ENERGY DEPENDENT 2625 1451 79 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2625 1451 80 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2625 1451 81 FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR2625 1451 82 THE OUTGOING NEUTRON; ISOTROPY IS ASSUMED FOR THE 2625 1451 83 ALPHA AND RESIDUAL). (N,XA) D-D EMISSION DATA HEAVILY 2625 1451 84 USED TO BENCHMARK THE CALCULATIONS. 2625 1451 85 6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2625 1451 86 NEUTRON, P, AND 53MN RESIDUAL, AND ENERGY DEPENDENT YIELD 2625 1451 87 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2625 1451 88 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2625 1451 89 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2625 1451 90 THE OUTGOING NEUTRON; ISOTROPY IS ASSUMED FOR THE PROTON 2625 1451 91 AND RESIDUAL). (N,XP) D-D EMISSION USED TO BENCHMARK THE 2625 1451 92 TNG CALCULATIONS. 2625 1451 93 6-51 THROUGH 6-57 INELASTIC SCATTERING EXCITING LEVELS; EACH 2625 1451 94 SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON 2625 1451 95 AND 54FE RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR 2625 1451 96 THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED 2625 1451 97 BY DWUCK (1) AND TNG (2,3,4)). 2625 1451 98 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2625 1451 99 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 54FE 2625 1451 100 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2625 1451 101 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; TNG 2625 1451 102 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2625 1451 103 FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE 2625 1451 104 OUTGOING NEUTRON). 2625 1451 105 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2625 1451 106 AND 54MN RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2625 1451 107 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2625 1451 108 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2625 1451 109 PRODUCT. (N,XP) D-D EMISSION DATA USED TO BENCHMARK THE 2625 1451 110 TNG CALCULATIONS. 2625 1451 111 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2625 1451 112 AND 51CR RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2625 1451 113 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2625 1451 114 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2625 1451 115 PRODUCT. (N,XA) D-D EMISSION DATA USED TO BENCHMARK THE 2625 1451 116 CALCULATIONS. 2625 1451 117 12-51 THROUGH 12-57 BRANCHING RATIOS FOR THE LEVELS ARE GIVEN. 2625 1451 118 12-102 (N,G) CAPTURE; BELOW 1 MEV FROM ENDF/B-V FOR NAT FE, BUT 2625 1451 119 ADJUSTED FOR ENERGY BALANCE. ABOVE 1 MEV FROM 2625 1451 120 TNG CALCULATIONS. 2625 1451 121 14-51 THROUGH 14-57 AND 14-102 GAMMA RAY ANGULAR DISTRIBUTIONS 2625 1451 122 ASSUMED TO BE ISOTROPIC. 2625 1451 123 15-102 (N,G) CAPTURE; AS IN 12-102 2625 1451 124 2625 1451 125 -------------------------------------------------------------- 2625 1451 126 UNCERTAINTY FILES 2625 1451 127 AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS 2625 1451 128 REQUIRED BY ENDF/B-VI. 2625 1451 129 2625 1451 130 33-1 UNCERTAINTIES ARE DERIVED FROM 1.E-5 TO 100 EV. FROM 100 2625 1451 131 EV TO 20 MEV THEY ARE EXPLICIT, USING LB=0,1 AND 8. 2625 1451 132 33-2 FROM 1.E-5 TO 100 EV, UNCERTAINTIES ARE EXPLICIT, BASED 2625 1451 133 UPON THERMAL UNCERTAINTY AND OTHER DATA. FROM 100 EV TO 2625 1451 134 20 MEV THE FILES ARE DERIVED. 2625 1451 135 33-3 FROM 1.E-5 TO 700 KEV UNCERTAINTIES ARE DERIVED. FROM 7002625 1451 136 KEV TO 20 MEV UNCERTAINTIES ARE EXPLICIT, USING LB=1 AND 82625 1451 137 33-4 UNCERTAINTIES ARE ALL DERIVED. 2625 1451 138 33-16 UNCERTAINTIES FOR (N,2N) ARE EXPLICIT, ESTIMATED FROM TNG 2625 1451 139 33-22 UNCERTAINTIES FOR (N,NA) ARE EXPLICIT, ESTIMATED FROM TNG 2625 1451 140 33-28 UNCERTAINTIES FOR (N,NP) ARE EXPLICIT, ESTIMATED FROM TNG 2625 1451 141 33-51-91 UNCERTAINTIES FOR INELASTIC SCATTERING ARE EXPLICIT, 2625 1451 142 BASED ON DATA AND CALCULATION UNCERTAINTIES. 2625 1451 143 33-102 UNCERTAINTIES ARE EXPLICIT, BASED ON THERMAL DATA AT LOW 2625 1451 144 ENERGIES, AND CALCULATED RESULTS ABOVE 700 KEV. 2625 1451 145 33-103 (N,P) COVARIANCES WERE TAKEN FROM THE GLUCS (8) 2625 1451 146 CALCULATION IN WHICH THIS REACTION WAS STUDIED 2625 1451 147 SIMULTANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2625 1451 148 SECTIONS (9). 2625 1451 149 33-104 UNCERTAINTIES ESTIMATED FROM DATA. 2625 1451 150 33-107 UNCERTAINTIES ESTIMATED FROM DATA. 2625 1451 151 2625 1451 152 REFERENCES: 2625 1451 153 (1) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2625 1451 154 COLORADO, UNPUBLISHED (1972). 2625 1451 155 (2) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2625 1451 156 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2625 1451 157 MOMENTUM," ORNL/TM-7042 (1980), AND NSE 100, 61 (1988). 2625 1451 158 (3) C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2625 1451 159 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2625 1451 160 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2625 1451 161 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2625 1451 162 NATIONAL LAB. 2625 1451 163 (4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2625 1451 164 STATISTICAL MODEL CODE", ORNL/TM-10093 (AUGUST, 1986). 2625 1451 165 (5) S. F. MUGHABGHAB, RESONANCE PARAMETERS AND THERMAL 2625 1451 166 CROSS SECTIONS, ACADEMIC PRESS, 1981. 2625 1451 167 (6) B. J. ALLEN, GEEL CONFERENCE, 1977. 2625 1451 168 (7) P.T. GUENTHER, D.L. SMITH, A.B. SMITH, AND J.F. WHALEN, ANN. 2625 1451 169 NUCL. ENERGY, 13 (11), 601-610 (1986). 2625 1451 170 (8) D.M. HETRICK AND C.Y. FU, "GLUCS: A GENERALIZED LEAST SQUARES2625 1451 171 PROGRAM FOR UPDATING CROSS SECTION EVALUATIONS WITH 2625 1451 172 CORRELATED DATA SETS," ORNL/TM-7341, ENDF-303 (OCT,1980). 2625 1451 173 (9) C.Y. FU AND D.M. HETRICK, "EXPERIENCE IN USING THE 2625 1451 174 COVARIANCES OF SOME ENDF/B-V DOSIMETRY CROSS SECTIONS: 2625 1451 175 PROPOSED IMPROVEMENTS AND ADDITIONS OF CROSS-REACTION 2625 1451 176 COVARIANCES," PROC. FOURTH ASTM-EURATOM SYMP. ON REACTOR 2625 1451 177 DOSIMETRY, GAITHERSBURG, MARYLAND, MARCH 22-26,1982, P877, 2625 1451 178 US NATIONAL BUREAU OF STANDARDS. 2625 1451 179 (10) S.M. GRIMES, R.C. HAIGHT, K.R. ALVAR, H.H. BARSCHALL, AND 2625 1451 180 R.R. BORCHERS, PHYS. REV., C19, 2127 (JUNE, 1979). 2625 1451 181 2625 1451 182 1 451 185 12625 1451 183 3 103 18 12625 1451 184 33 103 77 12625 1451 185 2625 1 0 186 2625 0 0 187 2.605400+4 5.347600+1 0 0 0 02625 3103 188 8.530000+4 8.530000+4 0 0 1 452625 3103 189 45 2 0 0 0 02625 3103 190 1.000000-5 0.000000+0 5.000000+5 0.000000+0 7.000000+5 0.000000+02625 3103 191 8.000000+5 5.797200-6 1.000000+6 1.566600-4 1.250000+6 9.526800-42625 3103 192 1.500000+6 2.761200-3 1.750000+6 5.747500-3 2.000000+6 2.422300-22625 3103 193 2.250000+6 4.145300-2 2.500000+6 6.455800-2 2.750000+6 1.014800-12625 3103 194 3.000000+6 1.701300-1 3.220000+6 1.941800-1 3.250000+6 1.828400-12625 3103 195 3.350000+6 1.944200-1 3.360000+6 1.965200-1 3.400000+6 1.984600-12625 3103 196 3.450000+6 1.955400-1 3.500000+6 1.974800-1 3.750000+6 2.455700-12625 3103 197 4.000000+6 2.951700-1 4.250000+6 3.005000-1 4.500000+6 3.264000-12625 3103 198 5.000000+6 4.025200-1 5.500000+6 4.331100-1 6.000000+6 4.733800-12625 3103 199 7.000000+6 4.814200-1 8.000000+6 4.822000-1 8.500000+6 4.834000-12625 3103 200 9.000000+6 4.842600-1 9.500000+6 4.810000-1 1.000000+7 4.776700-12625 3103 201 1.050000+7 4.740000-1 1.100000+7 4.700000-1 1.150000+7 4.650000-12625 3103 202 1.200000+7 4.512000-1 1.250000+7 4.304200-1 1.300000+7 4.097500-12625 3103 203 1.350000+7 3.828400-1 1.400000+7 3.416600-1 1.500000+7 2.697400-12625 3103 204 1.600000+7 2.154000-1 1.800000+7 1.502400-1 2.000000+7 1.181700-12625 3103 205 2625 3 0 206 2625 0 0 207 2.605400+4 5.347600+1 0 0 0 1262533103 208 0.000000+0 0.000000+0 0 103 0 4262533103 209 0.000000+0 0.000000+0 1 5 171 18262533103 210 1.000000-5 5.000000+5 1.000000+6 2.000000+6 3.000000+6 3.350000+6262533103 211 4.000000+6 5.000000+6 5.500000+6 6.000000+6 7.000000+6 9.500000+6262533103 212 1.000000+7 1.050000+7 1.200000+7 1.400000+7 1.800000+7 2.000000+7262533103 213 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533103 214 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533103 215 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 2.627900-2262533103 216 7.095400-3 4.993100-5 5.452900-5 6.494100-5 4.226800-4 3.274000-4262533103 217 3.823000-4 1.589300-4 1.463200-4 2.002600-4 1.975100-4 1.933600-4262533103 218 3.168600-4 1.292700-4 5.974000-4 5.574000-3 4.858300-4 4.784900-4262533103 219 4.596200-4 3.305300-4 3.026500-4 3.214400-4 2.374500-4 2.186700-4262533103 220 1.790700-4 1.730000-4 1.539900-4 2.022900-4 8.668200-5 1.577500-4262533103 221 1.570700-3 7.044100-4 6.733700-4 4.440200-4 3.972100-4 4.296600-4262533103 222 2.287100-4 2.085000-4 2.314900-4 2.264500-4 2.070100-4 5.179900-5262533103 223 3.546600-5 1.315200-4 1.548700-3 6.665700-4 4.413000-4 3.961600-4262533103 224 4.274700-4 2.306900-4 2.103200-4 2.310100-4 2.259900-4 2.063000-4262533103 225 5.270900-5 3.581100-5 1.293300-4 1.459400-3 4.329700-4 3.920000-4262533103 226 4.200500-4 2.350900-4 2.147100-4 2.299000-4 2.244400-4 2.028700-4262533103 227 5.586600-5 3.681600-5 1.226500-4 9.156900-4 4.262900-4 4.496200-4262533103 228 4.472200-4 4.136100-4 1.978100-4 1.857100-4 1.405800-4 2.346400-4262533103 229 1.004200-4-5.101400-6 7.318100-4 3.864400-4 3.970600-4 3.622000-4262533103 230 2.254700-4 2.169300-4 1.749400-4 1.699700-4 7.860400-5 2.577800-5262533103 231 8.243200-4 4.199700-4 3.932100-4 2.086900-4 1.980900-4 1.531600-4262533103 232 2.102800-4 9.217700-5 6.253400-6 6.496100-4 2.686900-4 3.412000-4262533103 233 3.317500-4 3.332800-4-8.017700-5-4.008500-6 1.883400-4 5.438700-4262533103 234 3.264700-4 3.159900-4 3.248000-4-7.198800-5 2.229000-6 1.795300-4262533103 235 7.820300-4 3.259500-4 3.195500-4 3.410600-4 1.520900-4-1.458400-4262533103 236 7.646200-4 2.337400-4 3.436400-4 1.534700-4-1.493900-4 8.503800-4262533103 237 4.054700-4 1.784700-4-2.000700-4 2.944900-3 4.262100-4 2.042100-3262533103 238 8.653200-4 7.363400-4 8.588900-3 0.000000+0 0.000000+0 0.000000+0262533103 239 0.000000+0 0.000000+0 18 4 72 36262533103 240 1.000000-5-1.000000+0 5.000000+5-3.100000-1 1.000000+6-4.600000-1262533103 241 2.000000+6-5.400000-1 3.000000+6-5.500000-1 3.350000+6-5.400000-1262533103 242 4.000000+6-4.700000-1 5.000000+6-1.000000+0 5.500000+6-1.000000+0262533103 243 6.000000+6-1.000000+0 7.000000+6-7.300000-1 9.500000+6-1.000000+0262533103 244 1.000000+7-1.000000+0 1.050000+7-7.700000-1 1.200000+7-6.300000-1262533103 245 1.400000+7-7.300000-1 1.800000+7-1.000000+0 2.000000+7 0.000000+0262533103 246 1.000000-5 0.000000+0 5.000000+5 1.621100-1 1.000000+6 7.465900-2262533103 247 2.000000+6 3.963200-2 3.000000+6 3.935300-2 3.350000+6 3.820200-2262533103 248 4.000000+6 3.026000-2 5.000000+6 2.705200-2 5.500000+6 2.871100-2262533103 249 6.000000+6 2.548700-2 7.000000+6 2.332100-2 9.500000+6 2.796500-2262533103 250 1.000000+7 2.765200-2 1.050000+7 2.916100-2 1.200000+7 5.426700-2262533103 251 1.400000+7 2.941600-2 1.800000+7 9.267700-2 2.000000+7 0.000000+0262533103 252 0.000000+0 0.000000+0 0 1 90 45262533103 253 1.000000-5 0.000000+0 5.000000+5 7.331800-3 6.000000+5 7.331800-3262533103 254 8.000000+5 7.331800-3 1.000000+6 2.307700-3 1.250000+6 2.307700-3262533103 255 1.500000+6 2.307700-3 1.750000+6 2.307700-3 2.000000+6 7.633700-4262533103 256 2.250000+6 7.633700-4 2.500000+6 7.633700-4 2.750000+6 7.633700-4262533103 257 3.000000+6 7.665900-4 3.220000+6 7.665900-4 3.250000+6 7.665900-4262533103 258 3.350000+6 7.092700-4 3.360000+6 7.092700-4 3.400000+6 7.092700-4262533103 259 3.450000+6 7.092700-4 3.500000+6 7.092700-4 3.750000+6 7.092700-4262533103 260 4.000000+6 3.873400-4 4.250000+6 3.873400-4 4.500000+6 3.873400-4262533103 261 5.000000+6 6.586300-4 5.500000+6 7.418900-4 6.000000+6 5.846500-4262533103 262 7.000000+6 3.573200-4 8.000000+6 3.573200-4 8.500000+6 3.573200-4262533103 263 9.000000+6 3.573200-4 9.500000+6 7.038300-4 1.000000+7 6.881500-4262533103 264 1.050000+7 5.893100-4 1.100000+7 5.893100-4 1.150000+7 5.893100-4262533103 265 1.200000+7 1.669800-3 1.250000+7 1.669800-3 1.300000+7 1.669800-3262533103 266 1.350000+7 1.669800-3 1.400000+7 5.685100-4 1.500000+7 5.685100-4262533103 267 1.600000+7 5.685100-4 1.800000+7 7.730000-3 2.000000+7 0.000000+0262533103 268 0.000000+0 0.000000+0 0 8 90 45262533103 269 1.000000-5 0.000000+0 5.000000+5 3.04200-19 6.000000+5 1.21330-17262533103 270 8.000000+5 2.73780-14 1.000000+6 6.29280-12 1.250000+6 2.32710-10262533103 271 1.500000+6 1.954900-9 1.750000+6 8.470000-9 2.000000+6 4.976800-8262533103 272 2.250000+6 1.457500-7 2.500000+6 3.535000-7 2.750000+6 8.734900-7262533103 273 3.000000+6 2.465400-6 3.220000+6 3.211700-6 3.250000+6 2.847500-6262533103 274 3.350000+6 2.978900-6 3.360000+6 3.043600-6 3.400000+6 3.103900-6262533103 275 3.450000+6 3.013300-6 3.500000+6 3.073400-6 3.750000+6 4.752500-6262533103 276 4.000000+6 3.749700-6 4.250000+6 3.886300-6 4.500000+6 4.585100-6262533103 277 5.000000+6 1.185700-5 5.500000+6 1.546300-5 6.000000+6 1.455700-5262533103 278 7.000000+6 9.201700-6 8.000000+6 9.031600-6 8.500000+6 8.791900-6262533103 279 9.000000+6 9.310500-6 9.500000+6 1.637700-5 1.000000+7 1.646400-5262533103 280 1.050000+7 1.370800-5 1.100000+7 1.371400-5 1.150000+7 1.363900-5262533103 281 1.200000+7 3.777000-5 1.250000+7 3.437100-5 1.300000+7 3.115000-5262533103 282 1.350000+7 2.719200-5 1.400000+7 7.373700-6 1.500000+7 4.596100-6262533103 283 1.600000+7 2.930800-6 1.800000+7 1.938700-5 2.000000+7 0.000000+0262533103 284 262533 0 285 2625 0 0 286 0 0 0 287 2.605600+4 5.545400+1 -1 0 0 12631 1451 1 0.000000+0 0.000000+0 0 0 0 62631 1451 2 1.000000+0 0.000000+0 0 0 10 62631 1451 3 0.000000+0 0.000000+0 0 0 178 32631 1451 4 26-FE- 56 ORNL EVAL-MAR89 FU, C.PEREY, HETRICK, F.PEREY 2631 1451 5 ORNL/TM-9964(1986) DIST-APR90 900413 2631 1451 6 ----ENDF/B-VI MATERIAL 2631 REVISION 0 2631 1451 7 -----INCIDENT NEUTRON DATA 2631 1451 8 ------ENDF-6 FORMAT 2631 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2631 1451 10 DOSIMETRY 2631 1451 11 ******************************************************************2631 1451 12 2631 1451 13 ORNL-4617(PE71) DESCRIBES ENDF/B-3. MAJOR REVISIONS FOR ENDF 2631 1451 14 /B-4 AND -5 ARE SUMMARIZED IN ORNL/TM-7523(FU80) AND ORNL/TM-8283 2631 1451 15 (FU82). THE PRESENT EVALUATION IS BASED ON THE CALCULATIONS 2631 1451 16 DESCRIBED IN ORNL/TM-9964(FU86). THE MAJOR IMPROVEMENTS OVER 2631 1451 17 ENDF-5 MOD3 ARE: 2631 1451 18 1. EVALUATION IS NOW ISOTOPIC. 2631 1451 19 2. ANALYSIS IS BASED ON A NEW TNG CODE MUCH IMPROVED OVER 2631 1451 20 THE VERSION USED FOR ENDF/B-5 FOR CALCULATING ENERGY-ANGLE 2631 1451 21 CORRELATED SPECTRA. 2631 1451 22 3. RESONANCE PARAMETERS ARE EXTENDED FROM 400 KEV TO 850 KEV 2631 1451 23 (PE89). 2631 1451 24 4. NEW DATA SETS (PA88,LA85,MA83,TA83,FU82,FR80,AR80,GR79, 2631 1451 25 HA89) ARE INCORPORATED. 2631 1451 26 5. INTEGRAL RESULTS (MA85,PA85,CR76,JO75) ARE CONSIDERED FOR 2631 1451 27 THE (N,N') EVALUATION ABOVE 3 MEV. 2631 1451 28 6. GAMMA-RAY PRODUCTION CROSS SECTIONS AND SPECTRA ARE GIVEN 2631 1451 29 FOR EACH REACTION,INCLUDING MT = 51-75, INSTEAD OF FOR THE 2631 1451 30 NONELASTIC. 2631 1451 31 7. RECOIL SPECTRA ARE GIVEN IN FILE-6. 2631 1451 32 8. CHARGED PARTICLE SPECTRA ARE GIVEN. 2631 1451 33 2631 1451 34 THE FOLLOWING DESCRIPTION FOR EACH FILE AND SECTION IS FOR 2631 1451 35 ENDF/B-VI: 2631 1451 36 2631 1451 37 FILE 2 RESONANCE PARAMETERS 2631 1451 38 REICH-MOORE PARAMETERS ARE GIVEN IN THE ENERGY 2631 1451 39 RANGE FROM 1.E-5 EV TO 850 KEV (PE88). 2631 1451 40 FILE 3 NEUTRON CROSS SECTIONS 2631 1451 41 SECTION 1 TOTAL INTERACTION 2631 1451 42 850 KEV TO 2 MEV -- HA89 FOR FE-56 2631 1451 43 2 MEV TO 20 MEV -- ENDF/B-V DATA FOR NATURAL FE USED 2631 1451 44 SECTION 2 ELASTIC SCATTERING 2631 1451 45 DERIVED BY SUBTRACTING THE NON-ELASTIC CROSS SECTION FROM THE 2631 1451 46 TOTAL CROSS SECTION. 2631 1451 47 SECTION 3 NON-ELASTIC INTERACTION 2631 1451 48 SUM OF ALL REACTION CROSS SECTIONS DESCRIBED BELOW. 2631 1451 49 SECTION 4 TOTAL INELASTIC SCATTERING 2631 1451 50 DERIVED BY ADDING INELASTIC SCATTERING CROSS SECTIONS FOR 2631 1451 51 EXCITING DISCRETE LEVELS AND THE CONTINUUM. 2631 1451 52 SECTION 16 N,2N REACTION -- SA71, FR80 2631 1451 53 SECTION 22 N,NA REACTION -- GR79 2631 1451 54 SECTION 28 N,NP REACTION -- GR79, AL61 2631 1451 55 SECTION 51 INELASTIC SCATTERING EXCITING FIRST LEVEL IN FE-56 2631 1451 56 .8611 MEV TO 1.5 MEV -- VO71,KI77. 2631 1451 57 1.5 MEV TO 2.122 MEV -- GI65 VO71 2631 1451 58 2.122 MEV TO 20 MEV -- BA66 KI70 MA71 RA70 PE69 SM73 2631 1451 59 SECTION 52 THROUGH 75 -- INELASTIC SCATTERING EXCITING LEVELS 2631 1451 60 BA66 MA71 RA70 KI70 2631 1451 61 SECTION 91 INELASTIC SCATTERING EXCITING THE CONTINUUM 2631 1451 62 GO66 KI70 KA73 BO63 SC65 CL72 MA69 HE75 TA83 PA8 2631 1451 63 SECTION 102 RADIATIVE CAPTURE -- HO69 ST70 GA76 AL76 LE76 2631 1451 64 1.E-5 EV TO 400 KEV -- COMPLETELY FROM RESONANCES 2631 1451 65 400 KEV TO 850 KEV -- COMBINATION OF RESONANCES AND FILE 3 2631 1451 66 850 KEV TO 20 MEV -- COMBINATION OF DATA AND CALCULATION 2631 1451 67 SECTION 103 N,P REACTION 2631 1451 68 LEAST-SQUARES ANALYSIS OF MANY DATA AND RATIOS (FU82) 2631 1451 69 UPDATED IN 1989 TO INCLUDE ENDF/B-VI STANDARDS 2631 1451 70 SECTION 104 N,D REACTION -- (AR80). 2631 1451 71 SECTION 105 N,T REACTION -- (AR80). 2631 1451 72 SECTION 106 N,HE-3 REACTION -- (AR80). 2631 1451 73 SECTION 107 N,A REACTION -- GR79,KI72 2631 1451 74 FILE 4 ANGULAR DISTRIBUTION OF SECONDARY NEUTRONS 2631 1451 75 ALL DISTRIBUTIONS IN FILE 4 ARE GIVEN IN THE CENTER OF MASS 2631 1451 76 SYSTEM IN THE LEGENDRE POLYNOMIAL REPRESENTATION. 2631 1451 77 SECTION 2 ELASTIC SCATTERING 2631 1451 78 .00001 EV TO 1.23 MEV -- BA68 CO63 KO63 PA64,KP77 2631 1451 79 1.23 MEV TO 4 MEV -- BA68 BE56 BE66 BO59 CR69 HO68,KP77 2631 1451 80 4 MEV TO 20 MEV -- BA68 KI70 PE70 2631 1451 81 SECTION 51 INELASTIC SCATTERING EXCITING FIRST LEVEL IN FE-56 2631 1451 82 BA66 GIL65 KI70 RA70 2631 1451 83 SECTION 52 THROUGH 75 -- INELASTIC SCATTERING EXCITING LEVELS 2631 1451 84 BA66 KI70 RA70 MA71 2631 1451 85 FILE 6 ENERGY-ANGLE DISTRIBUTIONS OF SECONDARY PARTICLES, RECOILS,2631 1451 86 AND GAMMA RAYS -- ALL SPECTRA, INCLUDING RECOILS, ARE BASED2631 1451 87 ON TNG (FU88) ANALYSIS OF THE DATA GIVEN BELOW AND ARE 2631 1451 88 GIVEN IN THE LAB SYSTEM 2631 1451 89 SECTION 16 N,2N REACTION 2631 1451 90 FOR NEUTRON: SA71, HE75, KA72, TA83, PA88 2631 1451 91 FOR GAMMA RAYS: DA70, DI73, CH76 2631 1451 92 SECTION 22 N,NA REACTION -- AS FOR (N,2N) 2631 1451 93 SECTION 28 N,NP REACTION -- AS FOR (N,2N) 2631 1451 94 SECTION 91 INELASTIC SCATTERING EXCITING THE CONTINUUM 2631 1451 95 FOR NEUTRONS: KI70 MA69 PE69 BO63 SC65 KA72 CL72 HA73 HE75 2631 1451 96 TA88 PA88 2631 1451 97 FOR GAMMA RAYS: DA70 DI73 CH76 LA85 2631 1451 98 SECTION 103 (N,P) REACTION 2631 1451 99 FOR PROTON: GR79 2631 1451 100 SECTION 107 (N,A) REACTION 2631 1451 101 FOR ALPHAS: GR79 2631 1451 102 FILE 12 MULTIPLICITIES OF GAMMA RAYS PRODUCED BY NEUTRON REACTIONS2631 1451 103 SECTION 51-75 INELASTIC SCATTERING EXCITING DISCRETE LEVELS 2631 1451 104 FROM BRANCHING RATIOS USED IN THE TNG ANALYSIS OF DA70,DI73, 2631 1451 105 CH76, LA85 2631 1451 106 SECTION 102 RADIATIVE CAPTURE 2631 1451 107 FROM EVALUATION BY WHITE, FU, AND YOST (WH73) 2631 1451 108 THESE GAMMA RAYS ARE REPRESENTATIVE OF NEUTRON ENERGY RANGES 2631 1451 109 AND EACH RANGE CONTAIN CONTRIBUTIONS OF S- AND P-WAVE 2631 1451 110 COMPONENTS OF FE-54 AND FE-56. THESE RESULTS ARE RETAINED 2631 1451 111 HERE AS WELL AS IN THE EVALUATIONS FOR FE-54, -57, AND -58. 2631 1451 112 FILE 14 ANGULAR DISTRIBUTIONS OF SECONDARY GAMMA RAYS 2631 1451 113 SECTIONS 51-75 AND 102 -- ASSUMED ISOTROPIC. 2631 1451 114 FILE 15 ENERGY DISTRIBUTIONS OF SECONDARY GAMMA RAYS 2631 1451 115 SECTION 102 RADIATIVE CAPTURE (WH73) 2631 1451 116 FILE 33 COVARIANCES FOR FILE-3 DATA 2631 1451 117 SECTION 103 WAS FROM GLUCS RUN OF 1989 (SEE 3-103) 2631 1451 118 2631 1451 119 REFERENCES 2631 1451 120 AL61 D. L. ALLAN, NUCL. PHYS. 24, 274(1961). 2631 1451 121 AL76 B.J.ALLEN ET AL.NUCL. PHYS. A269,408(1976). 2631 1451 122 AR80 E.D.ARTHUR AND P.G.YOUNG, LA-8626 (1980). 2631 1451 123 BA66 R. H. BASSEL ET. AL., COMPUTER CODE JULIE, ORNL(1966). 2631 1451 124 BA68 E. BARNARD ET. AL., PEL-180, PELINDABA, PRETORIA(1968). 2631 1451 125 BE56 J. R. BEYSTER ET. AL., PHYS. REV. 104, 1319(1956). 2631 1451 126 BE66 R. L. BECKER ET. AL., NUCL. PHYS. 89,154(1966). 2631 1451 127 BO59 N. A. 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AL., NSE 51, 496(1973). 2631 1451 181 2631 1451 182 1 451 185 12631 1451 183 3 103 19 12631 1451 184 33 103 125 12631 1451 185 2631 1 0 186 2631 0 0 187 2.605600+4 5.545400+1 0 0 0 02631 3103 188 -2.913000+6-2.913000+6 0 0 3 472631 3103 189 2 2 12 4 47 22631 3103 190 2.965500+6 0.000000+0 3.000000+6 7.001900-9 4.000000+6 6.001600-62631 3103 191 4.250000+6 3.201000-5 4.500000+6 1.403800-4 4.600000+6 2.406500-42631 3103 192 4.800000+6 5.775600-4 5.000000+6 1.069500-3 5.250000+6 2.294100-32631 3103 193 5.500000+6 4.798200-3 5.750000+6 8.479100-3 6.000000+6 1.273100-22631 3103 194 6.200000+6 1.645300-2 6.400000+6 1.917500-2 6.600000+6 2.375600-22631 3103 195 6.800000+6 2.758300-2 7.000000+6 2.728100-2 7.500000+6 3.534700-22631 3103 196 8.000000+6 4.173200-2 8.500000+6 4.767000-2 9.000000+6 5.514500-22631 3103 197 9.500000+6 5.907100-2 1.000000+7 6.690700-2 1.050000+7 7.572400-22631 3103 198 1.100000+7 8.241000-2 1.150000+7 9.069500-2 1.200000+7 1.033500-12631 3103 199 1.220000+7 1.056500-1 1.240000+7 1.093000-1 1.260000+7 1.116300-12631 3103 200 1.280000+7 1.143600-1 1.300000+7 1.152300-1 1.340000+7 1.169200-12631 3103 201 1.380000+7 1.166700-1 1.420000+7 1.135900-1 1.460000+7 1.093400-12631 3103 202 1.500000+7 1.042200-1 1.550000+7 9.512100-2 1.600000+7 8.625100-22631 3103 203 1.650000+7 7.978800-2 1.700000+7 7.162500-2 1.750000+7 6.627600-22631 3103 204 1.800000+7 6.129600-2 1.850000+7 5.681200-2 1.900000+7 5.264800-22631 3103 205 1.950000+7 5.096800-2 2.000000+7 5.055300-2 0.000000+0 0.000000+02631 3103 206 2631 3 0 207 2631 0 0 208 2.605600+4 5.545400+1 0 0 0 1263133103 209 0.000000+0 0.000000+0 0 103 0 4263133103 210 0.000000+0 0.000000+0 1 5 406 28263133103 211 1.000000-5 2.965500+6 3.000000+6 4.500000+6 4.600000+6 4.800000+6263133103 212 5.000000+6 5.750000+6 6.000000+6 6.200000+6 6.400000+6 6.600000+6263133103 213 6.800000+6 8.000000+6 8.500000+6 9.000000+6 9.500000+6 1.100000+7263133103 214 1.200000+7 1.300000+7 1.340000+7 1.380000+7 1.500000+7 1.550000+7263133103 215 1.700000+7 1.750000+7 1.800000+7 2.000000+7 0.000000+0 0.000000+0263133103 216 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 217 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 218 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 219 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 220 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 221 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 222 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 223 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 224 0.000000+0 0.000000+0 0.000000+0 3.516500-2 3.560200-3 3.560200-3263133103 225 2.926300-4 9.958600-4 2.042600-4 1.942400-4 2.044000-4 1.640400-4263133103 226 1.631400-4 1.507400-4 9.950600-5 1.281500-4 1.124400-4 1.109500-4263133103 227 8.050600-5 3.532100-5 2.097000-5 1.286600-5 1.572400-5 5.411900-6263133103 228 5.643300-6 5.168000-6 4.980600-6 4.767800-6 2.080700-2 7.205500-3263133103 229 5.705300-4 1.635300-3 3.792000-4 3.509200-4 3.642800-4 3.056100-4263133103 230 3.037000-4 2.771100-4 1.862600-4 2.407800-4 2.111200-4 2.085000-4263133103 231 1.511100-4 6.629700-5 3.943100-5 2.423400-5 2.969300-5 1.022000-5263133103 232 1.064500-5 9.746400-6 9.391200-6 8.991100-6 2.080700-2 5.705400-4263133103 233 1.635100-3 3.792000-4 3.509500-4 3.642700-4 3.056300-4 3.037000-4263133103 234 2.771100-4 1.862600-4 2.407800-4 2.111200-4 2.085100-4 1.511000-4263133103 235 6.629400-5 3.943400-5 2.423500-5 2.969300-5 1.022100-5 1.064500-5263133103 236 9.746700-6 9.391600-6 8.991300-6 9.393100-4 3.161400-4 4.764800-4263133103 237 3.636900-4 3.333900-4 3.929000-4 3.884000-4 3.260800-4 2.465900-4263133103 238 3.260400-4 2.847400-4 2.828100-4 2.032600-4 8.929700-5 5.365300-5263133103 239 3.332100-5 4.140900-5 1.421600-5 1.474200-5 1.348200-5 1.297000-5263133103 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1.535700-5 1.489500-5 1.418500-5 8.113200-4 3.912200-4263133103 266 2.990200-4 1.307800-4 7.729200-5 4.753300-5 5.883300-5 2.028300-5263133103 267 2.107200-5 1.931900-5 1.864200-5 1.782100-5 8.019200-4 3.533200-4263133103 268 1.350500-4 7.854100-5 4.836300-5 6.026300-5 2.067000-5 2.135300-5263133103 269 1.951800-5 1.873500-5 1.802500-5 1.321400-3 2.786500-4 1.322500-4263133103 270 7.022000-5 7.439000-5 2.539400-5 2.723200-5 2.504000-5 2.399200-5263133103 271 2.339200-5 1.411400-3 2.672300-4 1.160200-4 1.082900-4 3.762900-5263133103 272 4.132900-5 3.892000-5 3.957300-5 3.520200-5 1.010200-3 2.022800-4263133103 273 1.961700-4 5.416700-5 5.852900-5 5.376000-5 5.100300-5 5.022700-5263133103 274 5.177500-4 1.746600-4 6.090300-5 6.530100-5 5.913400-5 5.495600-5263133103 275 5.532500-5 4.536200-4 5.566200-5 6.230500-5 5.659300-5 5.270000-5263133103 276 5.312000-5 3.279800-4 8.159300-5 1.100600-4 9.873500-5 1.020100-4263133103 277 2.324800-4 1.430100-4 1.253600-4 1.382900-4 2.165400-4 1.276400-4263133103 278 1.570200-4 2.740800-4 1.206100-4 2.290400-4 0.000000+0 0.000000+0263133103 279 0.000000+0 0.000000+0 28 4 112 56263133103 280 1.000000-5 0.000000+0 2.965500+6-1.000000+0 3.000000+6-5.100000-1263133103 281 4.500000+6-1.000000+0 4.600000+6-1.000000+0 4.800000+6-1.000000+0263133103 282 5.000000+6-8.500000-1 5.750000+6-1.000000+0 6.000000+6-1.000000+0263133103 283 6.200000+6-1.000000+0 6.400000+6-1.000000+0 6.600000+6-1.000000+0263133103 284 6.800000+6-7.100000-1 8.000000+6-1.000000+0 8.500000+6-1.000000+0263133103 285 9.000000+6-1.000000+0 9.500000+6-5.600000-1 1.100000+7-7.400000-1263133103 286 1.200000+7-6.500000-1 1.300000+7-1.000000+0 1.340000+7-1.000000+0263133103 287 1.380000+7-7.000000-1 1.500000+7-1.000000+0 1.550000+7-4.400000-1263133103 288 1.700000+7-1.000000+0 1.750000+7-1.000000+0 1.800000+7-6.600000-1263133103 289 2.000000+7 0.000000+0 1.000000-5 0.000000+0 2.965500+6 0.000000+0263133103 290 3.000000+6 1.875200-1 4.500000+6 1.442500-1 4.600000+6 1.442500-1263133103 291 4.800000+6 3.064800-2 5.000000+6 1.233300-1 5.750000+6 3.161300-2263133103 292 6.000000+6 7.384900-2 6.200000+6 7.781200-2 6.400000+6 4.204600-2263133103 293 6.600000+6 4.255300-2 6.800000+6 4.333400-2 8.000000+6 3.398400-2263133103 294 8.500000+6 2.846200-2 9.000000+6 2.848400-2 9.500000+6 2.831800-2263133103 295 1.100000+7 3.635100-2 1.200000+7 3.756800-2 1.300000+7 3.178300-2263133103 296 1.340000+7 2.275400-2 1.380000+7 2.129800-2 1.500000+7 1.811000-2263133103 297 1.550000+7 1.524700-2 1.700000+7 1.471500-2 1.750000+7 1.655500-2263133103 298 1.800000+7 1.513400-2 2.000000+7 0.000000+0 0.000000+0 0.000000+0263133103 299 0.000000+0 0.000000+0 0 1 96 48263133103 300 1.000000-5 0.000000+0 2.965500+6 0.000000+0 3.000000+6 1.614100-2263133103 301 4.000000+6 1.614100-2 4.250000+6 1.614100-2 4.500000+6 1.872600-2263133103 302 4.600000+6 1.872700-2 4.800000+6 8.453800-4 5.000000+6 1.163700-2263133103 303 5.250000+6 1.163700-2 5.500000+6 1.163700-2 5.750000+6 8.994200-4263133103 304 6.000000+6 4.908300-3 6.200000+6 5.449200-3 6.400000+6 1.591100-3263133103 305 6.600000+6 1.629700-3 6.800000+6 1.199900-3 7.000000+6 1.199900-3263133103 306 7.500000+6 1.199900-3 8.000000+6 1.039400-3 8.500000+6 7.290500-4263133103 307 9.000000+6 7.301900-4 9.500000+6 4.041700-4 1.000000+7 4.041700-4263133103 308 1.050000+7 4.041700-4 1.100000+7 8.800300-4 1.150000+7 8.800300-4263133103 309 1.200000+7 8.256400-4 1.220000+7 8.256400-4 1.240000+7 8.256400-4263133103 310 1.260000+7 8.256400-4 1.280000+7 8.256400-4 1.300000+7 9.091500-4263133103 311 1.340000+7 4.659800-4 1.380000+7 2.857800-4 1.420000+7 2.857800-4263133103 312 1.460000+7 2.857800-4 1.500000+7 2.951900-4 1.550000+7 9.206300-5263133103 313 1.600000+7 9.206300-5 1.650000+7 9.206300-5 1.700000+7 1.948900-4263133103 314 1.750000+7 2.466700-4 1.800000+7 1.360500-4 1.850000+7 1.360500-4263133103 315 1.900000+7 1.360500-4 1.950000+7 1.360500-4 2.000000+7 0.000000+0263133103 316 0.000000+0 0.000000+0 0 8 96 48263133103 317 1.000000-5 0.000000+0 2.965500+6 0.000000+0 3.000000+6 8.79240-20263133103 318 4.000000+6 6.45970-14 4.250000+6 1.83760-12 4.500000+6 4.10030-11263133103 319 4.600000+6 1.20500-10 4.800000+6 3.13330-11 5.000000+6 1.478900-9263133103 320 5.250000+6 6.804700-9 5.500000+6 2.976700-8 5.750000+6 7.184900-9263133103 321 6.000000+6 8.839300-8 6.200000+6 1.639000-7 6.400000+6 6.500000-8263133103 322 6.600000+6 1.021900-7 6.800000+6 1.014400-7 7.000000+6 9.922900-8263133103 323 7.500000+6 1.665800-7 8.000000+6 2.011400-7 8.500000+6 1.840800-7263133103 324 9.000000+6 2.467200-7 9.500000+6 1.567000-7 1.000000+7 2.010300-7263133103 325 1.050000+7 2.575000-7 1.100000+7 6.640700-7 1.150000+7 8.043000-7263133103 326 1.200000+7 9.798800-7 1.220000+7 1.024000-6 1.240000+7 1.095900-6263133103 327 1.260000+7 1.143200-6 1.280000+7 1.199800-6 1.300000+7 1.341300-6263133103 328 1.340000+7 7.077800-7 1.380000+7 4.322200-7 1.420000+7 4.097000-7263133103 329 1.460000+7 3.796200-7 1.500000+7 3.562500-7 1.550000+7 9.255400-8263133103 330 1.600000+7 7.609800-8 1.650000+7 6.512000-8 1.700000+7 1.110900-7263133103 331 1.750000+7 1.203900-7 1.800000+7 5.679600-8 1.850000+7 4.879100-8263133103 332 1.900000+7 4.190100-8 1.950000+7 3.926900-8 2.000000+7 0.000000+0263133103 333 263133 0 334 2631 0 0 335 0 0 0 336 2.605800+4 5.743600+1 1 0 0 12637 1451 1 0.000000+0 0.000000+0 0 0 0 62637 1451 2 1.000000+0 0.000000+0 0 0 10 62637 1451 3 0.000000+0 0.000000+0 0 0 154 32637 1451 4 26-FE- 58 ORNL EVAL-NOV89 HETRICK,FU,N.M.LARSON 2637 1451 5 ORNL/TM- DIST-APR90 900417 2637 1451 6 ----ENDF/B-VI MATERIAL 2637 2637 1451 7 -----INCIDENT NEUTRON DATA 2637 1451 8 ------ENDF-6 FORMAT 2637 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2637 1451 10 DOSIMETRY 2637 1451 11 ******************************************************************2637 1451 12 THIS WORK EMPLOYED THE HAUSER-FESHBACH CODE TNG (1,2,3). 2637 1451 13 THE TNG CODE PROVIDES ENERGY AND ANGULAR DISTRIBUTIONS OF 2637 1451 14 PARTICLES EMITTED IN THE COMPOUND AND PRE-COMPOUND REACTIONS, 2637 1451 15 ENSURES CONSISTENCY AMONG ALL REACTIONS, AND MAINTAINS ENERGY 2637 1451 16 BALANCE. DETAILS PERTINENT TO THE CONTENTS OF THIS EVALUATION 2637 1451 17 WILL BE PUBLISHED AT A LATER DATE. 2637 1451 18 RESONANCE PARAMETERS AND THERMAL CAPTURE CROSS SECTION WERE 2637 1451 19 TAKEN FROM (4), AND ADJUSTED TO THE DATA OF BEER ET AL. (5) 2637 1451 20 THROUGH THE SAMMY CODE BY N.M. LARSON. SOME CAPTURE 2637 1451 21 WIDTHS WERE ADJUSTED DOWNWARD TO COMPROMISE WITH THE AVERAGED 2637 1451 22 DATA OF (6). AVERAGE CAPTURE DATA OF (7) WERE USED TO GENERATE 2637 1451 23 CAPTURE BACKGROUND BELOW 400 KEV AND FOR THE POINTWISE CROSS 2637 1451 24 SECTION'S FROM 400 KEV TO 2 MEV. THERMAL CAPTURE AND CAPTURE IN 2637 1451 25 THE 10.4 KEV RESONANCE WERE FURTHER ADJUSTED BY SAMMY TO AGREE 2637 1451 26 WITH THE ENDF/B-V DOSIMETRY EVALUATION (F. MANN ET AL.), WHICH WAS2637 1451 27 ADJUSTED TO AGREE WITH INTEGRAL DATA. LOW KEV SCATTERING 2637 1451 28 CROSS SECTIONS WERE BASED ON AP=0.72 GIVEN BY (8). 2637 1451 29 FOR TOTAL CROSS SECTIONS ABOVE 400 KEV AND CAPTURE CROSS 2637 1451 30 SECTIONS ABOVE 2 MEV, NATURAL FE DATA OF ENDF/B-V WERE USED FOR 2637 1451 31 THE PURPOSE OF MAKING SURE THAT THE TOTAL CROSS SECTION MINIMA 2637 1451 32 IN NAT FE IS MAINTAINED. 2637 1451 33 2637 1451 34 ----- DESCRIPTION OF FILES 2637 1451 35 (MF-MT) 2637 1451 36 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2637 1451 37 2-151 RESONANCE PARAMETERS; SEE ABOVE. 2637 1451 38 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2637 1451 39 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2637 1451 40 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2637 1451 41 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2637 1451 42 3-1 TOTAL CROSS SECTION; SEE ABOVE 2637 1451 43 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2637 1451 44 SUBTRACTING THE NONELASTIC FROM THE TOTAL. 2637 1451 45 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 3-28, 2637 1451 46 3-102, 3-103, AND 3-107. 2637 1451 47 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, AND 2637 1451 48 3-91. 2637 1451 49 3-16 (N,2N) CROSS SECTIONS WERE CALCULATED BY THE TNG 2637 1451 50 CODE (1,2,3). 2637 1451 51 3-22 (N,NA)+(N,AN) CROSS SECTIONS WERE CALCULATED BY THE TNG 2637 1451 52 CODE (1,2,3). 2637 1451 53 3-28 (N,NP)+(N,PN) CROSS SECTIONS WERE CALCULATED BY THE TNG 2637 1451 54 CODE (1,2,3). 2637 1451 55 3-51 TO 3-52 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE 2637 1451 56 FROM TNG (1,2,3). 2637 1451 57 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG 2637 1451 58 CALCULATED). 2637 1451 59 3-102 (N,G) CAPTURE CROSS SECTION; SEE ABOVE 2637 1451 60 3-103 (N,P) CROSS SECTIONS WERE CALCULATED BY THE TNG 2637 1451 61 CODE (1,2,3), BUT RENORMALIZED TO FIT THE DATA OF DYER 2637 1451 62 & HAMILTON (9) AND BAHAL & PEPELNIK (10) AT EN=14.5 MEV. 2637 1451 63 3-107 (N,A) CROSS SECTIONS WERE CALCULATED BY THE 2637 1451 64 TNG CODE (1,2,3), BUT RENORMALIZED TO FIT THE DATA OF 2637 1451 65 KNEFF ET AL. (11) AT EN = 14.8 MEV. 2637 1451 66 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR 2637 1451 67 ELASTIC SCATTERING ARE FROM ENDF/B-V. 2637 1451 68 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2637 1451 69 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2637 1451 70 PARAMETERS IN 2/151. 2637 1451 71 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2637 1451 72 NEUTRON AND 57FE RESIDUAL, AND ENERGY DEPENDENT YIELD 2637 1451 73 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2637 1451 74 RAY; TNG CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR2637 1451 75 EACH PRODUCT. ISOTROPY IS ASSUMED. 2637 1451 76 6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2637 1451 77 NEUTRON, ALPHA, AND 54CR RESIDUAL, AND ENERGY DEPENDENT 2637 1451 78 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2637 1451 79 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2637 1451 80 FOR EACH PRODUCT (ISOTROPY IS ASSUMED). 2637 1451 81 6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2637 1451 82 NEUTRON, P, AND 57MN RESIDUAL, AND ENERGY DEPENDENT YIELD 2637 1451 83 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2637 1451 84 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2637 1451 85 EACH PRODUCT (ISOTROPY IS ASSUMED). 2637 1451 86 6-51 THROUGH 6-52 INELASTIC SCATTERING EXCITING LEVELS; 2637 1451 87 ASSUMED ISOTROPIC. 2637 1451 88 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2637 1451 89 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 58FE 2637 1451 90 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2637 1451 91 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; TNG 2637 1451 92 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2637 1451 93 FOR EACH. ISOTROPY IS ASSUMED. 2637 1451 94 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2637 1451 95 AND 58MN RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2637 1451 96 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2637 1451 97 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2637 1451 98 PRODUCT. ISOTROPY IS ASSUMED. 2637 1451 99 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2637 1451 100 AND 55CR RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2637 1451 101 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2637 1451 102 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2637 1451 103 PRODUCT. ISOTROPY IS ASSUMED. 2637 1451 104 12-51 THROUGH 12-52 BRANCHING RATIOS FOR THE LEVELS ARE GIVEN. 2637 1451 105 12-102 (N,G) CAPTURE; BELOW 1 MEV FROM ENDF/B-V FOR NAT FE, BUT 2637 1451 106 ADJUSTED FOR ENERGY BALANCE. ABOVE 1 MEV FROM TNG 2637 1451 107 CALCULATIONS. 2637 1451 108 14-51 THROUGH 14-52 AND 14-102 GAMMA RAY ANGULAR DISTRIBUTIONS 2637 1451 109 ASSUMED TO BE ISOTROPIC. 2637 1451 110 15-102 (N,G) CAPTURE; AS IN 12-102. 2637 1451 111 2637 1451 112 -------------------------------------------------------------- 2637 1451 113 UNCERTAINTY FILES 2637 1451 114 AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS 2637 1451 115 REQUIRED BY ENDF/B-VI. 2637 1451 116 2637 1451 117 33-1 UNCERTAINTIES ARE DERIVED FROM 1.E-5 TO 100 EV. FROM 100 2637 1451 118 EV TO 20 MEV THEY ARE EXPLICIT, USING LB=0,1 AND 8. 2637 1451 119 33-2 FROM 1.E-5 TO 100 EV, UNCERTAINTIES ARE EXPLICIT, BASED 2637 1451 120 UPON THERMAL UNCERTAINTY AND OTHER DATA. FROM 100 EV TO 2637 1451 121 20 MEV THE FILES ARE DERIVED. 2637 1451 122 33-3 FROM 1.E-5 TO 400 KEV UNCERTAINTIES ARE DERIVED. FROM 4002637 1451 123 KEV TO 20 MEV UNCERTAINTIES ARE EXPLICIT, USING LB=1 AND 82637 1451 124 33-4 UNCERTAINTIES ARE ALL DERIVED. 2637 1451 125 33-16 UNCERTAINTIES FOR (N,2N) ARE EXPLICIT, ESTIMATED FROM TNG 2637 1451 126 33-22 UNCERTAINTIES FOR (N,NA) ARE EXPLICIT, ESTIMATED FROM TNG 2637 1451 127 33-28 UNCERTAINTIES FOR (N,NP) ARE EXPLICIT, ESTIMATED FROM TNG 2637 1451 128 33-51-91 UNCERTAINTIES FOR INELASTIC SCATTERING ARE EXPLICIT, 2637 1451 129 BASED ON DATA AND CALCULATION UNCERTAINTIES. 2637 1451 130 33-102 UNCERTAINTIES ARE EXPLICIT, BASED ON THERMAL DATA AT LOW 2637 1451 131 ENERGIES, DATA OF (7) FOR ENERGIES UP TO 2.0 MEV, AND 2637 1451 132 TNG AT ENERGIES ABOVE 2.0 MEV. 2637 1451 133 33-103 (N,P) COVARIANCES WERE ESTIMATED FROM DATA AND TNG. 2637 1451 134 33-107 UNCERTAINTIES ESTIMATED FROM DATA AND TNG. 2637 1451 135 2637 1451 136 REFERENCES: 2637 1451 137 (1) C.Y. FU, NSE 100, 61 (1988). 2637 1451 138 (2) C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2637 1451 139 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2637 1451 140 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2637 1451 141 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2637 1451 142 NATIONAL LAB. 2637 1451 143 (3) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2637 1451 144 STATISTICAL MODEL CODE", ORNL/TM-10093 (AUGUST, 1986). 2637 1451 145 (4) S. F. MUGHABGHAB ET AL, NEUTRON RESONANCE PARAMETERS AND 2637 1451 146 THERMAL CROSS SECTIONS, ACADEMIC PRESS, 1981 2637 1451 147 (5) H. BEER, LY DI HONG, AND F. KAPPELER, NUCL. SCI. ENG., 67, 2637 1451 148 184-189 (1978). 2637 1451 149 (6) B. J. ALLEN, J. OF PHYS. G6,381 (1980). 2637 1451 150 (7) TROFIMOV, ATOMNAYA ENERGIYA 58, 278 (1985). DATA TAKEN 2637 1451 151 FROM V. MCLANE, NEUTRON CROSS SECTION CURVES 2637 1451 152 (8) J. B. GARG, PHYS. REV. C18, 1141 (1978). 2637 1451 153 (9) N.C. DYER AND J.H. HAMILTON, J. INORG. NUCL. CHEM., 34, 1119-2637 1451 154 1120 (1972). 2637 1451 155 (10) B.M. BAHAL AND R. PEPELNIK, REPORT GKSS-84-E (1984). 2637 1451 156 (11) D.W. KNEFF, B.M. OLIVER, AND H. FARRAR,IV, NUCL. SCI. ENG., 2637 1451 157 92, 491-524 (1986). 2637 1451 158 1 451 161 12637 1451 159 2 151 75 12637 1451 160 3 102 11 12637 1451 161 2637 1 0 162 2637 0 0 163 2.605800+4 5.743600+1 0 0 1 02637 2151 164 2.605800+4 1.000000+0 0 0 1 02637 2151 165 1.000000-5 4.000000+5 1 3 0 12637 2151 166 0.000000+0 7.200000-1 1 0 3 32637 2151 167 5.743600+1 7.200000-1 0 0 96 162637 2151 168 -4.000000+4 5.000000-1 7.995702+3 6.495400+0 0.000000+0 0.000000+02637 2151 169 -9.998600+1 5.000000-1 1.095900-2 1.096400-2 0.000000+0 0.000000+02637 2151 170 1.038000+4 5.000000-1 2.482000+2 1.246400+0 0.000000+0 0.000000+02637 2151 171 4.355000+4 5.000000-1 6.130400+3 2.200000+0 0.000000+0 0.000000+02637 2151 172 6.720000+4 5.000000-1 8.648700+2 3.000000-1 0.000000+0 0.000000+02637 2151 173 9.149900+4 5.000000-1 9.095600+3 7.740000+0 0.000000+0 0.000000+02637 2151 174 1.212100+5 5.000000-1 2.398900+3 2.300000+0 0.000000+0 0.000000+02637 2151 175 1.786900+5 5.000000-1 2.676800+3 3.000000+0 0.000000+0 0.000000+02637 2151 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NEUTRON DATA 2725 1451 8 ------ENDF-6 FORMAT 2725 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2725 1451 10 DOSIMETRY 2725 1451 11 ******************************************************************2725 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2725 1451 13 * * * * * * *2725 1451 14 R.LAWSON(ANL),R.HOWERTON(LLNL),M.SUGIMOTO(JAERI) 2725 1451 15 2725 1451 16 1 451 21 12725 1451 17 2 151 124 12725 1451 18 3 102 15 12725 1451 19 3 107 18 12725 1451 20 33 107 108 12725 1451 21 2725 1 0 22 2725 0 0 23 2.705900+4 5.842690+1 0 0 1 02725 2151 24 2.705900+4 1.000000+0 0 0 1 02725 2151 25 1.000000-5 1.000000+5 1 2 0 02725 2151 26 3.500000+0 6.200000-1 0 0 2 02725 2151 27 5.842700+1 0.000000+0 0 0 702 1172725 2151 28 -5.000000+2 3.000000+0 4.819400+1 4.774700+1 4.470000-1 0.000000+02725 2151 29 -4.750000+2 4.000000+0 2.970600+0 2.523600+0 4.470000-1 0.000000+02725 2151 30 1.320000+2 4.000000+0 5.616700+0 5.146700+0 4.700000-1 0.000000+02725 2151 31 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2.656000-4 2.583000-4 2.474000-4 2.667000-4272533107 243 1.225000-3 1.078300-3 7.847000-4 4.025000-4 2.940000-4 2.684000-4272533107 244 2.684000-4 2.684000-4 2.684000-4 2.394000-4 1.606000-4 9.550000-5272533107 245 1.820000-5 9.550000-5 1.575000-4 2.352000-4 2.593000-4 2.593000-4272533107 246 2.593000-4 2.604000-4 2.656000-4 2.583000-4 2.474000-4 2.667000-4272533107 247 1.521000-3 1.007700-3 6.435000-4 2.620000-4 2.761000-4 2.761000-4272533107 248 2.761000-4 2.761000-4 2.889000-4 1.790000-4 9.120000-5 1.520000-5272533107 249 9.120000-5 1.755000-4 2.620000-4 2.667000-4 2.667000-4 2.667000-4272533107 250 2.659000-4 2.691000-4 2.558000-4 2.757000-4 2.476000-4 1.444000-3272533107 251 1.083000-3 7.660000-4 3.811000-4 2.914000-4 2.914000-4 2.914000-4272533107 252 2.599000-4 1.744000-4 1.037000-4 1.970000-5 1.037000-4 1.710000-4272533107 253 2.553000-4 2.815000-4 2.815000-4 2.815000-4 2.827000-4 2.884000-4272533107 254 2.804000-4 2.686000-4 2.895000-4 2.500000-3 1.988000-3 1.475000-3272533107 255 5.605000-4 3.540000-4 3.540000-4 3.420000-4 2.295000-4 1.170000-4272533107 256 1.950000-5 1.170000-4 2.250000-4 3.360000-4 3.420000-4 3.420000-4272533107 257 3.420000-4 3.410000-4 3.450000-4 3.230000-4 3.535000-4 3.175000-4272533107 258 3.136000-3 2.478000-3 1.585900-3 6.277000-4 4.295000-4 3.830000-4272533107 259 2.570000-4 1.310000-4 2.180000-5 1.310000-4 2.520000-4 3.763000-4272533107 260 4.149000-4 4.149000-4 4.149000-4 4.166000-4 4.250000-4 4.132000-4272533107 261 3.959000-4 4.267000-4 3.481000-3 2.541100-3 1.601200-3 6.613000-4272533107 262 4.371000-4 3.009000-4 1.610000-4 3.060000-5 1.610000-4 2.950000-4272533107 263 4.295000-4 4.371000-4 4.371000-4 4.371000-4 4.389000-4 4.478000-4272533107 264 4.354000-4 4.171000-4 4.495000-4 3.481000-3 2.541100-3 1.601200-3272533107 265 6.053000-4 3.009000-4 1.610000-4 3.068000-5 1.610000-4 2.950000-4272533107 266 4.295000-4 4.371000-4 4.371000-4 4.371000-4 4.389000-4 4.478000-4272533107 267 4.354000-4 4.171000-4 4.495000-4 3.481000-3 2.541100-3 1.546900-3272533107 268 4.814000-4 1.610000-4 3.060000-5 1.610000-4 2.950000-4 4.295000-4272533107 269 4.371000-4 4.371000-4 4.371000-4 4.389000-4 4.478000-4 4.354000-3272533107 270 4.171000-4 4.495000-4 3.481000-3 2.454900-3 1.384100-3 3.221000-4272533107 271 3.060000-5 1.610000-4 2.950000-4 4.295000-4 4.371000-4 4.371000-4272533107 272 4.371000-4 4.389000-4 4.478000-4 4.354000-4 4.171000-4 4.495000-4272533107 273 3.249000-3 2.151100-3 1.067000-3 3.700000-5 2.220000-5 1.995000-4272533107 274 4.149000-4 4.223000-4 4.223000-4 3.898000-4 3.887000-4 3.933000-4272533107 275 3.739000-4 4.029000-4 3.619000-4 2.601000-3 1.312700-3 8.610000-5272533107 276 -3.779000-4-1.785000-4 1.999000-4 2.907000-4 2.907000-4 2.907000-4272533107 277 2.845000-4 2.815000-4 2.927000-4 3.090000-4 2.590000-4 1.521000-3272533107 278 1.267000-4-1.216000-4-3.705000-4 2.180000-5 1.556000-4 1.556000-4272533107 279 1.556000-4 1.450000-4 1.345000-4 1.599000-4 1.575000-4 1.485000-4272533107 280 1.690000-4 1.267000-4 8.450000-5 3.640000-5 2.960000-5 2.960000-5272533107 281 2.960000-5 4.830000-5 4.480000-5 5.330000-5 5.250000-5 4.950000-5272533107 282 1.521000-3 1.287000-3 1.048300-3 3.112000-4 1.556000-4 1.556000-4272533107 283 1.450000-4 1.345000-4 1.599000-4 1.575000-4 1.485000-4 2.500000-3272533107 284 2.072000-3 1.311000-3 4.560000-4 2.850000-4 2.790000-4 2.760000-4272533107 285 2.870000-4 3.030000-4 3.175000-4 3.136000-3 2.330100-3 1.468300-3272533107 286 5.745000-4 5.902000-4 5.796000-4 5.969000-4 5.656000-4 5.689000-4272533107 287 3.249000-3 2.371700-3 1.494500-3 1.448900-3 1.533800-3 1.542400-3272533107 288 1.496800-3 1.520100-3 3.249000-3 2.371700-3 2.403100-3 2.399100-3272533107 289 2.383700-3 2.417900-3 2.388800-3 3.249000-3 3.251200-3 3.264300-3272533107 290 3.271800-3 3.223900-3 3.257500-3 3.844000-3 3.251200-3 3.253700-3272533107 291 3.256200-3 3.228300-3 4.761000-3 3.225000-3 3.275400-3 3.242300-3272533107 292 6.724000-3 3.229900-3 3.228300-3 1.020100-2 3.206700-3 1.612900-2272533107 293 272533 0 294 2725 0 0 295 0 0 0 296 2.805800+4 5.743800+1 -1 0 0 12825 1451 1 0.000000+0 0.000000+0 0 0 0 62825 1451 2 1.000000+0 0.000000+0 0 0 10 62825 1451 3 0.000000+0 0.000000+0 0 0 225 52825 1451 4 28-NI- 58 ORNL EVAL-OCT89 LARSON, CM PEREY, HETRICK, FU 2825 1451 5 ORNL/TM10219,ENDF-344 DIST-APR90 900216 2825 1451 6 ----ENDF/B-VI MATERIAL 2825 2825 1451 7 -----INCIDENT NEUTRON DATA 2825 1451 8 ------ENDF-6 FORMAT 2825 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2825 1451 10 DOSIMETRY 2825 1451 11 ******************************************************************2825 1451 12 THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE 2825 1451 13 DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1) 2825 1451 14 AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES 2825 1451 15 ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE 2825 1451 16 COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2825 1451 17 REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2825 1451 18 CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF 2825 1451 19 CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).2825 1451 20 2825 1451 21 ----- DESCRIPTION OF FILES 2825 1451 22 (MF-MT) 2825 1451 23 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2825 1451 24 2-151 RESONANCE PARAMETERS ARE USED EXCLUSIVELY TO PROVIDE THE 2825 1451 25 TOTAL, SCATTERING AND CAPTURE CROSS SECTIONS FROM 1.E-5 EV2825 1451 26 TO 812 KEV, EXCEPT FOR CAPTURE FROM 450-812 KEV. THE 2825 1451 27 RESONANCE PARAMETERS ARE TAKEN FROM AN 2825 1451 28 EXTENSIVE RESONANCE PARAMETER ANALYSIS OF TRANSMISSION, 2825 1451 29 SCATTERING AND CAPTURE DATA (6). NEGATIVE ENERGY 2825 1451 30 RESONANCES HAVE BEEN USED TO GIVE THE CORRECT THERMAL 2825 1451 31 VALUES. THE SCATTERING CROSS SECTION IS GIVEN COMPLETELY 2825 1451 32 BY THE RESONANCE PARAMETERS FROM 1.E-5 TO 812 KEV, BUT FOR2825 1451 33 THE CAPTURE CROSS SECTION, AVERAGE RESONANCE PARAMETERS 2825 1451 34 USED FROM 450 TO 812 KEV. THUS, THERE IS A BACKGROUND 2825 1451 35 CONTRIBUTION IN 3/102 FROM 450 TO 812 KEV TO ACCOUNT FOR 2825 1451 36 THE SMALL DIFFERENCE IN CAPTURE CROSS SECTION FROM THE 2825 1451 37 AVERAGE RESONANCE PARAMETERS AND THE AVERAGED EXPERIMENTAL2825 1451 38 DATA (6). THE REICH-MOORE CODE SAMMY (7) WAS USED FOR THE2825 1451 39 RESONANCE PARAMETER ANALYSIS. 2825 1451 40 THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE 2825 1451 41 RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 29.4 B, 2825 1451 42 ELASTIC SCATTERING 24.8 B, AND CAPTURE 4.62 B. 2825 1451 43 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2825 1451 44 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2825 1451 45 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2825 1451 46 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2825 1451 47 3-1 THE TOTAL CROSS SECTION WAS OBTAINED FROM .812 TO 20 MEV 2825 1451 48 BY ANALYSIS OF THE SAME HIGH RESOLUTION MEASUREMENT 2825 1451 49 USED FOR THE RESONANCE PARAMETER ANALYSIS IN 2/151. DATA 2825 1451 50 WERE BINNED AND COMPARED WITH ENERGY AVERAGED DATA FROM 2825 1451 51 ARGONNE (8) AND OTHER DATA FROM CSISRS, AND GOOD 2825 1451 52 AGREEMENT WAS FOUND. 2825 1451 53 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2825 1451 54 SUBTRACTING THE NONELASTIC FROM THE TOTAL 2825 1451 55 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 2825 1451 56 3-28, 3-102, 3-103, 3-104, AND 3-107. 2825 1451 57 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, .. 2825 1451 58 .., 3-58, AND 3-91 2825 1451 59 3-16 (N,2N) CROSS SECTIONS WERE TAKEN FROM THE EVALUATION 2825 1451 60 OF 58NI(N,2N)57NI CROSS SECTIONS BY PAVLIK AND 2825 1451 61 WINKLER (9). THESE CROSS SECTIONS WERE SLIGHTLY LARGER 2825 1451 62 THAN THE TNG CALCULATIONS. (SEE (N,NP) BELOW). 2825 1451 63 3-22 (N,NA) + (N,AN) CROSS SECTIONS WERE CALCULATED BY THE TNG 2825 1451 64 CODE (2,3,4,5). NO DATA AVAILABLE OTHER THAN A EMISSION. 2825 1451 65 3-28 (N,NP) + (N,PN) CROSS SECTIONS WERE CALCULATED BY THE TNG 2825 1451 66 CODE (2,3,4,5) BUT MODIFIED ABOVE EINC OF 13.0 MEV TO FIT 2825 1451 67 THE DATA OF PAVLIK ET AL. (10). THE MODIFICATION OF THE 2825 1451 68 TNG CALCULATION FOR (N,NP) CANCELED OUT THE DIFFERENCE 2825 1451 69 BETWEEN THE TNG CALCULATION FOR (N,2N) AND THE ACTUAL 2825 1451 70 (N,2N) USED (SEE ABOVE) SO THAT THE TNG CALCULATED 2825 1451 71 NONELASTIC CROSS DID NOT CHANGE. 2825 1451 72 3-51 TO 3-58 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE 2825 1451 73 FROM TNG (2,3,4,5) WHICH INCLUDES DIRECT INTERACTION 2825 1451 74 CROSS SECTIONS FROM DWUCK (1) CALCULATIONS. EXTENSIVE 2825 1451 75 COMPARISONS WITH EXPERIMENTAL DATA USED IN THE EVALUATION 2825 1451 76 PROCESS ARE SHOWN IN (5). 2825 1451 77 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG 2825 1451 78 CALCULATED). GUIDANCE FROM (N,XN) EMISSION DATA (5). 2825 1451 79 3-102 (N,G) CAPTURE CROSS SECTION IS PROVIDED ONLY BY RESONANCE 2825 1451 80 PARAMETERS FROM 1.E-5 EV TO 450 KEV. FROM 450 TO 812 KEV 2825 1451 81 A SMALL BACKGROUND FILE IS GIVEN IN 3/102 TO COMBINE WITH 2825 1451 82 THE 2/151 RESULTS. TNG CALCULATIONS WERE NORMALIZED 2825 1451 83 TO THE RESONANCE REGION RESULTS BETWEEN 100-812 KEV, 2825 1451 84 AND USED FROM 812 KEV TO 20 MEV. 2825 1451 85 3-103 (N,P) CROSS SECTIONS UP TO 6 MEV WERE TAKEN FROM THE 2825 1451 86 GLUCS (12) CALCULATION IN WHICH THIS REACTION WAS STUDIED 2825 1451 87 SIMULTANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2825 1451 88 SECTIONS (13). FROM 6 TO 13 MEV A LINE WAS DRAWN 2825 1451 89 THROUGH DATA (SEVERAL MEASURED DATA SETS DISAGREE AND A 2825 1451 90 COMPROMISE WAS CHOSEN). ABOVE 13 MEV, TNG CALC. WERE USED 2825 1451 91 WHICH COMPARED VERY WELL WITH RECENT MEASUREMENTS (5). 2825 1451 92 3-104 (N,D) CROSS SECTION - TAKEN FROM ENDF/B-V (11) WHICH 2825 1451 93 AGREED WELL WITH AVAILABLE DATA. 2825 1451 94 3-107 (N,A) CROSS SECTIONS WERE CALCULATED BY THE 2825 1451 95 TNG CODE (2,3,4,5) AND MODIFIED BETWEEN EINC FROM 7.0 TO 2825 1451 96 13 MEV TO BETTER FIT THE RATIO DATA OF WATTECAMPS (14). 2825 1451 97 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS FOR ELASTIC 2825 1451 98 SCATTERING WERE REVIEWED AND ADOPTED FROM ENDF/B-V (11). 2825 1451 99 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2825 1451 100 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2825 1451 101 PARAMETERS IN 2/151. 2825 1451 102 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2825 1451 103 NEUTRON AND 57NI RESIDUAL, AND ENERGY DEPENDENT YIELDS 2825 1451 104 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2825 1451 105 RAY; TNG CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR2825 1451 106 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2825 1451 107 THE OUTGOING NEUTRON). (N,XN) D-D EMISSION DATA HEAVILY 2825 1451 108 USED TO BENCHMARK THE TNG CALCULATIONS (5). 2825 1451 109 6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2825 1451 110 NEUTRON, ALPHA, AND 54FE RESIDUAL, AND ENERGY DEPENDENT 2825 1451 111 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2825 1451 112 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2825 1451 113 FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY 2825 1451 114 FOR THE OUTGOING NEUTRON; ISOTROPY IS ASSUMED FOR THE 2825 1451 115 ALPHA AND RESIDUAL). (N,XA) D-D EMISSION DATA HEAVILY USED2825 1451 116 TO BENCHMARK THE TNG CALCULATIONS (5). 2825 1451 117 6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2825 1451 118 NEUTRON, P, AND 57CO RESIDUAL, AND ENERGY DEPENDENT YIELD 2825 1451 119 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2825 1451 120 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2825 1451 121 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2825 1451 122 THE OUTGOING NEUTRON). (N,XP) D-D EMISSION DATA HEAVILY 2825 1451 123 USED TO BENCHMARK THE TNG CALCULATIONS (5). 2825 1451 124 6-51 THROUGH 6-58 INELASTIC SCATTERING EXCITING LEVELS; EACH 2825 1451 125 SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON 2825 1451 126 AND 58NI RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR 2825 1451 127 THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED 2825 1451 128 BY DWUCK (1) AND TNG (2,3,4,5)). EXTENSIVE COMPARISONS 2825 1451 129 WITH ANGULAR DISTRIBUTION DATA ARE GIVEN IN (5). 2825 1451 130 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2825 1451 131 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 58NI 2825 1451 132 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2825 1451 133 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; TNG 2825 1451 134 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2825 1451 135 FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE 2825 1451 136 OUTGOING NEUTRON). (N,XN) D-D EMISSION DATA HEAVILY USED 2825 1451 137 TO BENCHMARK THE TNG CALCULATIONS (5). 2825 1451 138 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2825 1451 139 AND 58CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2825 1451 140 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2825 1451 141 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2825 1451 142 PRODUCT. (N,XP) D-D EMISSION DATA HEAVILY USED TO BENCH- 2825 1451 143 MARK THE TNG CALCULATIONS (5). 2825 1451 144 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2825 1451 145 AND 55FE RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2825 1451 146 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2825 1451 147 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2825 1451 148 PRODUCT. (N,XA) D-D EMISSION DATA HEAVILY USED TO BENCH- 2825 1451 149 MARK THE TNG CALCULATIONS (5). 2825 1451 150 12-51 THROUGH 12-58 BRANCHING RATIOS FOR THE LEVELS, COMPILED 2825 1451 151 BY HETRICK ET AL. (5), ARE GIVEN. 2825 1451 152 12-102 (N,G) CAPTURE; MULTIPLICITIES WERE TNG CALCULATED. 2825 1451 153 14-51 THROUGH 14-58 AND 14-102 GAMMA RAY ANGULAR DISTRIBUTIONS 2825 1451 154 ASSUMED TO BE ISOTROPIC. 2825 1451 155 15-102 (N,G) CAPTURE; TNG CALCULATED. 2825 1451 156 -------------------------------------------------------------- 2825 1451 157 UNCERTAINTY FILES 2825 1451 158 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2825 1451 159 REQUIRED BY ENDF/B-VI FORMATS 2825 1451 160 2825 1451 161 33-1 TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 100 EV 2825 1451 162 EXPLICIT FROM 100 EV TO 20 MEV, USING LB=0,1 AND 8. 2825 1451 163 33-2 EXPLICIT FROM 1E-5 10 100 EV, DERIVED FROM 100EV TO 20 MEV2825 1451 164 33-3 DERIVED FROM 1.E-5 TO 450 KEV, EXPLICIT USING LB=1,8 FROM 2825 1451 165 450 KEV TO 20 MEV 2825 1451 166 33-4 DERIVED FROM THRESHOLD TO 20 MEV. 2825 1451 167 33-16 (N,2N) COVARIANCES WERE TAKEN FROM THE EVALUATION OF 2825 1451 168 PAVLIK AND WINKLER (9). 2825 1451 169 33-22 (N,NA) UNCERTAINTIES ESTIMATED FROM CALCULATION AND DATA 2825 1451 170 33-28 (N,NP) COVARIANCES - PAVLIK ET AL.'S (10) DATA USED AS A 2825 1451 171 GUIDE. 2825 1451 172 33-51-91 INELASTIC SCATTERING UNCERTAINTIES ESTIMATED FROM DATA 2825 1451 173 AND CALCULATIONS. 2825 1451 174 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2825 1451 175 ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CALCU- 2825 1451 176 LATIONS FROM TNG FROM 450 KEV TO 20 MEV. 2825 1451 177 33-103 (N,P) COVARIANCES WERE TAKEN FROM THE GLUCS (12) 2825 1451 178 CALCULATION IN WHICH THIS REACTION WAS STUDIED 2825 1451 179 SIMULTANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2825 1451 180 SECTIONS (13). 2825 1451 181 33-104 (N,D) COVARIANCES - CSISRS DATA USED AS A GUIDE. 2825 1451 182 33-107 (N,A) COVARIANCES - CSISRS DATA USED AS A GUIDE. 2825 1451 183 REFERENCES: 2825 1451 184 (1) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2825 1451 185 COLORADO, UNPUBLISHED (1972). 2825 1451 186 (2) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2825 1451 187 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2825 1451 188 MOMENTUM," ORNL/TM-7042 (1980). 2825 1451 189 (3) C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2825 1451 190 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2825 1451 191 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2825 1451 192 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2825 1451 193 NATIONAL LAB. 2825 1451 194 (4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2825 1451 195 STATISTICAL MODEL CODE", ORNL/TM-10093 (AUGUST, 1986). 2825 1451 196 (5) D.M. HETRICK, C.Y. FU, AND D.C. LARSON, "CALCULATED NEUTRON 2825 1451 197 -INDUCED CROSS SECTIONS FOR 58,60NI FROM 1 TO 20 MEV AND 2825 1451 198 COMPARISONS WITH EXPERIMENT," ORNL/TM-10219,ENDF-344 (1987). 2825 1451 199 (6) C.M. PEREY, F.G. PEREY, J.A. HARVEY, N.W. HILL, N.M. LARSON, 2825 1451 200 AND R.L. MACKLIN,"58NI+N TRANSMISSION, DIFFERENTIAL ELASTIC 2825 1451 201 SCATTERING AND CAPTURE MEASUREMENTS AND ANALYSIS FROM 5 TO 2825 1451 202 813 KEV", REPORT ORNL/TM-10841, IN PREPARATION 2825 1451 203 (7) N.M. LARSON AND F.G. PEREY,"USERS GUIDE FOR SAMMY:A COMPUTER 2825 1451 204 MODEL FOR MULTILEVEL R-MATRIX FITS TO NEUTRON DATA USING 2825 1451 205 BAYES' EQUATIONS", ORNL/TM-7485, OAK RIDGE NATIONAL LAB, 19802825 1451 206 "UPDATED USERS'GUIDE FOR SAMMY", ORNL/TM-9179, 1984, 2825 1451 207 ORNL/TM-9179R1, 1985, AND ORNL/TM-9179/R2, 1988. 2825 1451 208 (8) C. BUDTZ-JORGENSEN, P.T. GUENTHER, A.B. SMITH AND J.F. WHALEN2825 1451 209 "FAST-NEUTRON TOTAL AND SCATTERING CROSS SECTIONS OF 58NI", 2825 1451 210 ZEIT. PHYS A306,265,1982 AND REPORT ANL/NDM-61, 1982. 2825 1451 211 (9) A. PAVLIK AND G. WINKLER, "EVALUATION OF THE 58NI(N,2N)57NI 2825 1451 212 CROSS SECTIONS," INDC(AUS)-9/L, INTERNATIONAL NUCLEAR DATA 2825 1451 213 COMMITTEE, INTERNATIONAL ATOMIC ENERGY AGENCY, NUCLEAR DATA 2825 1451 214 SECTION (1983). 2825 1451 215 (10) A. PAVLIK, G. WINKLER, AND M. UHL, NUCL. SCI. ENG. 90, 2825 1451 216 186-202 (1985). 2825 1451 217 (11) M. DIVADEENAM,"NI ELEMENTAL NEUTRON INDUCED REACTION CROSS- 2825 1451 218 SECTION EVALUATION", BNL-NCS-51346, ENDF-294, MARCH 1979. 2825 1451 219 (12) D.M. HETRICK AND C.Y. FU, "GLUCS: A GENERALIZED LEAST SQUARES2825 1451 220 PROGRAM FOR UPDATING CROSS SECTION EVALUATIONS WITH 2825 1451 221 CORRELATED DATA SETS," ORNL/TM-7341, ENDF-303 (OCT,1980). 2825 1451 222 (13) C.Y. FU AND D.M. HETRICK, "EXPERIENCE IN USING THE 2825 1451 223 COVARIANCES OF SOME ENDF/B-V DOSIMETRY CROSS SECTIONS: 2825 1451 224 PROPOSED IMPROVEMENTS AND ADDITIONS OF CROSS-REACTION 2825 1451 225 COVARIANCES," PROC. FOURTH ASTM-EURATOM SYMP. ON REACTOR 2825 1451 226 DOSIMETRY, GAITHERSBURG, MARYLAND, MARCH 22-26,1982, P877, 2825 1451 227 US NATIONAL BUREAU OF STANDARDS. 2825 1451 228 (14) E. WATTECAMPS, GEEL, BELGIUM, PRIVATE COMMUNICATION (1987). 2825 1451 229 1 451 234 12825 1451 230 3 16 9 12825 1451 231 3 103 17 12825 1451 232 33 16 39 12825 1451 233 33 103 87 12825 1451 234 2825 1 0 235 2825 0 0 236 2.805800+4 5.743800+1 0 0 0 02825 3 16 237 -1.219600+7-1.219600+7 0 0 1 182825 3 16 238 18 2 0 0 0 02825 3 16 239 1.240800+7 0.000000+0 1.300000+7 4.541200-3 1.337500+7 1.077700-22825 3 16 240 1.360000+7 1.412000-2 1.380000+7 1.859400-2 1.400000+7 2.288100-22825 3 16 241 1.420000+7 2.658800-2 1.440000+7 3.148600-2 1.460000+7 3.685800-22825 3 16 242 1.485000+7 3.993600-2 1.550000+7 5.063500-2 1.625000+7 6.098800-22825 3 16 243 1.675000+7 6.516600-2 1.725000+7 6.760000-2 1.775000+7 7.063800-22825 3 16 244 1.850000+7 7.668000-2 1.950000+7 8.178500-2 2.000000+7 8.530000-22825 3 16 245 2825 3 0 246 2.805800+4 5.743800+1 0 0 0 02825 3103 247 4.020000+5 4.020000+5 0 0 1 412825 3103 248 41 2 0 0 0 02825 3103 249 1.000000-5 0.000000+0 5.000000+5 0.000000+0 1.000000+6 1.587700-32825 3103 250 1.500000+6 1.375400-2 2.000000+6 3.936300-2 2.500000+6 1.085400-12825 3103 251 3.000000+6 1.945700-1 3.500000+6 2.756300-1 4.000000+6 3.560700-12825 3103 252 4.500000+6 4.055400-1 5.000000+6 4.447300-1 5.500000+6 5.116300-12825 3103 253 6.000000+6 5.800000-1 6.500000+6 6.060000-1 7.000000+6 6.150000-12825 3103 254 7.500000+6 6.220000-1 8.000000+6 6.270000-1 8.500000+6 6.280000-12825 3103 255 9.000000+6 6.270000-1 9.500000+6 6.200000-1 1.000000+7 6.130000-12825 3103 256 1.050000+7 6.050000-1 1.100000+7 5.900000-1 1.150000+7 5.750000-12825 3103 257 1.200000+7 5.520000-1 1.250000+7 5.160700-1 1.300000+7 4.790000-12825 3103 258 1.350000+7 4.265300-1 1.400000+7 3.740700-1 1.450000+7 3.216000-12825 3103 259 1.500000+7 2.777000-1 1.550000+7 2.338000-1 1.600000+7 1.899000-12825 3103 260 1.650000+7 1.669700-1 1.700000+7 1.440300-1 1.750000+7 1.211000-12825 3103 261 1.800000+7 1.114240-1 1.850000+7 1.017480-1 1.900000+7 9.207200-22825 3103 262 1.950000+7 8.239600-2 2.000000+7 7.272000-2 0.000000+0 0.000000+02825 3103 263 2825 3 0 264 2825 0 0 265 2.805800+4 5.743800+1 0 0 0 1282533 16 266 0.000000+0 0.000000+0 0 16 0 2282533 16 267 0.000000+0 0.000000+0 1 5 171 18282533 16 268 1.000000-5 1.240800+7 1.300000+7 1.337500+7 1.360000+7 1.380000+7282533 16 269 1.400000+7 1.420000+7 1.440000+7 1.460000+7 1.485000+7 1.550000+7282533 16 270 1.625000+7 1.675000+7 1.725000+7 1.775000+7 1.850000+7 2.000000+7282533 16 271 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533 16 272 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533 16 273 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533 16 274 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533 16 275 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533 16 276 0.000000+0 0.000000+0 0.000000+0 2.787800-2 2.365500-3 1.659700-3282533 16 277 5.614300-4 5.913600-4 0.000000+0 4.048000-4 8.237000-4 2.692800-4282533 16 278 5.046000-3 2.539700-3 4.487900-3 2.814200-3 2.236900-3 3.178500-3282533 16 279 3.686400-3 4.985600-4 3.526400-4 3.379200-4 1.958400-4 2.796800-4282533 16 280 1.612800-4 1.958400-4 9.369500-4 4.262300-4 3.916700-4 4.243200-4282533 16 281 2.579300-4 3.302400-4 1.512900-3 4.280400-4 2.164800-4 2.091000-4282533 16 282 3.300400-4 1.476000-4 2.091000-4 4.251700-4 1.972100-4 3.763800-4282533 16 283 2.238600-4 1.779400-4 2.644500-4 7.568900-4 1.600800-4 1.626900-4282533 16 284 2.467800-4 9.396000-5 1.577600-4 1.061400-4 5.365000-5 8.874300-5282533 16 285 5.655000-5 4.495000-5 6.235000-5 5.184000-4 6.120100-5 1.269600-4282533 16 286 8.208000-5 8.568000-5 1.464000-4 8.880200-5 1.468800-4 1.216800-4282533 16 287 1.041600-4 1.032000-4 2.601000-4 1.290300-4 3.366000-5 8.092000-5282533 16 288 7.259000-5 2.516000-5 0.000000+0 2.652000-5 0.000000+0 0.000000+0282533 16 289 4.761000-4 7.452000-5 1.212100-4 8.418100-5 4.255000-5 7.038200-5282533 16 290 4.485000-5 3.565000-5 4.945000-5 2.916000-4 4.896000-5 2.635200-4282533 16 291 1.465200-4 2.203200-4 1.684800-4 1.339200-4 2.012400-4 2.601000-4282533 16 292 6.222100-5 3.145000-5 4.335100-5 2.652000-5 2.635000-5 2.924000-5282533 16 293 3.348900-3 1.128500-3 1.213300-3 1.284700-3 8.131400-4 1.154100-3282533 16 294 1.232100-3 6.038500-4 6.782100-4 4.817500-4 6.841400-4 2.340800-3282533 16 295 8.552800-4 8.221200-4 8.114200-4 1.368900-3 6.045100-4 8.720400-4282533 16 296 8.648900-4 6.531700-4 1.664100-3 0.000000+0 0.000000+0 0.000000+0282533 16 297 0.000000+0 0.000000+0 0 8 36 18282533 16 298 1.000000-5 0.000000+0 1.240800+7 0.000000+0 1.300000+7 6.388000-8282533 16 299 1.337500+7 4.757200-8 1.360000+7 3.351500-8 1.380000+7 2.907600-8282533 16 300 1.400000+7 3.015600-8 1.420000+7 2.043000-8 1.440000+7 5.244300-8282533 16 301 1.460000+7 4.401600-8 1.485000+7 4.609200-8 1.550000+7 9.540300-7282533 16 302 1.625000+7 5.092000-7 1.675000+7 1.104500-6 1.725000+7 6.950600-7282533 16 303 1.775000+7 4.795100-7 1.850000+7 1.087200-6 2.000000+7 9.338900-9282533 16 304 282533 0 305 2.805800+4 5.743800+1 0 0 0 1282533103 306 0.000000+0 0.000000+0 0 103 0 4282533103 307 0.000000+0 0.000000+0 1 5 231 21282533103 308 1.000000-5 5.000000+5 1.000000+6 1.500000+6 2.000000+6 2.500000+6282533103 309 3.000000+6 3.500000+6 4.000000+6 4.500000+6 6.000000+6 6.500000+6282533103 310 8.000000+6 9.000000+6 1.000000+7 1.100000+7 1.200000+7 1.250000+7282533103 311 1.300000+7 1.500000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0282533103 312 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533103 313 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0282533103 314 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 2.679400-2282533103 315 6.891700-3 3.064000-3-8.607900-4 2.447300-4 2.297400-4 2.220200-4282533103 316 2.340000-4 1.784000-4 1.670800-4 1.561600-4 1.539700-4 1.472700-4282533103 317 1.474700-4 1.366000-4 1.309700-4 1.309700-4 0.000000+0 0.000000+0282533103 318 1.687200-2 2.485000-3-6.981300-4 1.984900-4 1.863200-4 1.800600-4282533103 319 1.897800-4 1.446900-4 1.355000-4 1.266500-4 1.248700-4 1.194400-4282533103 320 1.196000-4 1.107800-4 1.062200-4 1.062200-4 0.000000+0 0.000000+0282533103 321 5.737100-3-1.611800-3 4.582400-4 4.301800-4 4.157200-4 4.381500-4282533103 322 3.340400-4 3.128300-4 2.924000-4 2.882900-4 2.757600-4 2.761300-4282533103 323 2.557700-4 2.452300-4 2.452400-4 0.000000+0 0.000000+0 4.752200-3282533103 324 7.286900-4 7.819300-4 8.329700-4 6.182100-4 5.331300-4 5.144800-4282533103 325 4.849200-4 5.040000-4 4.747200-4 4.514400-4 5.025100-4 5.288800-4282533103 326 5.288900-4 0.000000+0 0.000000+0 3.047100-3 2.223600-4 9.933200-4282533103 327 5.132800-4 4.849200-4 4.466700-4 4.476800-4 4.330200-4 4.148000-4282533103 328 4.190100-4 4.384300-4 4.484500-4 4.484600-4 0.000000+0 0.000000+0282533103 329 1.294600-3 5.208000-4 6.594100-4 4.467500-4 4.543000-4 4.243300-4282533103 330 4.308800-4 4.119400-4 3.966600-4 4.091200-4 4.155700-4 4.155600-4282533103 331 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2.500000+6 5.520100-2282533103 357 3.000000+6 3.598000-2 3.500000+6 3.834100-2 4.000000+6 3.395400-2282533103 358 4.500000+6 4.036900-2 6.000000+6 3.898100-2 6.500000+6 3.651100-2282533103 359 8.000000+6 3.689800-2 9.000000+6 3.889800-2 1.000000+7 4.559700-2282533103 360 1.100000+7 1.022100-1 1.200000+7 1.598800-1 1.250000+7 1.598800-1282533103 361 1.300000+7 1.850300-1 1.500000+7 1.683300-1 2.000000+7 0.000000+0282533103 362 0.000000+0 0.000000+0 0 1 82 41282533103 363 1.000000-5 0.000000+0 5.000000+5 2.411400-2 1.000000+6 1.518500-2282533103 364 1.500000+6 5.163400-3 2.000000+6 4.276900-3 2.500000+6 2.742400-3282533103 365 3.000000+6 1.165100-3 3.500000+6 1.323000-3 4.000000+6 1.037600-3282533103 366 4.500000+6 1.290700-3 5.000000+6 1.290700-3 5.500000+6 1.290700-3282533103 367 6.000000+6 1.367500-3 6.500000+6 6.718500-4 7.000000+6 6.718500-4282533103 368 7.500000+6 6.718500-4 8.000000+6 7.474300-4 8.500000+6 7.474300-4282533103 369 9.000000+6 8.987400-4 9.500000+6 8.987400-4 1.000000+7 1.440800-3282533103 370 1.050000+7 1.440800-3 1.100000+7 6.488100-3 1.150000+7 6.488100-3282533103 371 1.200000+7 2.300500-2 1.250000+7 2.300500-2 1.300000+7 6.470500-3282533103 372 1.350000+7 6.470500-3 1.400000+7 6.470500-3 1.450000+7 6.470500-3282533103 373 1.500000+7 5.355500-3 1.550000+7 5.355500-3 1.600000+7 5.355500-3282533103 374 1.650000+7 5.355500-3 1.700000+7 5.355500-3 1.750000+7 5.355500-3282533103 375 1.800000+7 5.355500-3 1.850000+7 5.355500-3 1.900000+7 5.355500-3282533103 376 1.950000+7 5.355500-3 2.000000+7 0.000000+0 0.000000+0 0.000000+0282533103 377 0.000000+0 0.000000+0 0 8 82 41282533103 378 1.000000-5 0.000000+0 5.000000+5 6.04710-15 1.000000+6 4.253200-9282533103 379 1.500000+6 1.085300-7 2.000000+6 7.363200-7 2.500000+6 3.589800-6282533103 380 3.000000+6 4.900900-6 3.500000+6 1.116800-5 4.000000+6 1.461700-5282533103 381 4.500000+6 2.358600-5 5.000000+6 2.836500-5 5.500000+6 3.754000-5282533103 382 6.000000+6 4.905700-5 6.500000+6 2.478700-5 7.000000+6 2.459900-5282533103 383 7.500000+6 2.544700-5 8.000000+6 2.884400-5 8.500000+6 2.687800-5282533103 384 9.000000+6 3.113400-5 9.500000+6 3.416300-5 1.000000+7 5.458400-5282533103 385 1.050000+7 5.212600-5 1.100000+7 2.275100-4 1.150000+7 2.180700-4282533103 386 1.200000+7 7.279600-4 1.250000+7 6.807500-4 1.300000+7 1.811800-4282533103 387 1.350000+7 1.574700-4 1.400000+7 1.317000-4 1.450000+7 1.038200-4282533103 388 1.500000+7 6.401800-5 1.550000+7 5.004400-5 1.600000+7 3.960900-5282533103 389 1.650000+7 3.202800-5 1.700000+7 2.674400-5 1.750000+7 2.170800-5282533103 390 1.800000+7 1.885400-5 1.850000+7 1.639700-5 1.900000+7 1.466800-5282533103 391 1.950000+7 1.321100-5 2.000000+7 0.000000+0 0.000000+0 0.000000+0282533103 392 282533 0 393 2825 0 0 394 0 0 0 395 2.806000+4 5.941600+1 -1 0 0 12831 1451 1 0.000000+0 0.000000+0 0 0 0 62831 1451 2 1.000000+0 0.000000+0 0 0 10 62831 1451 3 0.000000+0 0.000000+0 0 0 196 32831 1451 4 28-NI- 60 ORNL EVAL-OCT89 LARSON, CM PEREY, HETRICK, FU 2831 1451 5 ORNL/TM10219,ENDF-344 DIST-APR90 900220 2831 1451 6 ----ENDF/B-VI MATERIAL 2831 2831 1451 7 -----INCIDENT NEUTRON DATA 2831 1451 8 ------ENDF-6 FORMAT 2831 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2831 1451 10 DOSIMETRY 2831 1451 11 ******************************************************************2831 1451 12 THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE 2831 1451 13 DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1) 2831 1451 14 AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES 2831 1451 15 ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE 2831 1451 16 COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2831 1451 17 REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2831 1451 18 CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF 2831 1451 19 CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).2831 1451 20 2831 1451 21 ----- DESCRIPTION OF FILES 2831 1451 22 (MF-MT) 2831 1451 23 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2831 1451 24 2-151 RESONANCE PARAMETERS -- TAKEN FROM REF (6). THE NEGATIVE 2831 1451 25 ENERGY RESONANCES WERE ADJUSTED BY THE AUTHOR, C. M. PEREY2831 1451 26 TO GIVE THE PROPER THERMAL SCATTERING AND CAPTURE. USING 2831 1451 27 THESE MODIFIED RESULTS, NO BACKGROUND IS REQUIRED IN 3/1 2831 1451 28 IE, THE TOTAL, SCATTERING AND CAPTURE IS GIVEN COMPLETELY 2831 1451 29 BY THE RESONANCE PARAMETERS FROM 1.E-5 EV TO 450 KEV. 2831 1451 30 RESONANCE PARAMETERS ARE FROM A REICH-MOORE ANALYSIS WITH 2831 1451 31 THE CODE SAMMY (REF 7). 2831 1451 32 THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE 2831 1451 33 RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 3.9 B, 2831 1451 34 ELASTIC SCATTERING 0.98 B, AND CAPTURE 2.92 B. 2831 1451 35 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2831 1451 36 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2831 1451 37 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2831 1451 38 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2831 1451 39 3-1 TOTAL CROSS SECTION -- FROM 450 KEV TO 1611 KEV THE DATA 2831 1451 40 OF REF (8) WERE USED. FROM 1611 KEV TO 20 MEV THE DATA 2831 1451 41 OF REF (9) WERE USED. SEVERAL LOWER RESOLUTION DATA SETS 2831 1451 42 WERE ENERGY-BINNED AND CHECKED WITH BINNED DATA FROM 8,9. 2831 1451 43 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2831 1451 44 SUBTRACTING THE NONELASTIC FROM THE TOTAL 2831 1451 45 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 2831 1451 46 3-28, 3-102, 3-103, 3-104, AND 3-107. 2831 1451 47 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, .. 2831 1451 48 .., 3-61, AND 3-91 2831 1451 49 3-16 (N,2N) CROSS SECTIONS WERE CALCULATED BY THE TNG 2831 1451 50 CODE (2,3,4,5), NO 60NI DATA AVAILABLE. 2831 1451 51 3-22 (N,NA) CROSS SECTIONS WERE CALCULATED BY THE TNG 2831 1451 52 CODE (2,3,4,5). NO 60NI DATA AVAILABLE OTHER THAN A EMISS 2831 1451 53 3-28 (N,NP) CROSS SECTIONS WERE CALCULATED BY THE TNG 2831 1451 54 CODE (2,3,4,5). NO 60NI DATA AVAILABLE OTHER THAN P EMISS 2831 1451 55 3-51 TO 3-61 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE 2831 1451 56 FROM TNG (2,3,4,5) WHICH INCLUDES DIRECT INTERACTION 2831 1451 57 CROSS SECTIONS FROM DWUCK (1) CALCULATIONS. CALCULATIONS 2831 1451 58 COMPARED WITH EXTENSIVE EXPERIMENTAL DATA (5). 3-51 CROSS2831 1451 59 SECTIONS ABOVE 4.0 MEV ADJUSTED TO BETTER FIT RECENT 2831 1451 60 MEASURED DATA. 2831 1451 61 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG 2831 1451 62 CALCULATED). 2831 1451 63 3-102 (N,G) CAPTURE CROSS SECTION TAKEN FROM RESONANCE PARAMETER2831 1451 64 FROM 1.E-5 EV TO 450 KEV. TNG CALCULATIONS WERE 2831 1451 65 NORMALIZED TO THE AVERAGED CROSS SECTION OBTAINED FROM THE2831 1451 66 RESONANCE PARAMETERS FROM 100-450 KEV. THESE NORMALIZED 2831 1451 67 TNG RESULTS WERE USED FROM 450 KEV TO 20 MEV. 2831 1451 68 3-103 (N,P) CROSS SECTIONS FOR EN LESS THAN OR EQUAL TO 12 MEV 2831 1451 69 WERE TAKEN FROM A CURVE DRAWN THROUGH THE DATA OF VONACH 2831 1451 70 ET AL. (10). THIS CURVE WAS CONNECTED TO THE ANL DATA OF 2831 1451 71 GREENWOOD (11) FROM 14.5 TO 15.0 MEV. FROM 15.0 TO 20.0 2831 1451 72 MEV THE DATA OF PAULSEN AND LISKIEN (12) WERE USED 2831 1451 73 AS A GUIDE. THE NEW DATA BY VONACH ET AL. (10) 2831 1451 74 RESULT IN AN INTEGRAL VALUE CLOSE TO THE 2.39 VALUE OF 2831 1451 75 MANNHART (13). 2831 1451 76 3-104 (N,D) CROSS SECTION - ENDF/B-V (14) VALUES WERE NORMALIZED2831 1451 77 TO THE DATA OF GRIMES ET AL. (15). 2831 1451 78 3-107 (N,A) CROSS SECTIONS WERE CALCULATED BY THE 2831 1451 79 TNG CODE (2,3,4,5). LIMITED DATA OTHER THAN A EMISSION. 2831 1451 80 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR 2831 1451 81 ELASTIC SCATTERING ARE FROM ENDF/B-V. AGREE WELL WITH DATA2831 1451 82 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2831 1451 83 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2831 1451 84 PARAMETERS IN 2/151. 2831 1451 85 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2831 1451 86 NEUTRON AND 59NI RESIDUAL, AND ENERGY DEPENDENT YIELD 2831 1451 87 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2831 1451 88 RAY; TNG CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR2831 1451 89 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2831 1451 90 THE OUTGOING NEUTRON). COMPARES WELL WITH N EMISSION DATA2831 1451 91 WHEN COMBINED WITH N FROM 6/22, 6/28 AND 6/91. 2831 1451 92 6-22 (N,NA)+(N,AN); INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2831 1451 93 NEUTRON, ALPHA, AND 56FE RESIDUAL, AND ENERGY DEPENDENT 2831 1451 94 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2831 1451 95 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2831 1451 96 FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY 2831 1451 97 FOR THE OUTGOING NEUTRON; ISOTROPY IS ASSUMED FOR THE 2831 1451 98 ALPHA AND RESIDUAL). 2831 1451 99 6-28 (N,NP)+(N,PN); INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2831 1451 100 NEUTRON, P, AND 59CO RESIDUAL, AND ENERGY DEPENDENT YIELD 2831 1451 101 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2831 1451 102 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2831 1451 103 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2831 1451 104 THE OUTGOING NEUTRON). 2831 1451 105 6-51 THROUGH 6-61 INELASTIC SCATTERING EXCITING LEVELS; EACH 2831 1451 106 SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON 2831 1451 107 AND 60NI RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR 2831 1451 108 THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED 2831 1451 109 BY DWUCK (1) AND TNG (2,3,4,5)), COMPARED WITH DATA IN (5)2831 1451 110 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2831 1451 111 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 60NI 2831 1451 112 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2831 1451 113 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; TNG 2831 1451 114 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2831 1451 115 FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE 2831 1451 116 OUTGOING NEUTRON). 2831 1451 117 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2831 1451 118 AND 60CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2831 1451 119 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2831 1451 120 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2831 1451 121 PRODUCT. COMPARES WELL WITH P EMISSION DATA WHEN COMBINED2831 1451 122 WITH 6/28. 2831 1451 123 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2831 1451 124 AND 57FE RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2831 1451 125 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2831 1451 126 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2831 1451 127 PRODUCT. COMPARES WELL WITH A EMISSION DATA WHEN COMBINED2831 1451 128 WITH 6/22. 2831 1451 129 12-51 THROUGH 12-61 BRANCHING RATIOS FOR THE LEVELS, COMPILED 2831 1451 130 BY HETRICK ET AL. (5), ARE GIVEN. 2831 1451 131 12-102 (N,G) CAPTURE; MULTIPLICITIES WERE TNG CALCULATED. 2831 1451 132 14-51 THROUGH 14-58 AND 14-102 GAMMA RAY ANGULAR DISTRIBUTIONS 2831 1451 133 ASSUMED TO BE ISOTROPIC. 2831 1451 134 15-102 (N,G) CAPTURE; TNG CALCULATED 2831 1451 135 2831 1451 136 -------------------------------------------------------------- 2831 1451 137 UNCERTAINTY FILES 2831 1451 138 AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS 2831 1451 139 REQUIRED BY ENDF/B-VI. 2831 1451 140 2831 1451 141 33-1 UNCERTAINTIES ARE DERIVED FROM 1.E-5 TO 100 EV. FROM 100 2831 1451 142 EV TO 20 MEV THEY ARE EXPLICIT, USING LB=0,1 AND 8. 2831 1451 143 33-2 FROM 1.E-5 TO 100 EV, UNCERTAINTIES ARE EXPLICIT, BASED 2831 1451 144 UPON THERMAL UNCERTAINTY AND OTHER DATA. FROM 100 EV TO 2831 1451 145 20 MEV THE FILES ARE DERIVED. 2831 1451 146 33-3 FROM 1.E-5 TO 450 KEV UNCERTAINTIES ARE DERIVED. FROM 4502831 1451 147 KEV TO 20 MEV UNCERTAINTIES ARE EXPLICIT, USING LB=1 AND 82831 1451 148 33-4 UNCERTAINTIES ARE ALL DERIVED. 2831 1451 149 33-16 COVARIANCES WERE ESTIMATED AS NO DATA AVAILABLE. 2831 1451 150 33-22 UNCERTAINTIES FOR (N,NA) ARE EXPLICIT, ESTIMATED FROM TNG 2831 1451 151 33-28 UNCERTAINTIES FOR (N,NP) ARE EXPLICIT, ESTIMATED FROM DATA2831 1451 152 AND CALCULATION. 2831 1451 153 33-51-91 UNCERTAINTIES FOR INELASTIC SCATTERING ARE EXPLICIT, 2831 1451 154 BASED ON DATA AND CALCULATION UNCERTAINTIES. 2831 1451 155 33-102 UNCERTAINTIES ARE EXPLICIT, BASED ON THERMAL DATA AT LOW 2831 1451 156 ENERGIES, AND CALCULATED RESULTS ABOVE 450 KEV. 2831 1451 157 33-103 COVARIANCES - DATA OF VONACH ET AL. (10) AND CSISRS DATA 2831 1451 158 USED AS A GUIDE. 2831 1451 159 33-104 UNCERTAINTIES ESTIMATED, BASED ON DATA 2831 1451 160 33-107 UNCERTAINTIES EXPLICIT, BASED ON DATA AND TNG RESULTS. 2831 1451 161 2831 1451 162 REFERENCES: 2831 1451 163 (1) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2831 1451 164 COLORADO, UNPUBLISHED (1972). 2831 1451 165 (2) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2831 1451 166 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2831 1451 167 MOMENTUM," ORNL/TM-7042 (1980). 2831 1451 168 (3) C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2831 1451 169 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2831 1451 170 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2831 1451 171 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2831 1451 172 NATIONAL LAB. 2831 1451 173 (4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2831 1451 174 STATISTICAL MODEL CODE", ORNL/TM-10093 (AUGUST, 1986). 2831 1451 175 (5) D.M. HETRICK, C.Y. FU, AND D.C. LARSON, "CALCULATED NEUTRON 2831 1451 176 -INDUCED CROSS SECTIONS FOR 58,60NI FROM 1 TO 20 MEV AND 2831 1451 177 COMPARISONS WITH EXPERIMENT," ORNL/TM-10219,ENDF-344 (1987). 2831 1451 178 (6) C.M. PEREY, J.A. HARVEY, R.L. MACKLIN, F.G. PEREY AND R.R. 2831 1451 179 WINTERS, PHYS. REV. C27, 2556 (1983) 2831 1451 180 (7) N.M. LARSON AND F.G. PEREY,"USERS GUIDE FOR SAMMY:A COMPUTER 2831 1451 181 MODEL FOR MULTILEVEL R-MATRIX FITS TO NEUTRON DATA USING 2831 1451 182 BAYES' EQUATIONS, ORNL/TM-7485, OAK RIDGE NATIONAL LAB, 1980;2831 1451 183 "UPDATED USERS' GUIDE FOR SAMMY", ORNL/TM-9179, 1984, 2831 1451 184 ORNL/TM-9179/R1, 1985, AND ORNL/TM-9179/R2, 1988. 2831 1451 185 (8) J.A. HARVEY, ORELA, 80-M NE110 TA TARGET DATA, SENT TO NNDC 2831 1451 186 (9) J.A. HARVEY, ORELA, 80-M NE110 BE BLOCK DATA, SENT TO NNDC 2831 1451 187 (10) H. VONACH, M. WAGNER, AND R.C. HAIGHT, "NEUTRON CROSS 2831 1451 188 SECTIONS OF 58NI AND 60NI FOR 8-12 MEV NEUTRONS," PRIVATE 2831 1451 189 COMMUNICATION , 1989. 2831 1451 190 (11) L.R. GREENWOOD, DOE-ER-0046-21, 15,(MAY 1985) 2831 1451 191 (12) A. PAULSEN AND H. LISKIEN, NUKLEONIK 10, 91 (JULY 1967) 2831 1451 192 (13) W. MANNHART, PROC. FOURTH ASTM-EURATOM SYMP. ON REACTOR 2831 1451 193 DOSIMETRY, VOL. II, 637 (MARCH 22-26, 1982) 2831 1451 194 (14) M. DIVADEENAM, "NI ELEMENTAL NEUTRON INDUCED REACTION CROSS- 2831 1451 195 SECTION EVALUATION", BNL-NCS-51346, ENDF-294 (MARCH 1979). 2831 1451 196 (15) S.M. GRIMES, R.C. HAIGHT, K.R. ALVAR, H.H. BARSCHALL, AND 2831 1451 197 R.R. BORCHERS, PHYS. REV. C, 19(6), 2127 (JUNE 1979) 2831 1451 198 2831 1451 199 2831 1451 200 1 451 203 12831 1451 201 3 103 12 12831 1451 202 33 103 14 12831 1451 203 2831 1 0 204 2831 0 0 205 2.806000+4 5.941600+1 0 0 0 02831 3103 206 -2.042000+6-2.042000+6 0 0 1 252831 3103 207 25 2 0 0 0 02831 3103 208 2.076400+6 0.000000+0 3.000000+6 7.637400-8 3.500000+6 2.328300-52831 3103 209 4.000000+6 3.720600-4 4.500000+6 2.182500-3 5.000000+6 7.875900-32831 3103 210 5.500000+6 1.500000-2 6.000000+6 2.350000-2 6.500000+6 3.400000-22831 3103 211 7.000000+6 4.400000-2 7.500000+6 5.600000-2 8.000000+6 6.500000-22831 3103 212 8.500000+6 7.500000-2 9.000000+6 8.300000-2 9.500000+6 9.250000-22831 3103 213 1.000000+7 1.010000-1 1.100000+7 1.180000-1 1.200000+7 1.350000-12831 3103 214 1.300000+7 1.460000-1 1.350000+7 1.480000-1 1.400000+7 1.465000-12831 3103 215 1.450000+7 1.420200-1 1.600000+7 9.819000-2 1.750000+7 7.700000-22831 3103 216 2.000000+7 5.800000-2 0.000000+0 0.000000+0 0.000000+0 0.000000+02831 3103 217 2831 3 0 218 2831 0 0 219 2.806000+4 5.941600+1 0 0 0 1283133103 220 0.000000+0 0.000000+0 0 103 0 4283133103 221 0.000000+0 0.000000+0 0 0 6 3283133103 222 1.000000-5 0.000000+0 2.076400+6 1.00000-11 2.000000+7 0.000000+0283133103 223 0.000000+0 0.000000+0 0 1 6 3283133103 224 1.000000-5 0.000000+0 2.076400+6 5.000000-3 2.000000+7 0.000000+0283133103 225 0.000000+0 0.000000+0 0 1 16 8283133103 226 1.000000-5 0.000000+0 2.076400+6 7.200000-2 5.000000+6 1.800000-2283133103 227 7.000000+6 1.012500-2 1.000000+7 1.012500-2 1.200000+7 4.500000-3283133103 228 1.600000+7 1.800000-2 2.000000+7 0.000000+0 0.000000+0 0.000000+0283133103 229 0.000000+0 0.000000+0 0 8 16 8283133103 230 1.000000-5 0.000000+0 2.076400+6 4.66640-17 5.000000+6 1.240600-7283133103 231 7.000000+6 2.178000-6 1.000000+7 1.147600-5 1.200000+7 9.112500-6283133103 232 1.600000+7 1.928300-5 2.000000+7 0.000000+0 0.000000+0 0.000000+0283133103 233 283133 0 234 2831 0 0 235 0 0 0 236 2.906300+4 6.238900+1 1 0 0 12925 1451 1 0.000000+0 0.000000+0 0 0 0 62925 1451 2 1.000000+0 0.000000+0 0 0 10 62925 1451 3 0.000000+0 0.000000+0 0 0 228 72925 1451 4 29-CU- 63 ORNL EVAL-NOV89 HETRICK,FU,LARSON 2925 1451 5 ORNL/TM-9083,ENDF-337 DIST-APR90 900409 2925 1451 6 ----ENDF/B-VI MATERIAL 2925 2925 1451 7 -----INCIDENT NEUTRON DATA 2925 1451 8 ------ENDF-6 FORMAT 2925 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2925 1451 10 DOSIMETRY 2925 1451 11 ******************************************************************2925 1451 12 THIS WORK EMPLOYED SEVERAL NUCLEAR MODEL CODES INCLUDING THE 2925 1451 13 OPTICAL-MODEL CODE GENOA (1), THE DISTORTED WAVE BORN 2925 1451 14 APPROXIMATION (DWBA) PROGRAM DWUCK (2), AND THE HAUSER-FESHBACH 2925 1451 15 CODE TNG (3,4). THE TNG CODE PROVIDES ENERGY AND ANGULAR 2925 1451 16 DISTRIBUTIONS OF PARTICLES EMITTED IN THE COMPOUND AND PRE- 2925 1451 17 COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL REACTIONS, AND 2925 1451 18 MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE CONTENTS OF 2925 1451 19 THIS EVALUATION AND EXTENSIVE COMPARISONS OF CALCULATIONS WITH 2925 1451 20 EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5). 2925 1451 21 2925 1451 22 ----- DESCRIPTION OF FILES 2925 1451 23 (MF-MT) 2925 1451 24 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2925 1451 25 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB(6). POINT2925 1451 26 WISE RECONSTRUCTION COMPARED WITH DATA (7) SHOWED POORER 2925 1451 27 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2925 1451 28 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2925 1451 29 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2925 1451 30 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2925 1451 31 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2925 1451 32 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2925 1451 33 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2925 1451 34 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2925 1451 35 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2925 1451 36 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2925 1451 37 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2925 1451 38 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2925 1451 39 3-1 THE TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS 2925 1451 40 FROM 1.E-5 EV TO 99.5 KEV. FROM 1.E-5 TO 1 EV, -0.9B IS 2925 1451 41 GIVEN TO REDUCE THE ELASTIC AND GIVE THE CORRECT TOTAL 2925 1451 42 THERMAL CROSS SECTION (9.6 B). FROM 1 EV TO 170 EV THIS 2925 1451 43 GOES LINEARLY TO ZERO. A SMALL CONTRIBUTION FROM 3/102 IS2925 1451 44 REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151 2925 1451 45 REPRODUCES THE CAPTURE DATA. FROM 99.5 KEV TO 1.12 MEV 2925 1451 46 CU63 DATA FROM (7) IS USED, AFTER APPROPRIATE AVERAGING. 2925 1451 47 ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE. FROM 1.12 TO 2925 1451 48 4.0 MEV,NAT CU DATA OF PEREY (8) USED IN V5 IS RETAINED. 2925 1451 49 FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL (9) IS 2925 1451 50 AVERAGED AND USED. COMPARISONS FROM 1.2 TO 4.5 2925 1451 51 MEV WITH AVERAGED ARGONNE DATA FOR NAT CU (10) SHOW 2925 1451 52 1% AGREEMENT. 2925 1451 53 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2925 1451 54 SUBTRACTING THE NONELASTIC (3-3) FROM THE TOTAL. THE 2925 1451 55 THERMAL VALUE OF 5.1 B IS REPRODUCED. 2925 1451 56 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 2925 1451 57 3-28, 3-102, 3-103, 3-104, 3-106, AND 3-107. 2925 1451 58 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, .. 2925 1451 59 .., 3-72, AND 3-91 2925 1451 60 3-16 (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS (11) 2925 1451 61 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2925 1451 62 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2925 1451 63 SECTIONS (12). 2925 1451 64 3-22 (N,NA) CROSS SECTIONS WERE CALCULATED BY THE TNG 2925 1451 65 CODE (3,4,5). NO DATA AVAILABLE OTHER THAN A-EMISSION. 2925 1451 66 3-28 (N,NP) CROSS SECTIONS WERE CALCULATED BY THE TNG 2925 1451 67 CODE (3,4,5). DATA SCARCE FOR THIS REACTION (SEE (5)). 2925 1451 68 3-51 TO 3-72 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE 2925 1451 69 FROM TNG (3,4,5) WHICH INCLUDES DIRECT INTERACTION 2925 1451 70 CROSS SECTIONS FROM DWUCK (2) CALCULATIONS. EXTENSIVE 2925 1451 71 COMPARISONS WITH EXPERIMENTAL DATA USED IN THE EVALUATION 2925 1451 72 PROCESS ARE SHOWN IN (5). 2925 1451 73 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG 2925 1451 74 CALCULATED). GUIDANCE FROM (N,XN) EMISSION DATA (5). 2925 1451 75 3-102 (N,G) DATA TAKEN FROM RESONANCE PARAMETERS FROM 1.E-5 EVTO2925 1451 76 99.5 KEV. A SMALL BACKGROUND IS GIVEN HERE WHICH WHEN 2925 1451 77 ADDED TO THE RESONACE CONTRIBUTION REPRODUCES EXPERIMENTAL2925 1451 78 DATA,INCLUDING THE THERMAL VALUE OF 4.50B. V5 DATA FROM 2925 1451 79 99.5 KEV TO 20.0 MEV WERE REPLACED BY POINTS ON A CURVE 2925 1451 80 DRAWN THROUGH DATA FROM THE CSISRS LIBRARY (5,13); RESULTS2925 1451 81 COMPARABLE TO EYE GUIDE IN REF (14). 2925 1451 82 3-103 (N,P) CROSS SECTIONS WERE CALCULATED BY THE TNG 2925 1451 83 CODE (3,4,5). 2925 1451 84 3-104 (N,D) CROSS SECTION - SHAPE OF (N,P) CROSS SECTION USED 2925 1451 85 NORMALIZED TO AVAILABLE DATA (5). 2925 1451 86 3-106 (N,3HE) CROSS SECTION - SHAPE OF (N,A) CROSS SECTION 2925 1451 87 USED NORMALIZED TO AVAILABLE DATA (15). 2925 1451 88 3-107 (N,A) CROSS SECTIONS WERE TAKEN FROM THE GLUCS (11) 2925 1451 89 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2925 1451 90 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2925 1451 91 SECTIONS (12). 2925 1451 92 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR 2925 1451 93 ELASTIC SCATTERING ARE RESULTS OF FITTING DATA WITH GENOA.2925 1451 94 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2925 1451 95 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2925 1451 96 PARAMETERS IN 2/151. 2925 1451 97 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2925 1451 98 NEUTRON AND 62CU RESIDUAL, AND ENERGY DEPENDENT YIELD 2925 1451 99 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2925 1451 100 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2925 1451 101 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2925 1451 102 THE OUTGOING NEUTRON). (N,XN) AND (N,XG) D-D EMISSION 2925 1451 103 DATA HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS (5). 2925 1451 104 6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2925 1451 105 NEUTRON, ALPHA, AND 59CO RESIDUAL, AND ENERGY DEPENDENT 2925 1451 106 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2925 1451 107 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2925 1451 108 FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY 2925 1451 109 FOR THE OUTGOING NEUTRON). (N,XA) AND (N,XG) D-D EMISSION2925 1451 110 DATA HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS (5). 2925 1451 111 6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2925 1451 112 NEUTRON AND 62NI RESIDUAL, AND ENERGY DEPENDENT YIELD 2925 1451 113 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2925 1451 114 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2925 1451 115 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2925 1451 116 THE OUTGOING NEUTRON). (N,XP) AND (N,XG) D-D EMISSION 2925 1451 117 DATA HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS (5). 2925 1451 118 6-51 THROUGH 6-72 INELASTIC SCATTERING EXCITING LEVELS; EACH 2925 1451 119 SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON 2925 1451 120 AND 63CU RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR 2925 1451 121 THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED 2925 1451 122 BY DWUCK (2) AND TNG (3,4,5)). EXTENSIVE COMPARISONS 2925 1451 123 WITH ANGULAR DISTRIBUTION DATA ARE GIVEN IN REF. (5). 2925 1451 124 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2925 1451 125 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 63CU 2925 1451 126 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2925 1451 127 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; CALCULATED2925 1451 128 NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH (ANGULAR 2925 1451 129 DISTRIBUTIONS ARE GIVEN ONLY FOR THE OUTGOING NEUTRON). 2925 1451 130 (N,XN) AND (N,XG) D-D EMISSION DATA HEAVILY USED TO BENCH-2925 1451 131 MARK THE TNG CALCULATIONS (5). 2925 1451 132 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2925 1451 133 AND 63NI RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2925 1451 134 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2925 1451 135 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2925 1451 136 PRODUCT. (N,XP) AND (N,XG) D-D EMISSION DATA USED TO 2925 1451 137 BENCHMARK THE TNG CALCULATIONS (5). 2925 1451 138 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2925 1451 139 AND 60CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2925 1451 140 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2925 1451 141 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2925 1451 142 PRODUCT. (N,XA) AND (N,XG) D-D EMISSION DATA USED TO 2925 1451 143 BENCHMARK THE TNG CALCULATIONS (5). 2925 1451 144 12-51 THROUGH 12-72 BRANCHING RATIOS FOR THE LEVELS, COMPILED 2925 1451 145 BY HETRICK ET AL. (5) ARE GIVEN. 2925 1451 146 12-102 (N,G) CAPTURE; THERMAL VALUES TAKEN FROM DELFINI ET AL. 2925 1451 147 (16). HIGHER-ENERGY VALUES BASED ON TNG CALCULATIONS 2925 1451 148 USING PRIMARY BRANCHING RATIOS OF DELFINI ET AL. FOR 2925 1451 149 S-WAVES 2925 1451 150 14-51 THROUGH 14-72 GAMMA-RAY ANGULAR DISTRIBUTIONS ASSUMED TO 2925 1451 151 BE ISOTROPIC. 2925 1451 152 14-102 (N,G) CAPTURE; ISOTROPIC DISTRIBUTIONS ASSUMED 2925 1451 153 15-102 (N,G) CAPTURE; AS IN 12-102 2925 1451 154 -------------------------------------------------------------- 2925 1451 155 UNCERTAINTY FILES 2925 1451 156 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2925 1451 157 REQUIRED BY ENDF/B-VI FORMATS 2925 1451 158 2925 1451 159 33-1 TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV 2925 1451 160 EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8. 2925 1451 161 33-2 EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV2925 1451 162 33-3 DERIVED FROM 1.E-5 TO 99.5 KEV, EXPLICIT USING LB=1,8 FROM2925 1451 163 99.5 KEV TO 20 MEV. 2925 1451 164 33-4 DERIVED FROM THRESHOLD TO 20 MEV. 2925 1451 165 33-16 (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS (11) 2925 1451 166 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2925 1451 167 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2925 1451 168 SECTIONS (12). 2925 1451 169 33-22 (N,NA) UNCERTAINTIES ESTIMATED FROM TNG CALCULATION. 2925 1451 170 33-28 (N,NP) COVARIANCES ESTIMATED FROM TNG CALCULATION. 2925 1451 171 33-51-91 INELASTIC SCATTERING UNCERTAINTIES ESTIMATED FROM DATA 2925 1451 172 AND CALCULATIONS. 2925 1451 173 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2925 1451 174 ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 2925 1451 175 DATA (5,13,14) FROM 99.5 KEV TO 20 MEV. 2925 1451 176 33-103 (N,P) COVARIANCES - ESTIMATED AS NO DATA AVAILABLE. 2925 1451 177 33-104 (N,D) COVARIANCES - ESTIMATED, BASED ON DATA. 2925 1451 178 33-106 (N,3HE) COVARIANCES - ESTIMATED, BASED ON DATA. 2925 1451 179 33-107 (N,A) COVARIANCES WERE TAKEN FROM THE GLUCS (11) 2925 1451 180 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2925 1451 181 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2925 1451 182 SECTIONS (12). 2925 1451 183 REFERENCES: 2925 1451 184 (1) F.G. PEREY, COMPUTER CODE GENOA, ORNL, UNPUBLISHED (1967). 2925 1451 185 (2) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2925 1451 186 COLORADO, UNPUBLISHED (1972). 2925 1451 187 (3) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2925 1451 188 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2925 1451 189 MOMENTUM," ORNL/TM-7042 (1980). ALSO, 2925 1451 190 C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2925 1451 191 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2925 1451 192 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2925 1451 193 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2925 1451 194 NATIONAL LAB. 2925 1451 195 (4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2925 1451 196 STATISTICAL MODEL CODE," ORNL/TM-10093 (AUGUST, 1986). 2925 1451 197 (5) D.M. HETRICK, C.Y. FU, AND D.C. LARSON, "CALCULATED NEUTRON 2925 1451 198 -INDUCED CROSS SECTIONS FOR 63,65CU FROM 1 TO 20 MEV AND 2925 1451 199 COMPARISONS WITH EXPERIMENT," ORNL/TM-9083,ENDF-337 (1984). 2925 1451 200 (6) S. F. MUGHABGHAB, M. DIVADEENAM, AND N.E. HOLDEN, "NEUTRON 2925 1451 201 CROSS SECTIONS, VOL. 1, NEUTRON RESONANCE PARAMETERS AND 2925 1451 202 THERMAL CROSS SECTIONS, PART A, Z=1-60," ACADEMIC 2925 1451 203 PRESS (1981). 2925 1451 204 (7) M.S. PANDEY, J.B. GARG AND J.A. HARVEY, "NEUTRON TOTAL CROSS 2925 1451 205 SECTIONS AND RESONANCE PARAMETERS OF 63,65CU, I AND II", PHYS2925 1451 206 REV. C15 (1977) 600, AND PRIVATE COMMUNICATION. 2925 1451 207 (8) F.G. PEREY, PRIVATE COMMUNICATION, 1977. 2925 1451 208 (9) D.C. LARSON, "ORELA MEASUREMENTS TO MEET FUSION ENERGY 2925 1451 209 NEUTRON CROSS SECTION NEEDS," PROC. SYMP. ON CROSS-SECTIONS 2925 1451 210 FROM 10-50 MEV, BNL-NCS-51245, BROOKHAVEN NATIONAL 2925 1451 211 LAB. (JULY,1980). 2925 1451 212 (10) P. GUENTHER ET.AL, ARGONNE NATIONAL LAB DATA IN CSISRS WITH 2925 1451 213 AN/SAN 12869/002, AND W.P. POENITZ AND J.F. WHALEN, ANL/NDM802925 1451 214 IN CSISRS WITH AN/SAN 12853/ , BOTH DATED 1983 2925 1451 215 (11) D.M. HETRICK AND C.Y. FU, "GLUCS: A GENERALIZED LEAST SQUARES2925 1451 216 PROGRAM FOR UPDATING CROSS SECTION EVALUATIONS WITH 2925 1451 217 CORRELATED DATA SETS," ORNL/TM-7341, ENDF-303 (OCT,1980). 2925 1451 218 (12) C.Y. FU AND D.M. HETRICK, "EXPERIENCE IN USING THE 2925 1451 219 COVARIANCES OF SOME ENDF/B-V DOSIMETRY CROSS SECTIONS: 2925 1451 220 PROPOSED IMPROVEMENTS AND ADDITIONS OF CROSS-REACTION 2925 1451 221 COVARIANCES," PROC. FOURTH ASTM-EURATOM SYMP. ON REACTOR 2925 1451 222 DOSIMETRY, GAITHERSBURG, MARYLAND, MARCH 22-26,1982, P877, 2925 1451 223 US NATIONAL BUREAU OF STANDARDS. 2925 1451 224 (13) CSISRS LIBRARY, NATIONAL NUCLEAR DATA CENTER, BROOKHAVEN 2925 1451 225 NATIONAL LABORATORY, UPTON, N.Y. 11973. 2925 1451 226 (14) V. MCLANE, C.L. DUNFORD AND P.F. ROSE, "NEUTRON CROSS 2925 1451 227 SECTIONS, VOL. 2, NEUTRON CROSS SECTION CURVES", ACADEMIC 2925 1451 228 PRESS, 1988. 2925 1451 229 (15) S.M. QAIM, RADIOCHIMICA ACTA, 25, 13-16 (1978). 2925 1451 230 (16) M. G. DELFINI ET AL., NUCL. PHYS. A404, 250 (1983). 2925 1451 231 2925 1451 232 1 451 239 12925 1451 233 2 151 260 12925 1451 234 3 16 10 12925 1451 235 3 102 11 12925 1451 236 3 107 20 12925 1451 237 33 16 58 12925 1451 238 33 107 81 12925 1451 239 2925 1 0 240 2925 0 0 241 2.906300+4 6.238900+1 0 0 1 02925 2151 242 2.906300+4 1.000000+0 0 0 1 02925 2151 243 1.000000-5 9.950000+4 1 3 0 12925 2151 244 1.500000+0 6.700000-1 1 0 2 32925 2151 245 6.238900+1 6.700000-1 0 0 1296 2162925 2151 246 -1.870000+3 2.000000+0 9.280000+1 5.000000-1 0.000000+0 0.000000+02925 2151 247 -2.710000+2 1.000000+0 4.920000+0 5.000000-1 0.000000+0 0.000000+02925 2151 248 5.790000+2 2.000000+0 5.900000-1 4.850000-1 0.000000+0 0.000000+02925 2151 249 6.500000+2 1.000000+0 1.693000-2 3.500000-1 0.000000+0 0.000000+02925 2151 250 8.070000+2 2.000000+0 4.000000-3 5.000700-1 0.000000+0 0.000000+02925 2151 251 9.940000+2 2.000000+0 1.328000-2 5.000700-1 0.000000+0 0.000000+02925 2151 252 2.038000+3 1.000000+0 4.266700+1 5.000000-1 0.000000+0 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1.000000-5 0.000000+0 2.500000+6 1.798100-1 3.000000+6 1.241800-1292533107 647 3.500000+6 1.210500-1 4.000000+6 5.468900-2 4.500000+6 5.798400-2292533107 648 5.000000+6 4.704100-2 5.500000+6 4.866100-2 6.000000+6 4.414500-2292533107 649 7.500000+6 4.614400-2 7.750000+6 4.861700-2 8.000000+6 3.794500-2292533107 650 9.000000+6 3.809600-2 1.000000+7 4.892400-2 1.100000+7 7.718900-2292533107 651 1.200000+7 5.923200-2 1.300000+7 5.984700-2 2.000000+7 0.000000+0292533107 652 0.000000+0 0.000000+0 0 1 100 50292533107 653 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.742200+6 0.000000+0292533107 654 2.000000+6 0.000000+0 2.500000+6 2.909800-2 3.000000+6 1.387800-2292533107 655 3.500000+6 1.318700-2 4.000000+6 2.691800-3 4.500000+6 3.025900-3292533107 656 5.000000+6 1.991600-3 5.500000+6 2.131100-3 6.000000+6 1.105000-3292533107 657 6.500000+6 1.105000-3 7.000000+6 1.105000-3 7.500000+6 1.916300-3292533107 658 7.750000+6 2.127300-3 8.000000+6 1.295800-3 9.000000+6 1.306200-3292533107 659 1.000000+7 2.154200-3 1.100000+7 5.362400-3 1.200000+7 3.157600-3292533107 660 1.300000+7 6.447100-4 1.325000+7 6.447100-4 1.350000+7 6.447100-4292533107 661 1.375000+7 6.447100-4 1.400000+7 6.447100-4 1.425000+7 6.447100-4292533107 662 1.450000+7 6.447100-4 1.475000+7 6.447100-4 1.500000+7 6.447100-4292533107 663 1.525000+7 6.447100-4 1.550000+7 6.447100-4 1.575000+7 6.447100-4292533107 664 1.600000+7 6.447100-4 1.625000+7 6.447100-4 1.650000+7 6.447100-4292533107 665 1.675000+7 6.447100-4 1.700000+7 6.447100-4 1.725000+7 6.447100-4292533107 666 1.750000+7 6.447100-4 1.775000+7 6.447100-4 1.800000+7 6.447100-4292533107 667 1.825000+7 6.447100-4 1.850000+7 6.447100-4 1.875000+7 6.447100-4292533107 668 1.900000+7 6.447100-4 1.925000+7 6.447100-4 1.950000+7 6.447100-4292533107 669 1.975000+7 6.447100-4 2.000000+7 0.000000+0 0.000000+0 0.000000+0292533107 670 0.000000+0 0.000000+0 0 8 100 50292533107 671 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.742200+6 0.000000+0292533107 672 2.000000+6 0.000000+0 2.500000+6 4.06650-15 3.000000+6 1.78590-13292533107 673 3.500000+6 4.77250-12 4.000000+6 1.75710-11 4.500000+6 1.46500-10292533107 674 5.000000+6 4.82550-10 5.500000+6 2.339400-9 6.000000+6 3.761800-9292533107 675 6.500000+6 1.061500-8 7.000000+6 1.838800-8 7.500000+6 4.660700-8292533107 676 7.750000+6 6.646500-8 8.000000+6 4.891100-8 9.000000+6 9.647100-8292533107 677 1.000000+7 2.324600-7 1.100000+7 8.609500-7 1.200000+7 6.189100-7292533107 678 1.300000+7 1.372500-7 1.325000+7 1.406600-7 1.350000+7 1.380000-7292533107 679 1.375000+7 1.339200-7 1.400000+7 1.329900-7 1.425000+7 1.228400-7292533107 680 1.450000+7 1.156500-7 1.475000+7 1.078700-7 1.500000+7 1.001200-7292533107 681 1.525000+7 9.172200-8 1.550000+7 8.332100-8 1.575000+7 7.511000-8292533107 682 1.600000+7 6.749700-8 1.625000+7 6.006600-8 1.650000+7 5.321600-8292533107 683 1.675000+7 4.701600-8 1.700000+7 4.144400-8 1.725000+7 3.666600-8292533107 684 1.750000+7 3.250000-8 1.775000+7 2.886300-8 1.800000+7 2.563200-8292533107 685 1.825000+7 2.279900-8 1.850000+7 2.026600-8 1.875000+7 1.797600-8292533107 686 1.900000+7 1.582200-8 1.925000+7 1.391200-8 1.950000+7 1.213800-8292533107 687 1.975000+7 1.046900-8 2.000000+7 0.000000+0 0.000000+0 0.000000+0292533107 688 292533 0 689 2925 0 0 690 0 0 0 691 2.906500+4 6.437000+1 -1 0 0 12931 1451 1 0.000000+0 0.000000+0 0 0 0 62931 1451 2 1.000000+0 0.000000+0 0 0 10 62931 1451 3 0.000000+0 0.000000+0 0 0 231 32931 1451 4 29-CU- 65 ORNL EVAL-NOV89 HETRICK,FU,LARSON 2931 1451 5 ORNL/TM-9083,ENDF-337 DIST-APR90 900409 2931 1451 6 ----ENDF/B-VI MATERIAL 2931 2931 1451 7 -----INCIDENT NEUTRON DATA 2931 1451 8 ------ENDF-6 FORMAT 2931 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2931 1451 10 DOSIMETRY 2931 1451 11 ******************************************************************2931 1451 12 THIS WORK EMPLOYED SEVERAL NUCLEAR MODEL CODES INCLUDING THE 2931 1451 13 OPTICAL-MODEL CODE GENOA (1), THE DISTORTED WAVE BORN 2931 1451 14 APPROXIMATION (DWBA) PROGRAM DWUCK (2), AND THE HAUSER-FESHBACH 2931 1451 15 CODE TNG (3,4). THE TNG CODE PROVIDES ENERGY AND ANGULAR 2931 1451 16 DISTRIBUTIONS OF PARTICLES EMITTED IN THE COMPOUND AND PRE- 2931 1451 17 COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL REACTIONS, AND 2931 1451 18 MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE CONTENTS OF 2931 1451 19 THIS EVALUATION AND EXTENSIVE COMPARISONS OF CALCULATIONS WITH 2931 1451 20 EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5). 2931 1451 21 2931 1451 22 ----- DESCRIPTION OF FILES 2931 1451 23 (MF-MT) 2931 1451 24 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2931 1451 25 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB (6). POINT2931 1451 26 WISE RECONSTRUCTION COMPARED WITH DATA (7) SHOWED POORER 2931 1451 27 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2931 1451 28 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2931 1451 29 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2931 1451 30 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2931 1451 31 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2931 1451 32 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2931 1451 33 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2931 1451 34 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2931 1451 35 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2931 1451 36 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2931 1451 37 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2931 1451 38 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2931 1451 39 3-1 THE TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS 2931 1451 40 FROM 1.E-5 EV TO 99.5 KEV. THE THERMAL CROSS SECTION 2931 1451 41 (16.3B) IS REPRODUCED. A SMALL CONTRIBUTION FROM 3/102 IS2931 1451 42 REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151 2931 1451 43 REPRODUCES THE CAPTURE DATA. FROM 99.5 KEV TO 1.12 MEV 2931 1451 44 CU65 DATA FROM (7) IS USED, AFTER APPROPRIATE AVERAGING. 2931 1451 45 ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE. FROM 1.12 TO 2931 1451 46 4.0 MEV, NAT CU DATA OF PEREY (8) USED IN V5 IS RETAINED. 2931 1451 47 FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL (9) IS 2931 1451 48 AVERAGED AND USED. COMPARISONS FROM 1.2 TO 4.5 MEV 2931 1451 49 WITH AVERAGED ARGONNE DATA FOR NAT CU (10) SHOW 1% 2931 1451 50 AGREEMENT. LARSON'S DATA ABOVE 10 MEV WERE ADJUSTED 2931 1451 51 SLIGHTLY (<2 %) TO FIT 65CU DATA OF DYUMIN ET AL. (11). 2931 1451 52 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2931 1451 53 SUBTRACTING THE NONELASTIC FROM THE TOTAL. THE THERMAL 2931 1451 54 VALUE OF 14.1B IS REPRODUCED. 2931 1451 55 3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 2931 1451 56 3-28, 3-102, 3-103, 3-104, 3-105, 3-106, AND 3-107. 2931 1451 57 3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, .. 2931 1451 58 .., 3-63, AND 3-91 2931 1451 59 3-16 (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS (12) 2931 1451 60 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 61 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 62 SECTIONS (13). 2931 1451 63 3-22 (N,NA) CROSS SECTIONS WERE CALCULATED BY THE TNG 2931 1451 64 CODE (3,4,5) WHICH AGREES WELL WITH AVAILABLE DATA (5). 2931 1451 65 3-28 (N,NP) CROSS SECTIONS WERE CALCULATED BY THE TNG 2931 1451 66 CODE (3,4,5). DATA ARE SCARCE FOR THIS REACTION (5). 2931 1451 67 3-51 TO 3-63 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE 2931 1451 68 FROM TNG (3,4,5) WHICH INCLUDES DIRECT INTERACTION 2931 1451 69 CROSS SECTIONS FROM DWUCK (2) CALCULATIONS. EXTENSIVE 2931 1451 70 COMPARISONS WITH EXPERIMENTAL DATA USED IN THE EVALUATION 2931 1451 71 PROCESS ARE SHOWN IN (5). 2931 1451 72 3-91 INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG 2931 1451 73 CALCULATED). GUIDANCE FROM (N,XN) EMISSION DATA (5). 2931 1451 74 3-102 (N,G) CAPTURE CROSS SECTION IS OBTAINED FROM RESONANCE 2931 1451 75 PARAMETERS FROM 1.E-5 EV TO 99.5 KEV. A SMALL BACKGROUND 2931 1451 76 IS GIVEN HERE, WHICH WHEN ADDED TO THE RESONANCE 2931 1451 77 CONTRIBUTION REPRODUCES EXPERIMENTAL DATA, INCLUDING THE 2931 1451 78 THERMAL VALUE OF 2.17B. FROM 99.5 KEV TO 20 MEV, DATA 2931 1451 79 FROM THE CSISRS LIBRARY (14) WERE USED TO CONSTRUCT AN 2931 1451 80 AVERAGE CURVE THROUGH THE DATA; RESULTS ARE SIMILAR TO EYE2931 1451 81 GUIDE IN REF (15). 2931 1451 82 3-103 (N,P) CROSS SECTIONS WERE CALCULATED BY THE TNG 2931 1451 83 CODE (3,4,5) WHICH AGREES WELL WITH MEASURED DATA (5). 2931 1451 84 3-104 (N,D) CROSS SECTION - SHAPE OF (N,P) CROSS SECTION USED 2931 1451 85 NORMALIZED TO AVAILABLE DATA (5). 2931 1451 86 3-105 (N,T) CROSS SECTION - SHAPE OF (N,P) CROSS SECTION 2931 1451 87 USED NORMALIZED TO SYSTEMATICS OF QAIM AND STOCKLIN (16) 2931 1451 88 AT 14.6 MEV. 2931 1451 89 3-106 (N,3HE) CROSS SECTION - SHAPE OF (N,A) CROSS SECTION 2931 1451 90 USED NORMALIZED TO AVAILABLE DATA (17). 2931 1451 91 3-107 (N,A) CROSS SECTIONS WERE CALCULATED BY THE 2931 1451 92 TNG CODE (3,4,5). DATA ARE SCARCE AND DISCREPANT FOR 2931 1451 93 THIS REACTION (5). 2931 1451 94 4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR 2931 1451 95 ELASTIC SCATTERING ARE RESULTS OF FITTING DATA WITH GENOA.2931 1451 96 IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY 2931 1451 97 THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE 2931 1451 98 PARAMETERS IN 2/151. 2931 1451 99 6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2931 1451 100 NEUTRON AND 64CU RESIDUAL, AND ENERGY DEPENDENT YIELD 2931 1451 101 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2931 1451 102 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2931 1451 103 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2931 1451 104 THE OUTGOING NEUTRON). (N,XN) AND (N,XG) D-D EMISSION DATA2931 1451 105 HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 106 6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2931 1451 107 NEUTRON, ALPHA, AND 61CO RESIDUAL, AND ENERGY DEPENDENT 2931 1451 108 YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE 2931 1451 109 GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2931 1451 110 FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY 2931 1451 111 FOR THE OUTGOING NEUTRON). (N,XA) AND (N,XG) D-D EMISSION2931 1451 112 DATA HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 113 6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE 2931 1451 114 NEUTRON AND 64NI RESIDUAL, AND ENERGY DEPENDENT YIELD 2931 1451 115 BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA 2931 1451 116 RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR 2931 1451 117 EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR 2931 1451 118 THE OUTGOING NEUTRON). (N,XP) AND (N,XG) D-D EMISSION 2931 1451 119 DATA USED TO BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 120 6-51 THROUGH 6-63 INELASTIC SCATTERING EXCITING LEVELS; EACH 2931 1451 121 SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON 2931 1451 122 AND 65CU RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR 2931 1451 123 THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED 2931 1451 124 BY DWUCK (2) AND TNG (3,4,5)). COMPARISONS WITH ANGULAR 2931 1451 125 DISTRIBUTION DATA ARE GIVEN IN (5). 2931 1451 126 6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES 2931 1451 127 SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 65CU 2931 1451 128 RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG 2931 1451 129 CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; 2931 1451 130 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN 2931 1451 131 FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE 2931 1451 132 OUTGOING NEUTRON). (N,XN) AND (N,XG) D-D EMISSION DATA 2931 1451 133 RELIED UPON TO BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 134 6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P 2931 1451 135 AND 65NI RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2931 1451 136 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2931 1451 137 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2931 1451 138 PRODUCT. (N,XP) AND (N,XG) D-D EMISSION DATA USED TO 2931 1451 139 BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 140 6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A 2931 1451 141 AND 62CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED 2931 1451 142 ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY; 2931 1451 143 CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH 2931 1451 144 PRODUCT. (N,XA) AND (N,XG) D-D EMISSION DATA USED TO 2931 1451 145 BENCHMARK THE TNG CALCULATIONS (5). 2931 1451 146 12-51 THROUGH 12-63 BRANCHING RATIOS FOR THE LEVELS, COMPILED 2931 1451 147 BY HETRICK ET AL. (5) ARE GIVEN. 2931 1451 148 12-102 (N,G) CAPTURE; THERMAL VALUES TAKEN FROM DELFINI ET AL. 2931 1451 149 (18). HIGHER-ENERGY VALUES BASED ON TNG CALCULATIONS 2931 1451 150 USING PRIMARY BRANCHING RATIOS OF DELFINI ET AL. FOR 2931 1451 151 S-WAVES 2931 1451 152 14-51 THROUGH 14-63 GAMMA RAY ANGULAR DISTRIBUTIONS ASSUMED TO 2931 1451 153 BE ISOTROPIC. 2931 1451 154 14-102 (N,G) CAPTURE; ISOTROPIC DISTRIBUTIONS ASSUMED 2931 1451 155 15-102 (N,G) CAPTURE; AS IN 12-102 2931 1451 156 -------------------------------------------------------------- 2931 1451 157 UNCERTAINTY FILES 2931 1451 158 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2931 1451 159 REQUIRED BY ENDF/B-VI FORMATS 2931 1451 160 2931 1451 161 33-1 TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV 2931 1451 162 EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8. 2931 1451 163 33-2 EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV2931 1451 164 33-3 DERIVED FROM 1.E-5 TO 99.5 KEV, EXPLICIT USING LB=1,8 FROM2931 1451 165 99.5 KEV TO 20 MEV. 2931 1451 166 33-4 DERIVED FROM THRESHOLD TO 20 MEV. 2931 1451 167 33-16 (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS (12) 2931 1451 168 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 169 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 170 SECTIONS (13). 2931 1451 171 33-22 (N,NA) UNCERTAINTIES ESTIMATED FROM TNG AND DATA. 2931 1451 172 33-28 (N,NP) COVARIANCES ESTIMATED FROM TNG AND DATA. 2931 1451 173 33-51-91 INELASTIC SCATTERING UNCERTAINTIES ESTIMATED FROM DATA 2931 1451 174 AND CALCULATIONS. 2931 1451 175 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2931 1451 176 ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 2931 1451 177 DATA (5,14,15) FROM 99.5 KEV TO 20 MEV. 2931 1451 178 33-103 (N,P) COVARIANCES - ESTIMATED USING CSISRS DATA AS GUIDE. 2931 1451 179 33-104 (N,D) COVARIANCES - ESTIMATED, BASED ON DATA. 2931 1451 180 33-105 (N,T) COVARIANCES - ESTIMATED. 2931 1451 181 33-106 (N,3HE) COVARIANCES - ESTIMATED, BASED ON DATA. 2931 1451 182 33-107 (N,A) COVARIANCES WERE ESTIMATED BASED ON TNG AND DATA. 2931 1451 183 REFERENCES: 2931 1451 184 (1) F.G. PEREY, COMPUTER CODE GENOA, ORNL, UNPUBLISHED (1967). 2931 1451 185 (2) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2931 1451 186 COLORADO, UNPUBLISHED (1972). 2931 1451 187 (3) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND 2931 1451 188 PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR 2931 1451 189 MOMENTUM," ORNL/TM-7042 (1980). ALSO, 2931 1451 190 C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP 2931 1451 191 HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP. 2931 1451 192 NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y., 2931 1451 193 MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN 2931 1451 194 NATIONAL LAB. 2931 1451 195 (4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG 2931 1451 196 STATISTICAL MODEL CODE," ORNL/TM-10093 (AUGUST, 1986). 2931 1451 197 (5) D.M. HETRICK, C.Y. FU, AND D.C. LARSON, "CALCULATED NEUTRON 2931 1451 198 -INDUCED CROSS SECTIONS FOR 63,65CU FROM 1 TO 20 MEV AND 2931 1451 199 COMPARISONS WITH EXPERIMENT," ORNL/TM-9083,ENDF-337 (1984). 2931 1451 200 (6) S. F. MUGHABGHAB, M. DIVADEENAM, AND N.E. HOLDEN, "NEUTRON 2931 1451 201 CROSS SECTIONS, VOL. 1, NEUTRON RESONANCE PARAMETERS AND 2931 1451 202 THERMAL CROSS SECTIONS, PART A, Z=1-60," ACADEMIC 2931 1451 203 PRESS (1981). 2931 1451 204 (7) M.S. PANDEY, J.B. GARG AND J.A. HARVEY, "NEUTRON TOTAL CROSS 2931 1451 205 SECTIONS AND RESONANCE PARAMETERS OF 63,65CU, I AND II", PHYS2931 1451 206 REV. C15 (1977) 600, AND PRIVATE COMMUNICATION. 2931 1451 207 (8) F.G. PEREY, PRIVATE COMMUNICATION, 1977. 2931 1451 208 (9) D.C. LARSON, "ORELA MEASUREMENTS TO MEET FUSION ENERGY 2931 1451 209 NEUTRON CROSS SECTION NEEDS," PROC. SYMP. ON CROSS-SECTIONS 2931 1451 210 FROM 10-50 MEV, BNL-NCS-51245, BROOKHAVEN NATIONAL 2931 1451 211 LAB. (JULY,1980). 2931 1451 212 (10) P. GUENTHER ET.AL, ARGONNE NATIONAL LAB DATA IN CSISRS WITH 2931 1451 213 AN/SAN 12869/002, AND W.P. POENITZ AND J.F. WHALEN, ANL/NDM802931 1451 214 IN CSISRS WITH AN/SAN 12853/ , BOTH DATED 1983. 2931 1451 215 (11) A.I. DYUMIN, D.M. KAMINKER, G.N. POPOVA, & V.A. SMOLIN, 2931 1451 216 IZV.AKAD.NAUK SSSR, SER.FIZ. 36 (4), 852 (1972) 2931 1451 217 (12) D.M. HETRICK AND C.Y. FU, "GLUCS: A GENERALIZED LEAST SQUARES2931 1451 218 PROGRAM FOR UPDATING CROSS SECTION EVALUATIONS WITH 2931 1451 219 CORRELATED DATA SETS," ORNL/TM-7341, ENDF-303 (OCT,1980). 2931 1451 220 (13) C.Y. FU AND D.M. HETRICK, "EXPERIENCE IN USING THE 2931 1451 221 COVARIANCES OF SOME ENDF/B-V DOSIMETRY CROSS SECTIONS: 2931 1451 222 PROPOSED IMPROVEMENTS AND ADDITIONS OF CROSS-REACTION 2931 1451 223 COVARIANCES," PROC. FOURTH ASTM-EURATOM SYMP. ON REACTOR 2931 1451 224 DOSIMETRY, GAITHERSBURG, MARYLAND, MARCH 22-26,1982, P877, 2931 1451 225 US NATIONAL BUREAU OF STANDARDS. 2931 1451 226 (14) CSISRS LIBRARY, NATIONAL NUCLEAR DATA CENTER, BROOKHAVEN 2931 1451 227 NATIONAL LABORATORY, UPTON, N.Y. 11973. 2931 1451 228 (15) V. MCLANE, C.L. DUNFORD AND P.F. ROSE, "NEUTRON CROSS 2931 1451 229 SECTIONS, VOL. 2, NEUTRON CROSS SECTION CURVES", ACADEMIC 2931 1451 230 PRESS, 1988. 2931 1451 231 (16) S.M. QAIM AND G. STOCKLIN, NUCL. PHYS. A257 (1976). 2931 1451 232 (17) S.M. QAIM, RADIOCHIMICA ACTA, 25, 13-16 (1978). 2931 1451 233 (18) M. G. DELFINI ET AL., NUCL. PHYS. A404, 250 (1983). 2931 1451 234 2931 1451 235 1 451 238 12931 1451 236 3 16 10 12931 1451 237 33 16 63 12931 1451 238 2931 1 0 239 2931 0 0 240 2.906500+4 6.437000+1 0 0 0 02931 3 16 241 -9.910000+6-9.910000+6 0 0 1 202931 3 16 242 20 2 0 0 0 02931 3 16 243 1.006300+7 0.000000+0 1.050000+7 2.485100-2 1.100000+7 1.436700-12931 3 16 244 1.150000+7 2.978200-1 1.200000+7 4.646600-1 1.250000+7 6.190700-12931 3 16 245 1.300000+7 7.329500-1 1.350000+7 8.233200-1 1.400000+7 8.953100-12931 3 16 246 1.450000+7 9.488400-1 1.500000+7 9.825500-1 1.550000+7 1.013100+02931 3 16 247 1.600000+7 1.030500+0 1.650000+7 1.062800+0 1.700000+7 1.071100+02931 3 16 248 1.750000+7 1.075800+0 1.800000+7 1.091200+0 1.850000+7 1.076700+02931 3 16 249 1.900000+7 1.036800+0 2.000000+7 9.588500-1 0.000000+0 0.000000+02931 3 16 250 2931 3 0 251 2931 0 0 252 2.906500+4 6.437000+1 0 0 0 1293133 16 253 0.000000+0 0.000000+0 0 16 0 4293133 16 254 0.000000+0 0.000000+0 1 5 171 18293133 16 255 1.000000-5 1.006300+7 1.050000+7 1.100000+7 1.250000+7 1.400000+7293133 16 256 1.450000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7293133 16 257 1.850000+7 1.900000+7 1.920000+7 1.940000+7 1.960000+7 2.000000+7293133 16 258 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0293133 16 259 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0293133 16 260 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0293133 16 261 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0293133 16 262 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0293133 16 263 0.000000+0 0.000000+0 0.000000+0 1.170600-2 4.741300-4 4.284700-4293133 16 264 1.313100-4 1.106300-4 6.477000-5 6.417200-5 6.216900-5 6.343100-5293133 16 265 6.407700-5 8.524100-5 6.438800-5 7.086900-5 7.087400-5 8.306100-5293133 16 266 2.488000-3 4.064100-4 1.163400-4 9.510000-5 5.626300-5 5.558000-5293133 16 267 5.394900-5 5.491100-5 5.558000-5 7.326900-5 5.489000-5 6.083400-5293133 16 268 6.083000-5 7.134100-5 1.991200-3 1.930700-4 1.177700-4 7.578500-5293133 16 269 7.436800-5 7.324800-5 7.381700-5 7.499200-5 9.343300-5 7.354200-5293133 16 270 7.865700-5 7.865600-5 9.062800-5 8.238100-4 2.212500-4 1.427800-4293133 16 271 1.364600-4 1.342400-4 1.349100-4 1.367800-4 1.729100-4 1.303500-4293133 16 272 1.439400-4 1.439400-4 1.676400-4 2.733300-4 1.472200-4 1.462200-4293133 16 273 1.390100-4 1.414800-4 1.438200-4 2.011500-4 1.343200-4 1.624900-4293133 16 274 1.624900-4 1.965900-4 5.706800-4 1.858900-4 2.043800-4 1.847400-4293133 16 275 1.859800-4 1.535200-4 1.536800-4 1.405300-4 1.405500-4 1.422600-4293133 16 276 3.160300-4 1.768100-4 1.986800-4 1.742500-4 1.582500-4 1.711300-4293133 16 277 1.475800-4 1.475800-4 1.467700-4 3.522000-4 2.135300-4 1.888900-4293133 16 278 1.603100-4 1.798700-4 1.516900-4 1.516800-4 1.473200-4 1.180400-3293133 16 279 -7.011400-5 2.934400-4 2.097300-4 2.460200-4 2.460200-4 2.128100-4293133 16 280 5.144100-4 1.515000-4 1.968900-4 1.572400-4 1.572300-4 1.526400-4293133 16 281 1.451800-3 9.265200-5 5.591700-4 5.591700-4 4.462100-4 5.217100-4293133 16 282 2.367800-4 2.367800-4 2.042500-4 3.296600-3 1.219200-3 8.869900-4293133 16 283 3.296100-3 8.870500-4 2.088900-3 0.000000+0 0.000000+0 0.000000+0293133 16 284 0.000000+0 0.000000+0 18 4 72 36293133 16 285 1.000000-5 0.000000+0 1.006300+7-1.000000+0 1.050000+7-1.000000+0293133 16 286 1.100000+7-7.500000-1 1.250000+7-7.500000-1 1.400000+7-1.000000+0293133 16 287 1.450000+7-8.400000-1 1.600000+7-1.000000+0 1.650000+7-1.000000+0293133 16 288 1.700000+7-1.000000+0 1.750000+7-1.000000+0 1.800000+7-1.000000+0293133 16 289 1.850000+7-1.000000+0 1.900000+7-1.000000+0 1.920000+7-1.000000+0293133 16 290 1.940000+7-1.000000+0 1.960000+7-1.000000+0 2.000000+7 0.000000+0293133 16 291 1.000000-5 0.000000+0 1.006300+7 0.000000+0 1.050000+7 1.082000-1293133 16 292 1.100000+7 4.988000-2 1.250000+7 4.462300-2 1.400000+7 2.870200-2293133 16 293 1.450000+7 1.653300-2 1.600000+7 2.388900-2 1.650000+7 1.777700-2293133 16 294 1.700000+7 1.876700-2 1.750000+7 3.435700-2 1.800000+7 2.268100-2293133 16 295 1.850000+7 3.810300-2 1.900000+7 2.284100-2 1.920000+7 5.741600-2293133 16 296 1.940000+7 5.741200-2 1.960000+7 4.570400-2 2.000000+7 0.000000+0293133 16 297 0.000000+0 0.000000+0 0 1 48 24293133 16 298 1.000000-5 0.000000+0 1.006300+7 0.000000+0 1.050000+7 1.053600-2293133 16 299 1.100000+7 1.679400-3 1.150000+7 1.679400-3 1.200000+7 1.679400-3293133 16 300 1.250000+7 1.344100-3 1.300000+7 1.344100-3 1.350000+7 1.344100-3293133 16 301 1.400000+7 7.414300-4 1.450000+7 2.066400-4 1.500000+7 2.066400-4293133 16 302 1.550000+7 2.066400-4 1.600000+7 5.136100-4 1.650000+7 2.844300-4293133 16 303 1.700000+7 3.169800-4 1.750000+7 1.062300-3 1.800000+7 4.629700-4293133 16 304 1.850000+7 1.306700-3 1.900000+7 4.695400-4 1.920000+7 2.966900-3293133 16 305 1.940000+7 2.966500-3 1.960000+7 1.880000-3 2.000000+7 0.000000+0293133 16 306 0.000000+0 0.000000+0 0 8 48 24293133 16 307 1.000000-5 0.000000+0 1.006300+7 0.000000+0 1.050000+7 7.229600-7293133 16 308 1.100000+7 3.851700-6 1.150000+7 1.655100-5 1.200000+7 4.028900-5293133 16 309 1.250000+7 5.723600-5 1.300000+7 8.023000-5 1.350000+7 1.012300-4293133 16 310 1.400000+7 6.603500-5 1.450000+7 2.067100-5 1.500000+7 2.216600-5293133 16 311 1.550000+7 2.356500-5 1.600000+7 6.060200-5 1.650000+7 3.569700-5293133 16 312 1.700000+7 4.040600-5 1.750000+7 1.366100-4 1.800000+7 6.125200-5293133 16 313 1.850000+7 1.683100-4 1.900000+7 5.608100-5 1.920000+7 3.751700-4293133 16 314 1.940000+7 3.629800-4 1.960000+7 2.231600-4 2.000000+7 0.000000+0293133 16 315 293133 0 316 2931 0 0 317 0 0 0 318 4.911500+4 1.139200+2 1 0 0 14931 1451 1 0.000000+0 0.000000+0 0 0 0 64931 1451 2 1.000000+0 0.000000+0 0 0 10 64931 1451 3 0.000000+0 0.000000+0 0 0 107 94931 1451 4 49-IN-115 HEDL,ANL EVAL-MAR90 F.SCHMITTROTH,D.L.SMITH,S.CHIBA 4931 1451 5 DIST-JUN90 900329 4931 1451 6 ----ENDF/B-VI MATERIAL 4931 4931 1451 7 -----INCIDENT NEUTRON DATA 4931 1451 8 ------ENDF-6 FORMAT 4931 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4931 1451 10 DOSIMETRY 4931 1451 11 ******************************************************************4931 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 4931 1451 13 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 14 4931 1451 15 IN-115(N,N')IN-115M DOSIMETRY REACTION 4931 1451 16 4931 1451 17 EVALUATED BY S.CHIBA AND D.L. SMITH, ANL, JANUARY 1990 4931 1451 18 4931 1451 19 PRODUCTION OF THE ISOMER IN-115M BY THE (N,N') PROCESS IS 4931 1451 20 ROUTINELY EMPLOYED FOR NEUTRON DOSIMETRY APPLICATIONS. THIS 4931 1451 21 ISOMER IS THE FIRST-EXCITED STATE OF THE ISOTOPE IN-115 (336 KEV 4931 1451 22 EXCITATION ENERGY). THE REACTION THRESHOLD ENERGY IS 339 KEV. 4931 1451 23 THE ISOTOPIC ABUNDANCE OF IN-115 IN NATURAL INDIUM IS 95.7 4931 1451 24 PERCENT. THE HALF LIFE OF IN-115M IS 4.486 HOURS. THE DECAY 4931 1451 25 MODES ARE - BETA MINUS (5 PERCENT) ISOMERIC TRANSITION (95.0 4931 1451 26 PERCENT). THE NUMBER OF DECAY 336-KEV GAMMA RAYS EMITTED PER 4931 1451 27 DISINTEGRATION OF IN-115M IS 0.459. 4931 1451 28 4931 1451 29 DOCUMENTATION FOR THIS EVALUATION IS PROVIDED BY A.B. SMITH ET 4931 1451 30 AL., REPORT ANL/NDM-115, ARGONNE NATIONAL LABORATORY (1990). 4931 1451 31 EVALUATED RESULTS FOR THIS REACTION ARE BASED ON EXPERIMENTAL 4931 1451 32 DATA. NUCLEAR MODEL CALCULATIONS WERE USED TO ESTIMATE THE CROSS 4931 1451 33 SECTION SHAPE NEAR THRESHOLD. THIS EVALUATION WAS CARRIED OUT 4931 1451 34 USING THE STATISTICAL LEAST-SQUARES METHOD. THE CODE GMA, AS 4931 1451 35 DEVELOPED BY W. POENITZ (1981) AND LATER REVISED BY M. SUGIMOTO 4931 1451 36 (1987) AND S. CHIBA (1990), WAS EMPLOYED FOR THIS ANALYSIS. 4931 1451 37 4931 1451 38 * * * * * * *4931 1451 39 CAPTURE TO 54 MIN. ISOMERIC STATE DESCRIPTION MODIFIED BY 4931 1451 40 R. SCHENTER ON 2/20/84. THIS WAS DONE BY ADDING FILE9 WITH 4931 1451 41 Y=.79 AND PUTTING THE TOTAL CAPTURE WIDTH IN FILE2 AND THE TOTAL 4931 1451 42 CAPTURE CROSS SECTION IN FILE3. 4931 1451 43 P.C.TO NORMALIZATION+STANDARDS SUBCOMITTEE (N,N')D.L.SMITH MT=51 4931 1451 44 P.C.TO NORMALIZATION+STANDARDS SUBCOMITTEE (N,G)F.SCHMITTROTH 4931 1451 45 MF=2 MT=151 4931 1451 46 MF=2 MT=102 4931 1451 47 * * * * * 4931 1451 48 4931 1451 49 FILE INFORMATION 4931 1451 50 4931 1451 51 MF=1 MT=451 ATOMIC MASS FROM REF(1). 4931 1451 52 4931 1451 53 MF=2 MT=151 EVALUATION OF RESOLVED RESONANCE PARAMETERS 4931 1451 54 BASED ON NEW BNL-325, REF(2). 4931 1451 55 4931 1451 56 MF=3 MT= 51 EVALUATION OF THE 4.486 H ISOMER BASED ENTIRELY ON 4931 1451 57 REPORTED EXPERIMENTAL DATA.DOCUMENTATION AVAILABLE 4931 1451 58 AS ANL/NDM-26 BY D.L.SMITH REFERENCES 4931 1451 59 USED LISTED BELOW 10 TO 25. 4931 1451 60 MF=3 MT=102 VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA 4931 1451 61 SEE DOCUMENTATION 4931 1451 62 RADIATIVE NEUTRON CAPTURE TO THE IN 116M(54 MIN.) 4931 1451 63 STATE. 4931 1451 64 E GREATER THAN EH, 4931 1451 65 EVALUATION BASED ON EXPERIMENTAL DATA, REF.(3-7) 4931 1451 66 AND THEORETICAL CALCULATIONS, REF.(8,9). 4931 1451 67 E LESS THAT EH, 4931 1451 68 A 1/V COMPONENT WAS ADDED TO GIVE THE CORRECT 4931 1451 69 2200M/S CROSS SECTION TO THE 54 MIN. STATE(THE 4931 1451 70 2.2 SEC. STATE CROSS SECTION WAS INCLUDED). 4931 1451 71 RADIATIVE CAPTURE TO THE 2.2 SEC. STATE OF IN 116 WAS 4931 1451 72 INCLUDED AS PART OF THE CAPTURE TO THE 54 MIN. STATE 4931 1451 73 FOR BOTH THERMAL AND FAST ENERGIES. 4931 1451 74 RESULTS WERE DIVIDED BY .79 TO GIVE THE TOTAL CAP- 4931 1451 75 TURE CROSS SECTION IN FILE3. FILE9 COMBINED WITH 4931 1451 76 FILE3 IS REQUIRED TO GIVE CAPTURE TO 54 MIN. 4931 1451 77 ISOMERIC STATE. 4931 1451 78 4931 1451 79 2200M/S CAPTURE CROSS SECTION,BARNS TO THE 54 MIN. STATE 4931 1451 80 (FROM RESONANCE PARAMETERS) = 166.413 4931 1451 81 4931 1451 82 COMPUTED RESONANCE INTEGRAL =2587.345 4931 1451 83 4931 1451 84 REFERENCES. 4931 1451 85 1. A.H.WAPSTRA AND N.B.GOVE,NUCL. DATA TABLES,VOL.9,PART 1(1971).4931 1451 86 2. S.F.MUGHABGHAB AND D.I.GARBER,BNL-325,3RD ED.,VOL.1(1973) 4931 1451 87 3. H.A.GRENCH AND H.O.MENLOVE,PHYS.REV.,VOL.165(1968)1298. 4931 1451 88 4. H.O.MENLOVE,ET AL.,PHYS.REV.,VOL.163(1967)1299. 4931 1451 89 5. S.A.COX, PHYS.REV.,VOL.133(1964)B378. 4931 1451 90 6. A.E.JOHNSRUD,ET AL.,PHYS.REV.,VOL.116(1959)927. 4931 1451 91 7. G.PETO ET AL., J.NUCL.EN.,VOL.21(1967)797. 4931 1451 92 8. F.SCHMITTROTH, HEDL-TME 71-106(AUGUST 1971). 4931 1451 93 9. F.SCHMITTROTH, HEDL-TME 73-79,ENDF-195(NOVEMBER 1973). 4931 1451 94 10. D.L.SMITH ET.AL. ANL/NDM-14,1975 4931 1451 95 11. D.C.SANTRY AND J.P.BUTLER CAN.J.PHYS.54,757,1975 4931 1451 96 12. K.KOBAYASHI ET.AL. J.NUC.EN.27,741,73 4931 1451 97 13. A.PASZIT AND J.CSIKAI SOV.J.NUC.PHYS.15,232,1972 4931 1451 98 14. J.K.TEMPERLY AND D.E.BARNES BRL-1491,1970 4931 1451 99 15. P.DECOWSKI ET.AL.INR-1197 POLAND 1970 4931 1451 100 16. I.KIMURA ET.AL. J.NUC.SCI.TECH.JAP.6,485,69 4931 1451 101 17. R.C.BARRALL ET.AL. AFWL-TR-68-134,1969 4931 1451 102 18. H.ROETZER NUC.PHYS.A109,694,1968 4931 1451 103 19. B.MINETTI AND A.PAQUARETTI Z.PHYS.217,83,1968 4931 1451 104 20. H.A.GRENCH AND H.O.MENLOVE PHYS.REV.165,1298,1968 4931 1451 105 21. H.O.MENLOVE ET.AL. PHYS.REV.163,1308,1967 4931 1451 106 22. W.NAGEL AND A.H.W.ATEN JR. PHYSICA 31,1091,1965 4931 1451 107 23. A.A.ABEL AND C.GOODMAN PHYS.REV.93,197,1954 4931 1451 108 24. H.C.MARTIN AND B.C.DIVEN PHYS.REV.93,199,1954 4931 1451 109 25. S.G.COHEN NATURE 161,475,1948 4931 1451 110 4931 1451 111 1 451 120 04931 1451 112 2 151 145 04931 1451 113 3 102 16 04931 1451 114 8 4 4 14931 1451 115 8 102 5 04931 1451 116 9 102 4 04931 1451 117 10 4 14 14931 1451 118 33 102 6 04931 1451 119 40 4 87 14931 1451 120 4931 1 0 121 4931 0 0 122 4.911500+4 1.139200+2 0 0 1 04931 2151 123 4.911500+4 1.000000+0 0 0 1 04931 2151 124 1.000000-5 1.003800+3 1 1 0 04931 2151 125 4.500000+0 6.500000-1 0 0 2 04931 2151 126 1.139200+2 0.000000+0 0 0 534 894931 2151 127 1.457000+0 5.000000+0 7.503600-2 3.036400-3 7.200000-2 0.000000+04931 2151 128 3.860000+0 4.000000+0 8.135400-2 3.544400-4 8.100000-2 0.000000+04931 2151 129 9.120000+0 5.000000+0 8.157300-2 1.572700-3 8.000000-2 0.000000+04931 2151 130 1.210000+1 4.500000+0 1.401100-1 1.120000-4 1.400000-1 0.000000+04931 2151 131 2.273000+1 4.500000+0 8.204000-2 1.040000-3 8.100000-2 0.000000+04931 2151 132 2.300000+1 4.500000+0 8.088000-2 1.180000-3 7.970000-2 0.000000+04931 2151 133 3.960000+1 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1.975200-4 1.061300-4 9.587700-5 2.381900-4 2.588900-4493140 4 357 2.925200-4 4.489000-3 4.803200-4 6.511100-4 5.331900-4 5.680300-4493140 4 358 5.453800-4 5.185800-4 4.818600-4 4.317500-4 3.987800-4 4.052200-4493140 4 359 4.146000-4 4.100400-4 3.991200-4 4.020000-4 3.639400-4 3.411600-4493140 4 360 3.515500-4 3.690400-4 3.666200-4 1.945700-4 1.061300-4 9.587700-5493140 4 361 2.381900-4 2.532600-4 2.925200-4 4.489000-3 4.206300-4 5.270200-4493140 4 362 4.992800-4 5.057200-4 4.793900-4 4.735600-4 4.848100-4 4.389800-4493140 4 363 4.422000-4 4.529200-4 4.470200-4 4.361000-4 4.405900-4 3.995900-4493140 4 364 3.691700-4 3.825000-4 4.003900-4 3.971800-4 2.122600-4 1.149700-4493140 4 365 1.031100-4 2.593600-4 2.757700-4 3.162400-4 7.396000-3 5.024100-4493140 4 366 5.572800-4 5.250300-4 5.108400-4 4.692200-4 4.024800-4 3.880300-4493140 4 367 3.962900-4 4.024800-4 4.024800-4 3.876900-4 3.901000-4 3.660200-4493140 4 368 3.496800-4 3.632600-4 3.808100-4 3.790900-4 1.967700-4 1.078400-4493140 4 369 9.752400-5 2.377900-4 2.528400-4 2.942900-4 2.116000-3 5.216400-4493140 4 370 5.174100-4 4.926600-4 4.791400-4 4.669900-4 4.294600-4 4.349800-4493140 4 371 4.461100-4 4.416000-4 4.284900-4 4.327700-4 3.992800-4 3.705800-4493140 4 372 3.840500-4 4.040600-4 3.985400-4 2.084700-4 1.138500-4 1.030900-4493140 4 373 2.543800-4 2.743400-4 3.094000-4 2.916000-3 5.214800-4 5.005800-4493140 4 374 4.770900-4 4.529500-4 4.199000-4 4.263800-4 4.352400-4 4.302700-4493140 4 375 4.185500-4 4.212000-4 3.780000-4 3.529400-4 3.635300-4 3.810200-4493140 4 376 3.775700-4 2.019600-4 1.087000-4 9.914400-5 2.474300-4 2.676200-4493140 4 377 3.058600-4 1.369000-3 4.928400-4 4.737100-4 4.555400-4 4.235800-4493140 4 378 4.289000-4 4.386700-4 4.342300-4 4.221000-4 4.244600-4 3.843600-4493140 4 379 3.571200-4 3.687400-4 3.862800-4 3.824300-4 2.018700-4 1.098900-4493140 4 380 9.890100-5 2.455300-4 2.641800-4 3.012500-4 2.025000-3 4.673300-4493140 4 381 4.568400-4 4.363200-4 4.449600-4 4.527900-4 4.492800-4 4.367700-4493140 4 382 4.352400-4 3.855600-4 3.591000-4 3.712500-4 3.904200-4 3.847500-4493140 4 383 2.059200-4 1.113800-4 1.008500-4 2.524100-4 2.721600-4 3.079800-4493140 4 384 9.610000-4 4.542100-4 4.389600-4 4.471400-4 4.537800-4 4.531000-4493140 4 385 4.384900-4 4.382200-4 3.966800-4 3.687100-4 3.805600-4 3.995300-4493140 4 386 3.946300-4 2.127800-4 1.135500-4 1.029500-4 2.644900-4 2.838400-4493140 4 387 3.182500-4 1.296000-3 4.553300-4 4.648300-4 4.708100-4 4.674200-4493140 4 388 4.554000-4 4.510100-4 3.971500-4 3.707300-4 3.825400-4 4.017600-4493140 4 389 3.980900-4 2.122600-4 1.140500-4 1.040000-4 2.616500-4 2.812300-4493140 4 390 3.186000-4 5.760000-4 4.659800-4 4.705000-4 4.705900-4 4.559500-4493140 4 391 4.504300-4 4.011800-4 3.721000-4 3.849100-4 4.043500-4 3.985400-4493140 4 392 2.101400-4 1.140500-4 1.030300-4 2.597500-4 2.782100-4 3.143500-4493140 4 393 5.760000-4 4.842200-4 4.838400-4 4.692000-4 4.608000-4 4.052200-4493140 4 394 3.803000-4 3.928300-4 4.129900-4 4.067500-4 2.133100-4 1.156300-4493140 4 395 1.049800-4 2.635400-4 2.822400-4 3.200200-4 6.760000-4 4.885900-4493140 4 396 4.736200-4 4.661300-4 4.142300-4 3.892700-4 4.024000-4 4.230700-4493140 4 397 4.169400-4 2.230800-4 1.201200-4 1.088100-4 2.772900-4 2.992100-4493140 4 398 3.374800-4 5.760000-4 4.725100-4 4.665600-4 4.193300-4 3.930700-4493140 4 399 4.070900-4 4.276800-4 4.213400-4 2.185900-4 1.195900-4 1.075700-4493140 4 400 2.673400-4 2.862700-4 3.256800-4 5.290000-4 4.554000-4 3.967000-4493140 4 401 3.705800-4 3.825400-4 4.024100-4 3.959200-4 2.135300-4 1.146100-4493140 4 402 1.037100-4 2.671000-4 2.859400-4 3.229700-4 5.760000-4 4.065600-4493140 4 403 3.812200-4 3.928300-4 4.112600-4 4.067500-4 2.196500-4 1.188000-4493140 4 404 1.075700-4 2.730200-4 2.923200-4 3.285100-4 7.840000-4 4.415600-4493140 4 405 4.573800-4 4.394900-4 4.415600-4 2.303800-4 1.358300-4 1.209600-4493140 4 406 2.676500-4 2.869400-4 3.237900-4 1.444000-3 4.865500-4 4.473400-4493140 4 407 4.635200-4 2.658500-4 1.630200-4 1.456900-4 2.972000-4 3.192000-4493140 4 408 3.587200-4 1.089000-3 4.573800-4 4.715000-4 2.439400-4 1.535500-4493140 4 409 1.363200-4 2.633100-4 2.799700-4 3.193100-4 1.296000-3 4.801700-4493140 4 410 2.534400-4 1.496900-4 1.331600-4 2.900900-4 3.114700-4 3.525800-4493140 4 411 1.444000-3 2.574900-4 1.642700-4 1.456900-4 2.731800-4 2.904700-4493140 4 412 3.318200-4 1.936000-3 3.760700-4 3.314500-4 4.796900-4 4.841800-4493140 4 413 4.672800-4 1.089000-3 4.205500-4 2.111700-4 1.968100-4 1.674400-4493140 4 414 7.290000-4 2.090300-4 1.950500-4 1.688600-4 6.241000-3 9.887600-4493140 4 415 7.550800-4 7.056000-3 8.127800-4 1.392400-2 0.000000+0 0.000000+0493140 4 416 493140 0 417 4931 0 0 418 0 0 0 419 7.919700+4 1.952740+2 1 0 0 17925 1451 1 0.000000+0 0.000000+0 0 0 0 67925 1451 2 1.000000+0 0.000000+0 0 0 10 67925 1451 3 0.000000+0 0.000000+0 0 0 222 47925 1451 4 79-AU-197 LANL EVAL-JAN84 P.G.YOUNG 7925 1451 5 DIST-FEB90 900205 7925 1451 6 ----ENDF/B-VI MATERIAL 7925 7925 1451 7 -----INCIDENT NEUTRON DATA 7925 1451 8 ------ENDF-6 FORMAT 7925 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************7925 1451 10 DOSIMETRY 7925 1451 11 ******************************************************************7925 1451 12 7925 1451 13 *************** summary ******************************************7925 1451 14 7925 1451 15 A new evaluation of all neutron and gamma-ray data above the 7925 1451 16 resonance region is joined with the endf/b-v resolved resonance 7925 1451 17 region evaluation and with the Version VI standard cross section 7925 1451 18 for the (n,gamma) reaction below a neutron energy of 2.5 MeV. 7925 1451 19 7925 1451 20 *************** general description ******************************7925 1451 21 7925 1451 22 P.G.Young and E.D.Arthur 7925 1451 23 7925 1451 24 the new evaluation for files 3,4,5,12,13,14,15 is based on 7925 1451 25 statistical theory, hauser-feshbach, preequilibrium calculations 7925 1451 26 with the comnuc and gnash codes (ref1,2). deformed optical poten-7925 1451 27 tial of delaroche and ecis coupled-channel code were used to cal-7925 1451 28 culate neutron transmission coefficients and total and elastic 7925 1451 29 elastic cross sections (ref3,4). gamma-ray strength functions 7925 1451 30 were obtained by fitting morgan n,xg data (ref5) at 0.4 and 6.5 7925 1451 31 mev. calculated results were used for all major reactions except 7925 1451 32 total cross section. for total, the theoretical cross section 7925 1451 33 was used as prior in covariance analysis of experimental data 7925 1451 34 using glucs code (ref6). more details on experimental data used 7925 1451 35 are given below and in main reference for evaluation (ref 7). 7925 1451 36 7925 1451 37 *************** mf=2 resonance parameters ************************7925 1451 38 7925 1451 39 mt=151 resolved resonance parameters given from 1.0e-05ev 7925 1451 40 to 2 kev based on ref8 and references therein 7925 1451 41 and a bound level. some of the reson. spin assignments 7925 1451 42 from ref9. from 2 to 4.827 kev the parameters are based7925 1451 43 on macklin et al and hoffman et al normalized data. 7925 1451 44 see refs 10 and 11. 7925 1451 45 thermal cross sections are as follows: 7925 1451 46 capture = 98.71 b 7925 1451 47 scattering = 6.84 b 7925 1451 48 total = 105.55 b 7925 1451 49 the absorption resonance integral is 1559 b 7925 1451 50 7925 1451 51 *************** mf=3 smooth neutron cross sections ***************7925 1451 52 7925 1451 53 mt= 1 total cross section. based on glucs covariance analysis 7925 1451 54 using deformed optical model calculation as the prior and 7925 1451 55 experimental data from refs 12-22, 29 for fitting. 7925 1451 56 mt= 2 elastic cross section. difference of mt=1 and sum of 7925 1451 57 all nonelastic cross sections. closely approximates theore- 7925 1451 58 tical results. 7925 1451 59 mt= 4 inelastic cross section. sum of mt=51-63, 91. 7925 1451 60 mt= 16 (n,2n) cross section. theoretical calculation used. 7925 1451 61 in good agreement with exp. below 23 mev. see refs 23-25. 7925 1451 62 mt= 17 (n,3n) cross section. theoretical calculation used. 7925 1451 63 in good agreement with exp. at all energies (refs 24,25). 7925 1451 64 mt= 37 (n,4n) cross section. theoretical calculation used. 7925 1451 65 in reasonable agreement with data of ref 25. 7925 1451 66 mt=51-63 (n,nprime) cross sections to levels. except for mt=53 7925 1451 67 and 56, all are from compound-nucleus calculations with the 7925 1451 68 comnuc code. mt=53 and 56 also include direct reaction com- 7925 1451 69 ponents from ecis calculations (mt53 and 56 are the 5/2+ 7925 1451 70 and 7/2+ members of the ground state rotational band) and 7925 1451 71 extend to 30 mev. mt=51,52,54,55,57-63 are zeroed above 6 mev.7925 1451 72 mt= 91 inelastic continuum cross section. from gnash theoreti- 7925 1451 73 cal calculations. includes (n,gn) component from 0.1 to 7925 1451 74 2.0 mev. conventional (n,ng) continuum starts at 1.2236 7925 1451 75 mev. q-value has no significance except corresponds to thres.7925 1451 76 mt=102 (n,gamma) cross section. below 2.5 MeV, adopted the 7925 1451 77 ENDF/B-VI standard cross section (Ref.30,31) down to the 7925 1451 78 resonance region. At higher energies, the theoretical cal- 7925 1451 79 culations were adjusted to agree with experimental data. A 7925 1451 80 semi-direct component normalized to an average of experimental7925 1451 81 data at 14 MeV was included above En = 6 MeV. 7925 1451 82 at higher energies, use theoretical calculations, which agree 7925 1451 83 reasonably with available exp. data. above 5 mev, calculation7925 1451 84 includes semi-direct component normalized to average of 7925 1451 85 14 mev data. 7925 1451 86 mt=103 (n,p) cross section. adopted endf/b-v with smooth 7925 1451 87 extrapolation to 30 mev. based on exp data of ref 26. 7925 1451 88 mt=107 (n,alpha) cross section. adopted endf/b-v with smooth 7925 1451 89 extrapolation to 30 mev. based on data of ref 26. 7925 1451 90 7925 1451 91 *************** mf=4 neutron angular distributions ***************7925 1451 92 7925 1451 93 mt= 2 elastic scattering. legendre coefficients obtained by 7925 1451 94 combining ecis direct reaction calculations with comnuc com- 7925 1451 95 pound nucleus results. 7925 1451 96 mt= 16 (n,2n) angular distribution. used kalbach-mann (ref 27) 7925 1451 97 semi-empirical shape averaged over the emitted neutron 7925 1451 98 spectrum at each incident neutron energy. 7925 1451 99 mt= 17 (n,3n) angular distribution. same comment as mt=16. 7925 1451 100 mt= 37 (n,4n) angular distribution. same comment as mt=16. 7925 1451 101 mt=51-63 (n,nprime) level angular distributions. legendre coef 7925 1451 102 -ficients obtained from comnuc compound nucleus calculations. 7925 1451 103 for mt=53 and 56, ecis direct reaction results were combined 7925 1451 104 with the compound nucleus calculations, 7925 1451 105 mt= 91 (n,nprime) continuum. same comment as for mt=16. 7925 1451 106 7925 1451 107 *************** mf=5 neutron energy distributions ****************7925 1451 108 7925 1451 109 mt= 16 (n,2n) tabulated distribution from gnash calculations. 7925 1451 110 mt= 17 (n,3n) tabulated distribution from gnash calculations. 7925 1451 111 mt= 37 (n,4n) tabulated distribution from gnash calculations. 7925 1451 112 mt= 91 (n,nprime) continuum tabulated distribution obtained from7925 1451 113 gnash calculation. 7925 1451 114 7925 1451 115 *************** mf=8 radioactive decay data **********************7925 1451 116 7925 1451 117 mt= 16 decay data for the 10 hour metastable sixth excited state7925 1451 118 in au-196. endf/b-v data adopted without change. 7925 1451 119 7925 1451 120 *************** mf=10 radioactive nuclide cross sections *********7925 1451 121 7925 1451 122 mt= 16 production cross section for the 10-hour metastable sixth7925 1451 123 excited state of au-196 through (n,2n) reactions. endf/b-v 7925 1451 124 data adopted, with smooth extrapolation to 30 mev. 7925 1451 125 7925 1451 126 *************** mf=12 photon multiplicities **********************7925 1451 127 7925 1451 128 mt=102 (n,gamma) yield at low energies obtained by requiring 7925 1451 129 energy conservation with mf=15,mt=102 results. beginning 7925 1451 130 near 10 kev, gnash results used. 7925 1451 131 7925 1451 132 *************** mf=13 photon cross sections **********************7925 1451 133 7925 1451 134 mt= 4 gamma-ray production cross sections obtained from gnash 7925 1451 135 calculations for continua regions and from comnuc for 7925 1451 136 discrete levels. ecis was used to calculate direct react- 7925 1451 137 tion contributions for 3rd and 6th levels of au-197. 7925 1451 138 mt= 16 gamma-ray production cross sections obtained from gnash 7925 1451 139 calculations at all incident neutron energies. 7925 1451 140 mt= 17 gamma-ray production cross sections obtained from gnash 7925 1451 141 calculations at all incident neutron energies. 7925 1451 142 mt= 37 gamma-ray production cross sections obtained from gnash 7925 1451 143 calculations at all incident neutron energies. 7925 1451 144 7925 1451 145 *************** mf=14 photon angular distributions ***************7925 1451 146 7925 1451 147 mt= 4 photons from inelastic scattering assumed isotropic. 7925 1451 148 mt= 16 photons from (n,2n) reactions assumed isotropic. 7925 1451 149 mt= 17 photons from (n,3n) reactions assumed isotropic. 7925 1451 150 mt= 37 photons from (n,4n) reactions assumed isotropic. 7925 1451 151 mt=102 photons from (n,gamma) reactions assumed isotropic. 7925 1451 152 7925 1451 153 *************** mf=15 photon energy distributions ****************7925 1451 154 7925 1451 155 mt= 4 inelastic scattering photon tabulated distributions 7925 1451 156 obtained from gnash calculations for continua regions and 7925 1451 157 from comnuc for discrete levels. direct contributions for 7925 1451 158 mt=53 and mt=56 obtained from ecis calculations. 7925 1451 159 mt= 16 (n,2n) photon tabulated distributions obtained from 7925 1451 160 gnash calculations. 7925 1451 161 mt= 17 (n,3n) photon tabulated distributions obtained from 7925 1451 162 gnash calculations. 7925 1451 163 mt= 37 (n,4n) photon tabulated distributions obtained from 7925 1451 164 gnash calculations. 7925 1451 165 mt=102 (n,gamma) tabulated thermal distribution obtained from 7925 1451 166 experimental data of ref 28. thermal spectrum linearly inter-7925 1451 167 polated to gnash calculation at 10 kev. gnash results used 7925 1451 168 at higher energies. 7925 1451 169 7925 1451 170 *************** mf=33 neutron cross section covariances **********7925 1451 171 7925 1451 172 mt= 1 total cross section covariance from glucs analysis. 7925 1451 173 7925 1451 174 *************** references ***************************************7925 1451 175 7925 1451 176 1. c.l.dunford. ai-aec-12931(1970) 7925 1451 177 2. p.g.young, e.d.arthur, la-6947 (1977). 7925 1451 178 3. j.p.delaroche, harwell conference (1978)p.366. 7925 1451 179 4. j.raynal, iaea smr-9/8 (1970). 7925 1451 180 5. g.l.morgan, e.newman, ornl-tm-4973 (1975). 7925 1451 181 6. d.m.hetrick, c.y.fu, ornl/tm-7341 (1980). 7925 1451 182 7. p.g.young, e.d.arthur, in la- -pr (1984)p.12. 7925 1451 183 8. s.f.mughabghab and d.i.garber bnl-325,3rd edn,vol i(1973). 7925 1451 184 9. a.lottin and a.jain conf on nuclear structure study with 7925 1451 185 neutrons,budapest,1972 p34 and private communication. 7925 1451 186 10. r.macklin et al. phys. rev/c 11,1270(1975) and private 7925 1451 187 communication. 7925 1451 188 11. m.m. hoffman et al. 75washington conf., 2, 868(1971) knox. 7925 1451 189 12. w.poenitz et al., nuc.sci.eng. 78, 333(1981). 7925 1451 190 13. d.g.foster jr., d.glasgow, phys.rev. c3, 576(1971). 7925 1451 191 14. k.k.seth,phys.letters,16,306(1965). 7925 1451 192 15. s.c.snowdon, phys.rev. 90, 615(1953). 7925 1451 193 16. j.f.whalen,anl-7210,16(1966). 7925 1451 194 17. n.nereson, phys.rev. 94, 1678(1954). 7925 1451 195 18. a.bratenahl et al., phys.rev. 110, 927(1958). 7925 1451 196 19. j.p.conner,phys.rev.109,1268(1958). 7925 1451 197 20. j.h.coon,phys.rev.88,562(1952). 7925 1451 198 21. j.m.peterson,phys.rev.120,521(1960). 7925 1451 199 22. e.g.bilpuch,private communication(1959). 7925 1451 200 23. j.frehaut et al, proc. 10-50 mev conf, bnl-ncs-51245 (1980) 7925 1451 201 page 399. 7925 1451 202 24. l.r.veeser et al, phys.rev. c16, 1792(1977). 7925 1451 203 25. b.p.bayhurst et al, phys.rev. c12, 451(1975). 7925 1451 204 26. r.j.prestwood and b.p.bayhurst,phys.rev.121,1438(1961). 7925 1451 205 27. c.kalbach and f.mann, bnl-ncs-5/245,p.689 (1980). 7925 1451 206 28. v.j.orphan et al, ga-10248 (1970). 7925 1451 207 29. d.c.larson, proc. 10-50 mev conf, bnl-ncs-51245 (1980) p.277.7925 1451 208 30. A.Carlson et al., Nuc.Data for Basic & Applied Science, 7925 1451 209 Santa Fe, NM (1985) p.1429. 7925 1451 210 31. W.Poenitz, ANL-West, personnal communication (1989). 7925 1451 211 7925 1451 212 ******************************************************************7925 1451 213 ***************** PROGRAM FIXUP (VERSION 86-2) ******************7925 1451 214 *RECONSTRUCTED MT NUMBERS 7925 1451 215 4 =+( 51, 91) 7925 1451 216 103 =+(700,718) 7925 1451 217 104 =+(720,738) 7925 1451 218 105 =+(740,758) 7925 1451 219 106 =+(760,778) 7925 1451 220 107 =+(780,798) 7925 1451 221 101 =+(102,114) 7925 1451 222 27 =+( 18, 18)+(101,101) 7925 1451 223 3 =+( 4, 4)+( 6, 9)+( 16, 17)+( 22, 37) 7925 1451 224 19 =+( 18, 18)-( 20, 21)-( 38, 38) 7925 1451 225 1 =+( 2, 3) 7925 1451 226 1 451 230 17925 1451 227 2 151 268 07925 1451 228 3 16 25 17925 1451 229 3 102 44 17925 1451 230 7925 1 0 231 7925 0 0 232 7.919700+4 1.952740+2 0 0 1 07925 2151 233 7.919700+4 1.000000+0 0 0 1 07925 2151 234 1.000000-5 5.000000+3 1 2 0 07925 2151 235 1.500000+0 9.800000-1 0 0 1 07925 2151 236 1.952740+2 0.000000+0 0 0 1578 2637925 2151 237 -3.380000+1 2.000000+0 2.562000-1 1.562000-1 1.000000-1 0.000000+07925 2151 238 4.906000+0 2.000000+0 1.377000-1 1.520000-2 1.225000-1 0.000000+07925 2151 239 4.645000+1 1.000000+0 1.241300-1 1.300000-4 1.240000-1 0.000000+07925 2151 240 5.810000+1 1.000000+0 1.164000-1 4.400000-3 1.120000-1 0.000000+07925 2151 241 6.030000+1 2.000000+0 1.980000-1 6.800000-2 1.300000-1 0.000000+07925 2151 242 7.840000+1 1.000000+0 1.467000-1 1.670000-2 1.300000-1 0.000000+07925 2151 243 1.070000+2 2.000000+0 1.278000-1 7.800000-3 1.200000-1 0.000000+07925 2151 244 1.223000+2 2.000000+0 1.248000-1 8.000000-4 1.240000-1 0.000000+07925 2151 245 1.442000+2 1.000000+0 1.289000-1 8.900000-3 1.200000-1 0.000000+07925 2151 246 1.513000+2 2.000000+0 1.420000-1 2.200000-2 1.200000-1 0.000000+07925 2151 247 1.629000+2 1.000000+0 1.800000-1 5.000000-2 1.300000-1 0.000000+07925 2151 248 1.649000+2 2.000000+0 1.153000-1 9.300000-3 1.060000-1 0.000000+07925 2151 249 1.903000+2 1.000000+0 1.740000-1 4.400000-2 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2.100000-1 0.000000+07925 2151 439 3.637000+3 2.000000+0 5.740000-1 4.500000-1 1.240000-1 0.000000+07925 2151 440 3.651000+3 2.000000+0 1.280000-1 4.000000-3 1.240000-1 0.000000+07925 2151 441 3.671000+3 1.000000+0 2.240000-1 1.000000-1 1.240000-1 0.000000+07925 2151 442 3.691000+3 2.000000+0 2.650000-1 1.410000-1 1.240000-1 0.000000+07925 2151 443 3.695000+3 1.000000+0 1.320000-1 8.000000-3 1.240000-1 0.000000+07925 2151 444 3.709000+3 1.000000+0 1.770000-1 5.300000-2 1.240000-1 0.000000+07925 2151 445 3.727000+3 2.000000+0 5.370000-1 4.130000-1 1.240000-1 0.000000+07925 2151 446 3.744000+3 1.000000+0 1.960000-1 7.200000-2 1.240000-1 0.000000+07925 2151 447 3.759000+3 1.000000+0 1.820000-1 5.800000-2 1.240000-1 0.000000+07925 2151 448 3.762000+3 2.000000+0 1.470000-1 2.300000-2 1.240000-1 0.000000+07925 2151 449 3.791000+3 1.000000+0 1.290000-1 5.000000-3 1.240000-1 0.000000+07925 2151 450 3.807000+3 2.000000+0 3.410000-1 2.170000-1 1.240000-1 0.000000+07925 2151 451 3.841000+3 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3.370000-1 2.130000-1 1.240000-1 0.000000+07925 2151 477 4.364000+3 2.000000+0 3.240000-1 2.000000-1 1.240000-1 0.000000+07925 2151 478 4.388000+3 1.000000+0 4.580000-1 3.340000-1 1.240000-1 0.000000+07925 2151 479 4.422000+3 2.000000+0 5.790000-1 4.550000-1 1.240000-1 0.000000+07925 2151 480 4.435000+3 2.000000+0 2.190000-1 9.500000-2 1.240000-1 0.000000+07925 2151 481 4.454000+3 1.000000+0 2.270000-1 1.030000-1 1.240000-1 0.000000+07925 2151 482 4.521000+3 2.000000+0 5.640000-1 4.400000-1 1.240000-1 0.000000+07925 2151 483 4.537000+3 1.000000+0 5.390000-1 4.150000-1 1.240000-1 0.000000+07925 2151 484 4.540000+3 2.000000+0 2.260000-1 1.020000-1 1.240000-1 0.000000+07925 2151 485 4.550000+3 1.000000+0 2.070000-1 8.300000-2 1.240000-1 0.000000+07925 2151 486 4.572000+3 2.000000+0 6.080000-1 4.840000-1 1.240000-1 0.000000+07925 2151 487 4.590000+3 2.000000+0 2.240000-1 1.000000-1 1.240000-1 0.000000+07925 2151 488 4.611000+3 1.000000+0 1.940000-1 7.000000-2 1.240000-1 0.000000+07925 2151 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07925 3 16 505 8.100700+6 0.000000+0 8.250000+6 3.440350-3 8.699100+6 1.268210-17925 3 16 506 8.750000+6 1.475590-1 9.000000+6 2.875000-1 9.250000+6 4.600190-17925 3 16 507 9.500000+6 6.304000-1 9.750000+6 7.894770-1 1.000000+7 9.434000-17925 3 16 508 1.025000+7 1.088120+0 1.050000+7 1.224000+0 1.075000+7 1.349590+07925 3 16 509 1.100000+7 1.462000+0 1.125000+7 1.560850+0 1.150000+7 1.648000+07925 3 16 510 1.175000+7 1.728680+0 1.200000+7 1.793000+0 1.225000+7 1.841960+07925 3 16 511 1.250000+7 1.890000+0 1.275000+7 1.937910+0 1.300000+7 1.985000+07925 3 16 512 1.325000+7 2.025650+0 1.350000+7 2.060000+0 1.375000+7 2.092600+07925 3 16 513 1.400000+7 2.122000+0 1.425000+7 2.147710+0 1.450000+7 2.171000+07925 3 16 514 1.475000+7 2.191980+0 1.500000+7 2.211000+0 1.525000+7 2.226380+07925 3 16 515 1.550000+7 2.237000+0 1.575000+7 2.239450+0 1.600000+7 2.232000+07925 3 16 516 1.625000+7 2.215170+0 1.650000+7 2.188000+0 1.675000+7 2.151710+07925 3 16 517 1.700000+7 2.103000+0 1.725000+7 2.044530+0 1.750000+7 1.981920+07925 3 16 518 1.775000+7 1.915100+0 1.800000+7 1.844000+0 1.850000+7 1.689600+07925 3 16 519 1.900000+7 1.536000+0 1.950000+7 1.396960+0 2.000000+7 1.267000+07925 3 16 520 2.050000+7 1.144200+0 2.100000+7 1.030890+0 2.150000+7 9.293830-17925 3 16 521 2.200000+7 8.420000-1 2.250000+7 7.693610-1 2.300000+7 7.096780-17925 3 16 522 2.350000+7 6.616560-1 2.400000+7 6.240000-1 2.450000+7 5.912630-17925 3 16 523 2.500000+7 5.602710-1 2.550000+7 5.331430-1 2.600000+7 5.120000-17925 3 16 524 2.650000+7 4.946300-1 2.700000+7 4.777080-1 2.750000+7 4.616820-17925 3 16 525 2.800000+7 4.470000-1 2.850000+7 4.331260-1 2.900000+7 4.194450-17925 3 16 526 2.950000+7 4.062920-1 3.000000+7 3.940000-1 0.000000+0 0.000000+07925 3 16 527 7925 3 0 528 7.919700+4 1.952740+2 0 0 0 07925 3102 529 6.340000+4 6.340000+4 0 0 1 1227925 3102 530 122 2 0 0 0 07925 3102 531 1.000000-5 0.000000+0 2.530000-2 0.000000+0 5.000000+3 0.000000+07925 3102 532 5.000000+3 2.093650+0 6.000000+3 1.866200+0 7.000000+3 1.668600+07925 3102 533 8.000000+3 1.455000+0 9.000000+3 1.273000+0 1.000000+4 1.172070+07925 3102 534 1.500000+4 8.918000-1 2.000000+4 6.963000-1 2.400000+4 6.396000-17925 3102 535 3.000000+4 5.863000-1 4.000000+4 5.001000-1 4.500000+4 4.570000-17925 3102 536 5.000000+4 4.350500-1 5.500000+4 4.131000-1 6.000000+4 3.963500-17925 3102 537 6.500000+4 3.796000-1 7.000000+4 3.665500-1 7.500000+4 3.535000-17925 3102 538 8.000000+4 3.417500-1 8.500000+4 3.300000-1 9.000000+4 3.225000-17925 3102 539 9.500000+4 3.150000-1 1.000000+5 3.099000-1 1.200000+5 2.921000-17925 3102 540 1.500000+5 2.719000-1 1.700000+5 2.617000-1 1.800000+5 2.567000-17925 3102 541 1.900000+5 2.534000-1 2.000000+5 2.502000-1 2.100000+5 2.469000-17925 3102 542 2.200000+5 2.445000-1 2.300000+5 2.415000-1 2.350000+5 2.403000-17925 3102 543 2.400000+5 2.388000-1 2.450000+5 2.374000-1 2.500000+5 2.360000-17925 3102 544 2.600000+5 2.331000-1 2.700000+5 2.299000-1 2.800000+5 2.141000-17925 3102 545 3.000000+5 1.999000-1 3.250000+5 1.877000-1 3.500000+5 1.778000-17925 3102 546 3.750000+5 1.689000-1 4.000000+5 1.614000-1 4.250000+5 1.538000-17925 3102 547 4.500000+5 1.462000-1 4.750000+5 1.389000-1 5.000000+5 1.324000-17925 3102 548 5.200000+5 1.270000-1 5.400000+5 1.236000-1 5.700000+5 1.186000-17925 3102 549 6.000000+5 1.084000-1 6.500000+5 1.002000-1 7.000000+5 9.640000-27925 3102 550 7.500000+5 9.280000-2 8.000000+5 8.970000-2 8.500000+5 8.690000-27925 3102 551 9.000000+5 8.430000-2 9.400000+5 8.250000-2 9.600000+5 8.180000-27925 3102 552 9.800000+5 8.100000-2 1.000000+6 8.030000-2 1.100000+6 7.720000-27925 3102 553 1.250000+6 7.290000-2 1.400000+6 6.940000-2 1.600000+6 6.650000-27925 3102 554 1.800000+6 5.960000-2 2.000000+6 5.340000-2 2.200000+6 4.330000-27925 3102 555 2.400000+6 3.600000-2 2.500000+6 3.340000-2 2.750000+6 2.904190-27925 3102 556 3.000000+6 2.480330-2 3.250000+6 2.068920-2 3.500000+6 1.711460-27925 3102 557 3.750000+6 1.433960-2 4.000000+6 1.205200-2 4.250000+6 1.091040-27925 3102 558 4.500000+6 9.576000-3 4.750000+6 8.393720-3 5.000000+6 7.314300-37925 3102 559 5.500000+6 6.202050-3 6.000000+6 4.719000-3 6.500000+6 3.574800-37925 3102 560 7.000000+6 2.415000-3 7.500000+6 9.986730-4 8.000000+6 8.180000-47925 3102 561 8.500000+6 9.384900-4 9.000000+6 1.080000-3 9.500000+6 1.240300-37925 3102 562 1.000000+7 1.400000-3 1.050000+7 1.549700-3 1.100000+7 1.690000-37925 3102 563 1.150000+7 1.787400-3 1.200000+7 1.830000-3 1.250000+7 1.829400-37925 3102 564 1.300000+7 1.720000-3 1.350000+7 1.521100-3 1.400000+7 1.320000-37925 3102 565 1.450000+7 1.124200-3 1.500000+7 9.289400-4 1.550000+7 7.382300-47925 3102 566 1.600000+7 5.560000-4 1.650000+7 4.099100-4 1.700000+7 3.145000-47925 3102 567 1.750000+7 2.540900-4 1.800000+7 2.130000-4 1.900000+7 1.439300-47925 3102 568 2.000000+7 9.830000-5 2.100000+7 7.171000-5 2.200000+7 5.210000-57925 3102 569 2.300000+7 3.975500-5 2.400000+7 3.070000-5 2.500000+7 2.494600-57925 3102 570 2.600000+7 1.986900-5 2.700000+7 1.600000-5 2.800000+7 1.307500-57925 3102 571 2.900000+7 1.069900-5 3.000000+7 9.200000-6 0.000000+0 0.000000+07925 3102 572 7925 3 0 573 7925 0 0 574 0 0 0 575 9.023200+4 2.300400+2 1 1 0 19040 1451 1 0.000000+0 1.000000+0 0 0 0 69040 1451 2 1.000000+0 0.000000+0 0 0 10 69040 1451 3 0.000000+0 0.000000+0 0 0 244 49040 1451 4 90-TH-232 BNL,ANL+ EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+ 9040 1451 5 DIST-FEB90 900205 9040 1451 6 ----ENDF/B-VI MATERIAL 9040 9040 1451 7 -----INCIDENT NEUTRON DATA 9040 1451 8 ------ENDF-6 FORMAT 9040 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9040 1451 10 DOSIMETRY 9040 1451 11 ******************************************************************9040 1451 12 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 9040 1451 13 * * * * * * *9040 1451 14 9040 1451 15 PRINCIPAL EVALUATORS- J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH9040 1451 16 AND J.F.WHALEN(ANL) 9040 1451 17 R.J.HOWERTON(LLL) 9040 1451 18 B.R.LEONARD ET.AL.(BNW) 9040 1451 19 G.DE SAUSSURE,R.L.MACKLIN AND R.GWIN(ORNL) 9040 1451 20 D.K. OLSEN (ORNL) 9040 1451 21 THIS EVALUATION ASSEMBLED BY M.R.BHAT(BNL) USING THE FOLLOWING 9040 1451 22 EVALUATIONS 9040 1451 23 9040 1451 24 NU-BAR(PROMPT) - R.GWIN(ORNL)(REF 4) 9040 1451 25 DELAYED NEUTRONS -M.BRADY AND T.ENGLAND (REF 5) 9040 1451 26 ENERGY RELEASE IN FISSION - R.SHER ET.AL.(STANFORD)(REF 6) 9040 1451 27 THERMAL RANGE - B.R.LEONARD JR.ET.AL.(REF 2) WITH CHANGES BY9040 1451 28 M.R.BHAT(REF 31) 9040 1451 29 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) 9040 1451 30 9040 1451 31 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN 9040 1451 32 (REF 3)WITH CHANGES BY M.R.BHAT(REF 31) 9040 1451 33 SMOOTH CROSS-SECTIONS 50KEV TO 20MEV - 9040 1451 34 TOTAL - A.B.SMITH(ANL)(REF1) 9040 1451 35 FISSION - W.P.POENITZ(ANL)(REF1) 9040 1451 36 CAPTURE - W.P.POENITZ(ANL)(REF1) 9040 1451 37 INELASTIC AND OTHER CROSS-SECTIONS - J.W.MEADOWS ET.AL.(ANL)9040 1451 38 (REF1)9040 1451 39 FISSION NEUTRON SPECTRUM - M.R.BHAT(BNL) (REF 31) 9040 1451 40 GAMMA-RAY PRODUCTION - R.J.HOWERTON(LLL) 9040 1451 41 9040 1451 42 MF = 1 9040 1451 43 9040 1451 44 MT=452(NU-BAR TOTAL) CONSISTENT WITH MT=455 AND MT=456 9040 1451 45 MT=455(DELAYED NEUTRON YIELDS) (REF 5) 9040 1451 46 MT=456(PROMPT NU-BAR) BASED ON CF-252 NU-BAR(PROMPT)=3.757 9040 1451 47 EVALUATION BY GWIN(ORNL) BASED ON A STRAIGHT LINE LEAST- 9040 1451 48 SQUARES FIT TO THE AVAILABLE DATA LISTED IN REF.4. 9040 1451 49 MT=458(ENERGY RELEASE IN FISSION)EVALUATION BY R.SHER ET.AL(STAN-9040 1451 50 FORD)(REF 6) 9040 1451 51 9040 1451 52 MF = 2 9040 1451 53 9040 1451 54 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) RESONANCE 9040 1451 55 PARAMETERS ARE UNCERTAINTY WEIGHTED AVERAGE OF AVAILABLE 9040 1451 56 DATA AND THE TWO BOUND RESONANCES OF REF 2 WITH MF=3 FILES 9040 1451 57 FOR TRUNCATION, MISSED P-WAVES, 1/V CAPTURE, AND SMOOTH 9040 1451 58 CONNECTION TO THE END OF THE THERMAL REGION CROSS SECTIONS 9040 1451 59 AT 5 EV. AVERAGE CAPTURE WIDTH = 2.44 X 10-2 EV, S-WAVE 9040 1451 60 STRENGTH FUNCTION = 0.826 X 10-4, P-WAVE STRENGTH FUNCTION 9040 1451 61 ASSUMED = 1.6 X 10-4, 5 TO 4000 EV INFINITLY DILUTE CAPTURE 9040 1451 62 RESONANCE INTEGRAL = 81.58 B. 9040 1451 63 9040 1451 64 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN(REF3)9040 1451 65 EVALUATION OF CAPTURE CROSS-SECTION USED FROM 4 TO 25KEV. 9040 1451 66 ABOVE 25KEV THIS WAS JOINED SMOOTHLY TO THE POENITZ EVALUAT-9040 1451 67 ION FROM 40KEV TO 50KEV. THESE WERE THEN FITTED USING THE 9040 1451 68 CODE UR(REF30) TO GIVE THE UNRESOLVED RESONANCE 9040 1451 69 PARAMETERS. IN THIS FIT THE S- AND D-WAVE PARAMETERS WERE 9040 1451 70 KEPT CONSTANT AT THE VALUES GIVEN BY DE SAUSSURE AND MACKLIN9040 1451 71 THE P-WAVE GAMMA WIDTH WAS INCREASED TO 2.52E-02EV AND THE 9040 1451 72 P-WAVE REDUCED NEUTRON WIDTH VARIED TO FIT THE CAPTURE 9040 1451 73 CROSS-SECTION. 9040 1451 74 THE CAPTURE CROSS-SECTION INPUT USED TO EXTRACT THE 9040 1451 75 UNRESOLVED RESONANCE PARAMETERS IN FILE 2/151 ARE GIVEN HERE9040 1451 76 FROM 4KEV TO 50KEV AT 1KEV INTERVAL. THESE ARE - 9040 1451 77 1.130,0.982,0.908,0.850,0.803,0.764,0.730,0.702,0.677,0.655,9040 1451 78 0.636,0.618,0.601,0.585,0.571,0.557,0.545,0.533,0.521,0.511,9040 1451 79 0.500,0.491,0.482,0.473,0.465,0.459,0.452,0.448,0.443,0.439,9040 1451 80 0.435,0.431,0.429,0.426,0.424,0.422,0.420,0.420,0.420,0.420,9040 1451 81 0.417,0.414,0.411,0.407,0.401,0.395,0.390. 9040 1451 82 9040 1451 83 MF = 3 9040 1451 84 9040 1451 85 THERMAL REGION(1.0E-05 TO 5.0EV) - MAINLY BASED ON LEONARD(REF 2)9040 1451 86 EVALUATION WITH THE FOLLOWING CHANGES. THE CAPTURE CROSS- 9040 1451 87 SECTION IN REF 2 NORMALISED TO 7.40B AT 2.53E-02 EV AND USED9040 1451 88 FROM 1.0E-05EV TO 2.53E-02EV. FROM 2.6E-02EV TO 14EV THE FIT9040 1451 89 TO THEIR CAPTURE DATA BY CHRIEN AND LIOU(REF 32) USED. THE 9040 1451 90 SCATTERING CROSS-SECTION WAS CHANGED TO ACCOMMODATE CHANGES 9040 1451 91 IN CAPTURE USING TOTAL SLIGHTLY MODIFIED FROM THAT GIVEN IN 9040 1451 92 REF 2. THESE SMALL CHANGES ARE DESCRIBED IN REF 31. 9040 1451 93 2.53E-02EV CROSS-SECTION CAPTURE 7.4 B 9040 1451 94 DILUTE CAPTURE RESONANCE INTEGRAL 86.1 B 9040 1451 95 (0.5 TO 2.0E+07 EV) 9040 1451 96 TOTAL CROSS-SECTION FROM 50KEV TO 20MEV - A.B.SMITH(ANL)(REF1) 9040 1451 97 MAINLY BASED ON REF9-13 FROM 0.025 TO 1 MEV,FROM 1 TO 5MEV 9040 1451 98 BASED ON REF 9-11,14-16.FROM 5 TO 15MEV EVALUATION BASED ON 9040 1451 99 REF11,15,17.FROM 15 TO 20MEV EVALUATION BASED ON NUCLEAR 9040 1451 100 MODEL CALCULATIONS DESCRIBED IN REF18 AND 19.TOTAL CROSS- 9040 1451 101 SECTION MODIFIED BY ABOUT 3PER-CENT BETWEEN 50 TO 150KEV 9040 1451 102 JOIN SMOOTHLY WITH THAT CALCULATED FROM THE UNRESOLVED 9040 1451 103 RESONANCE PARAMETERS BELOW 50KEV. 9040 1451 104 ELASTIC SCATTERING CROSS-SECTION FROM 50KEV TO 20MEV OBTAINED AS 9040 1451 105 THE DIFFERENCE BETWEEN TOTAL AND OTHER PARTIAL CROSSSECTIONS9040 1451 106 FISSION CROSS-SECTION UP TO 20MEV - W.P.POENITZ(REF1) DETAILS OF 9040 1451 107 THE EVALUATION AND THE DATA SETS USED ARE IN REF1. 9040 1451 108 FISSION CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 109 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 110 CAPTURE CROSS-SECTION 50KEV -20MEV - W.P.POENITZ. DETAILS OF THE 9040 1451 111 EVALUATION AND THE DATA SETS USED ARE GIVEN IN REF1. 9040 1451 112 CAPTURE CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 113 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 114 INELASTIC SCATTERING CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF 1) 9040 1451 115 EVALUATION BASED MAINLY ON THE DATA IN REF 20-25. INTER- 9040 1451 116 POLATED BETWEEN EXPERIMENTAL DATA AND EXTRAPOLATED TO 20MEV 9040 1451 117 USING COUPLED CHANNEL NUCLEAR MODEL CALCULATIONS OF REF26. 9040 1451 118 (N,2N) AND (N,3N) CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF1) 9040 1451 119 EXPERIMENTAL DATA USED(REF1) RENORMALISED TO A COMMON SET OF9040 1451 120 STANDARDS BEFORE EVALUATION. 9040 1451 121 9040 1451 122 MF = 4 9040 1451 123 9040 1451 124 ELASTIC ANGULAR DISTRIBUTIONS - J.W.MEADOWS ET.AL.(REF1) 9040 1451 125 EVALUATION BASED ON EXPERIMENTAL DATA TO 2.5MEV(REF1) AND 9040 1451 126 NUCLEAR MODEL CALCULATIONS FROM 2.5 TO 20MEV(REF26) 9040 1451 127 INELASTIC ANGULAR DISTRIBUTIONS - J.W.MEADOWS ET.AL.(REF1) 9040 1451 128 ANGULAR DISTRIBUTIONS BASED ON COUPLED CHANNEL NUCLEAR MODEL9040 1451 129 CALCULATED AND EXPERIMENTAL DATA FOR INCIDENT ENERGIES LESS 9040 1451 130 OR EQUAL TO 2.5MEV AND ON MODEL CALCULATIONS(REF 26) FROM 9040 1451 131 2.5 TO 20MEV. 9040 1451 132 MT=16,17,18 ASSUMED TO BE ISOTROPIC 9040 1451 133 9040 1451 134 MF = 5 9040 1451 135 9040 1451 136 MT=16,17 - J.W.MEADOWS ET.AL.(REF 1 ) 9040 1451 137 THESE DETERMINED USING THE STATISTICAL MODEL OF SEGEV(REF27)9040 1451 138 WITH A PRECOMPOUND COMPONENT 9040 1451 139 MT=18 THE VARIATION OF THE MEAN ENERGY OF THE FISSION SPECTRUM 9040 1451 140 WAS CALCULATED USING THE METHOD OF HOWERTON AND DOYAS(REF28)9040 1451 141 AND NORMALISED TO THE 14.3MEV DATA OF VASIL'EV ET.AL(REF29) 9040 1451 142 TO GIVE A MEAN ENERGY OF 2.255MEV AT 14MEV AND THE 9040 1451 143 CORRESPONDING WATT SPECTRUM SHAPE PARAMETERS OBTAINED. 9040 1451 144 MT=91 - J.W.MEADOWS ET.AL.(REF1) 9040 1451 145 EVALUATION BASED ON THE MEASURED EVAPORATION TEMPERATURES OF9040 1451 146 REF23 AND 25. IN ADDITION, A HARDER PRECOMPOUND COMPONENT 9040 1451 147 VARYING FROM 0.0 AT 6MEV TO 20PER-CENT AT 20MEV PATTERNED 9040 1451 148 AFTER U-238 WAS ADDED 9040 1451 149 MT=455 (REF 5) 9040 1451 150 9040 1451 151 MF = 12 9040 1451 152 MT = 18 9040 1451 153 FOR ALL INCIDENT ENERGIES THE MULTIPLICITY OF PHOTONS FROM 9040 1451 154 FISSION WAS DERIVED FROM THE DATA OF REF. 34 WHICH REPORTED THE 9040 1451 155 PHOTON SPECTRUM AND TOTAL PHOTON ENERGY FOR THERMAL FISSION OF 9040 1451 156 U-235. IT WAS ASSUMED THAT THE MULTIPLICITY IS INDEPENDENT OF 9040 1451 157 INCIDENT NEUTRON ENERGY. 9040 1451 158 MT = 102 9040 1451 159 FOR ALL INCIDENT NEUTRON ENERGIES PHOTON PRODUCTION FROM 9040 1451 160 NEUTRON CAPTURE IS REPRESENTED BY AN ENERGY DEPENDENT MULTIPLI- 9040 1451 161 CITY TO BE APPLIED TO THE (N,GAMMA) CROSS SECTION AND BY AN 9040 1451 162 ENERGY INDEPENDENT SPECTRUM. THE SPECTRUM USED WAS BASED ON AN 9040 1451 163 UNDOCUMENTED MEASURED SPECTRUM FOR U-238 WITH A MINOR ADJUSTMENT9040 1451 164 FOR SMALL Q-VALUE DIFFERENCE BETWEEN TH-232 AND U-238. THE 9040 1451 165 AVERAGE ENERGY OF TE ASSUMED SPECTRUM WAS THEN DIVIDED INTO THE 9040 1451 166 Q-VALUE TO OBTAIN A MULTIPLICITY AT ZERO NEUTRON ENERGY. THE 9040 1451 167 MULTIPLICITY AT 20 MEV WAS OBTAINED (REF 35) BY USE OF THE 9040 1451 168 FORMULA M(E) = M0(EN+Q)/Q WHERE M0 IS THE MULTIPLICITY AT ZERO 9040 1451 169 NEUTRON ENERGY AS DESCRIBED ABOVE. 9040 1451 170 MF = 13 9040 1451 171 MT = 3 9040 1451 172 EXPLICIT REPRESENTATION OF THREE PHOTONS (.04971, .1632 AND 9040 1451 173 .3344 MEV) FROM INELASTIC SCATTERING WAS DERIVED FROM FILE 3 9040 1451 174 DATA AND KNOWN BRANCHING RATIOS. FOR INCIDENT NEUTRON ENERGIES 9040 1451 175 GT .7251 MEV THE METHOD OF REF. 36 WAS USED TO CALCULATE PHOTON 9040 1451 176 PRODUCTION CROSS SECTIONS AND SPECTRA FROM ALL REACTIONS EXCEPT 9040 1451 177 PHOTONS FROM THE FIRST THREE INELASTIC GROUPS, NEUTRON CAPTURE 9040 1451 178 AND NEUTRON-INDUCED FISSION. 9040 1451 179 MF = 14 9040 1451 180 MT = 3,18,102 ALL PHOTONS WERE ASSUMED TO BE ISOTROPIC 9040 1451 181 MF = 15 9040 1451 182 MT = 3 9040 1451 183 PHOTON SPECTRA WERE OBTAINED USING THE METHOD OF REF. 36. 9040 1451 184 MT = 18 9040 1451 185 THE MEASURED PHOTON SPECTRUM OF REF. 34 WAS USED FOR ALL 9040 1451 186 INCIDENT NEUTRON ENERGIES BECAUSE THERE ARE NO EXPERIMENTAL DATA9040 1451 187 FOR TH-232. 9040 1451 188 MT = 102 9040 1451 189 THE SECTRUM USED WAS BASED ON AN UNDOCUMENTED MEASURED 9040 1451 190 PHOTON SPECTRUM AT THERMAL NEUTRON ENERGY FOR U-238 WITH A MINOR9040 1451 191 ADJUSTMENT FOR THE SMALL Q-VALUE DIFFERENCE BETWEEN U-238 AND TH9040 1451 192 -232. IT WAS ASSUMED THAT THE SPECTRUM REMAINED UNCHANGED FOR 9040 1451 193 ALL INCIDENT NEUTRON ENERGIES. 9040 1451 194 MF = 31 9040 1451 195 MT = 452 BY R. GWIN (REF. 4) 9040 1451 196 MF = 33 9040 1451 197 MT = 18,102 BASED ON ERROR ESTIMATES IN THESE CROSS SECTIONS IN 9040 1451 198 REF. 32,1 9040 1451 199 9040 1451 200 PRINCIPAL REFERENCES 9040 1451 201 9040 1451 202 1 J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH,J.F.WHALEN AND 9040 1451 203 R.J.HOWERTON ANL/NDM-35(FEB.1978) 9040 1451 204 2 D.F.NEWMAN,B.R.LEONARD JR.ET.AL. EPRI NP-222,MAY1977 9040 1451 205 3 G.DE SAUSSURE AND R.L.MACKLIN ORNL/TM-6161(ENDF-255)DEC.1977 9040 1451 206 4 R.GWIN ORNL/TM-6245(ENDF-262)MAY1978. 9040 1451 207 5 M.C.BRADY AND T.R.ENGLAND NSE 103,129(1989) 9040 1451 208 **DATA INSERTED INTO FILE AT BNL** 9040 1451 209 6. R.SHER + C.BECK EPRI 1771/81 + REV. 1/83 + PC TO MAGURNO 2/83 9040 1451 210 7 S.F.MUGHABGHAB AND D.I.GARBER,BNL-325(3RD EDN)1973 9040 1451 211 8 H.DERRIEN,NEANDC(E)163U(1975) 9040 1451 212 9 J.F.WHALEN AND A.B.SMITH,NUC.SCI.AND ENG.67,129(1978) 9040 1451 213 10J.MEADOWS,A.SMITH AND J.WHALEN,PRIV.COMM.1977 9040 1451 214 11U.FASOLI ET.AL.NUCL.PHYS.A151,369(1970) 9040 1451 215 12M.DIVADEENAM ET.AL. DISS.ABS.28,3834(1968) 9040 1451 216 13C.UTTLEY ET.AL.66PARIS,1,165,1966 9040 1451 217 14L.GREEN ET.AL.PROC.CONF.ON NUCL.CROSS-SECTIONS AND TECHNOLOGY, 9040 1451 218 KNOXVILLE,1,325(1971) 9040 1451 219 15D.FOSTER ET.AL.PRIV.COMM(1967) 9040 1451 220 16R.BATCHELOR ET.AL.NUCL.PHYS.65,236(1956) 9040 1451 221 17J.COON ET.AL.PHYS.REV.88,562(1952) 9040 1451 222 18P.GUENTHER ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(19779040 1451 223 19C.PHILIS ET.AL. ANL/NDM-28(1977) 9040 1451 224 20A.SMITH PHYS.REV.126,718(1962) 9040 1451 225 21W.MCMURRAY ET.AL.SOUTHERN UNIV.NUCLEAR INST.REPORT,SUNI-41(19759040 1451 226 22J.HAOUAT ETAL.PROC.INER.CONF.ON INTERACTIONS OF NEUTRONS WITH 9040 1451 227 NUCLEI,CONF-760715(1976) 9040 1451 228 23R.BATCHELOR AND J.TOWLE,NUCL.PHYS.65,236(1965) 9040 1451 229 24R.BATCHELOR AND J.TOWLE,PROC.PHYS.SOC.73,193(1959) 9040 1451 230 25A.SMITH,PRIV.COMM,1970 AND 1977 9040 1451 231 26A.SMITH ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(1977) 9040 1451 232 27M.SEGEV.ET.AL.TRANS.AM.NUCL.SOC,22,679(1975)AND PRIV.COMM(1976)9040 1451 233 28R.J.HOWERTON AND R.J.DOYAS,NUCL.SCI AND ENG,46,414(1971) 9040 1451 234 29YU.A.VASIL'EV ET.AL.PHYSICS OF NUCLEAR FISSION,N.A.PERFILOV AND9040 1451 235 V.P.EISMONT(EDS),ISRAEL PROGRAM OF SCI.TRANS.JERUSALEM(1964) 9040 1451 236 30E.M.PENNINGTON,PRIVATE COMMUNICATION(1973) 9040 1451 237 31M.R.BHAT,ENDF-268(TO BE PUBLISHED) 9040 1451 238 32R.E.CHRIEN AND H.LIOU(1978) TO BE PUBLISHED 9040 1451 239 33W.P.POENITZ,ANL/NDM-45(1978) TO BE PUBLISHED 9040 1451 240 34R.W.PEELE AND F.C.MAIENSCHEIN, NUCL.SCI.ENG. 40, 485 (1970) 9040 1451 241 35R.J. HOWERTON, D.E. CULLEN, R.C. HAIGHT, M.H. MACGREGOR, S.T. 9040 1451 242 PERKINS, AND E.F. PLECHARTY, "THE LLL EVALUATED NUCLEAR DATA 9040 1451 243 LIBRARY (ENDL): EVALUATION TECHNIQUES, REACTION INDEX, AND 9040 1451 244 DESCRIPTIONS OF INDIVIDUAL EVALUATIONS," UCRL-50400, VOL. 15, 9040 1451 245 PART A, LAWERENCE LIVERMORE LABORATORY (1975). 9040 1451 246 36S.T.PERKINS,R.C.HAIGHT AND R.J.HOWERTON,NUCL.SCI.ENG.57,1(1975)9040 1451 247 37D.K.OLSEN, ORNL/TM-8056(1982), ENDF-319. 9040 1451 248 1 451 252 19040 1451 249 2 151 691 09040 1451 250 3 18 37 09040 1451 251 33 18 9 09040 1451 252 9040 1 0 253 9040 0 0 254 9.023200+4 2.300400+2 0 0 1 09040 2151 255 9.023200+4 1.000000+0 0 0 2 09040 2151 256 5.000000+0 4.000000+3 1 2 0 09040 2151 257 0.000000+0 9.720000-1 0 0 2 09040 2151 258 2.300400+2 0.000000+0 0 0 1458 2439040 2151 259 -2.220000+1 5.000000-1 2.746600-2 6.266000-3 2.120000-2 0.000000+09040 2151 260 -2.952000+0 5.000000-1 2.212800-2 9.277000-4 2.120000-2 0.000000+09040 2151 261 2.180600+1 5.000000-1 2.716000-2 2.060000-3 2.510000-2 0.000000+09040 2151 262 2.346400+1 5.000000-1 3.045000-2 3.850000-3 2.660000-2 0.000000+09040 2151 263 5.951900+1 5.000000-1 2.780000-2 3.800000-3 2.400000-2 0.000000+09040 2151 264 6.923200+1 5.000000-1 6.620000-2 4.320000-2 2.300000-2 0.000000+09040 2151 265 1.130300+2 5.000000-1 3.663000-2 1.313000-2 2.350000-2 0.000000+09040 2151 266 1.208700+2 5.000000-1 4.651000-2 2.251000-2 2.400000-2 0.000000+09040 2151 267 1.291900+2 5.000000-1 2.258100-2 3.381000-3 1.920000-2 0.000000+09040 2151 268 1.543600+2 5.000000-1 2.459900-2 1.986000-4 2.440000-2 0.000000+09040 2151 269 1.703900+2 5.000000-1 8.655000-2 6.125000-2 2.530000-2 0.000000+09040 2151 270 1.927200+2 5.000000-1 3.840000-2 1.690000-2 2.150000-2 0.000000+09040 2151 271 1.994000+2 5.000000-1 2.964000-2 1.024000-2 1.940000-2 0.000000+09040 2151 272 2.195200+2 5.000000-1 2.444400-2 4.364000-5 2.440000-2 0.000000+09040 2151 273 2.212900+2 5.000000-1 5.475000-2 3.025000-2 2.450000-2 0.000000+09040 2151 274 2.516500+2 5.000000-1 5.788000-2 3.178000-2 2.610000-2 0.000000+09040 2151 275 2.632300+2 5.000000-1 4.382000-2 2.222000-2 2.160000-2 0.000000+09040 2151 276 2.859100+2 5.000000-1 5.353000-2 3.103000-2 2.250000-2 0.000000+09040 2151 277 3.056200+2 5.000000-1 5.088000-2 2.888000-2 2.200000-2 0.000000+09040 2151 278 3.094700+2 5.000000-1 2.445600-2 5.641000-5 2.440000-2 0.000000+09040 2151 279 3.291100+2 5.000000-1 1.013800-1 7.448000-2 2.690000-2 0.000000+09040 2151 280 3.420000+2 5.000000-1 6.039000-2 3.889000-2 2.150000-2 0.000000+09040 2151 281 3.653600+2 5.000000-1 5.046000-2 2.606000-2 2.440000-2 0.000000+09040 2151 282 3.694900+2 5.000000-1 5.011000-2 2.571000-2 2.440000-2 0.000000+09040 2151 283 4.011000+2 5.000000-1 3.552000-2 1.112000-2 2.440000-2 0.000000+09040 2151 284 4.029200+2 5.000000-1 2.450600-2 1.059000-4 2.440000-2 0.000000+09040 2151 285 4.210300+2 5.000000-1 2.490200-2 5.025000-4 2.440000-2 0.000000+09040 2151 286 4.544700+2 5.000000-1 2.562700-2 1.227000-3 2.440000-2 0.000000+09040 2151 287 4.627700+2 5.000000-1 8.847000-2 6.407000-2 2.440000-2 0.000000+09040 2151 288 4.890500+2 5.000000-1 8.297000-2 5.857000-2 2.440000-2 0.000000+09040 2151 289 5.002700+2 5.000000-1 2.445000-2 4.972000-5 2.440000-2 0.000000+09040 2151 290 5.106400+2 5.000000-1 2.870700-2 4.307000-3 2.440000-2 0.000000+09040 2151 291 5.288000+2 5.000000-1 3.851000-2 1.411000-2 2.440000-2 0.000000+09040 2151 292 5.405600+2 5.000000-1 2.543000-2 1.030000-3 2.440000-2 0.000000+09040 2151 293 5.701300+2 5.000000-1 5.246000-2 2.806000-2 2.440000-2 0.000000+09040 2151 294 5.784300+2 5.000000-1 2.659900-2 2.199000-3 2.440000-2 0.000000+09040 2151 295 5.944300+2 5.000000-1 2.451800-2 1.178000-4 2.440000-2 0.000000+09040 2151 296 5.985700+2 5.000000-1 3.508000-2 1.068000-2 2.440000-2 0.000000+09040 2151 297 6.182000+2 5.000000-1 2.922700-2 4.827000-3 2.440000-2 0.000000+09040 2151 298 6.569800+2 5.000000-1 7.334000-2 4.894000-2 2.440000-2 0.000000+09040 2151 299 6.656300+2 5.000000-1 5.120000-2 2.680000-2 2.440000-2 0.000000+09040 2151 300 6.755500+2 5.000000-1 2.341000-1 2.097000-1 2.440000-2 0.000000+09040 2151 301 6.878500+2 5.000000-1 7.586000-2 5.146000-2 2.440000-2 0.000000+09040 2151 302 7.014700+2 5.000000-1 3.942000-2 1.502000-2 2.440000-2 0.000000+09040 2151 303 7.132600+2 5.000000-1 5.369000-2 2.929000-2 2.440000-2 0.000000+09040 2151 304 7.245300+2 5.000000-1 2.452600-2 1.258000-4 2.440000-2 0.000000+09040 2151 305 7.415000+2 5.000000-1 2.147000-1 1.903000-1 2.440000-2 0.000000+09040 2151 306 7.720500+2 5.000000-1 2.452800-2 1.277000-4 2.440000-2 0.000000+09040 2151 307 7.790900+2 5.000000-1 3.716000-2 1.276000-2 2.440000-2 0.000000+09040 2151 308 8.046400+2 5.000000-1 2.019000-1 1.775000-1 2.440000-2 0.000000+09040 2151 309 8.293300+2 5.000000-1 2.464900-2 2.487000-4 2.440000-2 0.000000+09040 2151 310 8.373300+2 5.000000-1 2.594000-2 1.540000-3 2.440000-2 0.000000+09040 2151 311 8.427600+2 5.000000-1 5.368000-2 2.928000-2 2.440000-2 0.000000+09040 2151 312 8.667600+2 5.000000-1 3.870000-2 1.430000-2 2.440000-2 0.000000+09040 2151 313 8.905800+2 5.000000-1 6.350000-2 3.910000-2 2.440000-2 0.000000+09040 2151 314 9.070200+2 5.000000-1 2.629700-2 1.897000-3 2.440000-2 0.000000+09040 2151 315 9.275800+2 5.000000-1 2.476200-2 3.623000-4 2.440000-2 0.000000+09040 2151 316 9.437199+2 5.000000-1 6.973000-2 4.533000-2 2.440000-2 0.000000+09040 2151 317 9.632000+2 5.000000-1 3.116400-2 6.764000-3 2.440000-2 0.000000+09040 2151 318 9.834699+2 5.000000-1 6.120000-2 3.680000-2 2.440000-2 0.000000+09040 2151 319 9.910900+2 5.000000-1 1.108000-1 8.640000-2 2.440000-2 0.000000+09040 2151 320 1.011100+3 5.000000-1 1.462000-1 1.218000-1 2.440000-2 0.000000+09040 2151 321 1.030000+3 5.000000-1 2.461900-2 2.192000-4 2.440000-2 0.000000+09040 2151 322 1.039700+3 5.000000-1 3.525000-2 1.085000-2 2.440000-2 0.000000+09040 2151 323 1.065000+3 5.000000-1 2.977100-2 5.371000-3 2.440000-2 0.000000+09040 2151 324 1.077800+3 5.000000-1 3.502000-2 1.062000-2 2.440000-2 0.000000+09040 2151 325 1.093600+3 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0.000000+09040 2151 884 3.900000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 885 4.000000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 886 4.100000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 887 4.200000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 888 4.300000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 889 4.400000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 890 4.500000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 891 4.600000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 892 4.700000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 893 4.800000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 894 4.900000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 895 5.000000+4 8.500000+0 0.000000+0 7.500000-4 2.130000-2 0.000000+09040 2151 896 2.500000+0 0.000000+0 2 0 288 479040 2151 897 0.000000+0 0.000000+0 0.000000+0 1.000000+0 0.000000+0 0.000000+09040 2151 898 4.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 899 5.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 900 6.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 901 7.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 902 8.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 903 9.000000+3 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 904 1.000000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 905 1.100000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 906 1.200000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 907 1.300000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 908 1.400000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 909 1.500000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 910 1.600000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 911 1.700000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 912 1.800000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 913 1.900000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 914 2.000000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 915 2.100000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 916 2.200000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 917 2.300000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 918 2.400000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 919 2.500000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 920 2.600000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 921 2.700000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 922 2.800000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 923 2.900000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 924 3.000000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 925 3.100000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 926 3.200000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 927 3.300000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 928 3.400000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 929 3.500000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 930 3.600000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 931 3.700000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 932 3.800000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 933 3.900000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 934 4.000000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 935 4.100000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 936 4.200000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 937 4.300000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 938 4.400000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 939 4.500000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 940 4.600000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 941 4.700000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 942 4.800000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 943 4.900000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 944 5.000000+4 5.700000+0 0.000000+0 5.000000-4 2.130000-2 0.000000+09040 2151 945 9040 2 0 946 9040 0 0 947 9.023200+4 2.300400+2 0 0 0 09040 3 18 948 1.884700+8 1.884700+8 0 0 1 1019040 3 18 949 101 2 0 0 0 09040 3 18 950 1.000000-5 0.000000+0 5.000000+5 0.000000+0 6.000000+5 1.100000-59040 3 18 951 7.000000+5 3.400000-5 7.500000+5 9.100000-5 7.750000+5 1.350000-49040 3 18 952 8.000000+5 1.350000-4 8.500000+5 1.840000-4 9.000000+5 5.250000-49040 3 18 953 9.300000+5 9.860001-4 9.500000+5 1.320000-3 9.600000+5 1.170000-39040 3 18 954 9.900000+5 1.080000-3 1.000000+6 1.250000-3 1.030000+6 1.770000-39040 3 18 955 1.050000+6 1.860000-3 1.100000+6 2.100000-3 1.150000+6 2.850000-39040 3 18 956 1.200000+6 5.230000-3 1.250000+6 8.020000-3 1.300000+6 1.910000-29040 3 18 957 1.320000+6 2.410000-2 1.350000+6 4.020000-2 1.400000+6 5.630000-29040 3 18 958 1.440000+6 6.530000-2 1.450000+6 6.530000-2 1.460000+6 6.430000-29040 3 18 959 1.500000+6 7.240000-2 1.550000+6 1.010000-1 1.575000+6 1.140000-19040 3 18 960 1.585000+6 1.190000-1 1.600000+6 1.150000-1 1.610000+6 1.140000-19040 3 18 961 1.625000+6 1.110000-1 1.650000+6 9.560000-2 1.675000+6 8.350000-29040 3 18 962 1.700000+6 8.660000-2 1.710000+6 9.270000-2 1.725000+6 9.370000-29040 3 18 963 1.750000+6 8.360000-2 1.775000+6 7.560000-2 1.800000+6 7.770000-29040 3 18 964 1.840000+6 8.980000-2 1.850000+6 1.010000-1 1.865000+6 9.280000-29040 3 18 965 1.885000+6 1.110000-1 1.900000+6 1.190000-1 1.950000+6 1.140000-19040 3 18 966 2.000000+6 1.190000-1 2.050000+6 1.160000-1 2.150000+6 1.300000-19040 3 18 967 2.200000+6 1.330000-1 2.250000+6 1.320000-1 2.300000+6 1.250000-19040 3 18 968 2.350000+6 1.160000-1 2.400000+6 1.140000-1 2.450000+6 1.160000-19040 3 18 969 2.500000+6 1.170000-1 2.750000+6 1.260000-1 2.900000+6 1.250000-19040 3 18 970 3.000000+6 1.330000-1 3.050000+6 1.390000-1 3.250000+6 1.380000-19040 3 18 971 3.500000+6 1.420000-1 3.750000+6 1.400000-1 4.000000+6 1.430000-19040 3 18 972 4.500000+6 1.420000-1 5.000000+6 1.440000-1 5.200000+6 1.460000-19040 3 18 973 5.400000+6 1.430000-1 5.600000+6 1.380000-1 5.700000+6 1.390000-19040 3 18 974 5.800000+6 1.390000-1 5.900000+6 1.440000-1 6.000000+6 1.520000-19040 3 18 975 6.100000+6 1.720000-1 6.200000+6 1.920000-1 6.400000+6 2.390000-19040 3 18 976 6.500000+6 2.660000-1 7.000000+6 3.720000-1 7.200000+6 3.890000-19040 3 18 977 7.400000+6 3.950000-1 7.800000+6 3.660000-1 8.000000+6 3.590000-19040 3 18 978 8.500000+6 3.380000-1 9.000000+6 3.240000-1 9.500000+6 3.200000-19040 3 18 979 1.000000+7 3.220000-1 1.100000+7 3.040000-1 1.200000+7 3.040000-19040 3 18 980 1.250000+7 3.120000-1 1.300000+7 3.220000-1 1.400000+7 3.430000-19040 3 18 981 1.460000+7 3.880000-1 1.500000+7 4.030000-1 1.550000+7 4.270000-19040 3 18 982 1.600000+7 4.500000-1 1.700000+7 4.760000-1 1.800000+7 4.690000-19040 3 18 983 1.900000+7 5.160000-1 2.000000+7 5.460000-1 0.000000+0 0.000000+09040 3 18 984 9040 3 0 985 9040 0 0 986 9.023200+4 2.300400+2 0 0 0 1904033 18 987 0.000000+0 0.000000+0 0 18 0 2904033 18 988 0.000000+0 0.000000+0 0 1 6 3904033 18 989 1.000000-5 0.000000+0 6.000000+5 1.600000-3 2.000000+7 0.000000+0904033 18 990 0.000000+0 0.000000+0 0 1 24 12904033 18 991 1.000000-5 0.000000+0 6.000000+5 1.024000-1 8.000000+5 1.440000-2904033 18 992 9.000000+5 1.000000-2 1.000000+6 4.900000-3 1.200000+6 3.600000-3904033 18 993 1.400000+6 1.600000-3 1.600000+6 3.600000-3 3.000000+6 1.600000-3904033 18 994 4.000000+6 3.600000-3 6.000000+6 1.000000-2 2.000000+7 0.000000+0904033 18 995 904033 0 996 9040 0 0 997 0 0 0 998 9.223500+4 2.330250+2 1 1 0 19228 1451 1 0.000000+0 1.000000+0 0 0 0 69228 1451 2 1.000000+0 0.000000+0 0 0 10 69228 1451 3 0.000000+0 0.000000+0 0 0 393 49228 1451 4 92-U -235 ORNL,LANL EVAL-APR89 L.W.WESTON,P.G.YOUNG,W.P.POENITZ 9228 1451 5 DIST-JUN90 900605 9228 1451 6 ----ENDF/B-VI MATERIAL 9228 9228 1451 7 -----INCIDENT NEUTRON DATA 9228 1451 8 ------ENDF-6 FORMAT 9228 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9228 1451 10 DOSIMETRY 9228 1451 11 ******************************************************************9228 1451 12 *****************************************************************9228 1451 13 PRINCIPAL EVALUATORS: 9228 1451 14 9228 1451 15 THERMAL PARAMETERS: STANDARDS COMMITTEE OF CSEWG. 9228 1451 16 9228 1451 17 RESOLVED RESONANCE REGION: (0-2250 EV) L.C.LEAL(U.TENN.), 9228 1451 18 G.DESAUSSURE(ORNL), R.B.PEREZ(U.TENN.), N.M.LARSON(ORNL), 9228 1451 19 M.S.MOORE(LANL), AND R.Q. WRIGHT(ORNL). 9228 1451 20 9228 1451 21 UNRESOLVED RESONANCE REGION AND FILE 3 BELOW 100 KEV AND THE 9228 1451 22 CAPTURE CROSS SECTION ABOVE 100 KEV: 9228 1451 23 W.P.POENITZ(ANL) AND L.W.WESTON(ORNL) 9228 1451 24 9228 1451 25 FISSION CROSS SECTION ABOVE 100 KEV: W.P.POENITZ(ANL) AND THE 9228 1451 26 STANDARDS COMMITTEE OF CSEWG. 9228 1451 27 9228 1451 28 MODEL CALCULATIONS AND FITS ABOVE 100 KEV: P.G.YOUNG(LANL) AND 9228 1451 29 R.E.MACFARLANE(LANL) AND E.D.ARTHUR(LANL). 9228 1451 30 9228 1451 31 COVARIANCE FILES: R.W.PEELLE(ORNL). 9228 1451 32 9228 1451 33 **********mf=1 descriptive and nubar information*****************9228 1451 34 9228 1451 35 mt=452 total nubar. sum of mt=455 and 456. 9228 1451 36 mt=455 delayed neutron yields. england (en89) 9228 1451 37 mt=456 prompt neutron yields. at neutron energies below 1 kev, 9228 1451 38 taken from the evaluation of frehaut (renormalized to 9228 1451 39 match the the u235 thermal standard), which indicates 9228 1451 40 a constant value. Above 1 kev, an new covariance 9228 1451 41 analysis of all available experimental data was per- 9228 1451 42 formed using the GLUCS analysis code (he80). The data 9228 1451 43 were obtained from the NNDC and were renormalized 9228 1451 44 using the ENDF/B-VI standards. A smooth curve was 9228 1451 45 drawn through the results of the covariance analysis. 9228 1451 46 9228 1451 47 9228 1451 48 **********mf=2 resolved and unresolved resonance region**********9228 1451 49 mt=151 9228 1451 50 the 235-u neutron cross sections are described with reich-moore 9228 1451 51 type resolved resonance parameters up to 2250 ev. the resolved 9228 1451 52 range is divided into 10 regions, each described by its own set 9228 1451 53 of resonance parameters as shown in the table below. 9228 1451 54 9228 1451 55 subdivisions of the resolved resonance region. 9228 1451 56 energy region (ev) number of resonances 9228 1451 57 9228 1451 58 0- 110 236 9228 1451 59 110- 300 384 9228 1451 60 300- 500 272 9228 1451 61 500- 750 361 9228 1451 62 750-1000 303 9228 1451 63 9228 1451 64 1000-1250 368 9228 1451 65 1250-1500 334 9228 1451 66 1500-1750 302 9228 1451 67 1750-2000 280 9228 1451 68 2000-2250 502 9228 1451 69 9228 1451 70 the resonance parameters were obtained by fitting experimental9228 1451 71 data with the resonance analysis code sammy (1). the partial 9228 1451 72 cross section measurements were all renormalized to the 2200 m/s 9228 1451 73 values of the endf/b-vi standards committee (2). the energy scale 9228 1451 74 of all the data sets were aligned on the energy scale of the 80-m 9228 1451 75 flight path trans. meas. of harvey et al.(3).the length of that 9228 1451 76 flight path has been measured with great accuracy (4). 9228 1451 77 9228 1451 78 1. thermal region 9228 1451 79 in the thermal region the following measurements were given9228 1451 80 most weight: the trans. measurement of r.r. spencer et al(5); the9228 1451 81 recent fission cross section measurements of r.gwin et al.(6), 9228 1451 82 of schrack(7) and wagemans et al(8); the endf/b-v capture cross 9228 1451 83 section below 0.5 ev renorm. at 2200 m/s to the value proposed by9228 1451 84 the endf/b-vi standards comm(2). a value of 9.938 f was obtained9228 1451 85 for the effect. scatt. radius from the consistent fit of these 9228 1451 86 data. in the following table, the 2200 m/s values of the cross 9228 1451 87 section computed with the parameters of this evaluation, using 9228 1451 88 the leal-hwang method (9), are compared to the values proposed by9228 1451 89 the endf/b-vi standards committee: 9228 1451 90 9228 1451 91 2200 m/s values of the cross sections (b). 9228 1451 92 9228 1451 93 **** this evaluation **** endf/b-vi 9228 1451 94 cross section at 0k at 300k standards 9228 1451 95 9228 1451 96 total 698.42 698.22 698.67+/-1.71 9228 1451 97 scattering 15.48 15.52 15.46+/-1.06 9228 1451 98 absorbtion 682.94 682.70 683.22+/-1.34 9228 1451 99 fission 584.18 583.98 584.25+/-1.11 9228 1451 100 capture 98.76 98.72 98.96+/-0.74 9228 1451 101 9228 1451 102 2. energy region 0-110 ev. 9228 1451 103 9228 1451 104 in the resonance region up to 110 ev, the following meas- 9228 1451 105 urements were given most weight: the transmission measurements 9228 1451 106 of harvey et al.(3); the fission measurements of gwin et al.(6), 9228 1451 107 of schrack (7), of wagemans and deruytter (10) and of weston 9228 1451 108 and todd(11); the spin-separated fission cross-section data of 9228 1451 109 moore et al.(12) from the analysis of the polarized neut. polar- 9228 1451 110 ized target measurement of keyworth et al.(13). the following 9228 1451 111 tables provide a comparison of the fission and capture cross 9228 1451 112 sections integrated over several energy intervals as obtained 9228 1451 113 from different data sets and from this evaluation: 9228 1451 114 9228 1451 115 comparison of fission integrals (b ev) 9228 1451 116 9228 1451 117 this schrack gwin+ wagemans+ weston+ 9228 1451 118 interval (ev) evaluation (7) (6) (10) (11) 9228 1451 119 9228 1451 120 0.0206-0.06239 19.16 19.18 19.26 19.26+/-0.08 9228 1451 121 7.8- 11.0 246.02 239.41 247.4 246.+/-2.5 9228 1451 122 9228 1451 123 0.5- 10.0 404. 397. 406. 406. 9228 1451 124 10.0- 50.0 1805. 1796. 1838.5 1838. 9228 1451 125 50.0-100.0 1582. 1586. 1632. 1647.5 1601.9 9228 1451 126 100.0-110.0 187. 186. 183. 190.6 188. 9228 1451 127 9228 1451 128 9228 1451 129 comparison of capture integrals (b ev) 9228 1451 130 9228 1451 131 inter.(ev) this eval. de saussure et al(14) perez et al(15) 9228 1451 132 9228 1451 133 0.5- 10.0 224.9 231.6 (not measured) 9228 1451 134 10.0- 50.0 1162 1178 1252 9228 1451 135 50.0-100.0 685 721 747 9228 1451 136 100.0-110.0 152 158 176 9228 1451 137 9228 1451 138 9228 1451 139 3. energy region 110-2250 ev. 9228 1451 140 9228 1451 141 above 110 ev, the resonances cannot be fully resolved. 9228 1451 142 the resonance parm. provided in this eval. represent well the 9228 1451 143 high resolution thick sample trans. meas. of harvey et al. (3) 9228 1451 144 and the fission measurement of weston and todd (11,18). the 9228 1451 145 assignment of spins to the resonance structures is also roughly 9228 1451 146 consistent with the spin-separated fission data of moore et al. 9228 1451 147 (12). in the following table, the fission cross sections 9228 1451 148 averaged over decimal intervals between 100 and 2250 ev from 9228 1451 149 this evaluation is compared with values from endf-b/v (16) and 9228 1451 150 with values proposed by the endf/b-vi standards committee (2). 9228 1451 151 9228 1451 152 9228 1451 153 comparison of average evaluated fission cross sections (b). 9228 1451 154 9228 1451 155 energy this proposed 9228 1451 156 interval eval. endf/b-v standard schrack(7) weston+ 9228 1451 157 (ev) (18) 9228 1451 158 9228 1451 159 100- 200 20.64 20.71 21.14+/-0.09 20.91 20.48 9228 1451 160 200- 300 20.04 20.21 20.67+/-0.10 20.05 20.15 9228 1451 161 300- 400 12.80 12.90 13.14+/-0.07 13.21 12.86 9228 1451 162 400- 500 13.43 13.46 13.79+/-0.07 13.83 13.53 9228 1451 163 500- 600 14.95 14.86 15.19+/-0.08 14.63 14.77 9228 1451 164 600- 700 11.49 11.35 11.47+/-0.06 11.46 11.27 9228 1451 165 700- 800 10.88 10.90 11.14+/-0.06 10.79 10.79 9228 1451 166 800- 900 8.32 8.02 8.25+/-0.04 7.82 7.95 9228 1451 167 900-1000 7.22 7.34 7.53+/-0.04 7.18 7.30 9228 1451 168 1000-2000 7.12 7.20 7.35+/-0.04 7.04 9228 1451 169 9228 1451 170 9228 1451 171 in the following table the fission and capture resonance 9228 1451 172 integrals obtained from this eval. are compared to the values 9228 1451 173 obtained for endf/b-v and to the values evaluated from integral 9228 1451 174 measurement (19). the few percent contributions above 2250 ev 9228 1451 175 for this evaluation were estimated to be equal to the values 9228 1451 176 obtained from endf/b-v. 9228 1451 177 9228 1451 178 the ratio of the capt. resonance integral to the fiss. resonance 9228 1451 179 integral, 0.477, is lower than the value 0.513 +/-0.015 obtained 9228 1451 180 from direct measurements (17). 9228 1451 181 9228 1451 182 9228 1451 183 comparison of fission and capture resonance integrals (b) 9228 1451 184 9228 1451 185 ****** fission ****** ****** capture ***** 9228 1451 186 9228 1451 187 e (ev) this eval. endf/b-v this eval. endf/b-v 9228 1451 188 9228 1451 189 0.5- 5 85.06 85.27 25.50 24.66 9228 1451 190 5- 50 109.09 111.68 74.67 78.49 9228 1451 191 50- 110 25.36 25.74 11.36 11.87 9228 1451 192 110- 300 20.81 20.93 8.46 10.67 9228 1451 193 300- 500 6.70 6.73 2.50 2.95 9228 1451 194 500- 750 5.23 5.29 1.83 1.98 9228 1451 195 750- 1000 2.46 2.37 1.18 1.27 9228 1451 196 1000- 1250 1.94 1.83 0.94 0.85 9228 1451 197 1250- 1500 1.32 1.33 0.52 0.49 9228 1451 198 1500- 1750 0.95 0.99 0.53 0.37 9228 1451 199 1750- 2000 0.89 0.91 0.34 0.39 9228 1451 200 2000- 2250 0.60 0.65 0.37 0.30 9228 1451 201 9228 1451 202 2250ev-20 mev (18.20) 18.20 (4.68) 4.68 9228 1451 203 9228 1451 204 0.5 ev-20 mev 278.61 281.92 132.88 138.97 9228 1451 205 9228 1451 206 from reference 19 275+/-5 144+/-6 9228 1451 207 9228 1451 208 references for resolved resonance region. 9228 1451 209 9228 1451 210 1.n.m.larson, ornl/tm-9719/r1,(1985); also n.m.larson, f.g.perey 9228 1451 211 "resonance parameter analysis with sammy," int. conf. nuclear 9228 1451 212 data for science and tech., may 30-june 3, 1988, mito, japan. 9228 1451 213 2.a.carlson et al.,"results of the endf/b-vi standards eval- 9228 1451 214 uation," private communication 31 aug 1987. 9228 1451 215 3.j.a.harvey et al.,"high-resolution neutron trans. measurements 9228 1451 216 on 235-u, 238-u, and 239-pu, "inter. conf. nucl. data for 9228 1451 217 science and technology, may 30-june 3, 1988, mito, japan. 9228 1451 218 4.d.c.larson and n.m.larson, ornl/tm-9097 (1985). 9228 1451 219 5.r.r.spencer et al. nucl.sci.eng. 96,318 (1987). 9228 1451 220 6.r.gwin et al.,nucl.sci.eng. 88,37 (1984) 9228 1451 221 7.r.c.schrack,"measurement of the 235-u(n,f) reaction from 9228 1451 222 thermal to 1 kev, "internat. conf. nuclear data for science 9228 1451 223 and technology, may 30-june 3, mito, japan. 9228 1451 224 8.c.wagemans et al.,"subthermal fiss. cross-section measurements 9228 1451 225 for 233u, 235u, and 239pu, "internat. conf. nuclear data for 9228 1451 226 science and technology, may 30-june 3, 1988, mito, japan. 9228 1451 227 9.l.c.leal and r.n.hwang,tansao 55,1-341 (1987). 9228 1451 228 10.c.wagemans and a.j.deruytter,p499 in proc.of nuclear data for 9228 1451 229 basic and applied science, santa fe, new mexico, may 13-17, 9228 1451 230 1985 vol. i (1986). 9228 1451 231 11.l.w.weston and j.h.todd,nucl.sci.eng. 88,567 (1984). 9228 1451 232 12.m.s.moore et al.,phys.rev.c18,1328 (1978). 9228 1451 233 13.g.a.keyworth et al.,phys.rev.lett. 31,1077 (1973). 9228 1451 234 14.g.de saussure et al.,ornl/tm-1804 (1967). 9228 1451 235 15.r.b.perez et al.,nucl.sci.eng. 53,46 (1973) 9228 1451 236 16.m.r.bhat, bnl-ncs-51184 (endf-248) (1980) 9228 1451 237 17.j.hardy,"235u resonance fission integral and alpha based on 9228 1451 238 integral measurements," endf-300,section b.1 (1979). 9228 1451 239 18.l.w.weston and j.h.todd, private communication (1988). these 9228 1451 240 measurements done on a flight path of 80 m. have a much better 9228 1451 241 resolution than those of ref.11 and have been used above 110 ev9228 1451 242 19.s.f.mughabghab,"neutron cross sections",voli,partb (1984) 9228 1451 243 9228 1451 244 -----------------------------------------------------------------9228 1451 245 unresolved resonance region: 9228 1451 246 the unresolved resonance region was derived by a fitacs 9228 1451 247 (fritz froehner code) fit by l.w.weston to the standards 9228 1451 248 committee recommendation for the fission cross section and9228 1451 249 new capture evaluation based on newer alpha measurements 9228 1451 250 (see anl-83-4 supplement). these results were then fit 9228 1451 251 with ures (ed pennington code) so endf would reproduce. 9228 1451 252 unresolved resonance region extends from 2.25 to 25 kev 9228 1451 253 and is used only for self shielding calculations. dilute 9228 1451 254 cross sections are taken from file 3 which shows 9228 1451 255 experimentally observed structure carried over from 9228 1451 256 version 5 up to 100 kev. 9228 1451 257 9228 1451 258 **********mf=3 smooth cross sections*****************************9228 1451 259 model calculations from lanl: p.g.young,r.e.macfarlane,e.d.arthur9228 1451 260 the evaluation above 100 kev is based on a detailed theoretical 9228 1451 261 analysis utilizing the available experimental data. coupled 9228 1451 262 channel optical model calculations with the ecis code (ra70) 9228 1451 263 were used to provide the total, elastic, and inelastic cross 9228 1451 264 sections to the first 3 members of the ground state rotational 9228 1451 265 band, as well as neutron elastic and inelastic angular distri- 9228 1451 266 butions to the rotational levels. the ecis code was also 9228 1451 267 used to calculate neutron transmission coefficients. hauser- 9228 1451 268 feshbach statistical theory calculations were carried out with 9228 1451 269 the gnash (ar88, yo77) and comnuc (du70) code systems, including 9228 1451 270 preequilibrium and fission. dwba calculations were performed 9228 1451 271 with the dwuck code (ku70) for several vibrational levels, using 9228 1451 272 b(el) values inferred from (d,d') data on u234, u235, u238, as 9228 1451 273 well as coulomb excitation measurements. a weak coupling 9228 1451 274 model (pe69) was used to apply the u234 and u238 results to 9228 1451 275 states in u235. 9228 1451 276 a preliminary description of the analysis was given at the mito 9228 1451 277 conference (yo88). 9228 1451 278 9228 1451 279 9228 1451 280 mt=1 sum of partial cross sections from 2.25 to 100 kev. 9228 1451 281 0.10 to 20 mev, obtained from a covariance analysis of 9228 1451 282 available experimental data, using an initial or prior 9228 1451 283 cross section from the coupled-channel optical model 9228 1451 284 analysis. experimental data used include fo71, ve80, 9228 1451 285 bo71, po81, gr73, sc74, po83, pe60, wh65, ca73, ga60, 9228 1451 286 and br58. the glucs code was used for analysis (he80).9228 1451 287 mt=2 unchanged from version 5 from 2.25 to 100 kev. 9228 1451 288 0.12 to 20 mev, based on subtraction of mt=4,16,17,18,37, 9228 1451 289 and 102 from mt=1. 9228 1451 290 mt=4 sum of mt=51-91 9228 1451 291 mt=16 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 292 mt=17 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 293 mt=18 2.2 to 120 kev average values from simultaneous evaluation9228 1451 294 of w.p.poenitz with structure carried over from version 5.9228 1451 295 renormalizaion was over 3 ranges per decade. 9228 1451 296 0.10 to 20 mev is standards evaluation by csewg standards 9228 1451 297 committee. 9228 1451 298 mt=19 (n,f) first-chance fission cross section. 9228 1451 299 ratio of first-chance to total fission obtained from 9228 1451 300 gnash calculations. 9228 1451 301 mt=20 (n,nf) second-chance fission cross section. 9228 1451 302 ratio of second-chance to total fission obtained from 9228 1451 303 gnash calculations. 9228 1451 304 mt=21 (n,2nf) third-chance fission cross section. 9228 1451 305 ratio of third-chance to total fission obtained from 9228 1451 306 gnash calculations. 9228 1451 307 mt=37 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 308 mt=38 (n,3nf) fourth-chance fission cross section. 9228 1451 309 mt=51,52,54,55,57,58,60-62,64-72,74-76 threshold to 6.0 mev, 9228 1451 310 compound-nucleus reaction theory calculations with 9228 1451 311 width fluctuation corrections using the comnuc code. 9228 1451 312 mt=53,56,59,63 thr. to 20 mev, coupled-channel optical model 9228 1451 313 calculations (9/2-, 11/2-, 13/2-, 15/2- members of the 9228 1451 314 ground-state rotational band), plus compound-nucleus 9228 1451 315 contributions from comnuc calculations. 9228 1451 316 mt=73,77-84 threshold to 20 mev, distorted wave born approx- 9228 1451 317 imation calculations with dwuck code. these levels are 9228 1451 318 composites of l=2 and l=3 vibrational states. the l=2 9228 1451 319 states are mt=78-80,82 and the l=3 states are mt=73,77, 9228 1451 320 81,83,84. 9228 1451 321 mt=91 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 322 note that the mt=77-84 vibrational states lie in the 9228 1451 323 mt=91 continuum region. 9228 1451 324 mt=102 2.25 to 1000 kev capture is based on newer alpha 9228 1451 325 measurements (see anl-83-4 supplement) and fission from 9228 1451 326 standards committee recomendations. most of the 9228 1451 327 structure below 100 kev is from fission. 1-20 mev is 9228 1451 328 based on renormalized comnuc/gnash calculation, with a 9228 1451 329 semi-direct component added above a few mev. 9228 1451 330 9228 1451 331 **********mf=4 neutron angular distributions******************** 9228 1451 332 9228 1451 333 mt=2 elastic scattering angular distribution based on ecis 9228 1451 334 coupled-channel calculations, with a compound elastic 9228 1451 335 component from comnuc included below 6 mev. 9228 1451 336 mt=51,52,54,55,57,58,60-62,64-72,74-76 threshold to 6.0 mev, 9228 1451 337 compound-nucleus reaction theory calculations with 9228 1451 338 width fluctuation corrections using the comnuc code. 9228 1451 339 mt=53,56,59,63 thr. to 20 mev, coupled-channel optical model 9228 1451 340 calculations (9/2-, 11/2-, 13/2-, 15/2- members of the 9228 1451 341 ground-state rotational band), plus compound-nucleus 9228 1451 342 contributions from comnuc calculations. 9228 1451 343 mt=73,77-84 threshold to 20 mev, distorted wave born approx. 9228 1451 344 imation calculations with dwuck code. 9228 1451 345 9228 1451 346 ************mf=5 neutron energy distributions******************* 9228 1451 347 9228 1451 348 mf=18 composite neutron energy distributions from fission. 9228 1451 349 based on calculations by d.madland (ma88) using madland-9228 1451 350 nix formalism. the calculations include the first-, 9228 1451 351 second-, and third-chance fission neutron components. 9228 1451 352 the calculations end at 15 mev; the 20-mev spectrum 9228 1451 353 is simply a duplication of the 15-mev spectrum. 9228 1451 354 tabulated distribution law (lf=1) is used. 9228 1451 355 mt=455 england (en89) 9228 1451 356 9228 1451 357 ************mf=6 correlated energy-angle distributions********** 9228 1451 358 9228 1451 359 mt=16 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 360 neutron distributions only. 9228 1451 361 mt=17 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 362 neutron distributions only. 9228 1451 363 mt=37 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 364 neutron distributions only. 9228 1451 365 mt=91 gnash hauser-feshbach statistical/preequilibrium calc. 9228 1451 366 neutron distributions only. 9228 1451 367 9228 1451 368 ******************references************************************ 9228 1451 369 9228 1451 370 ar88 e.d.arthur, la-ur-88-382 (1988). 9228 1451 371 bo72 k.boeckhoff et al., j.nuc.en.26,91(1972). 9228 1451 372 br58 a.bratenahl et al., phys.rev.110,927(1958). 9228 1451 373 ca73 j.cabe et al., cea-r-4524 (1973). 9228 1451 374 du70 c.l.dunford, ai-aec-12931 (1970). 9228 1451 375 en89 t.r.england etal, la-11151-ms(88), la_11534t(89); 9228 1451 376 laur-88-4118 to be published nse(89). 9228 1451 377 fo71 d.foster & d.glasgow, phys.rev.c3,576(1971). 9228 1451 378 ku70 p.d.kunz, dwuck: a distorted-wave born approximation 9228 1451 379 program, unpublished report. 9228 1451 380 fr86 j.frehaut, neandc(e) 238/l (1986). 9228 1451 381 gr73 l.green et al., usndc-9 (1973) p.170 9228 1451 382 he80 d.hetrick & c.y.fu, ornl/tm-7341 (1980). 9228 1451 383 pe60 j.peterson et al., phys.rev.120, 521(1960). 9228 1451 384 pe69 r.j.peterson, ann.phys. 53, 40 (1069). 9228 1451 385 po81 w.poenitz et al., nuc.sci.eng.78, 333(1981). 9228 1451 386 po83 w.poenitz et al., anl-ndm-80, 1983. 9228 1451 387 ra70 j.raynal,iaea smr-9/8 (1970). 9228 1451 388 sc74 r.schwartz et al., nuc.sci.engr.54,322(1974). 9228 1451 389 ut66 c.uttley et al., paris conf. (1966) v1, p165. 9228 1451 390 ve80 v.vertebnyj et al., yfi-16,8(1973). 9228 1451 391 wh65 w.whalen et al., anl-7110 (1965) p.15. 9228 1451 392 yo77 p.g.young & e.d.arthur, la-6947 (1977). 9228 1451 393 yo88 p.g.young & e.d.arthur, nuc.data for sci.& tech., mito 9228 1451 394 conference (1988) p.603. 9228 1451 395 9228 1451 396 **************************************************************** 9228 1451 397 1 451 401 19228 1451 398 2 151 3474 19228 1451 399 3 18 114 19228 1451 400 33 18 78 19228 1451 401 9228 1 0 402 9228 0 0 403 9.223500+4 2.330250+2 0 0 1 09228 2151 404 9.223500+4 1.000000+0 0 1 11 09228 2151 405 1.000000-5 1.100000+2 1 3 0 09228 2151 406 3.500000+0 1.001760+0 0 0 1 29228 2151 407 2.330250+2 0.000000+0 0 0 1416 2369228 2151 408 -9.997600+1 3.000000+0 3.771300-3 3.980000-2 1.861800-1 4.700200-19228 2151 409 -4.621000+0 4.000000+0 9.278500-3 3.979600-2 1.704700-1-2.438000-29228 2151 410 -2.226500+0 3.000000+0 1.631700-4 3.979900-2-8.733200-1-1.168600+09228 2151 411 -4.354300-1 4.000000+0 1.208700-4 3.255400-2-3.095000-4-1.208000-19228 2151 412 2.774500-1 3.000000+0 3.941500-6 3.435200-2 8.494600-2-3.268200-29228 2151 413 1.136100+0 4.000000+0 1.415000-5 3.672300-2 5.884300-3 1.097200-19228 2151 414 2.036000+0 3.000000+0 9.170400-6 3.780400-2-1.025800-2-1.175900-59228 2151 415 2.770700+0 4.000000+0 9.821900-7 2.953800-2 1.058700-1-2.275400-29228 2151 416 3.147100+0 3.000000+0 2.518000-5 3.987300-2-7.765900-2 2.007000-29228 2151 417 3.617000+0 4.000000+0 4.248600-5 3.626100-2-4.339400-2 1.199200-29228 2151 418 4.850500+0 4.000000+0 5.332100-5 3.625200-2 4.048800-5-4.182500-39228 2151 419 5.433800+0 4.000000+0 2.169200-5 5.506400-2-1.159100-1-3.317000-19228 2151 420 6.203300+0 3.000000+0 7.374800-5 4.789000-2-1.316700-1 6.013900-29228 2151 421 6.390300+0 4.000000+0 2.313200-4 3.739700-2 1.092100-2-5.420100-59228 2151 422 7.083600+0 4.000000+0 1.107100-4 3.700700-2 1.187600-2 2.275400-29228 2151 423 7.652400+0 3.000000+0 3.523600-6 4.821400-2 1.244500-1 1.059600-19228 2151 424 8.762900+0 4.000000+0 9.405500-4 3.358200-2 3.233200-2-7.313700-29228 2151 425 8.957300+0 3.000000+0 1.205900-4 4.708400-2-7.953400-2 2.731100-19228 2151 426 9.270900+0 4.000000+0 1.143400-4 3.473500-2 6.474600-2 9.839100-39228 2151 427 9.745600+0 3.000000+0 4.285800-5 4.180000-2 1.498500-1-5.449600-29228 2151 428 1.016700+1 4.000000+0 4.844300-5 3.291100-2-1.330800-2-4.689500-29228 2151 429 1.077900+1 3.000000+0 1.109300-5 3.721000-2-9.523600-2-1.723800-19228 2151 430 1.166600+1 4.000000+0 5.078800-4 3.778000-2 2.459700-3 2.915800-39228 2151 431 1.233000+1 4.000000+0 1.054500-4 1.544000-2 8.889900-2 2.024600-29228 2151 432 1.239900+1 3.000000+0 1.247700-3 3.902400-2-2.769900-3 2.619100-29228 2151 433 1.285800+1 4.000000+0 6.367200-5 3.844300-2 9.249100-3 1.006700-19228 2151 434 1.327300+1 4.000000+0 4.428800-5 3.289100-2-6.135500-3 1.334800-19228 2151 435 1.369100+1 3.000000+0 4.191600-5 3.140800-2-1.899800-2-9.233200-29228 2151 436 1.394800+1 3.000000+0 6.072500-4 3.175500-2-4.128300-1 9.229200-29228 2151 437 1.405200+1 4.000000+0 1.752600-5 3.063500-2-1.152100-1 4.084000-19228 2151 438 1.455100+1 3.000000+0 1.120400-4 3.887300-2 1.049800-2 6.061500-39228 2151 439 1.541500+1 4.000000+0 2.083400-4 4.043300-2-2.287600-2 3.281900-29228 2151 440 1.608700+1 4.000000+0 3.318900-4 3.604700-2 1.695100-2 5.397800-39228 2151 441 1.664700+1 4.000000+0 2.447200-4 3.228700-2 2.851800-2 8.313000-29228 2151 442 1.798600+1 4.000000+0 8.356200-5 2.821000-2 1.873300-2-1.394000-19228 2151 443 1.804000+1 3.000000+0 2.916900-4 3.430300-2 8.581700-2-3.224700-29228 2151 444 1.899400+1 4.000000+0 5.762100-5 3.517300-2-1.108100-2 1.085000-39228 2151 445 1.919300+1 3.000000+0 2.400100-4 3.114500-2-1.759600-1 1.561100-19228 2151 446 1.929100+1 4.000000+0 2.611000-3 3.808400-2-5.847400-3 5.683200-29228 2151 447 2.016600+1 3.000000+0 8.860100-5 3.344500-2-8.319500-3-1.061300-19228 2151 448 2.017900+1 4.000000+0 7.727000-6 3.359200-2 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9.212100-4 3.500000-2 2.000000-2-2.000000-29228 2151 1776 8.250000+2 3.000000+0 1.876200-3 3.792300-2 3.595300-2-3.590800-29228 2151 1777 8.256000+2 4.000000+0 1.710200-3 2.843700-2 7.295700-2-7.247300-29228 2151 1778 8.265000+2 3.000000+0 9.488100-4 3.500000-2 2.000000-2-2.000000-29228 2151 1779 8.274000+2 4.000000+0 7.487600-4 3.500000-2 2.000000-2-2.000000-29228 2151 1780 8.280000+2 3.000000+0 1.533200-3 2.615800-2 3.784900-2-3.781500-29228 2151 1781 8.286000+2 4.000000+0 2.912100-3 3.847800-2 5.924400-2-7.057000-29228 2151 1782 8.298900+2 4.000000+0 3.527400-3 3.850900-2 4.556700-2-3.299000-29228 2151 1783 8.309000+2 3.000000+0 1.006900-3 3.500000-2 2.000000-2-2.000000-29228 2151 1784 8.324900+2 3.000000+0 5.716200-3 4.571500-2 1.394000-2-1.449600-29228 2151 1785 8.332000+2 4.000000+0 6.908700-4 3.500000-2 2.000000-2-2.000000-29228 2151 1786 8.341000+2 3.000000+0 2.528300-3 3.786600-2 1.665500-2-1.737000-29228 2151 1787 8.348000+2 4.000000+0 2.122100-3 3.180300-2 3.443200-2-3.842000-29228 2151 1788 8.356000+2 4.000000+0 2.199100-3 2.155300-2 4.270400-2-4.748900-29228 2151 1789 8.369600+2 3.000000+0 1.132100-2 3.359800-2 4.671500-2-4.869300-29228 2151 1790 8.376000+2 4.000000+0 5.792200-3 3.969500-2 1.677000-2-7.995100-29228 2151 1791 8.384000+2 3.000000+0 5.274300-3 3.344200-2 2.299800-2-2.441200-29228 2151 1792 8.396000+2 3.000000+0 8.538900-4 3.774800-2 3.194900-2-3.195700-29228 2151 1793 8.402300+2 4.000000+0 7.524300-3 2.951000-2 1.516400-2-8.474100-39228 2151 1794 8.420100+2 4.000000+0 6.045900-3 3.672500-2-1.516700-2 4.594500-29228 2151 1795 8.431900+2 3.000000+0 9.241800-3 3.160100-2 1.737200-1-1.740700-19228 2151 1796 8.446000+2 3.000000+0 5.953100-4 3.500000-2 2.000000-2-2.000000-29228 2151 1797 8.452000+2 4.000000+0 5.309300-3 3.502100-2 1.218900-1-1.240800-19228 2151 1798 8.458000+2 3.000000+0 3.856800-3 3.200300-2-1.548400-2 9.515100-39228 2151 1799 8.463000+2 4.000000+0 5.110700-3 3.500000-2-1.177100-1 2.169700-29228 2151 1800 8.472000+2 4.000000+0 5.984200-3 3.500000-2 3.365500-1-8.722700-29228 2151 1801 8.475600+2 3.000000+0 1.358400-2 3.271500-2 1.328800-1-1.352700-19228 2151 1802 8.486000+2 4.000000+0 3.497800-3 3.479600-2 1.545700-1-1.385200-19228 2151 1803 8.496000+2 3.000000+0 1.039800-4 3.806000-2 2.277600-2-2.277700-29228 2151 1804 8.499000+2 4.000000+0 9.049800-4 3.774100-2 8.419400-2-8.623500-29228 2151 1805 8.508000+2 4.000000+0 1.381800-3 3.500000-2 2.000000-2-2.000000-29228 2151 1806 8.511700+2 3.000000+0 6.729100-3 3.272100-2 2.813700-2-3.048600-29228 2151 1807 8.523000+2 4.000000+0 7.019100-4 3.888300-2 3.115600-2-3.141300-29228 2151 1808 8.528000+2 4.000000+0 6.794700-3 6.894000-2 3.251600-2-3.845100-29228 2151 1809 8.549000+2 3.000000+0 4.818100-3 3.439800-2 3.670100-2-3.754300-29228 2151 1810 8.561000+2 4.000000+0 6.763900-4 3.785900-2 2.532700-2-2.588500-29228 2151 1811 8.567000+2 3.000000+0 2.993500-3 3.622600-2 2.334200-2-2.409800-29228 2151 1812 8.577000+2 3.000000+0 1.365000-3 3.500000-2 2.000000-2-2.000000-29228 2151 1813 8.586000+2 4.000000+0 4.408200-3 4.004400-2 2.439700-1-2.172400-19228 2151 1814 8.595000+2 4.000000+0 1.458300-3 3.500000-2 2.000000-2-2.000000-29228 2151 1815 8.601000+2 3.000000+0 6.854900-3 3.403000-2 2.015400-1-2.016800-19228 2151 1816 8.614500+2 4.000000+0 2.481900-2 7.656000-2 2.227500-2-1.010800-19228 2151 1817 8.626800+2 4.000000+0 1.433900-2 3.549200-2 1.044700-1-6.783600-29228 2151 1818 8.650500+2 3.000000+0 1.129200-2 3.829900-2 1.490000-2-1.469700-29228 2151 1819 8.661700+2 4.000000+0 1.267100-2 5.897500-2 9.291100-2 1.141600-29228 2151 1820 8.677500+2 4.000000+0 2.593300-2 2.070500-1 1.926200-1-1.429800-19228 2151 1821 8.684000+2 3.000000+0 7.002400-3 8.644400-2 1.690800-1-1.687500-19228 2151 1822 8.690000+2 3.000000+0 2.915400-3 3.500000-2 1.852800-2-3.757600-29228 2151 1823 8.705000+2 4.000000+0 6.233400-4 3.500000-2 2.000000-2-2.000000-29228 2151 1824 8.713000+2 3.000000+0 3.308200-3 3.861200-2 3.579800-2-3.667300-29228 2151 1825 8.719000+2 4.000000+0 1.788700-3 3.891800-2 2.961500-2-2.759300-29228 2151 1826 8.724000+2 3.000000+0 1.248600-3 3.500000-2 2.000000-2-2.000000-29228 2151 1827 8.734000+2 4.000000+0 1.586400-3 3.708300-2 4.799500-2-4.647300-29228 2151 1828 8.747000+2 4.000000+0 1.545800-3 3.500000-2 2.000000-2-2.000000-29228 2151 1829 8.754500+2 3.000000+0 1.443500-2 8.083700-2 2.139400-1-2.176700-19228 2151 1830 8.761000+2 3.000000+0 1.250700-3 3.500000-2 2.000000-2-2.000000-29228 2151 1831 8.774000+2 4.000000+0 1.447400-3 3.518700-2 4.010000-2-3.810400-29228 2151 1832 8.781000+2 4.000000+0 1.352400-3 3.500000-2 2.000000-2-2.000000-29228 2151 1833 8.790000+2 4.000000+0 5.979400-3 2.706800-2 6.770100-2-5.309900-29228 2151 1834 8.798000+2 4.000000+0 4.224100-3 3.979200-2 3.201100-2-2.218600-29228 2151 1835 8.804000+2 3.000000+0 4.161300-3 4.055600-2 4.499400-2-4.519000-29228 2151 1836 8.810000+2 4.000000+0 6.730500-3 3.932400-2 1.722400-1-1.639700-19228 2151 1837 8.822000+2 3.000000+0 1.914600-3 3.500000-2 2.000000-2-2.000000-29228 2151 1838 8.829000+2 3.000000+0 6.634900-3 2.741100-2 7.339900-2-7.365500-29228 2151 1839 8.837900+2 4.000000+0 9.585600-3 4.685800-3 5.814700-2-2.175100-29228 2151 1840 8.844000+2 4.000000+0 3.640700-3 2.841800-2 3.098100-2-8.116900-49228 2151 1841 8.851000+2 3.000000+0 9.322500-3 3.647200-2 1.425000-1-1.425800-19228 2151 1842 8.867000+2 4.000000+0 2.814900-3 3.474600-2 8.001700-2-3.833800-29228 2151 1843 8.873000+2 4.000000+0 5.409100-4 3.500000-2 2.000000-2-2.000000-29228 2151 1844 8.878000+2 3.000000+0 4.179300-4 3.785400-2 2.372000-2-2.371900-29228 2151 1845 8.883000+2 4.000000+0 6.479600-4 3.500000-2 2.000000-2-2.000000-29228 2151 1846 8.888000+2 4.000000+0 1.005500-3 3.691800-2 2.365200-2-2.360700-29228 2151 1847 8.900000+2 4.000000+0 4.500800-4 3.775600-2 2.545800-2-2.542400-29228 2151 1848 8.911000+2 3.000000+0 1.897500-3 3.654500-2 2.341400-2-2.341300-29228 2151 1849 8.919000+2 4.000000+0 9.314500-4 3.763800-2 2.142700-2-2.139500-29228 2151 1850 8.926900+2 3.000000+0 5.761200-3 2.034100-2 5.070000-2-5.132300-29228 2151 1851 8.940000+2 4.000000+0 1.721200-3 3.287800-2 4.162000-2-4.150300-29228 2151 1852 8.953000+2 3.000000+0 1.806400-4 3.747900-2 2.474000-2-2.473800-29228 2151 1853 8.958000+2 4.000000+0 5.239700-3 2.943300-2 2.239400-2-2.135600-29228 2151 1854 8.963000+2 3.000000+0 7.831700-3 3.570400-2 7.393600-3-6.654900-39228 2151 1855 8.971600+2 4.000000+0 2.371700-2 5.966300-2 1.901600-1-1.232000-19228 2151 1856 8.985600+2 3.000000+0 1.220000-2 4.864300-2 6.432500-2-6.531800-29228 2151 1857 8.990000+2 3.000000+0 1.244300-3 3.500000-2 2.000000-2-2.000000-29228 2151 1858 8.998000+2 4.000000+0 9.942800-3 3.794100-2 1.076500-1 6.679500-29228 2151 1859 9.009700+2 4.000000+0 1.081100-2 6.145800-2 4.378500-2-4.391100-29228 2151 1860 9.020000+2 3.000000+0 4.076700-3 3.850000-2-4.585400-2-4.256000-29228 2151 1861 9.029600+2 4.000000+0 1.224200-2 6.552700-2 1.882800-2-1.058800-19228 2151 1862 9.037000+2 4.000000+0 7.636200-4 3.500000-2 2.000000-2-2.000000-29228 2151 1863 9.044000+2 3.000000+0 4.407800-3 3.850000-2 7.309000-2-7.299600-29228 2151 1864 9.049000+2 4.000000+0 1.091400-3 3.500000-2 2.000000-2-2.000000-29228 2151 1865 9.059900+2 3.000000+0 1.428500-2 6.779600-2 8.703800-2-8.490800-29228 2151 1866 9.070000+2 4.000000+0 3.395000-3 3.850000-2 7.217300-2-3.041600-29228 2151 1867 9.083000+2 3.000000+0 3.245900-3 3.850000-2 6.594200-4-9.873000-49228 2151 1868 9.087400+2 4.000000+0 9.688200-3 6.133500-2 1.004900-1-4.552300-29228 2151 1869 9.102000+2 4.000000+0 2.069500-3 3.850000-2 3.651600-2-2.867600-29228 2151 1870 9.106300+2 4.000000+0 7.685800-3 5.929000-2 6.586100-2-3.706200-29228 2151 1871 9.124000+2 4.000000+0 8.985800-4 3.850000-2 3.826600-2-3.830800-29228 2151 1872 9.132000+2 3.000000+0 8.592700-4 3.500000-2 2.000000-2-2.000000-29228 2151 1873 9.142300+2 3.000000+0 6.610000-3 3.850000-2 5.767900-2-5.877900-29228 2151 1874 9.160500+2 4.000000+0 3.266800-2 1.639800-1 1.296900-1 3.857100-29228 2151 1875 9.176000+2 4.000000+0 3.819000-3 3.850000-2 2.659100-2-1.711800-29228 2151 1876 9.188000+2 4.000000+0 9.084000-4 3.500000-2 2.000000-2-2.000000-29228 2151 1877 9.201800+2 4.000000+0 8.804600-3 4.828900-2 2.016400-3-4.230800-29228 2151 1878 9.211000+2 3.000000+0 1.163400-3 3.500000-2 2.000000-2-2.000000-29228 2151 1879 9.230500+2 3.000000+0 3.206700-2 7.474200-2 1.040700-1-1.063000-19228 2151 1880 9.241000+2 4.000000+0 2.677500-3 3.500000-2 3.771600-2-9.087500-39228 2151 1881 9.246000+2 4.000000+0 1.556500-2 5.083200-2 1.849200-2-1.038300-19228 2151 1882 9.253000+2 4.000000+0 1.499300-3 3.850000-2 1.438800-2-1.444200-29228 2151 1883 9.264000+2 3.000000+0 5.035500-3 2.791500-2 1.056900-1-1.048700-19228 2151 1884 9.278500+2 4.000000+0 1.285800-3 3.500000-2 2.000000-2-2.000000-29228 2151 1885 9.284000+2 4.000000+0 4.246400-3 3.789400-2 3.188900-3-3.057800-29228 2151 1886 9.295500+2 3.000000+0 2.101100-2 1.818200-2-2.461500-2 8.434000-39228 2151 1887 9.318400+2 3.000000+0 2.841500-2 6.019600-2 4.260000-2-1.616200-29228 2151 1888 9.345000+2 4.000000+0 8.293700-3 6.681500-2 5.690000-2-7.179400-29228 2151 1889 9.350000+2 3.000000+0 4.375500-3 3.850000-2 3.703400-2-4.115100-29228 2151 1890 9.358000+2 3.000000+0 1.271500-3 3.500000-2 2.000000-2-2.000000-29228 2151 1891 9.365000+2 4.000000+0 3.535900-4 3.850000-2 3.791900-2-3.798500-29228 2151 1892 9.367000+2 4.000000+0 2.191600-4 3.850000-2 2.946400-2-2.948800-29228 2151 1893 9.381000+2 3.000000+0 1.208800-3 3.850000-2 1.854400-2-1.854200-29228 2151 1894 9.390000+2 4.000000+0 1.380600-3 3.850000-2 2.151300-2-2.149700-29228 2151 1895 9.402000+2 4.000000+0 2.851300-3 3.850000-2 4.962100-3-9.355600-39228 2151 1896 9.410000+2 4.000000+0 1.279000-3 3.500000-2 2.000000-2-2.000000-29228 2151 1897 9.419100+2 3.000000+0 2.830000-2 7.541900-2 1.200300-1-1.370900-19228 2151 1898 9.446900+2 3.000000+0 1.263300-2 3.850000-2 4.850200-2 7.013100-39228 2151 1899 9.452000+2 4.000000+0 1.773200-3 3.850000-2 1.728900-2-1.730600-29228 2151 1900 9.461000+2 3.000000+0 3.649600-3 3.850000-2 3.260800-2-3.043400-29228 2151 1901 9.473100+2 4.000000+0 1.583000-2 9.400200-2 2.082600-1-2.102600-19228 2151 1902 9.478000+2 3.000000+0 3.725900-3 3.500000-2 5.052400-2-4.961100-29228 2151 1903 9.490000+2 4.000000+0 4.733800-3 3.850000-2 5.689700-2-6.617300-29228 2151 1904 9.496000+2 3.000000+0 3.162800-3 3.850000-2 4.141200-2-4.447600-29228 2151 1905 9.506000+2 4.000000+0 1.338200-3 3.500000-2 2.000000-2-2.000000-29228 2151 1906 9.514900+2 3.000000+0 1.401900-2 5.346000-2 1.256200-1-1.263600-19228 2151 1907 9.527000+2 4.000000+0 4.356700-3 3.850000-2 7.935200-3-1.799400-19228 2151 1908 9.536000+2 3.000000+0 3.418300-3 3.850000-2 4.813600-2-4.921000-29228 2151 1909 9.555000+2 3.000000+0 1.559400-3 3.500000-2 2.000000-2-2.000000-29228 2151 1910 9.569000+2 3.000000+0 5.854400-3 3.850000-2-4.635200-2 7.457300-29228 2151 1911 9.572300+2 4.000000+0 1.229400-2 6.794800-2 2.077300-1-2.129300-19228 2151 1912 9.586000+2 4.000000+0 1.385800-3 3.500000-2 2.000000-2-2.000000-29228 2151 1913 9.590000+2 3.000000+0 9.601200-4 3.500000-2 2.000000-2-2.000000-29228 2151 1914 9.596000+2 3.000000+0 6.824500-3 3.850000-2 5.189100-2-5.354900-29228 2151 1915 9.610700+2 4.000000+0 1.377400-2 6.300800-2-4.808000-3-1.444900-29228 2151 1916 9.619000+2 4.000000+0 2.776500-3 3.850000-2 6.748700-3-7.264900-39228 2151 1917 9.621000+2 4.000000+0 1.157000-3 3.500000-2 2.000000-2-2.000000-29228 2151 1918 9.625000+2 4.000000+0 1.008700-3 3.500000-2 2.000000-2-2.000000-29228 2151 1919 9.636000+2 4.000000+0 1.325800-3 3.850000-2 1.688000-2-1.688000-29228 2151 1920 9.653600+2 3.000000+0 7.764000-3 3.867300-2 3.505300-2-3.543200-29228 2151 1921 9.662000+2 3.000000+0 1.133800-3 3.500000-2 2.000000-2-2.000000-29228 2151 1922 9.670000+2 3.000000+0 2.394400-3 3.850000-2 7.458000-3-7.699800-39228 2151 1923 9.677199+2 4.000000+0 4.793700-3 3.850000-2 3.936200-2-4.346400-29228 2151 1924 9.684000+2 4.000000+0 1.036600-3 3.500000-2 2.000000-2-2.000000-29228 2151 1925 9.691600+2 4.000000+0 7.229500-3 5.271300-2-9.499400-4-3.629400-29228 2151 1926 9.699000+2 3.000000+0 2.846300-3 3.850000-2 2.607700-2-2.686600-29228 2151 1927 9.711000+2 4.000000+0 2.505800-3 3.850000-2 3.679000-2-2.788100-29228 2151 1928 9.723000+2 3.000000+0 5.181700-3 3.850000-2 2.306000-2-2.526200-29228 2151 1929 9.732000+2 3.000000+0 5.114500-3 3.850000-2 2.354400-2-2.548700-29228 2151 1930 9.738000+2 4.000000+0 1.005600-3 3.500000-2 2.000000-2-2.000000-29228 2151 1931 9.747000+2 4.000000+0 5.682800-3 1.751900-2 9.191500-2-9.383800-29228 2151 1932 9.753000+2 3.000000+0 3.540200-3 3.850000-2 2.125900-2-2.755900-29228 2151 1933 9.767000+2 3.000000+0 1.990900-3 3.500000-2 2.000000-2-2.000000-29228 2151 1934 9.780500+2 3.000000+0 2.032600-2 4.881400-2 1.048800-1-1.041600-19228 2151 1935 9.784000+2 4.000000+0 1.452800-3 3.850000-2 1.684800-2-1.684800-29228 2151 1936 9.798000+2 3.000000+0 6.445700-3 3.850000-2 3.641300-2-3.340100-29228 2151 1937 9.805000+2 3.000000+0 8.454100-3 2.633800-2 7.112200-2-7.092800-29228 2151 1938 9.819000+2 4.000000+0 3.622700-3 3.850000-2 3.047100-2-3.169500-29228 2151 1939 9.827000+2 4.000000+0 2.893200-3 3.500000-2 3.401300-2-3.307500-29228 2151 1940 9.832000+2 4.000000+0 2.147200-3 3.500000-2 3.118000-2-3.057400-29228 2151 1941 9.836900+2 4.000000+0 8.537900-3 4.071100-2 7.702100-2-7.815000-29228 2151 1942 9.843000+2 3.000000+0 5.948600-4 3.500000-2 2.000000-2-2.000000-29228 2151 1943 9.849100+2 4.000000+0 9.185900-3 3.850000-2 6.873700-2-7.386100-29228 2151 1944 9.863000+2 4.000000+0 2.264900-3 3.850000-2 4.157800-2-4.036000-29228 2151 1945 9.866600+2 3.000000+0 3.758600-3 2.175800-2 5.201300-2-5.212100-29228 2151 1946 9.871000+2 4.000000+0 1.653100-3 3.850000-2 3.178300-2-3.092700-29228 2151 1947 9.887000+2 4.000000+0 8.250400-4 3.850000-2 2.763300-2-2.763300-29228 2151 1948 9.893000+2 3.000000+0 1.717100-3 3.850000-2 2.009200-2-2.009200-29228 2151 1949 9.908200+2 4.000000+0 8.128000-3 1.730100-2 2.857800-2-2.548100-29228 2151 1950 9.922000+2 4.000000+0 1.248500-3 3.500000-2 2.000000-2-2.000000-29228 2151 1951 9.928900+2 4.000000+0 4.951400-3 3.850000-2-1.740400-2-8.816200-29228 2151 1952 9.932000+2 3.000000+0 2.997100-3 3.850000-2 6.637000-3-6.946100-39228 2151 1953 9.940000+2 4.000000+0 8.927500-4 3.500000-2 2.000000-2-2.000000-29228 2151 1954 9.952000+2 3.000000+0 1.373600-4 3.850000-2 2.186800-2-2.186800-29228 2151 1955 9.967000+2 4.000000+0 8.793100-4 3.850000-2 2.477600-2-2.477600-29228 2151 1956 9.980300+2 4.000000+0 9.582400-3 2.165300-2 1.082100-2-1.494800-29228 2151 1957 1.000700+3 4.000000+0 1.874300-3 3.850000-2-1.199100-1 4.834000-29228 2151 1958 1.001000+3 3.000000+0 3.022300-2 1.385100-1 1.440900-1-1.438000-19228 2151 1959 1.001500+3 3.000000+0 3.702300-3 3.850000-2 1.081700-2-1.063400-29228 2151 1960 1.001700+3 4.000000+0 5.672800-4 3.850000-2 4.983300-2-4.966300-29228 2151 1961 1.002600+3 4.000000+0 2.822500-4 3.850000-2 4.314900-2-4.310300-29228 2151 1962 1.003600+3 4.000000+0 5.536500-4 3.850000-2 1.472300-2-1.466100-29228 2151 1963 1.004300+3 3.000000+0 2.720500-3 3.850000-2 9.625100-2-9.627900-29228 2151 1964 1.005000+3 4.000000+0 3.939300-3 3.850000-2 7.194000-2-7.317600-29228 2151 1965 1.005700+3 3.000000+0 1.257600-2 9.675800-2 2.234600-2-2.062800-29228 2151 1966 1.006000+3 3.000000+0 6.898900-4 3.850000-2 4.003300-2-4.003300-29228 2151 1967 1.006300+3 4.000000+0 1.386300-3 3.850000-2 4.582300-2-4.600700-29228 2151 1968 1.007500+3 4.000000+0 1.558300-2 2.345200-1 1.034200-1-1.072700-19228 2151 1969 1.007900+3 3.000000+0 5.662800-3 3.850000-2-3.345100-1 3.344300-19228 2151 1970 1.008600+3 4.000000+0 6.792300-4 3.850000-2-8.357800-4 8.440200-49228 2151 1971 1.009400+3 4.000000+0 5.034000-4 3.850000-2 2.081900-2-2.085500-29228 2151 1972 1.010100+3 3.000000+0 1.997200-3 3.850000-2-1.269300-1 1.269300-19228 2151 1973 1.010800+3 3.000000+0 5.431500-4 3.850000-2 8.490000-3-8.490000-39228 2151 1974 1.011100+3 4.000000+0 1.806000-2 7.488500-2 2.039200-2-1.816700-29228 2151 1975 1.012300+3 3.000000+0 1.946800-3 3.850000-2 7.381200-3-7.381300-39228 2151 1976 1.000000+3 1.250000+3 1 3 0 09228 2151 1977 3.500000+0 1.001760+0 0 0 1 29228 2151 1978 2.330250+2 0.000000+0 0 0 2208 3689228 2151 1979 9.871000+2 4.000000+0 1.384200-3 3.857900-2 2.654000-2-2.663200-29228 2151 1980 9.887000+2 4.000000+0 7.863000-4 3.878700-2 2.739400-2-2.761300-29228 2151 1981 9.893000+2 3.000000+0 1.053300-3 3.895100-2 1.924000-2-2.003600-29228 2151 1982 9.908200+2 4.000000+0 1.104100-2 3.452800-2 1.924000-2-3.468600-29228 2151 1983 9.928900+2 4.000000+0 7.637300-3 5.364900-2 4.689800-2-1.264300-29228 2151 1984 9.932000+2 3.000000+0 4.698500-3 4.414200-2 4.735500-2-1.799500-29228 2151 1985 9.952000+2 3.000000+0 3.386300-4 3.936100-2 1.390300-2-2.185300-29228 2151 1986 9.967000+2 4.000000+0 1.177600-3 4.163400-2 6.876200-3-2.466100-29228 2151 1987 9.980300+2 4.000000+0 1.270800-2 9.152100-2 7.784700-2-5.433000-29228 2151 1988 1.000700+3 4.000000+0 1.068700-3 3.993600-2-1.374000-1-4.205700-39228 2151 1989 1.001000+3 3.000000+0 3.009000-2 1.457900-1 1.510200-1-1.333700-19228 2151 1990 1.001500+3 3.000000+0 8.350500-3 4.287400-2 6.584700-2-1.201700-29228 2151 1991 1.001700+3 4.000000+0 5.580300-4 3.894000-2 5.757200-2-4.964300-29228 2151 1992 1.002600+3 4.000000+0 2.539500-3 3.906300-2 7.336000-2-4.308800-29228 2151 1993 1.003600+3 4.000000+0 1.142600-3 3.877400-2 4.625900-2-1.462500-29228 2151 1994 1.004300+3 3.000000+0 3.767300-3 3.898400-2 9.159900-2-9.610400-29228 2151 1995 1.005000+3 4.000000+0 4.426300-3 3.911500-2 3.560300-2-5.345500-29228 2151 1996 1.005700+3 3.000000+0 1.473100-2 9.744400-2 1.671000-2-1.237100-29228 2151 1997 1.006000+3 3.000000+0 8.457500-4 3.855600-2 4.267700-2-4.004000-29228 2151 1998 1.006300+3 4.000000+0 3.346400-3 3.862200-2 4.526100-2-4.600100-29228 2151 1999 1.007500+3 4.000000+0 2.452600-2 2.351300-1 1.588000-1-1.109200-19228 2151 2000 1.007900+3 3.000000+0 1.073500-3 3.846400-2-3.415500-1 3.426000-19228 2151 2001 1.008600+3 4.000000+0 3.218400-3 3.786200-2-4.027600-2 1.596400-39228 2151 2002 1.009400+3 4.000000+0 3.539700-4 3.852400-2 2.480800-2-2.086200-29228 2151 2003 1.010100+3 3.000000+0 2.144500-3 3.838900-2-1.253300-1 1.269300-19228 2151 2004 1.010800+3 3.000000+0 3.518200-4 3.868400-2 9.379900-3-8.476300-39228 2151 2005 1.011100+3 4.000000+0 1.812200-2 7.830000-2 2.917700-2-1.791100-29228 2151 2006 1.012300+3 3.000000+0 2.154200-3 3.920600-2-9.941700-4-7.522900-39228 2151 2007 1.012800+3 4.000000+0 2.843000-4 3.870100-2 1.152900-2-7.640400-39228 2151 2008 1.013400+3 3.000000+0 1.251800-3 3.890900-2 1.420900-2-2.406300-29228 2151 2009 1.013900+3 4.000000+0 3.443800-3 3.920500-2 2.360900-3-1.801100-29228 2151 2010 1.014600+3 3.000000+0 2.844200-2 6.527100-2 1.966900-2-1.573500-29228 2151 2011 1.015400+3 4.000000+0 4.152900-3 3.842200-2 1.662300-2-5.397400-39228 2151 2012 1.015800+3 3.000000+0 3.242500-3 3.851900-2 8.106500-2-2.023400-29228 2151 2013 1.016400+3 4.000000+0 1.516400-3 3.854200-2 6.075200-2-6.234700-39228 2151 2014 1.017700+3 3.000000+0 1.519200-2 4.337600-2 6.228500-3-1.282800-29228 2151 2015 1.018400+3 4.000000+0 1.909500-3 4.017300-2 5.608800-3-1.554500-29228 2151 2016 1.019400+3 3.000000+0 2.224900-3 4.027000-2 3.046500-1-6.783600-39228 2151 2017 1.019700+3 4.000000+0 2.799300-3 4.086000-2 7.577000-3-2.066500-29228 2151 2018 1.020000+3 3.000000+0 8.495100-3 2.150500-1 9.441200-3-3.547100-29228 2151 2019 1.020700+3 3.000000+0 7.238300-3 1.424800-1 6.586600-2-7.973800-29228 2151 2020 1.021400+3 4.000000+0 3.584100-3 4.036200-2 2.427800-2-6.215800-39228 2151 2021 1.022000+3 4.000000+0 5.473500-4 3.901000-2 6.210200-2-6.068600-29228 2151 2022 1.022500+3 3.000000+0 6.343200-3 4.255800-2 1.883200-2-2.016500-29228 2151 2023 1.023100+3 4.000000+0 1.641700-3 3.899600-2 7.000400-2-6.304200-29228 2151 2024 1.024000+3 4.000000+0 7.255100-4 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18 4069 0.000000+0 0.000000+0 9543 18 1 0922833 18 4070 0.000000+0 0.000000+0 0 3 0 0922833 18 4071 2.000000+5 2.000000+7 9228 18 4 2922833 18 4072 2.000000+5 1.000000+0 2.000000+7 0.000000+0 0.000000+0 0.000000+0922833 18 4073 922833 0 4074 9228 0 0 4075 0 0 0 4076 9.223800+4 2.360060+2 1 1 0 19237 1451 1 0.000000+0 1.000000+0 0 0 0 69237 1451 2 1.000000+0 0.000000+0 0 0 10 69237 1451 3 0.000000+0 0.000000+0 0 0 281 49237 1451 4 92-U -238 ORNL,LANL+ EVAL-NOV89 L.W.WESTON,P.G.YOUNG,W.POENITZ 9237 1451 5 DIST-JUN90 900112 9237 1451 6 ----ENDF/B-VI MATERIAL 9237 9237 1451 7 -----INCIDENT NEUTRON DATA 9237 1451 8 ------ENDF-6 FORMAT 9237 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9237 1451 10 DOSIMETRY 9237 1451 11 ******************************************************************9237 1451 12 9237 1451 13 ******************************************************************9237 1451 14 9237 1451 15 PRINCIPAL EVALUATORS 9237 1451 16 RESOLVED RESONANCE REGION: (10-5 to 10,000 eV) 9237 1451 17 ORNL-G.DeSaussure, D.K.Olsen, Roger Macklin 9237 1451 18 HARWELL- M.C.Moxon, M.G.Sowerby 9237 1451 19 UNRESOLVED RESONANCE REGION: (10 TO 149 keV) 9237 1451 20 F.H.Froehner(KFK) and W.P.Poenitz(ANL) 9237 1451 21 REACTIONS ABOVE 149 keV: 9237 1451 22 MODEL CODE CALCULATIONS: (LANL)-P.G.Young and R.E.MacFarlane 9237 1451 23 FISSION: W.P.Poenitz(ANL) 9237 1451 24 NU-BAR DELAYED AND DELAYED NEUTRON SPECTRA: 9237 1451 25 KAISER AND CARPENTER (Ka78) 9237 1451 26 GAMMA PRODUCTION FILES: R. HOWERTON (LLL) 9237 1451 27 UNCERTAINTY FILES: L.W.WESTON (ORNL) 9237 1451 28 9237 1451 29 *****************************************************************9237 1451 30 9237 1451 31 ENDF/B-VI EVALUATION ABOVE THE UNRESOLVED RESONANCE REGION 9237 1451 32 P.G.Young and R.E.MacFarlane (LANL) 9237 1451 33 9237 1451 34 Above the unresolved resonance region, new evaluations were 9237 1451 35 performed of the neutron total, (n,2n), (,n,3n), (n,4n), (n,f), 9237 1451 36 (n,nf), (n,2nf), (n,3nf), and (n,gamma) cross sections as well as9237 1451 37 prompt nubar. The elastic and inelastic data from ENDF/B-V were 9237 1451 38 carried over for Version VI. 9237 1451 39 To provide the new data, coupled channel optical model calcu- 9237 1451 40 lations were performed with the ECIS code (Ra70) for the lowest 9237 1451 41 3 members of the U238 ground state rotational band. These calcu- 9237 1451 42 lations were used to provide initial (prior) values for a covar- 9237 1451 43 iance analysis of the total cross section and to provide neutron 9237 1451 44 transmission coefficients for nuclear reaction theory 9237 1451 45 calculations with the GNASH (Ar88,Yo77) and COMNUC (Du70) Hauser-9237 1451 46 Feshbach statistical/fission/preequilibrium codes. These theory 9237 1451 47 calculations were used to provide the MF=6 neutron distributions 9237 1451 48 from the (n,2n), (n,3n), and (n,4n) reactions as well as prior 9237 1451 49 values for covariance analyses of the cross sections for those 9237 1451 50 reactions. Additionally, the above analyses plus DWBA calcu- 9237 1451 51 lations were used to check the ENDF/B-V evaluation of elastic 9237 1451 52 and inelastic scattering. While some differences were found, 9237 1451 53 the earlier work was generally found to be reliable, and we 9237 1451 54 decided to carry over the ENDF/B-V data because of the effort 9237 1451 55 taken to match experimental data, both at lower energies and at 9237 1451 56 14 MeV. 9237 1451 57 9237 1451 58 **********MF=1 DESCRIPTIVE AND NUBAR INFORMATION*****************9237 1451 59 9237 1451 60 MT=452 Total Nubar. Sum of MT=455 and 456. 9237 1451 61 MT=455 Delayed Neutron Yields. Kaiser and Carpenter (Ka78) 9237 1451 62 MT=456 Prompt Neutron Yields. Taken from Frehaut (Fr86) 9237 1451 63 9237 1451 64 **********MF=2 RESONANCE REGION**********************************9237 1451 65 RESOLVED RESONANCE REGION: 9237 1451 66 The resolved resonance region determines the thermal cross 9237 1451 67 sections. The thermal capture is that adopted by the Standards9237 1451 68 Committee of CSEWG (2.7081 with an uncer. of 0.0095 barns). 9237 1451 69 Thermal (2200 m/s) constants are: 9237 1451 70 Total Cross Section: 12.068 barns 9237 1451 71 Capture Cross Section: 2.709 9237 1451 72 The resolved resonance region has been extended to 10 keV. 9237 1451 73 The evaluation was from the work of D.K.Olsen(ORNL), Roger 9237 1451 74 Macklin(ORNL), M.C.Moxon(Harwell), and M.G.Sowerby(Harwell) 9237 1451 75 UNRESOLVED RESONANCE REGION: 10 to 149 keV 9237 1451 76 The unresolved resonance region was evaluated by F.H.Froehner 9237 1451 77 (KFK) with his code FITACS which does simultaneous fit to 9237 1451 78 experimental data. The unresolved resonance region is to be 9237 1451 79 used only for self shielding calculations and File 3 should be 9237 1451 80 used to determine the dilute cross sections. 9237 1451 81 9237 1451 82 *********** MF=3 SMOOTH CROSS SECTIONS **************************9237 1451 83 9237 1451 84 MT=1 Neutron Total Cross Section 9237 1451 85 10 TO 149 KEV, EVALUATION BY F.H.FROEHNER (KFK) WHICH IS BASED 9237 1451 86 ON FITACS FIT TO EXPERIMENTAL DATA. 9237 1451 87 0.25 to 20 MeV, based on a covariance analysis of the 9237 1451 88 available exp. data, especially Fo71, Sc74, Po81, Po83, Ha73, 9237 1451 89 Br58, Ca73, Pe60, Wh71, Ba65, Ut66, Sh79, and Li79. Similar 9237 1451 90 data exclusions as used in the covariance analysis by Smith et 9237 1451 91 al.(Sm82) were incorporated but additional data were included. 9237 1451 92 Prior cross section for the analysis was taken from the 9237 1451 93 coupled-channel analysis, described above. The GLUCS code sys. 9237 1451 94 was used for the covariance analysis (He80). 9237 1451 95 MT=16 (n,2n) Cross Section 9237 1451 96 Threshold to 20 MeV, based on a covariance analysis of 9237 1451 97 available data (Ba66,Fr80a,Ko80,Yo78,Ma69,Ma72,Ka79,La73,Ro57, 9237 1451 98 Ph56,Pe61,Ve78,An58,Ry80), similar to MT=1. Prior cross sec- 9237 1451 99 tion obtained from GNASH analysis, with approx.20% uncertainty.9237 1451 100 MT=17 (n,3n) Cross Section 9237 1451 101 Threshold to 20 MeV, based on a covariance analysis 9237 1451 102 similar to MT=16. Exp. data used in the analysis were Fr80b, 9237 1451 103 Wh62,Ro57,Ma69,Al61, and Ve78. 9237 1451 104 MT=18 Fission Cross Section (Total) 9237 1451 105 0.01 TO 0.3 EVALUATED BY L.W.WESTON FROM DI80 AND SL77. 9237 1451 106 0.3 to 20 MeV, taken directly from the simultaneous 9237 1451 107 standards analysis (Ca85,Po89). 9237 1451 108 MT=19 First-Chance Fission (n,f) Cross Section 9237 1451 109 0.3 t0 20 MeV, based on the ratio of first-chance to 9237 1451 110 total fission from ENDF/B-V using the present MT=18 sigma. 9237 1451 111 MT=20 Second-Chance Fission (n,nf) Cross Section 9237 1451 112 5.0 to 20 MeV, based on the ratio of second-chance to 9237 1451 113 total fission from ENDF/B-V using the present MT=18 sigma. 9237 1451 114 MT=21 Third-Chance Fission (n,2nf) Cross Section 9237 1451 115 12 to 20 MeV, based on the ratio of third-chance to 9237 1451 116 total fission from ENDF/B-V using the present MT=18 sigma. 9237 1451 117 MT=37 (n,4n) Cross Section 9237 1451 118 Threshold to 20 MeV, taken directly from the GNASH 9237 1451 119 analysis described above. 9237 1451 120 MT=38 Fourth-Chance Fission (n,3nf) Cross Section 9237 1451 121 18 to 20 MeV, based on the ratio of fourth-chance to 9237 1451 122 total fission from ENDF/B-V using the present MT=18 sigma. 9237 1451 123 MT=102 Radiative Capture Cross Section 9237 1451 124 10 TO 149 KEV, EVALUATION BY F.H.FROEHNER (KFK) WHICH IS BASED 9237 1451 125 ON FITACS FIT TO THE EXPERIMENTAL DATA. 9237 1451 126 0.10 to 20 MeV, taken directly from simultaneous 9237 1451 127 standards analysis (Ca85,Po89). 9237 1451 128 9237 1451 129 ************MF=6 CORRELATED ENERGY-ANGLE DISTRIBUTIONS********** 9237 1451 130 9237 1451 131 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9237 1451 132 Neutron distributions only. 9237 1451 133 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9237 1451 134 Neutron distributions only. 9237 1451 135 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9237 1451 136 Neutron distributions only. 9237 1451 137 9237 1451 138 ******************REFERENCES************************************ 9237 1451 139 9237 1451 140 Al61 K.Allen et al., J.Nuc.En.14,100(1961). 9237 1451 141 An58 G.Antropov et al., A.E. 5,456(1958). 9237 1451 142 Ar88 E.D.Arthur, LA-UR-88-382 (1988). 9237 1451 143 Ba66 D.Barr, (LANL) pers.communication to R.Howerton (1966). 9237 1451 144 Ba65 R.Batchelor et al., Nuc.Phys.65,236(1965). 9237 1451 145 Br58 A.Bratenahl et al., Phys.Rev.110,927(1958). 9237 1451 146 Ca73 J.Cabe et al., CEA-R-4524 (1973). 9237 1451 147 Ca85 A.Carlson et al., Nuc.Data for Basic & Applied Science, 9237 1451 148 Santa Fe, NM (1985) p.1429. 9237 1451 149 DI80 F.C.DIFILLIPPO ET AL., PR/C 21, 1400. 9237 1451 150 Du70 C.L.Dunford, AI-AEC-12931 (1970). 9237 1451 151 Fo71 D.Foster & D.Glasgow, Phys.Rev.C3,576(1971). 9237 1451 152 Fr80a J.Frehaut et al., Nucl.Sci.Eng.74,29(1980). 9237 1451 153 Fr80b J.Frehaut et al., BNL-NCS-512457 (1980) p399. 9237 1451 154 Fr86 J.Frehaut, NEANDC(E) 238/L (1986). 9237 1451 155 Ha73 S.Hayes et al., Nuc.Sci.Eng.50,243(1973). 9237 1451 156 He80 D.Hetrick & C.Y.Fu, ORNL/TM-7341 (1980). 9237 1451 157 Ka78 R.Kaiser & S.Carpenter (ANL-West) personal communication. 9237 1451 158 Ko80 N.Kornilov et al., ZFK-410 (1980)p68. 9237 1451 159 La73 J.Landrum et al., Phys.Rev.C8,1938(1973).]9238 1451 9237 1451 160 Li79 P.Lisowski et al., (LANL WNR measurement)pers.comm.(1979). 9237 1451 161 Ma69 D.Mather, AWRE-O-47 (1969). 9237 1451 162 Ma72 D.Mather et al., AWRE-O-72 (1972). 9237 1451 163 Pe60 J.Peterson et al., Phys.Rev.120, 521(1960). 9237 1451 164 Pe61 J.Perkin, J.Nuc.En.14,69(1961). 9237 1451 165 Ph56 J.Phillips, AERE-NP/R-2033 (1956). 9237 1451 166 Po81 W.Poenitz et al., Nuc.Sci.Eng.78, 333(1981). 9237 1451 167 Po83 W.Poenitz et al., ANL-NDM-80, 1983. 9237 1451 168 Po89 W.Poenitz, (ANL-West) personal communication (1989). 9237 1451 169 Ra70 J.Raynal,IAEA SMR-9/8 (1970). 9237 1451 170 Ro57 L.Rosen et al., LA-2111 (1958). 9237 1451 171 Ry80 T.Ryves, J.Phys.G 6,771(1980). 9237 1451 172 Sc74 R.Schwartz et al., Nuc.Sci.Engr.54,322(1974). 9237 1451 173 Sh78 R.Shamu et al., personal communication, 1978. 9237 1451 174 SL77 R.SLOVACEK ET AL., NSE 62, 455 9237 1451 175 Sm82 A.Smith et al., ANL/NDM-74 (1982). 9237 1451 176 Ut66 C.Uttley et al., Paris Conf. (1966) v1, p165. 9237 1451 177 Ve78 L.Veeser & E.Arthur, Harwell Nuc.Data Conf. (1978)p1054. 9237 1451 178 Wh62 P.White et al., J.Nuc.En.A/B 16,261(1962). 9237 1451 179 Wh71 J.Whalen et al., ANL-7710 (1971) p9. 9237 1451 180 Yo77 P.G.Young & E.D.Arthur, LA-6947 (1977). 9237 1451 181 Yo78 Chou You-Pu, HSJ-77091 (1978). 9237 1451 182 9237 1451 183 *****************************************************************9237 1451 184 DESCRIPTION OF DATA CARRIED OVER FROM ENDF/B-V 9237 1451 185 9237 1451 186 Principal Evaluators -- ENDF/B-V 9237 1451 187 E.M.Pennington (ANL), A.B.Smith (ANL), W.P.Poenitz (ANL) 9237 1451 188 R.Howerton (LLL) - gamma production files 9237 1451 189 contributing evaluators 9237 1451 190 R.Kaiser and S.Carpenter (ANL-west) - nu-bar delayed and delayed 9237 1451 191 neutron spectra 9237 1451 192 MF = 3 smooth cross sections 9237 1451 193 Cross sections above 45 kev were evaluated by A.Smith. 9237 1451 194 more details are provided under each reaction type below. Most 9237 1451 195 of the evaluation above 45 kev is described in ANL/NDM-32. 9237 1451 196 MT = 2 (elastic scattering cross section) 9237 1451 197 Elastic scattering cross sections and angular distributions 9237 1451 198 were evaluated using data referenced in ANL/NDM-32. Coupled- 9237 1451 199 channel optical model calculations assisted in the analysis 9237 1451 200 (ref.9). Below about 1 mev, the elastic cross section was 9237 1451 201 determined for consistency with the total and partial non- 9237 1451 202 elastics. Above 1 mev, the total and elastic cross sections 9237 1451 203 determine the non-elastic. The angular distributions are 9237 1451 204 entirely experimental below 1.8 MeV, and calculated at higher 9237 1451 205 energies. At most energies the ENDF/B-V elastic cross section 9237 1451 206 is smaller than ENDF/B-IV. 9237 1451 207 MT = 4,51-77,91 (total inelastic, inelastic to 27 levels, and 9237 1451 208 continuum inelastic) 9237 1451 209 The evaluation is a correlation of theory (ref.9) and expt. 9237 1451 210 Experimental data referenced in ANL/NDM-32 are considered. 9237 1451 211 Levels at higher energies are composites of actual levels. The 9237 1451 212 total inelastic cross section is generally much higher than in 9237 1451 213 ENDF/B-IV, except near threshold and over a region below 1 MeV.9237 1451 214 The individual levels have tails extending to much higher 9237 1451 215 energies than in ENDF/B-IV, resulting in a much smaller 9237 1451 216 continuum. Thus inelastic scattering at high incident energies 9237 1451 217 leads to higher average final energies in ENDF/B-V than in 9237 1451 218 ENDF/B-IV. This tends to counteract the effect on the flux 9237 1451 219 spectrum of having higher total inelastic in ENDF/B-V than in 9237 1451 220 ENDF/B-IV. In performing the inelastic evaluation, one revision9237 1451 221 was made in the first draft (ref.10) to give improved agreement9237 1451 222 with calculations for ZPR-6-7. However, no changes outside the 9237 1451 223 estimated uncertainties were made. 9237 1451 224 MF = 4 angular distributions 9237 1451 225 MT = 2 (elastic) 9237 1451 226 The elastic angular distributions are expressed as f(l) coeff. 9237 1451 227 in the center of mass system at 31 energies. They were obtained9237 1451 228 as described above under MF=3, MT=2 above 45 kev. Below 45 kev,9237 1451 229 ENDF/B-IV was used. The 20 MeV f(l) were modified in order to 9237 1451 230 avoid negative excursions caused by the restriction to NL=20. 9237 1451 231 MT = 18,19,20,21,38,91 9237 1451 232 Angular distributions were taken as isotropic in lab system. 9237 1451 233 MT = 51-77 (inelastic levels) 9237 1451 234 The angular distributions for the inelastic levels are given as9237 1451 235 probabilities in the C.-of-M. system. They were calculated with9237 1451 236 the model described in ref.9. comparisons with expt. are given 9237 1451 237 in ANL/NDM-32. In ENDF/B-IV all but the four highest pseudo 9237 1451 238 levels were assumed isotropic. 9237 1451 239 MF = 5 secondary energy dist. 9237 1451 240 MT = 18 (total fission) 9237 1451 241 The total fission spectrum is an energy-dependent watt spectrum9237 1451 242 (LF=11) with parameters chosen to give the same average energy 9237 1451 243 as the combination of MT=19,20,21,38. 9237 1451 244 MT = 19 (first chance fission) 9237 1451 245 An energy-dependent Watt spectrum was determined using methods 9237 1451 246 similar to those of ref.15. A Watt spectrum has more neutrons 9237 1451 247 at intermediate energies, and fewer at low and high energies 9237 1451 248 than a Maxwellian of the same average energy. 9237 1451 249 MT = 20,21,38 (2nd,3rd,4th chance fission) 9237 1451 250 A combination of a Watt spectrum and 1,2, or 3 LF=9 laws for 9237 1451 251 MT=20,21,38, respectively, is used. 9237 1451 252 MT = 91 (inelastic continuum) lf=1 by A.Smith 9237 1451 253 MT = 455 (delayed neutron spectra) ref.1. 9237 1451 254 MF = 12 9237 1451 255 MT = 18 (fission) and 102 (n,gamma) multiplicities are given 9237 1451 256 MF = 13 9237 1451 257 MT = 3 (non-elastic) 9237 1451 258 Photon production cross sections are given above inel. thresh. 9237 1451 259 MF = 14 9237 1451 260 MT = 3,18,102 photon angular distributions are isotropic. 9237 1451 261 MF = 15 9237 1451 262 MT = 3,18,102 9237 1451 263 Continuous photon energy spectra are present. Description 9237 1451 264 of the evaluation of MF=12-15 is included in ANL/NDM-32 9237 1451 265 REFERENCES 9237 1451 266 1. R.E.Kaiser and S.G.Carpenter (ANL-west) priv. comm. (Mar.78) 9237 1451 267 **data inserted into file at BNL by R. Kinsey 4/18/78** 9237 1451 268 2. S.A.Cox ANL/NDM-5 (1974) 9237 1451 269 3. C.Besant et al sem. fast pulsed reactors CONF-760111 (1976) 9237 1451 270 4. F.Manero and V.Konshin at. energy review 10 no.4 (1972) 9237 1451 271 5. M.Soleilhac revised data received from NNCSC. 9237 1451 272 6. B.Nurpeisov et al Sov. At. Energy translation 807 (Mar,1976) 9237 1451 273 7. G.deSaussure et al. PNE 3,87 (1979) 9237 1451 274 8. D.Olsen et al Nuc. Sci. Eng. 62,479 (1977) 9237 1451 275 9. P.Guenther,D.Havel,A.Smith ANL-NDM-22 (1976) 9237 1451 276 10. E.Pennington,W.Poenitz,A.Smith TANS 26,591 (1977) 9237 1451 277 11. W.P.Poenitz ANL/NDM-45 (1979) 9237 1451 278 12. W.P.Poenitz and A.B.Smith,ed. ANL-76-90 (1976) 9237 1451 279 13. R.Slovacek et al Nucl. Sci. Eng. 62,455 (1977) 9237 1451 280 14. M.Caner,M.Segev,W.Yiftah Nucl. Sci. Eng. 59,395 (1976) 9237 1451 281 15. E.Kujawski and L.Stewart TANS 24,453 (1976) 9237 1451 282 16. L.W.Weston p.c. to B.A.Magurno Nov.12,1982 9237 1451 283 17. Sher +Beck EPRI NP-1771/81 +rev. 1/83 +pc to Magurno 2/83 9237 1451 284 **************************************************************** 9237 1451 285 1 451 289 19237 1451 286 2 151 2049 19237 1451 287 3 18 49 19237 1451 288 33 18 11 09237 1451 289 9237 1 0 290 9237 0 0 291 9.223800+4 2.360060+2 0 0 1 09237 2151 292 9.223800+4 1.000000+0 0 0 2 09237 2151 293 1.000000-5 1.000000+4 1 3 0 09237 2151 294 0.000000+0 9.428482-1 0 0 2 29237 2151 295 2.360060+2 0.000000+0 0 0 4806 8019237 2151 296 -4.133000+2 5.000000-1 2.033000-1 2.300000-2 0.000000+0 0.000000+09237 2151 297 -3.933000+2 5.000000-1 8.259900-2 2.300000-2 0.000000+0 0.000000+09237 2151 298 -3.733000+2 5.000000-1 8.047500-2 2.300000-2 0.000000+0 0.000000+09237 2151 299 -3.533000+2 5.000000-1 7.828800-2 2.300000-2 0.000000+0 0.000000+09237 2151 300 -3.333000+2 5.000000-1 1.830000-1 2.300000-2 0.000000+0 0.000000+09237 2151 301 -3.133000+2 5.000000-1 7.372700-2 2.300000-2 0.000000+0 0.000000+09237 2151 302 -2.933000+2 5.000000-1 7.133200-2 2.300000-2 0.000000+0 0.000000+09237 2151 303 -2.733000+2 5.000000-1 3.306400-1 2.300000-2 0.000000+0 0.000000+09237 2151 304 -2.533000+2 5.000000-1 6.629200-2 2.300000-2 0.000000+0 0.000000+09237 2151 305 -2.333000+2 5.000000-1 2.443800-1 2.300000-2 0.000000+0 0.000000+09237 2151 306 -2.133000+2 5.000000-1 6.083500-2 2.300000-2 0.000000+0 0.000000+09237 2151 307 -1.933000+2 5.000000-1 1.390000-1 2.300000-2 0.000000+0 0.000000+09237 2151 308 -1.733000+2 5.000000-1 5.260000-2 2.300000-2 0.000000+0 0.000000+09237 2151 309 -1.533000+2 5.000000-1 2.480000-1 2.300000-2 0.000000+0 0.000000+09237 2151 310 -1.333000+2 5.000000-1 2.310000-2 2.300000-2 0.000000+0 0.000000+09237 2151 311 -1.133000+2 5.000000-1 4.160000-2 2.300000-2 0.000000+0 0.000000+09237 2151 312 -9.330000+1 5.000000-1 1.930000-2 2.300000-2 0.000000+0 0.000000+09237 2151 313 -7.330000+1 5.000000-1 1.700000-2 2.300000-2 0.000000+0 0.000000+09237 2151 314 -5.330000+1 5.000000-1 7.300000-3 2.300000-2 0.000000+0 0.000000+09237 2151 315 -3.330000+1 5.000000-1 8.086000-4 2.300000-2 8.203000-7 0.000000+09237 2151 316 6.674000+0 5.000000-1 1.493000-3 2.300000-2 9.990000-9 0.000000+09237 2151 317 2.087100+1 5.000000-1 1.026000-2 2.291000-2 5.420000-8 0.000000+09237 2151 318 3.668200+1 5.000000-1 3.413000-2 2.289000-2 9.770000-9 0.000000+09237 2151 319 6.603200+1 5.000000-1 2.460000-2 2.336000-2 5.265000-8 0.000000+09237 2151 320 8.074900+1 5.000000-1 1.865000-3 2.300000-2 6.049000-8 0.000000+09237 2151 321 1.025600+2 5.000000-1 7.170000-2 2.342000-2 1.286000-8 0.000000+09237 2151 322 1.169020+2 5.000000-1 2.549000-2 2.299000-2 0.000000+0 0.000000+09237 2151 323 1.456600+2 5.000000-1 8.470000-4 2.300000-2 0.000000+0 0.000000+09237 2151 324 1.652900+2 5.000000-1 3.367000-3 2.300000-2 0.000000+0 0.000000+09237 2151 325 1.896700+2 5.000000-1 1.732000-1 2.238000-2 3.573000-8 0.000000+09237 2151 326 2.085100+2 5.000000-1 5.111000-2 2.394000-2 8.347000-8 0.000000+09237 2151 327 2.373800+2 5.000000-1 2.716000-2 2.454000-2 3.542000-8 0.000000+09237 2151 328 2.572160+2 5.000000-1 3.400000-5 2.300000-2 0.000000+0 0.000000+09237 2151 329 2.736600+2 5.000000-1 2.578000-2 2.210000-2 0.000000+0 0.000000+09237 2151 330 2.910000+2 5.000000-1 1.687000-2 2.212000-2 0.000000+0 0.000000+09237 2151 331 3.113200+2 5.000000-1 1.089000-3 2.300000-2 0.000000+0 0.000000+09237 2151 332 3.477900+2 5.000000-1 8.200000-2 2.251000-2 2.320000-7 0.000000+09237 2151 333 3.534500+2 5.000000-1 2.400000-5 2.300000-2 0.000000+0 0.000000+09237 2151 334 3.769300+2 5.000000-1 1.129000-3 2.300000-2 1.400000-7 0.000000+09237 2151 335 3.976200+2 5.000000-1 6.207000-3 2.300000-2 0.000000+0 0.000000+09237 2151 336 4.102400+2 5.000000-1 2.009000-2 2.390000-2 0.000000+0 0.000000+09237 2151 337 4.340700+2 5.000000-1 9.787000-3 2.288000-2 0.000000+0 0.000000+09237 2151 338 4.631700+2 5.000000-1 5.613000-3 2.280000-2 1.346000-6 0.000000+09237 2151 339 4.784300+2 5.000000-1 4.126000-3 2.300000-2 2.290000-7 0.000000+09237 2151 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6.603730+3 1.500000+0 7.207600-4 2.300000-2 0.000000+0 0.000000+09237 2151 1817 6.612960+3 1.500000+0 2.573500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1818 6.622140+3 1.500000+0 1.229400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1819 6.637951+3 5.000000-1 7.979200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1820 6.640860+3 1.500000+0 5.788000-5 2.300000-2 0.000000+0 0.000000+09237 2151 1821 6.660750+3 1.500000+0 1.670000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1822 6.673070+3 1.500000+0 2.556200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1823 6.688000+3 1.500000+0 1.726500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1824 6.700790+3 1.500000+0 1.826000-3 2.300000-2 0.000000+0 0.000000+09237 2151 1825 6.703993+3 5.000000-1 9.886500-5 2.300000-2 0.000000+0 0.000000+09237 2151 1826 6.719370+3 5.000000-1 5.645400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1827 6.721380+3 1.500000+0 2.129400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1828 6.729993+3 1.500000+0 2.814000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1829 6.731960+3 5.000000-1 7.477000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1830 6.741210+3 1.500000+0 9.938100-4 2.300000-2 0.000000+0 0.000000+09237 2151 1831 6.754010+3 5.000000-1 6.885800-3 2.350000-2 0.000000+0 0.000000+09237 2151 1832 6.768180+3 1.500000+0 1.063800-2 2.300000-2 0.000000+0 0.000000+09237 2151 1833 6.775890+3 5.000000-1 4.997400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1834 6.778330+3 1.500000+0 1.680800-4 2.300000-2 0.000000+0 0.000000+09237 2151 1835 6.791360+3 1.500000+0 2.666800-4 2.300000-2 0.000000+0 0.000000+09237 2151 1836 6.795960+3 5.000000-1 4.590400-2 2.350000-2 0.000000+0 0.000000+09237 2151 1837 6.816570+3 5.000000-1 1.065700-2 2.350000-2 0.000000+0 0.000000+09237 2151 1838 6.819980+3 1.500000+0 1.581400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1839 6.834660+3 1.500000+0 1.546300-4 2.300000-2 0.000000+0 0.000000+09237 2151 1840 6.836243+3 5.000000-1 1.167100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1841 6.848350+3 1.500000+0 4.919800-3 2.350000-2 0.000000+0 0.000000+09237 2151 1842 6.869870+3 1.500000+0 1.850000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1843 6.871550+3 5.000000-1 3.382200-2 2.350000-2 0.000000+0 0.000000+09237 2151 1844 6.887760+3 5.000000-1 1.140100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1845 6.901110+3 1.500000+0 2.181300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1846 6.902870+3 5.000000-1 2.691200-2 2.350000-2 0.000000+0 0.000000+09237 2151 1847 6.911140+3 1.500000+0 7.063500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1848 6.915743+3 1.500000+0 8.408900-3 2.350000-2 0.000000+0 0.000000+09237 2151 1849 6.917420+3 5.000000-1 1.532500-2 2.350000-2 0.000000+0 0.000000+09237 2151 1850 6.926650+3 1.500000+0 4.237700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1851 6.933790+3 5.000000-1 1.585900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1852 6.941370+3 1.500000+0 4.211700-3 2.350000-2 0.000000+0 0.000000+09237 2151 1853 6.954020+3 5.000000-1 1.245700-2 2.300000-2 0.000000+0 0.000000+09237 2151 1854 6.975931+3 1.500000+0 3.231700-4 2.300000-2 0.000000+0 0.000000+09237 2151 1855 6.989640+3 1.500000+0 1.643800-3 2.300000-2 0.000000+0 0.000000+09237 2151 1856 6.994750+3 1.500000+0 3.181700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1857 7.003870+3 5.000000-1 5.170900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1858 7.015250+3 1.500000+0 3.175200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1859 7.015681+3 1.500000+0 1.126300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1860 7.035940+3 5.000000-1 1.805600-2 2.300000-2 0.000000+0 0.000000+09237 2151 1861 7.048750+3 1.500000+0 3.473400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1862 7.054610+3 1.500000+0 1.115900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1863 7.076420+3 1.500000+0 4.186300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1864 7.088570+3 1.500000+0 1.544600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1865 7.096410+3 5.000000-1 7.584100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1866 7.108610+3 1.500000+0 1.101300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1867 7.119980+3 5.000000-1 5.510600-4 2.300000-2 0.000000+0 0.000000+09237 2151 1868 7.127860+3 5.000000-1 3.781100-2 2.350000-2 0.000000+0 0.000000+09237 2151 1869 7.134000+3 1.500000+0 3.846900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1870 7.169270+3 5.000000-1 2.222700-2 2.350000-2 0.000000+0 0.000000+09237 2151 1871 7.174400+3 5.000000-1 1.501400-2 2.300000-2 0.000000+0 0.000000+09237 2151 1872 7.183702+3 1.500000+0 1.738000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1873 7.199020+3 1.500000+0 1.791400-2 2.350000-2 0.000000+0 0.000000+09237 2151 1874 7.212550+3 1.500000+0 6.207400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1875 7.215230+3 5.000000-1 1.893500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1876 7.226181+3 1.500000+0 1.576900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1877 7.234010+3 1.500000+0 3.590500-4 2.300000-2 0.000000+0 0.000000+09237 2151 1878 7.242030+3 5.000000-1 9.319300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1879 7.262223+3 1.500000+0 2.584800-3 2.300000-2 0.000000+0 0.000000+09237 2151 1880 7.272480+3 1.500000+0 5.561300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1881 7.281300+3 5.000000-1 1.854700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1882 7.291270+3 5.000000-1 4.628100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1883 7.297620+3 1.500000+0 5.442600-4 2.300000-2 0.000000+0 0.000000+09237 2151 1884 7.303980+3 1.500000+0 2.571700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1885 7.321050+3 1.500000+0 2.770300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1886 7.332810+3 1.500000+0 3.072700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1887 7.340420+3 1.500000+0 9.359800-3 2.350000-2 0.000000+0 0.000000+09237 2151 1888 7.362760+3 1.500000+0 1.537500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1889 7.365710+3 5.000000-1 1.748300-2 2.350000-2 0.000000+0 0.000000+09237 2151 1890 7.368110+3 5.000000-1 7.343500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1891 7.382590+3 1.500000+0 5.217000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1892 7.386980+3 1.500000+0 1.328000-3 2.300000-2 0.000000+0 0.000000+09237 2151 1893 7.403650+3 1.500000+0 7.175100-4 2.300000-2 0.000000+0 0.000000+09237 2151 1894 7.414870+3 5.000000-1 6.379900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1895 7.416432+3 1.500000+0 2.028400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1896 7.422250+3 1.500000+0 2.026600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1897 7.430411+3 1.500000+0 3.270000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1898 7.438530+3 1.500000+0 2.520000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1899 7.453090+3 1.500000+0 8.180700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1900 7.464110+3 5.000000-1 1.419900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1901 7.469050+3 1.500000+0 3.639100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1902 7.489480+3 1.500000+0 7.946300-4 2.300000-2 0.000000+0 0.000000+09237 2151 1903 7.505380+3 5.000000-1 3.488900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1904 7.512250+3 1.500000+0 8.595600-5 2.300000-2 0.000000+0 0.000000+09237 2151 1905 7.515100+3 1.500000+0 8.521500-5 2.300000-2 0.000000+0 0.000000+09237 2151 1906 7.523744+3 5.000000-1 2.625600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1907 7.530070+3 1.500000+0 3.210600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1908 7.534830+3 5.000000-1 1.629500-2 2.350000-2 0.000000+0 0.000000+09237 2151 1909 7.548661+3 1.500000+0 9.792600-4 2.300000-2 0.000000+0 0.000000+09237 2151 1910 7.552030+3 5.000000-1 2.356200-2 2.350000-2 0.000000+0 0.000000+09237 2151 1911 7.583661+3 1.500000+0 4.674700-4 2.300000-2 0.000000+0 0.000000+09237 2151 1912 7.601580+3 5.000000-1 3.331800-3 2.300000-2 0.000000+0 0.000000+09237 2151 1913 7.603280+3 1.500000+0 9.950900-3 2.350000-2 0.000000+0 0.000000+09237 2151 1914 7.631780+3 1.500000+0 3.988500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1915 7.639770+3 1.500000+0 4.494900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1916 7.660960+3 5.000000-1 3.320300-2 2.350000-2 0.000000+0 0.000000+09237 2151 1917 7.677474+3 1.500000+0 1.934100-2 2.350000-2 0.000000+0 0.000000+09237 2151 1918 7.685703+3 1.500000+0 1.945100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1919 7.695030+3 1.500000+0 4.369400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1920 7.706350+3 5.000000-1 2.506700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1921 7.707830+3 1.500000+0 3.960600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1922 7.713750+3 1.500000+0 4.317700-4 2.300000-2 0.000000+0 0.000000+09237 2151 1923 7.722641+3 1.500000+0 4.972300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1924 7.724610+3 5.000000-1 1.149300-2 2.350000-2 0.000000+0 0.000000+09237 2151 1925 7.739974+3 5.000000-1 1.579600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1926 7.745920+3 5.000000-1 1.061100-2 2.300000-2 0.000000+0 0.000000+09237 2151 1927 7.756860+3 5.000000-1 2.042900-2 2.350000-2 0.000000+0 0.000000+09237 2151 1928 7.769810+3 1.500000+0 1.009500-2 2.300000-2 0.000000+0 0.000000+09237 2151 1929 7.773840+3 5.000000-1 2.947200-2 2.350000-2 0.000000+0 0.000000+09237 2151 1930 7.789230+3 1.500000+0 4.108400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1931 7.793310+3 1.500000+0 9.089300-4 2.300000-2 0.000000+0 0.000000+09237 2151 1932 7.795320+3 5.000000-1 3.350400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1933 7.819820+3 1.500000+0 4.425700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1934 7.836590+3 5.000000-1 1.699500-4 2.300000-2 0.000000+0 0.000000+09237 2151 1935 7.851260+3 5.000000-1 6.115000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1936 7.863953+3 1.500000+0 1.414200-4 2.300000-2 0.000000+0 0.000000+09237 2151 1937 7.865110+3 1.500000+0 2.893200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1938 7.876662+3 5.000000-1 1.495600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1939 7.880995+3 1.500000+0 2.043700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1940 7.882370+3 5.000000-1 5.083100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1941 7.884330+3 1.500000+0 8.324400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1942 7.892370+3 1.500000+0 3.819700-4 2.300000-2 0.000000+0 0.000000+09237 2151 1943 7.902170+3 1.500000+0 1.776400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1944 7.904170+3 5.000000-1 2.612100-2 2.350000-2 0.000000+0 0.000000+09237 2151 1945 7.910891+3 1.500000+0 1.373500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1946 7.916530+3 5.000000-1 1.041300-2 2.300000-2 0.000000+0 0.000000+09237 2151 1947 7.925770+3 1.500000+0 3.879300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1948 7.933480+3 1.500000+0 2.869400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1949 7.934954+3 5.000000-1 3.602000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1950 7.941210+3 5.000000-1 2.351300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1951 7.943270+3 1.500000+0 5.363600-3 2.350000-2 0.000000+0 0.000000+09237 2151 1952 7.956183+3 1.500000+0 8.910000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1953 7.958250+3 1.500000+0 2.860800-3 2.300000-2 0.000000+0 0.000000+09237 2151 1954 7.965530+3 1.500000+0 4.874300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1955 7.984170+3 5.000000-1 6.781000-3 2.300000-2 0.000000+0 0.000000+09237 2151 1956 7.986933+3 1.500000+0 2.209800-2 2.350000-2 0.000000+0 0.000000+09237 2151 1957 7.990412+3 1.500000+0 3.542600-4 2.300000-2 0.000000+0 0.000000+09237 2151 1958 8.001860+3 5.000000-1 1.933500-2 2.350000-2 0.000000+0 0.000000+09237 2151 1959 8.018570+3 1.500000+0 1.839000-3 2.300000-2 0.000000+0 0.000000+09237 2151 1960 8.028704+3 1.500000+0 8.395900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1961 8.066380+3 5.000000-1 6.704300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1962 8.077475+3 1.500000+0 1.321900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1963 8.086204+3 5.000000-1 3.142100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1964 8.098662+3 1.500000+0 1.315300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1965 8.102912+3 1.500000+0 1.115200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1966 8.109820+3 5.000000-1 1.172900-3 2.300000-2 0.000000+0 0.000000+09237 2151 1967 8.114080+3 1.500000+0 7.149100-4 2.300000-2 0.000000+0 0.000000+09237 2151 1968 8.122630+3 1.500000+0 2.303500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1969 8.132210+3 1.500000+0 8.085900-4 2.300000-2 0.000000+0 0.000000+09237 2151 1970 8.151210+3 5.000000-1 9.260700-3 2.300000-2 0.000000+0 0.000000+09237 2151 1971 8.152850+3 1.500000+0 1.552800-2 2.350000-2 0.000000+0 0.000000+09237 2151 1972 8.168662+3 1.500000+0 3.286200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1973 8.171621+3 5.000000-1 7.484100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1974 8.185892+3 1.500000+0 2.381500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1975 8.191290+3 5.000000-1 1.957300-2 2.350000-2 0.000000+0 0.000000+09237 2151 1976 8.214162+3 5.000000-1 4.017000-4 2.300000-2 0.000000+0 0.000000+09237 2151 1977 8.233300+3 5.000000-1 4.793600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1978 8.267970+3 1.500000+0 4.749200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1979 8.273930+3 5.000000-1 7.942700-3 2.350000-2 0.000000+0 0.000000+09237 2151 1980 8.295413+3 1.500000+0 3.737600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1981 8.303826+3 5.000000-1 6.480600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1982 8.339060+3 1.500000+0 7.457100-4 2.300000-2 0.000000+0 0.000000+09237 2151 1983 8.343220+3 1.500000+0 5.218100-3 2.300000-2 0.000000+0 0.000000+09237 2151 1984 8.352954+3 5.000000-1 4.693500-3 2.300000-2 0.000000+0 0.000000+09237 2151 1985 8.372163+3 5.000000-1 1.497000-2 2.350000-2 0.000000+0 0.000000+09237 2151 1986 8.379140+3 1.500000+0 1.226300-3 2.300000-2 0.000000+0 0.000000+09237 2151 1987 8.383050+3 5.000000-1 1.580400-2 2.350000-2 0.000000+0 0.000000+09237 2151 1988 8.384736+3 5.000000-1 1.580100-2 2.350000-2 0.000000+0 0.000000+09237 2151 1989 8.394250+3 1.500000+0 1.221400-3 2.300000-2 0.000000+0 0.000000+09237 2151 1990 8.416140+3 1.500000+0 6.689200-3 2.300000-2 0.000000+0 0.000000+09237 2151 1991 8.418670+3 5.000000-1 1.743600-2 2.350000-2 0.000000+0 0.000000+09237 2151 1992 8.439246+3 1.500000+0 1.260100-4 2.300000-2 0.000000+0 0.000000+09237 2151 1993 8.459180+3 5.000000-1 6.928400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1994 8.489156+3 1.500000+0 2.091400-4 2.300000-2 0.000000+0 0.000000+09237 2151 1995 8.493406+3 5.000000-1 2.122100-2 2.350000-2 0.000000+0 0.000000+09237 2151 1996 8.494286+3 1.500000+0 1.990500-2 2.350000-2 0.000000+0 0.000000+09237 2151 1997 8.500000+3 5.000000-1 3.924300-2 2.300000-2 0.000000+0 0.000000+09237 2151 1998 8.509130+3 1.500000+0 5.643600-3 2.300000-2 0.000000+0 0.000000+09237 2151 1999 8.511371+3 5.000000-1 1.424000-2 2.350000-2 0.000000+0 0.000000+09237 2151 2000 8.525670+3 5.000000-1 6.045700-3 2.350000-2 0.000000+0 0.000000+09237 2151 2001 8.529000+3 1.500000+0 5.634400-3 2.300000-2 0.000000+0 0.000000+09237 2151 2002 8.537500+3 5.000000-1 4.537600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2003 8.553906+3 1.500000+0 3.635600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2004 8.556496+3 5.000000-1 3.659900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2005 8.558220+3 1.500000+0 1.659600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2006 8.563705+3 5.000000-1 6.232500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2007 8.577576+3 5.000000-1 1.917300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2008 8.580180+3 1.500000+0 3.625100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2009 8.582790+3 5.000000-1 1.883800-2 2.300000-2 0.000000+0 0.000000+09237 2151 2010 8.591486+3 1.500000+0 1.648300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2011 8.603390+3 1.500000+0 1.749400-4 2.300000-2 0.000000+0 0.000000+09237 2151 2012 8.608000+3 5.000000-1 1.500800-2 2.350000-2 0.000000+0 0.000000+09237 2151 2013 8.612540+3 1.500000+0 2.817500-2 2.350000-2 0.000000+0 0.000000+09237 2151 2014 8.621286+3 1.500000+0 1.638100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2015 8.639246+3 5.000000-1 1.454000-2 2.350000-2 0.000000+0 0.000000+09237 2151 2016 8.643920+3 1.500000+0 4.588900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2017 8.655640+3 1.500000+0 1.870000-2 2.350000-2 0.000000+0 0.000000+09237 2151 2018 8.665040+3 5.000000-1 9.954600-4 2.300000-2 0.000000+0 0.000000+09237 2151 2019 8.697446+3 1.500000+0 4.565600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2020 8.708826+3 5.000000-1 2.511500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2021 8.716970+3 5.000000-1 1.857000-2 2.300000-2 0.000000+0 0.000000+09237 2151 2022 8.721122+3 1.500000+0 6.263300-4 2.300000-2 0.000000+0 0.000000+09237 2151 2023 8.726460+3 5.000000-1 5.259700-4 2.300000-2 0.000000+0 0.000000+09237 2151 2024 8.728246+3 1.500000+0 1.601200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2025 8.743116+3 1.500000+0 1.596100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2026 8.758156+3 1.500000+0 6.145400-4 2.300000-2 0.000000+0 0.000000+09237 2151 2027 8.761000+3 5.000000-1 6.042000-3 2.300000-2 0.000000+0 0.000000+09237 2151 2028 8.763326+3 1.500000+0 8.175400-3 2.350000-2 0.000000+0 0.000000+09237 2151 2029 8.768170+3 1.500000+0 1.587400-3 2.300000-2 0.000000+0 0.000000+09237 2151 2030 8.783746+3 1.500000+0 2.531100-3 2.350000-2 0.000000+0 0.000000+09237 2151 2031 8.790946+3 1.500000+0 2.166400-3 2.300000-2 0.000000+0 0.000000+09237 2151 2032 8.792955+3 5.000000-1 1.617600-2 2.350000-2 0.000000+0 0.000000+09237 2151 2033 8.798455+3 1.500000+0 6.498700-3 2.300000-2 0.000000+0 0.000000+09237 2151 2034 8.810930+3 1.500000+0 1.572500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2035 8.823156+3 1.500000+0 1.178000-3 2.300000-2 0.000000+0 0.000000+09237 2151 2036 8.841122+3 5.000000-1 7.641500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2037 8.842800+3 1.500000+0 4.502000-3 2.300000-2 0.000000+0 0.000000+09237 2151 2038 8.853116+3 1.500000+0 7.461900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2039 8.862520+3 1.500000+0 5.811000-4 2.300000-2 0.000000+0 0.000000+09237 2151 2040 8.865560+3 5.000000-1 5.044800-3 2.350000-2 0.000000+0 0.000000+09237 2151 2041 8.876826+3 5.000000-1 1.707200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2042 8.879455+3 1.500000+0 8.406800-3 2.350000-2 0.000000+0 0.000000+09237 2151 2043 8.884746+3 1.500000+0 1.059900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2044 8.898180+3 1.500000+0 5.429700-3 2.350000-2 0.000000+0 0.000000+09237 2151 2045 8.901236+3 1.500000+0 4.476100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2046 8.909196+3 1.500000+0 3.491700-3 2.300000-2 0.000000+0 0.000000+09237 2151 2047 8.912560+3 5.000000-1 5.839500-3 2.350000-2 0.000000+0 0.000000+09237 2151 2048 8.919750+3 5.000000-1 9.214100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2049 8.928210+3 1.500000+0 1.531500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2050 8.943590+3 1.500000+0 2.793200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2051 8.949890+3 1.500000+0 8.399000-3 2.300000-2 0.000000+0 0.000000+09237 2151 2052 8.957786+3 5.000000-1 2.204100-2 2.300000-2 0.000000+0 0.000000+09237 2151 2053 8.959550+3 1.500000+0 2.113900-2 2.350000-2 0.000000+0 0.000000+09237 2151 2054 8.973246+3 1.500000+0 1.515600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2055 8.991220+3 1.500000+0 1.023900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2056 8.993706+3 1.500000+0 5.415900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2057 9.002414+3 1.500000+0 5.902900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2058 9.013236+3 5.000000-1 4.285600-2 2.350000-2 0.000000+0 0.000000+09237 2151 2059 9.014866+3 1.500000+0 6.388300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2060 9.022786+3 1.500000+0 3.416200-2 2.350000-2 0.000000+0 0.000000+09237 2151 2061 9.024220+3 5.000000-1 1.112900-2 2.300000-2 0.000000+0 0.000000+09237 2151 2062 9.033600+3 1.500000+0 5.396700-3 2.300000-2 0.000000+0 0.000000+09237 2151 2063 9.055040+3 5.000000-1 4.092300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2064 9.060696+3 1.500000+0 1.130600-2 2.300000-2 0.000000+0 0.000000+09237 2151 2065 9.071550+3 5.000000-1 3.671400-2 2.350000-2 0.000000+0 0.000000+09237 2151 2066 9.073446+3 1.500000+0 8.292100-3 2.350000-2 0.000000+0 0.000000+09237 2151 2067 9.076590+3 1.500000+0 2.449300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2068 9.079496+3 5.000000-1 3.595700-2 2.350000-2 0.000000+0 0.000000+09237 2151 2069 9.088250+3 1.500000+0 2.444800-3 2.300000-2 0.000000+0 0.000000+09237 2151 2070 9.103420+3 1.500000+0 1.469300-3 2.300000-2 0.000000+0 0.000000+09237 2151 2071 9.105456+3 5.000000-1 1.198400-2 2.300000-2 0.000000+0 0.000000+09237 2151 2072 9.111786+3 1.500000+0 2.047600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2073 9.123831+3 1.500000+0 3.402600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2074 9.138956+3 5.000000-1 2.355600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2075 9.151050+3 1.500000+0 3.877300-2 2.350000-2 0.000000+0 0.000000+09237 2151 2076 9.152970+3 5.000000-1 1.765000-2 2.350000-2 0.000000+0 0.000000+09237 2151 2077 9.161906+3 1.500000+0 1.738500-3 2.300000-2 0.000000+0 0.000000+09237 2151 2078 9.172770+3 1.500000+0 9.251400-3 2.300000-2 0.000000+0 0.000000+09237 2151 2079 9.183020+3 5.000000-1 1.763400-2 2.300000-2 0.000000+0 0.000000+09237 2151 2080 9.186736+3 1.500000+0 1.815000-2 2.350000-2 0.000000+0 0.000000+09237 2151 2081 9.193920+3 1.500000+0 1.920000-3 2.300000-2 0.000000+0 0.000000+09237 2151 2082 9.204206+3 5.000000-1 1.473200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2083 9.207560+3 1.500000+0 1.914900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2084 9.213210+3 1.500000+0 4.665200-4 2.300000-2 0.000000+0 0.000000+09237 2151 2085 9.228050+3 1.500000+0 4.615900-4 2.300000-2 0.000000+0 0.000000+09237 2151 2086 9.238390+3 1.500000+0 2.870200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2087 9.244866+3 5.000000-1 1.443800-3 2.300000-2 0.000000+0 0.000000+09237 2151 2088 9.246390+3 1.500000+0 7.552400-3 2.350000-2 0.000000+0 0.000000+09237 2151 2089 9.250696+3 1.500000+0 1.031100-3 2.300000-2 0.000000+0 0.000000+09237 2151 2090 9.283831+3 1.500000+0 4.304200-3 2.300000-2 0.000000+0 0.000000+09237 2151 2091 9.291790+3 5.000000-1 7.144700-3 2.300000-2 0.000000+0 0.000000+09237 2151 2092 9.304060+3 1.500000+0 4.294900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2093 9.311390+3 1.500000+0 4.291600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2094 9.325020+3 1.500000+0 9.089600-4 2.300000-2 0.000000+0 0.000000+09237 2151 2095 9.327170+3 5.000000-1 1.403100-2 2.350000-2 1.120000-4 0.000000+09237 2151 2096 9.341430+3 1.500000+0 4.237700-4 2.300000-2 0.000000+0 0.000000+09237 2151 2097 9.357576+3 1.500000+0 1.377600-3 2.300000-2 0.000000+0 0.000000+09237 2151 2098 9.365290+3 1.500000+0 4.266900-3 2.300000-2 0.000000+0 0.000000+09237 2151 2099 9.375500+3 1.500000+0 8.583600-4 2.350000-2 0.000000+0 0.000000+09237 2151 2100 9.379036+3 5.000000-1 5.433700-3 2.300000-2 0.000000+0 0.000000+09237 2151 2101 9.403560+3 1.500000+0 2.802600-3 2.300000-2 0.000000+0 0.000000+09237 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0.000000+09237 2151 2305 5.000000+4 1.068700+1 7.183650-2 1.316960-3 2.323690-2 0.000000+09237 2151 2306 5.500000+4 1.056900+1 1.000490-1 1.301680-3 2.329150-2 0.000000+09237 2151 2307 6.000000+4 1.045210+1 1.205790-1 1.286570-3 2.334590-2 0.000000+09237 2151 2308 7.000000+4 1.022260+1 1.509830-1 1.256970-3 2.345590-2 0.000000+09237 2151 2309 8.000000+4 9.998350+0 1.736300-1 1.228140-3 2.356620-2 0.000000+09237 2151 2310 9.000000+4 9.779200+0 1.916700-1 1.200040-3 2.367730-2 0.000000+09237 2151 2311 1.000000+5 9.565060+0 2.065770-1 1.172670-3 2.378880-2 0.000000+09237 2151 2312 1.200000+5 9.151280+0 2.300500-1 1.120040-3 2.401360-2 0.000000+09237 2151 2313 1.400000+5 8.756100+0 2.478480-1 1.070090-3 2.424070-2 0.000000+09237 2151 2314 1.490290+5 8.583590+0 2.545650-1 1.048390-3 2.434410-2 0.000000+09237 2151 2315 2.500000+0 0.000000+0 5 0 144 239237 2151 2316 0.000000+0 0.000000+0 2.000000+0 1.000000+0 0.000000+0 0.000000+09237 2151 2317 1.000000+4 8.299250+0 0.000000+0 1.027740-3 2.280600-2 0.000000+09237 2151 2318 1.200000+4 8.262180+0 0.000000+0 1.022880-3 2.282740-2 0.000000+09237 2151 2319 1.400000+4 8.225280+0 0.000000+0 1.018050-3 2.284870-2 0.000000+09237 2151 2320 1.600000+4 8.188530+0 0.000000+0 1.013240-3 2.287000-2 0.000000+09237 2151 2321 1.800000+4 8.151940+0 0.000000+0 1.008450-3 2.289140-2 0.000000+09237 2151 2322 2.000000+4 8.115530+0 0.000000+0 1.003690-3 2.291280-2 0.000000+09237 2151 2323 2.300000+4 8.061240+0 0.000000+0 9.966010-4 2.294500-2 0.000000+09237 2151 2324 2.600000+4 8.007320+0 0.000000+0 9.895640-4 2.297720-2 0.000000+09237 2151 2325 3.000000+4 7.936130+0 0.000000+0 9.802820-4 2.302050-2 0.000000+09237 2151 2326 3.500000+4 7.847890+0 0.000000+0 9.687930-4 2.307440-2 0.000000+09237 2151 2327 4.000000+4 7.760740+0 0.000000+0 9.574650-4 2.312850-2 0.000000+09237 2151 2328 4.500000+4 7.674530+0 0.000000+0 9.462730-4 2.318260-2 0.000000+09237 2151 2329 4.509020+4 7.672980+0 0.000000+0 9.460720-4 2.318350-2 0.000000+09237 2151 2330 5.000000+4 7.589370+0 5.101460-2 9.352360-4 2.323690-2 0.000000+09237 2151 2331 5.500000+4 7.505260+0 7.104670-2 9.243480-4 2.329150-2 0.000000+09237 2151 2332 6.000000+4 7.421930+0 8.562190-2 9.135810-4 2.334590-2 0.000000+09237 2151 2333 7.000000+4 7.258440+0 1.072040-1 8.924990-4 2.345590-2 0.000000+09237 2151 2334 8.000000+4 7.098690+0 1.232750-1 8.719600-4 2.356620-2 0.000000+09237 2151 2335 9.000000+4 6.942590+0 1.360730-1 8.519530-4 2.367730-2 0.000000+09237 2151 2336 1.000000+5 6.790070+0 1.466460-1 8.324610-4 2.378880-2 0.000000+09237 2151 2337 1.200000+5 6.495360+0 1.632840-1 7.949790-4 2.401360-2 0.000000+09237 2151 2338 1.400000+5 6.213960+0 1.758910-1 7.594130-4 2.424070-2 0.000000+09237 2151 2339 1.490290+5 6.091120+0 1.806460-1 7.439600-4 2.434410-2 0.000000+09237 2151 2340 9237 2 0 2341 9237 0 0 2342 9.223800+4 2.360058+2 0 0 0 09237 3 18 2343 1.980600+8 1.980600+8 0 0 1 1389237 3 18 2344 138 2 0 0 0 09237 3 18 2345 1.000000-5 0.000000+0 4.000000+3 0.000000+0 1.000000+4 0.000000+09237 3 18 2346 1.000000+4 9.700000-5 1.200000+4 9.744840-5 1.400000+4 9.782910-59237 3 18 2347 1.600000+4 9.614560-5 1.800000+4 9.285120-5 2.000000+4 9.000000-59237 3 18 2348 2.300000+4 8.451800-5 2.600000+4 7.998470-5 3.000000+4 7.500000-59237 3 18 2349 3.500000+4 6.831050-5 4.000000+4 6.300000-5 4.500000+4 6.150470-59237 3 18 2350 4.509020+4 6.147960-5 5.000000+4 6.019720-5 5.500000+4 5.903840-59237 3 18 2351 6.000000+4 5.800000-5 7.000000+4 5.707790-5 8.000000+4 5.629090-59237 3 18 2352 9.000000+4 5.560580-5 1.000000+5 5.500000-5 1.200000+5 5.679310-59237 3 18 2353 1.400000+5 5.835470-5 1.490290+5 5.900000-5 1.600000+5 6.083960-59237 3 18 2354 1.700000+5 6.245490-5 1.800000+5 6.401710-5 1.900000+5 6.553090-59237 3 18 2355 2.000000+5 6.700000-5 2.100000+5 6.945760-5 2.200000+5 7.188470-59237 3 18 2356 2.300000+5 7.428310-5 2.350000+5 7.547210-5 2.400000+5 7.665440-59237 3 18 2357 2.450000+5 7.783030-5 2.500000+5 7.900000-5 2.600000+5 8.363940-59237 3 18 2358 2.700000+5 8.836070-5 2.800000+5 9.316230-5 3.000000+5 1.030000-49237 3 18 2359 3.800000+5 2.329100-4 4.000000+5 2.513100-4 4.200000+5 2.757700-49237 3 18 2360 4.300000+5 2.757700-4 4.400000+5 2.830300-4 4.500000+5 2.849200-49237 3 18 2361 4.600000+5 2.907900-4 4.700000+5 3.037200-4 5.000000+5 3.764200-49237 3 18 2362 5.500000+5 6.295200-4 5.800000+5 6.907800-4 5.900000+5 7.588000-49237 3 18 2363 6.000000+5 8.225500-4 6.200000+5 9.276700-4 6.400000+5 1.127760-39237 3 18 2364 6.500000+5 1.239150-3 6.600000+5 1.293840-3 6.800000+5 1.574290-39237 3 18 2365 7.000000+5 1.716510-3 7.500000+5 2.573770-3 7.800000+5 3.578210-39237 3 18 2366 8.000000+5 4.470280-3 8.500000+5 7.168360-3 8.800000+5 1.077044-29237 3 18 2367 9.000000+5 1.362465-2 9.300000+5 1.549431-2 9.500000+5 1.653853-29237 3 18 2368 9.700000+5 1.582249-2 1.000000+6 1.398000-2 1.020000+6 1.570000-29237 3 18 2369 1.030000+6 1.692639-2 1.050000+6 2.000000-2 1.080000+6 2.700000-29237 3 18 2370 1.100000+6 3.004000-2 1.130000+6 3.600000-2 1.140000+6 3.810000-29237 3 18 2371 1.150000+6 3.920000-2 1.170000+6 4.024741-2 1.200000+6 4.208724-29237 3 18 2372 1.230000+6 4.030000-2 1.240000+6 4.000000-2 1.250000+6 3.912000-29237 3 18 2373 1.280000+6 5.020000-2 1.300000+6 6.500000-2 1.350000+6 1.118813-19237 3 18 2374 1.400000+6 1.855000-1 1.450000+6 2.822391-1 1.480000+6 3.310000-19237 3 18 2375 1.500000+6 3.560000-1 1.525000+6 3.805000-1 1.550000+6 3.990000-19237 3 18 2376 1.575000+6 4.125000-1 1.600000+6 4.226000-1 1.700000+6 4.550000-19237 3 18 2377 1.800000+6 4.820000-1 1.900000+6 5.070000-1 2.000000+6 5.250000-19237 3 18 2378 2.100000+6 5.355000-1 2.200000+6 5.391000-1 2.400000+6 5.373000-19237 3 18 2379 2.600000+6 5.328000-1 2.800000+6 5.270000-1 3.000000+6 5.160000-19237 3 18 2380 3.200000+6 5.210000-1 3.600000+6 5.354000-1 4.000000+6 5.483000-19237 3 18 2381 4.500000+6 5.496000-1 4.700000+6 5.470000-1 5.000000+6 5.405000-19237 3 18 2382 5.300000+6 5.430000-1 5.500000+6 5.500000-1 5.800000+6 5.731000-19237 3 18 2383 6.000000+6 6.153000-1 6.200000+6 6.859000-1 6.500000+6 8.257000-19237 3 18 2384 6.700000+6 8.870000-1 7.000000+6 9.403000-1 7.300000+6 9.680000-19237 3 18 2385 7.500000+6 9.807000-1 7.750000+6 9.910000-1 8.000000+6 9.935000-19237 3 18 2386 8.500000+6 1.000000+0 9.000000+6 9.979000-1 1.000000+7 9.868000-19237 3 18 2387 1.100000+7 9.830000-1 1.150000+7 9.830000-1 1.200000+7 9.850000-19237 3 18 2388 1.300000+7 1.013000+0 1.400000+7 1.130000+0 1.450000+7 1.155000+09237 3 18 2389 1.500000+7 1.198000+0 1.600000+7 1.259300+0 1.700000+7 1.256100+09237 3 18 2390 1.800000+7 1.249300+0 1.900000+7 1.295400+0 2.000000+7 1.352100+09237 3 18 2391 9237 3 0 2392 9237 0 0 2393 9.223800+4 2.360060+2 0 0 0 1923733 18 2394 0.000000+0 0.000000+0 0 18 0 2923733 18 2395 0.000000+0 0.000000+0 0 1 32 16923733 18 2396 1.000000-5 6.250000-2 1.000000+0 0.000000+0 4.000000+3 9.000000-2923733 18 2397 3.000000+5 1.102500-2 8.000000+5 5.625000-3 1.600000+6 6.800000-4923733 18 2398 2.500000+6 8.810000-4 3.000000+6 6.800000-4 4.000000+6 6.800000-4923733 18 2399 5.000000+6 6.760000-4 6.000000+6 1.521000-3 8.000000+6 1.024000-3923733 18 2400 1.000000+7 8.410000-4 1.200000+7 1.369000-3 1.400000+7 4.096000-3923733 18 2401 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0923733 18 2402 0.000000+0 0.000000+0 0 1 4 2923733 18 2403 1.000000-5 2.250000-4 2.000000+7 0.000000+0 0.000000+0 0.000000+0923733 18 2404 923733 0 2405 9237 0 0 2406 0 0 0 2407 9.323700+4 2.350118+2 1 1 0 19346 1451 1 0.000000+0 1.000000+0 0 0 0 69346 1451 2 1.000000+0 0.000000+0 0 0 10 69346 1451 3 0.000000+0 0.000000+0 0 0 217 39346 1451 4 93-NP-237 LANL EVAL-APR90 P.YOUNG, E.ARTHUR, F.MANN 9346 1451 5 DIST-JUN90 900531 9346 1451 6 ----ENDF/B-VI MATERIAL 9346 9346 1451 7 -----INCIDENT NEUTRON DATA 9346 1451 8 ------ENDF-6 FORMAT 9346 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9346 1451 10 DOSIMETRY 9346 1451 11 ******************************************************************9346 1451 12 9346 1451 13 **************************************************************** 9346 1451 14 9346 1451 15 SUMMARY OF ENDF/B-VI EVALUATION 9346 1451 16 9346 1451 17 Principal evaluators: P.G.Young, E.D.Arthur, F.M.Mann, with 9346 1451 18 significant contributions from D.G.Madland, M.Bozoian, and 9346 1451 19 R.E.MacFarlane. 9346 1451 20 9346 1451 21 The evaluation below about 8 keV is based on the analysis of 9346 1451 22 Derrien (De80) and is the same as JEF-2 (De89). 9346 1451 23 9346 1451 24 The evaluation above 8 keV is based on a detailed theoretical 9346 1451 25 analysis utilizing the available experimental data. Coupled 9346 1451 26 channel optical model calculations with the ECIS code (Ra70) 9346 1451 27 were used to provide the total, elastic, and inelastic cross 9346 1451 28 sections to the first 3 members of the ground state rotational 9346 1451 29 band, as well as neutron elastic and inelastic angular distri- 9346 1451 30 butions to the rotational levels. The optical potential of 9346 1451 31 Lagrange (Ha80) was extended to higher energies based on ENDF/B- 9346 1451 32 VI analyses for U235, U238, Pu239. The ECIS calculation was also 9346 1451 33 used to provide neutron transmission coefficients. Hauser- 9346 1451 34 Feshbach statistical theory calculations were carried out with 9346 1451 35 the GNASH (Ar88, Yo77) and COMNUC (Du70) code systems, including 9346 1451 36 preequilibrium and fission. DWBA calculations were performed 9346 1451 37 with the DWUCK code (Ku70) for several vibrational levels, using 9346 1451 38 B(El) values inferred from (d,d') data on various actinides, as 9346 1451 39 well as Coulomb excitation measurements. Based on systematics, 9346 1451 40 fictitious l=2 and l=3 vibrational states were located near 9346 1451 41 excitation energy 1 MeV, containing all the vibrational 9346 1451 42 strength. 9346 1451 43 9346 1451 44 A summary of the theoretical methods and evaluation up to 9346 1451 45 5 MeV is given in reference Ar85. 9346 1451 46 9346 1451 47 **********MF=1 Descriptive and Nubar Information*****************9346 1451 48 9346 1451 49 MT=452 Total Nubar. Sum of MT=455 and 456. 9346 1451 50 MT=455 Delayed Neutron Yields. Based on the analysis and 9346 1451 51 evaluation of England for ENDF/B-VI (En89). 9346 1451 52 MT=456 Prompt Neutron Yields. Based on smooth curve through 9346 1451 53 experimental data of Ma83,Ve78,Fr82, after renormaliza- 9346 1451 54 tion for ENDF/B-VI standards. Results agree closely 9346 1451 55 with values from Madland-Nix theory (Ma84). 9346 1451 56 MT=458 Energy from Fission. Based on revised Sher (Sh83). 9346 1451 57 9346 1451 58 **********MF=2 Resonance Region**********************************9346 1451 59 9346 1451 60 MT=151 Resolved Resonance Parameters. Derrien evaluation (De80, 9346 1451 61 De89) used for resolved resonance region, which spans 9346 1451 62 the range 1.0E-5 to 150 eV. 9346 1451 63 9346 1451 64 **********MF=3 Smooth Cross Sections*****************************9346 1451 65 9346 1451 66 MT=1 Neutron Total Cross Section. Derrien evaluation (De80, 9346 1451 67 De89) used from 1.0E-5 eV to 0.008 MeV. Above 0.008 MeV,9346 1451 68 coupled channel optical model calculations (ECIS code) 9346 1451 69 with modified Lagrange (Ha80) potential used as prior 9346 1451 70 in a covariance analysis of measurement by Lisowski et 9346 1451 71 al. from 3 to 20 MeV (Li90). 9346 1451 72 MT=2 All Energies, based on subtraction of MT=4,16,17,18,37, 9346 1451 73 and 102 from MT=1. 9346 1451 74 MT=4 Sum of MT=51-91 9346 1451 75 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 76 For 9 MeV and below, the experimental data of Kornilov et 9346 1451 77 al. (Ko85) are used. 9346 1451 78 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 79 MT=18 Fission Cross Section. Derrien evaluation (De80,De89) 9346 1451 80 used to 0.8 MeV. Empirical fit mainly to experimental 9346 1451 81 data of Lisowski et al.(Li88) and Meadows (Me83,Me88) 9346 1451 82 in the energy range 0.9-20 MeV, but with consideration 9346 1451 83 of the data of Te86, Ka85, Ca82, Wh67, Za84, and Va82. 9346 1451 84 MT=19 (n,f) first-chance fission cross section. 9346 1451 85 Ratio of first-chance to total fission obtained from 9346 1451 86 GNASH calculations. 9346 1451 87 MT=20 (n,nf) second-chance fission cross section. 9346 1451 88 Ratio of second-chance to total fission obtained from 9346 1451 89 GNASH calculations. 9346 1451 90 MT=21 (n,2nf) third-chance fission cross section. 9346 1451 91 Ratio of third-chance to total fission obtained from 9346 1451 92 GNASH calculations. 9346 1451 93 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 94 MT=38 (n,3nf) fourth-chance fission cross section. 9346 1451 95 Ratio of fourth-chance to total fission obtained from 9346 1451 96 GNASH calculations. 9346 1451 97 MT=51,53 Threshold to 20 MeV, coupled-channel optical model 9346 1451 98 calculations (7/2+ and 9/2+ members of the K=5/2 ground 9346 1451 99 state rotational band) using the ECIS code. Compound 9346 1451 100 nucleus contributions, obtained from COMNUC calcula- 9346 1451 101 tions, are also included. 9346 1451 102 MT=52,54-79 Threshold to 6.0 MeV, compound nucleus reaction 9346 1451 103 theory (COMNUC code) calculation with width fluctuation 9346 1451 104 corrections. Cross section assumed zero above 6 MeV. 9346 1451 105 MT=80 Threshold to 20 MeV, distorted wave Born approximation 9346 1451 106 (DWUCK code) calculations with the DWUCK code for 9346 1451 107 l=2 vibrational states. Systematics in actinide region 9346 1451 108 used to infer B(E2) values for normalizing DWBA 9346 1451 109 calculations. 9346 1451 110 MT=81 Threshold to 20 MeV, distorted wave Born approximation 9346 1451 111 (DWUCK code) calculations with the DWUCK code for 9346 1451 112 l=3 vibrational states. Systematics in actinide region 9346 1451 113 used to infer B(E3) values for normalizing DWBA 9346 1451 114 calculations. 9346 1451 115 MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 116 Note that the MT=80,81 vibrational states lie in the 9346 1451 117 MT=91 continuum region. 9346 1451 118 MT=102 Radiative Capture. From 1.0e-5 eV to 5 MeV, based on 9346 1451 119 Derrien (De80,De89) evaluation, which follows exp. 9346 1451 120 data of Weston (We81). Above 5 MeV, ENDF/B-V evaluation 9346 1451 121 is retained. 9346 1451 122 9346 1451 123 **********MF=4 Neutron Angular Distributions******************** 9346 1451 124 9346 1451 125 MT=2 Elastic scattering angular distribution based on ECIS 9346 1451 126 coupled-channel calculations, with a compound elastic 9346 1451 127 component from COMNUC included below 6 MeV. 9346 1451 128 MT=51-55,57 Thres. to 20 MeV, Coupled-channel optical model 9346 1451 129 calculations plus compound-nucleus contributions. 9346 1451 130 MT=52,54-79 Threshold to 6.0 MeV, COMNUC compound 9346 1451 131 nucleus reaction theory calculation with width fluctua- 9346 1451 132 tions. Cross sections assumed zero from 6-20 MeV. 9346 1451 133 MT=80,81 Thres. to 20 MeV, Distorted wave Born approximation 9346 1451 134 calculations with DWUCK code. 9346 1451 135 9346 1451 136 ************MF=5 Neutron Energy Distributions******************* 9346 1451 137 9346 1451 138 MF=18 Composite neutron energy distributions from fission. 9346 1451 139 Based on calculations by D.Madland (Ma84) using Madland-9346 1451 140 Nix formalism. The calculations include the first-, 9346 1451 141 second-, and third-chance fission neutron components. 9346 1451 142 The data are represented using LAW 12. 9346 1451 143 MT=455 Provided by Tal England (En89). 9346 1451 144 9346 1451 145 ************MF=6 Correlated Energy-Angle Distributions********** 9346 1451 146 9346 1451 147 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 148 Updated Kalbach-Mann systematics used for specifying 9346 1451 149 neutron distributions (Ka87). Only neutrons given. 9346 1451 150 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 151 Updated Kalbach-Mann systematics used for specifying 9346 1451 152 neutron distributions (Ka87). Only neutrons given. 9346 1451 153 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 154 Updated Kalbach-Mann systematics used for specifying 9346 1451 155 neutron distributions (Ka87). Only neutrons given. 9346 1451 156 MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9346 1451 157 Updated Kalbach-Mann systematics used for specifying 9346 1451 158 neutron distributions (Ka87). Only neutrons given. 9346 1451 159 9346 1451 160 ****************MF=8 Radioactive Nuclide Data******************* 9346 1451 161 9346 1451 162 MT=16 Information on ground and 77-keV isomeric states. 9346 1451 163 9346 1451 164 ***********MF=8 Multiplicities for Radioactive Nuclides********* 9346 1451 165 9346 1451 166 MT=16 Branching ratios calculated by Gardner & Gardner (Ga89) 9346 1451 167 were used. 9346 1451 168 9346 1451 169 ****************MF=12,13,14,15 Photon Production**************** 9346 1451 170 9346 1451 171 The evaluation of photon production was performed by Fred Mann 9346 1451 172 (WHC). Photon production for discrete inelastic scattering 9346 1451 173 (mt = 51 - 79) uses transition probability arrays. Gamma-ray 9346 1451 174 branching for the excited states of Np237 was obtained from the 9346 1451 175 Nuclear Data Sheets. Photon production for neutron capture was 9346 1451 176 taken from the ENDF/B-VI evaluation of Pu-240 (which has a similar9346 1451 177 Q value). Photon production for fission was taken from the 9346 1451 178 ENDF/B-VI evaluation of U235. The photon spectrum for non elastic9346 1451 179 events (mt=3) was taken from the ENDF/B-VI evaluation of U235. 9346 1451 180 The photon yield was adjusted to conserve energy. 9346 1451 181 9346 1451 182 ******************References************************************ 9346 1451 183 9346 1451 184 Ar85 E.D.Arthur et al., LA-10288-PR (1985)p.13. 9346 1451 185 Ar88 E.D.Arthur, LA-UR-88-382 (1988). 9346 1451 186 Ca82 M.Cance et al., Antwerp Nuc.Data Conf., p.51 (1982). 9346 1451 187 De80 H.Derrien, INDC(FR)-42 (1980). 9346 1451 188 De89 H.Derrien, personal communication (July 7, 1989). 9346 1451 189 Du70 C.L.Dunford, AI-AEC-12931 (1970). 9346 1451 190 En89 M.C.Brady & T.R.England,NSE 103,129(1989). 9346 1451 191 Fr82 J.Frehaut et al., Int.Conf.on Nucl.Data for Sci.& Tech., 9346 1451 192 Antwerp, 6-10 Sept.1982, p.78. 9346 1451 193 Ga89 G.Gardner & M.Gardner, Personal Communication (1989) and 9346 1451 194 UCAR-10062-83/1 (1983) p.51. 9346 1451 195 Ha80 G.Haouat,Ch.Lagrange, et al., Nucl.Cross Sect. for Tech., 9346 1451 196 Knoxville,Tenn., Oct. 22-26, 1979. 9346 1451 197 Ka85 K.Kanda et al., JAERI-M-85-035, 220 (1985). 9346 1451 198 Ka87 C.Kalbach, LA-UR-87-4139 (1987) [to be pub.in Phys.Rev.C]. 9346 1451 199 Ko85 N.V.Kornilov et al., Sov.J.At.En.58,131(1985). 9346 1451 200 Ku70 P.D.Kunz, DWUCK: A Distorted-Wave Born Approximation 9346 1451 201 Program, unpublished report. 9346 1451 202 Li88 P.W.Lisowski et al., Nucl.Data for Sci.& Tech. Conf., Mito 9346 1451 203 Japan, May 30 - June 3, 1988. 9346 1451 204 Li90 P.W.Lisowski et al., Personal Communication. 9346 1451 205 Ma83 V.V.Malinovsky et al., YK 1,50 (1983). 9346 1451 206 Ma84 D.G.Madland, personal communication (1984). 9346 1451 207 Me83 J.W.Meadows et al., Nuc.Sci.Eng.85,271(1983). 9346 1451 208 Me88 J.W.Meadows et al., Ann.Nuc.En.15,421(1988). 9346 1451 209 Ra70 J.Raynal,IAEA SMR-9/8 (1970). 9346 1451 210 Sh83 R.Sher & Beck, EPRI NP1771/81 + Rev1/83 + personal 9346 1451 211 communication to Magurno (Feb., 1983). 9346 1451 212 Ve78 L.Veeser et al., Phys.Rev.C17, 385 (1978). 9346 1451 213 Te86 H.Terayama, NETU-47 (1986). 9346 1451 214 We81 L.Weston et al., Nuc.Sci.Eng.79,184(1981). 9346 1451 215 Wh67 P.White, Jour.Nuc.Ener.21, 671 (1967). 9346 1451 216 Va82 M.Varnagy, Nuc.Inst.Meth. 196, 465 (1982). 9346 1451 217 Za84 K.Zasadny, Tr.Am.Nuc.Soc. 47, 425 (1984). 9346 1451 218 Yo77 P.G.Young & E.D.Arthur, LA-6947 (1977). 9346 1451 219 9346 1451 220 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582 5.250000+6 1.442241+0 5.300000+6 1.439511+0 5.350000+6 1.439314+09346 3 18 583 5.400000+6 1.441105+0 5.450000+6 1.444375+0 5.500000+6 1.448642+09346 3 18 584 5.550000+6 1.453895+0 5.600000+6 1.461479+0 5.700000+6 1.486412+09346 3 18 585 5.750000+6 1.505077+0 5.800000+6 1.528818+0 5.900000+6 1.591260+09346 3 18 586 6.000000+6 1.660739+0 6.100000+6 1.720998+0 6.200000+6 1.776434+09346 3 18 587 6.250000+6 1.803230+0 6.300000+6 1.829062+0 6.400000+6 1.877752+09346 3 18 588 6.500000+6 1.924258+0 6.600000+6 1.968192+0 6.700000+6 2.008960+09346 3 18 589 6.750000+6 2.027979+0 6.800000+6 2.045994+0 6.900000+6 2.078806+09346 3 18 590 7.000000+6 2.107741+0 7.100000+6 2.133240+0 7.200000+6 2.155626+09346 3 18 591 7.250000+6 2.165752+0 7.300000+6 2.175221+0 7.400000+6 2.192353+09346 3 18 592 7.500000+6 2.207671+0 7.600000+6 2.221286+0 7.700000+6 2.232980+09346 3 18 593 7.750000+6 2.238043+0 7.800000+6 2.242552+0 7.900000+6 2.249827+09346 3 18 594 8.000000+6 2.255494+0 8.100000+6 2.259339+0 8.200000+6 2.260375+09346 3 18 595 8.250000+6 2.259555+0 8.300000+6 2.257699+0 8.400000+6 2.250505+09346 3 18 596 8.500000+6 2.240245+0 8.600000+6 2.228945+0 8.700000+6 2.217506+09346 3 18 597 8.750000+6 2.211995+0 8.800000+6 2.206751+0 8.900000+6 2.197400+09346 3 18 598 9.000000+6 2.188693+0 9.250000+6 2.168189+0 9.500000+6 2.151158+09346 3 18 599 9.750000+6 2.137745+0 1.000000+7 2.129200+0 1.025000+7 2.121739+09346 3 18 600 1.050000+7 2.114313+0 1.075000+7 2.107009+0 1.100000+7 2.099902+09346 3 18 601 1.125000+7 2.093053+0 1.150000+7 2.086565+0 1.175000+7 2.081664+09346 3 18 602 1.200000+7 2.078382+0 1.225000+7 2.076125+0 1.250000+7 2.074368+09346 3 18 603 1.275000+7 2.072999+0 1.300000+7 2.074517+0 1.325000+7 2.079220+09346 3 18 604 1.350000+7 2.086802+0 1.375000+7 2.096991+0 1.400000+7 2.111111+09346 3 18 605 1.425000+7 2.137832+0 1.450000+7 2.168353+0 1.475000+7 2.190883+09346 3 18 606 1.500000+7 2.206203+0 1.525000+7 2.219687+0 1.550000+7 2.229985+09346 3 18 607 1.575000+7 2.235866+0 1.600000+7 2.236222+0 1.625000+7 2.233362+09346 3 18 608 1.650000+7 2.230353+0 1.675000+7 2.227180+0 1.700000+7 2.223829+09346 3 18 609 1.725000+7 2.220290+0 1.750000+7 2.216557+0 1.775000+7 2.212833+09346 3 18 610 1.800000+7 2.209198+0 1.825000+7 2.205630+0 1.850000+7 2.202109+09346 3 18 611 1.875000+7 2.198615+0 1.900000+7 2.195132+0 1.925000+7 2.191643+09346 3 18 612 1.950000+7 2.188133+0 1.975000+7 2.184590+0 2.000000+7 2.181000+09346 3 18 613 9346 3 0 614 9346 0 0 615 0 0 0 616 9.423900+4 2.369986+2 1 1 0 19437 1451 1 0.000000+0 1.000000+0 0 0 0 69437 1451 2 1.000000+0 0.000000+0 0 0 10 69437 1451 3 0.000000+0 0.000000+0 0 0 566 39437 1451 4 94-Pu-239 LANL EVAL-APR89 P.YOUNG, L.WESTON, W.POENITZ 9437 1451 5 DIST-JUN90 900531 9437 1451 6 ----ENDF/B-VI MATERIAL 9437 9437 1451 7 -----INCIDENT NEUTRON DATA 9437 1451 8 ------ENDF-6 FORMAT 9437 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9437 1451 10 DOSIMETRY 9437 1451 11 ******************************************************************9437 1451 12 9437 1451 13 **************************************************************** 9437 1451 14 9437 1451 15 SUMMARY OF ENDF/B-VI EVALUATION 9437 1451 16 P.G.Young, L.W.Weston, and W.P.Poenitz 9437 1451 17 9437 1451 18 1. Resonance region evaluation performed by H.Derrien and 9437 1451 19 G.de Saussure ORNL/TM-10986 (January, 1989). NEW PARAMETER 9437 1451 20 SET EXTENDING TO 2 KEV SUBSTITUTED NOV. 89. 9437 1451 21 2. Prompt nubar evaluation obtained from work of E.Fort (Nuc. 9437 1451 22 Sci.Eng.99, 375 (1988). 9437 1451 23 3. Delayed nubar evaluation (En89). 9437 1451 24 4. Energy range above unresolved resonances (0.03 - 20 MeV) was 9437 1451 25 evaluated by P.Young, using results from W.Poenitz, G.Hale, 9437 1451 26 R.Peelle, and A.Carlson from the simultaneous standards 9437 1451 27 evaluation (Santa Fe Conf.(1985)p.1429), and with contri- 9437 1451 28 butions from R.MacFarlane and L.Weston. More details are 9437 1451 29 given below, and full documentation will be issued later. 9437 1451 30 9437 1451 31 *************************************************** 9437 1451 32 THERMAL REGION 9437 1451 33 *************************************************** 9437 1451 34 9437 1451 35 THE REICH-MOORE RESONANCE PARAMETERS HAVE BEEN OBTAINED FROM THE 9437 1451 36 ANALYSIS OF SEVERAL TRANSMISSION,FISSION,ABSORPTION AND CAPTURE 9437 1451 37 EXPERIMENTAL DATA(1,4,5,7,10). THIS SET OF RESONANCE PARAMETERS 9437 1451 38 ALSO CONSIDERED THE 1988 FISSION DATA OF WESTON AND TODD. THE 9437 1451 39 THERMAL REGION WAS THEN REFITTED AND THE FOLLOWING WERE OBTAINED 9437 1451 40 FOR THE 0.0253 CROSS SECTIONS. 9437 1451 41 9437 1451 42 EVALUATION PROPOSED 9437 1451 43 293K STANDARD VALUES(BARN)(11) 9437 1451 44 **************************************************************** 9437 1451 45 FISSION 747.08 747.99+-1.87 9437 1451 46 CAPTURE 271.39 271.43+-2.14 9437 1451 47 SCATTERING 8.00 7.88+-0.97 9437 1451 48 TOTAL 1025.79 1027.30+-5.00 9437 1451 49 **************************************************************** 9437 1451 50 9437 1451 51 THE EXPERIMENTAL FISSION CROSS-SECTIONS WERE RENORMALIZED TO THE 9437 1451 52 VALUE OF 748.0 BARNS AT 0.0253 EV IN AGREEMENT WITH THE PROPOSED 9437 1451 53 STANDARD VALUE AND WITH THE UP-DATED ABSOLUTE VALUE OF DERUYTER 9437 1451 54 (12).THE OTHER EXPERIMENTAL ABSOLUTE VALUE IS THE 1025.0+-6.0 9437 1451 55 BARNS OBTAINED BY SPENCER(10) FOR THE TOTAL CROSS-SECTION.FITTING 9437 1451 56 THE RENORMALIZED EXPERIMENTAL FISSION,CAPTURE AND ABSORPTION DATA 9437 1451 57 AND SPENCER EXPERIMENTAL TRANSMISSION OVER THE ENERGY RANGE 0.02 9437 1451 58 EV TO 7.0 EV ONE OBTAINS,AT 0.0253 EV,270.17 BARNS FOR THE CAPTURE9437 1451 59 CROSS-SECTION AND 747.19 BARNS FOR THE FISSION CROSS-SECTION ,IN 9437 1451 60 VERY GOOD AGREEMENT WITH THE PROPOSED STANDARD VALUES. 9437 1451 61 9437 1451 62 THE WESTON AND TODD 1988 FISSION DATA, OBTAINED ON A 80 M FLIGHT 9437 1451 63 PATH WITH A RESOLUTION COMPARABLE TO THE HARVEY TRANSMISSION 9437 1451 64 RESOLUTION, WERE INCLUDED IN THE SAMMY FIT EXPERIMENTAL DATA BASE.9437 1451 65 DESPITE THE DIFFICULTIES ENCOUNTERED WITH A QUITE LARGE RESIDUAL 9437 1451 66 BACKGROUND IN THE NEW WESTON DATA, THE NEW SET OF RESONANCE 9437 1451 67 PARAMETERS HAS IMPROVED COMPARED TO THE PREVIOUS ONE. DUE TO THE 9437 1451 68 IMPROVEMENT OF THE RESOLUTION IN THE FISSION EXPERIMENT MORE 9437 1451 69 RESONANCES WERE IDENTIFIED IN THE HIGH ENERGY RANGE OF THE DATA 9437 1451 70 AND THE FISSION WIDTHS ARE MORE ACCURATE. 9437 1451 71 9437 1451 72 A REICH-MOORE SAMMY FIT OF HARVEY TRANSMISSION DATA AND WESTON 9437 1451 73 1988 FISSION DATA WAS PERFORMED IN THE ENERGY RANGE 1 KEV TO 2 9437 1451 74 KEV. PRELIMINARY RESULTS ARE GIVEN IN THE FILE. A BACKGROUND 9437 1451 75 CONTRIBUTION IN FILE 3 MUST BE ADDED. FURTHER ANALYSIS SHOULD BE 9437 1451 76 PERFORMED TO OBTAIN MORE ACCURATE SET OF RESONANCE PARAMETERS IN 9437 1451 77 1 TO 2 KEV ENERGY RANGE. HOWEVER, THE PRESENT SET OF DATA SHOULD 9437 1451 78 BE MOST USEFUL FOR THE CALCULATION OF THE SELF SHIELDING IN THIS 9437 1451 79 ENERGY RANGE. 9437 1451 80 THE SCATTERING CROSS-SECTION IS 0.91 BARN LARGER THAN THE PROPOSED9437 1451 81 STANDARD.IT CORRESPONDS TO A RADIUS OF 9.46+-0.20 F OBTAINED IN 9437 1451 82 THE ANALYSIS OF THE TRANMISSION DATA UP TO 1 KEV.TO OBTAIN THE 9437 1451 83 THE PROPOSED STANDARD VALUE ONE SHOULD USE A RADIUS OF 9.11 F. 9437 1451 84 9437 1451 85 ONE SHOULD ALSO NOTE THAT THE 293 DEGREES K CROSS-SECTIONS CALCUL-9437 1451 86 ATED AT 0.0253 EV DEPEND ON THE WAY THE DOPPLER BROADENING CALCUL-9437 1451 87 ATION IS PERFORMED.FOR INSTANCE USING A GAUSSIAN BROADENING FUNC- 9437 1451 88 TION WILL GIVE A FISSION CROSS-SECTION ABOUT 2.5 BARNS LARGER THAN9437 1451 89 THE ONE OBTAINED FROM THE ACCURATE CALCULATION WHICH CONSERVES THE9437 1451 90 1/V CROSS SECTION.THIS DIFFERENCE IS OF THE SAME ORDER OF 9437 1451 91 MAGNITUDE AS THE ACCURACY ON THE PROPOSED STANDARD FISSION CROSS 9437 1451 92 SECTION.THE VALUES GIVEN IN THE ABOVE TABLE WERE OBTAINED FROM AN 9437 1451 93 ACCURATE CALCULATION METHOD (SPECIAL SAMMY OPTION FOR THE THERMAL 9437 1451 94 ENERGY RANGE). 9437 1451 95 9437 1451 96 THE FOLLOWING TABLE SHOWS THE EXPERIMENTAL CROSS-SECTIONS AVERAGED9437 1451 97 OVER TWO ENERGY INTERVALS AND COMPARED TO THE CALCULATED VALUES: 9437 1451 98 9437 1451 99 AVERAGE CROSS-SECTIONS(BARN) 9437 1451 100 **************************************************************** 9437 1451 101 REFERENCES(1-10) 0.02 TO 0.06 EV 0.02 TO 0.65 EV 9437 1451 102 **************************************************************** 9437 1451 103 EXP CALC (293K) EXP CALC (293K) 9437 1451 104 GWIN71 FISS 631.41 843.71 9437 1451 105 GWIN76 FISS 631.41 838.39 9437 1451 106 GWIN84 FISS(*) 631.41 631.45(+0.01%) 837.18 839.01(+0.22%) 9437 1451 107 DERUYTER70 FISS 631.41 859.43 9437 1451 108 WAGEMANS80 FISS 631.41 862.56 9437 1451 109 GWIN71 CAPTURE 243.84 242.84(-0.41%) 524.75 518.13(-1.26%) 9437 1451 110 GWIN76 ABSORPT(*) 875.90 874.29(-0.18%) 1359.96 1357.14(-0.21%) 9437 1451 111 SPENCER84 TOT(*) 883.20 882.99(-0.02%) 1361.69 1368.8 (+0.52%) 9437 1451 112 **************************************************************** 9437 1451 113 (*)THESE DATA HAD THE LARGIEST WEIGHT IN THE THERMAL FIT.THE VA- 9437 1451 114 LUES BETWEEN THE PARENTHESES GIVE THE PERCENTAGE DEVIATION BETWEEN9437 1451 115 THE CALCULATED DATA AND THE EXPERIMENTAL DATA. 9437 1451 116 ***************************************************************** 9437 1451 117 9437 1451 118 THE VALUE OF 631.4 BARNS FOR ALL THE AVERAGE EXPERIMENTAL FISSION 9437 1451 119 CROSS-SECTIONS IN THE ENERGY RANGE 0.02 EV TO 0.06 EV CORRESPONDS 9437 1451 120 TO THE RENORMALIZATION OF ALL THE FISSION EXPERIMENTS TO 748.+-1. 9437 1451 121 BARNS AT 0.0253 EV.ORNL DATA ARE CONSISTENT WITHIN 0.8% OVER THE 9437 1451 122 0.3 EV RESONANCE.DERUYTER AND WAGEMANS DATA ARE ABOUT 2.5 % LARGER9437 1451 123 AND WERE NOT INCLUDED IN THE FIT.WHEN NORMALIZED ON THE STANDARD 9437 1451 124 VALUE AT 0.0253 EV ,GWIN76 ABSORPTION AGREES WITH THE ABSORPTION 9437 1451 125 OBTAINED FROM SPENCER TOTAL WITHIN 0.7% OVER THE RESONANCE.THE 9437 1451 126 PRESENT EVALUATION IS ESSENTIALLY THE RESULT OF A CONSISTENT ANA- 9437 1451 127 LYSIS OF ALL THE AVAILABLE ORNL DATA WITH A LARGER WEIGHT ON GWIN 9437 1451 128 1984 FISSION,GWIN 1976 ABSORPTION AND SPENCER TRANSMISSION DATA. 9437 1451 129 9437 1451 130 *************************************************** 9437 1451 131 THE RESOLVED RESONANCE REGION 9437 1451 132 *************************************************** 9437 1451 133 9437 1451 134 THE REICH-MOORE RESONANCE PARAMETERS ARE GIVEN IN THE ENERGY RAN- 9437 1451 135 GE UP TO 2 KEV. FOUR NEGATIVE ENERGY RESONANCES AND SIX FICTI- 9437 1451 136 TIOUS RESONANCES ABOVE 2 KEV ARE USED TO REPRESENT THE EFFECT OF 9437 1451 137 THE EXTERNAL REGION RESONANCES.A CONSTANT VALUE OF 9.41 F COULD 9437 1451 138 BE USED FOR THE SCATTERING RADIUS R' FOR THE CALCULATION OF THE 9437 1451 139 CROSS-SECTION IN THE ENTIRE ENERGY RANGE ANALYSED. 9437 1451 140 THE SET OF RESONANCE PARAMETERS IS ESSENTIALLY THE RESULT OF A BA-9437 1451 141 YESIAN REICH-MOORE ANALYSIS(SAMMY)(13) OF THE HARVEY TRANSMIS- 9437 1451 142 SION DATA(9) AND GWIN AND WESTON 1984 FISSION DATA(7,8).THE HARVEY9437 1451 143 DATA WERE TAKEN ON A 80 M FLIGHT PATH AT LIQUID NITROGEN TEMPERA- 9437 1451 144 TURE WITH A RESOLUTION GOOD ENOUGH TO SEPARATE MORE THAN 80 % OF 9437 1451 145 THE RESONANCE UP TO 1 KEV. BLONS 1973 DATA(3), WHICH HAVE BETTER 9437 1451 146 RESOLUTION THAN 84 WESTON DATA, WERE USED TO IDENTIFY NARROW FIS- 9437 1451 147 SION RESONANCES IN THE HIGH ENERGY REGION.A PRELIMINARY CORRELATED9437 1451 148 FIT ON HARVEY ,WESTON AND BLONS DATA,ALLOWING THE ADJUSTMENT OF 9437 1451 149 THE NORMALIZATION COEFFICIENTS AND OF THE BACKGROUND CORRECTIONS, 9437 1451 150 HAS SHOWN THAT NO SUCH ADJUSTMENT WAS NECESSARY TO HAVE CONSIS- 9437 1451 151 TANCY BETWEEN HARVEY DATA AND WESTON DATA.BLONS NEEDED A LARGE RE-9437 1451 152 ADJUSTMENT OF THE BACKGROUND AND OF THE NORMALISATION.THE FINAL 9437 1451 153 FIT WAS PERFORMED BY USING ONLY HARVEY,GWIN 1984(E < 30 EV) AND 9437 1451 154 WESTON 1984(E > 10 EV) AS INPUT TO SAMMY. 9437 1451 155 IF ONE COMPARES THE MOST RECENT EXPERIMENTAL FISSION DATA,IN THE 9437 1451 156 RESOLVED RESONANCE REGION ABOVE 100 EV,ONE FINDS THAT WESTON DATA 9437 1451 157 ARE THE SMALLEST AND BLONS DATA THE LARGEST.THE FOLLOWING TABLE 9437 1451 158 SHOWS THE PROPOSED ENDF/B-6 STANDARD VALUES(11) AND THE VALUES 9437 1451 159 OBTTAINED FROM THE RESONANCE PARAMETERS ,AVERAGED IN THE SAME 9437 1451 160 ENERGY INTERVALS: 9437 1451 161 9437 1451 162 CROSS-SECTIONS IN BARN 9437 1451 163 *********************************************************** 9437 1451 164 ENERGY CALCUL WESTON84 BLONS73 PROPOSED 9437 1451 165 (KEV) (293K) STANDARD 9437 1451 166 *********************************************************** 9437 1451 167 0.1-0.2 18.159 18.095(-3.1%) 18.93(+1.4%) 18.66+-0.13 9437 1451 168 0.2-0.3 17.318 17.441(-2.7%) 17.79(-0.5%) 17.88+-0.12 9437 1451 169 0.3-0.4 8.116 8.130(-3.6%) 8.91(+5.7%) 8.43+-0.06 9437 1451 170 0.4-0.5 9.318 9.337(-2.5%) 9.71(+1.5%) 9.57+-0.07 9437 1451 171 0.5-0.6 14.944 15.170(-2.6%) 15.51(-0.3%) 15.56+-0.11 9437 1451 172 0.6-0.7 4.337 4.192(-6.4%) 4.63(+3.8%) 4.46+-0.04 9437 1451 173 0.7-0.8 5.340 5.385(-4.5%) 5.94(+5.5%) 5.63+-0.04 9437 1451 174 0.8-0.9 4.636 4.765(-4.5%) 5.11(+2.6%) 4.98+-0.04 9437 1451 175 0.9-1.0 8.050 8.165(-1.7%) 8.57(+3.3%) 8.30+-0.07 9437 1451 176 *********************************************************** 9437 1451 177 0.1-1.0 10.024 10.075(-3.1%) 10.57(+1.7%) 10.39 9437 1451 178 *********************************************************** 9437 1451 179 9437 1451 180 THE VALUES BETWEEN PARENTHESES ARE THE PERCENTAGE DEVIATION FROM 9437 1451 181 THE STANDARD DATA.THE CALCULATED CROSS-SECTIONS AGREE WITHIN 0.5% 9437 1451 182 ON AVERAGE WITH WESTON VALUES AND ARE ABOUT 3.5% SMALLER THAN THE 9437 1451 183 PROPOSED STANDARD VALUES.THEY ARE ONLY 1.8% SMALLER THAN ENDF/B-5 9437 1451 184 OVER THE ENERGY RANGE 0.1 TO 1.0 KEV.THE AUTHORS OF THE PRESENT 9437 1451 185 EVALUATION HAVE THE FEELING THAT MANY OF THE EXPERIMENTAL FISSION 9437 1451 186 CROSS-SECTIONS SUFFER OF AN UNDERESTIMATION OF THE EXPERIMENTAL 9437 1451 187 BACKGROUND LEADING TO SYSTEMATIC OVERESTIMATE OF THE CROSS-SECTION9437 1451 188 THE AUTHORS OF ENDF/B-6 STANDARD EVALUATION HAVE PROBABLY NOT CON-9437 1451 189 SIDERED THE BACKGROUND PROBLEMS IN THIS WAY.THE VERY SMALL ERRORS 9437 1451 190 THEY OBTAINED ARE DUE TO A STATISTICAL PROCESSING OF A LARGE A- 9437 1451 191 MOUNT OF EXPERIMENTAL DATA.BLONS FISSION CROSS-SECTIONS ARE AN EX-9437 1451 192 AMPLE OF DATA IN WHICH THE UNDERESTIMATION OF THE EXPERIMENTAL 9437 1451 193 BACKGROUND COULD BE VERY IMPORTANT(ABOUT 3 BARNS AT 40 EV AND 0.4 9437 1451 194 BARN AT 1 KEV). 9437 1451 195 9437 1451 196 GWIN 1971 AND GWIN 1976 ABSORPTION DATA(1,4) WERE NOT INCLUDED IN 9437 1451 197 THE R-MATRIX ANALYSIS OF THE RESOLVED RESONANCE REGION.ACCURATE 9437 1451 198 ABSORPTION CROSS-SECTIONS SHOULD BE CALCULATED BY THE PARAMETERS 9437 1451 199 OBTAINED FROM THE ANALYSIS OF THE TRANSMISSION DATA.THE FOLLOWING 9437 1451 200 TABLE SHOWS AVERAGE CALCULATED VALUES OF THE CAPTURE,ABSORPTION 9437 1451 201 AND ALPHA COMPARED TO GWIN 1976 DATA: 9437 1451 202 9437 1451 203 CROSS-SECTIONS IN BARN 9437 1451 204 ******************************************************* 9437 1451 205 ENERGY(EV) CALC. VALUES (293K) GWIN 1976 DATA 9437 1451 206 ******************************************************* 9437 1451 207 CAPT ABSORP ALPHA ABSORP ALPHA 9437 1451 208 7.3- 16.0 79.08 194.99 0.68 208.00 0.74(*) 9437 1451 209 16.0- 37.5 21.12 44.83 0.89 46.50 0.89(*) 9437 1451 210 37.5- 50.0 47.90 68.18 2.36 83.15 2.96(*) 9437 1451 211 50.0-100.0 34.56 91.59 0.61 92.84 0.63 9437 1451 212 100.0-200.0 15.78 33.94 0.87 33.66 0.87 9437 1451 213 200.0-300.0 16.01 33.33 0.92 34.69 0.94 9437 1451 214 300.0-400.0 9.83 17.94 1.21 18.31 1.16 9437 1451 215 400.0-500.0 4.07 13.39 0.44 13.56 0.44 9437 1451 216 500.0-600.0 11.12 26.06 0.74 26.54 0.72 9437 1451 217 600.0-700.0 6.51 10.84 1.50 11.57 1.54 9437 1451 218 700.0-800.0 5.01 10.35 0.94 10.52 0.97 9437 1451 219 800.0-900.0 3.78 8.42 0.82 9.30 0.82 9437 1451 220 900.0-999.9 5.39 13.44 0.67 13.23 0.70 9437 1451 221 ****************************************************** 9437 1451 222 (*) GWIN 1971 DATA 9437 1451 223 9437 1451 224 BELOW 500 EV, THE CALCULATED ABSORPTION VALUES AGREE QUITE WELL 9437 1451 225 WITH GWIN 1976 EXPERIMENTAL DATA AND ARE ON AVERAGE 2.9% LOWER 9437 1451 226 ABOVE 500 EV .THE CALCULATED ALPHA VALUES ARE ON AVERAGE 1.7% 9437 1451 227 LOWER ABOVE 500 EV. 9437 1451 228 9437 1451 229 9437 1451 230 UNRESOLVED RESONANCE REGION******************* 9437 1451 231 9437 1451 232 THE AVERAGE RESONANCE PRAMETERS ARE GIVEN IN THE ENERGY RANGE 9437 1451 233 2 KEV TO 30 KEV FOR 74 ENERGY POINTS.THEY WERE OBTAINED BY U- 9437 1451 234 -SING THE CADARACHE STATISTICAL CODE FISINGA TO FIT THE GROSS 9437 1451 235 STRUCTURE OF THE SACLAY EXPERIMENTAL TOTAL CROSS SECTIONS(17) 9437 1451 236 BELOW 4 KEV AND OF SELECTED EXPERIMENTAL FISSION CROSS SECTIONS9437 1451 237 NORMALIZED TO ENDF/B-VI STANDARD EVALUATION(11).ABOVE 4 KEV NO 9437 1451 238 HIGH RESOLUTION TOTAL CROSS SECTION DATA ARE AVAILABLE;AVERAGE 9437 1451 239 TOTAL CROSS SECTIONS WERE CALCULATED TO BE CONSISTENT WITH THE 9437 1451 240 STASTISTICAL PARAMATERS OBTAINED IN THE RESOLVED RESONANCE RE- 9437 1451 241 GION(14) AND WITH THE OPTICAL MODEL PARAMETERS OF LAGRANGE AND 9437 1451 242 MADLAND(15) OBTAINED BY FITTING THE EXPERIMENTAL DATA IN THE 9437 1451 243 HIGH ENERGY RANGE.A VALUE OF 9.46 F WAS USED FOR THE EFFECTIVE 9437 1451 244 RADIUS. THE VALUES OBTAINED FOR ALPHA ARE CONSISTENT WITH THE 9437 1451 245 EXPERIMENTAL DATA. 9437 1451 246 THE COMPETITIVE WIDTH IS NOT USED FOR THE INELASTIC SCATTE- 9437 1451 247 RING CROSS SECTION .IN EACH ENERGY POINT OF THE UNRESOLVED RE- 9437 1451 248 GION,THE NEUTRON WIDTH CORRESPONDS ONLY TO THE ELASTIC SCATTE- 9437 1451 249 RING CROSS SECTION . THE INELASTIC SCATTERING CROSS SECTION 9437 1451 250 SHOULD BE FOUND IN THE FILE 3 . 9437 1451 251 THE CROSS SECTIONS OBTAINED AT 0K BY PROCESSING THE EVALUATED9437 1451 252 FILE BY NJOY-87.1, ARE GIVEN IN THE FOLLOWING TABLE,'FISS' FOR 9437 1451 253 THE FISSION VALUES AND 'CAPT' FOR THE CAPTURE VALUES: 9437 1451 254 9437 1451 255 ENERGY(KEV) CROSS SECTIONS ENERGY(KEV) CROSS SECTIONS 9437 1451 256 (BARN) (BARN) 9437 1451 257 FISS CAPT FISS CAPT 9437 1451 258 2.050 1.879 3.137 11.750 1.812 1.042 9437 1451 259 2.150 3.119 3.315 12.250 1.900 0.968 9437 1451 260 2.250 2.691 2.800 12.750 1.864 0.947 9437 1451 261 2.350 3.436 3.331 13.250 1.858 0.917 9437 1451 262 2.450 4.280 2.456 13.750 1.715 0.942 9437 1451 263 2.550 2.725 2.754 14.250 1.492 0.948 9437 1451 264 2.650 3.103 3.425 14.750 1.797 0.854 9437 1451 265 2.750 4.169 2.010 15.250 1.883 0.797 9437 1451 266 2.850 4.126 2.077 15.750 1.697 0.843 9437 1451 267 2.950 3.362 3.710 16.250 1.801 0.782 9437 1451 268 3.050 3.017 1.998 16.750 1.628 0.824 9437 1451 269 3.150 4.896 1.934 17.250 1.498 0.819 9437 1451 270 3.250 3.954 2.277 17.750 1.862 0.701 9437 1451 271 3.350 1.710 2.166 18.250 1.711 0.736 9437 1451 272 3.450 2.198 2.572 18.750 1.632 0.748 9437 1451 273 3.550 2.214 1.885 19.250 1.738 0.694 9437 1451 274 3.650 2.394 2.948 19.750 1.743 0.677 9437 1451 275 3.750 3.067 1.624 20.500 1.672 0.679 9437 1451 276 3.850 3.556 2.122 21.500 1.646 0.661 9437 1451 277 3.950 2.931 2.397 22.500 1.472 0.697 9437 1451 278 4.125 2.114 2.270 23.500 1.632 0.619 9437 1451 279 4.375 2.509 2.129 24.500 1.636 0.597 9437 1451 280 4.625 2.772 1.715 25.500 1.547 0.607 9437 1451 281 4.875 1.980 2.186 26.500 1.628 0.562 9437 1451 282 5.125 2.406 1.916 27.500 1.544 0.572 9437 1451 283 5.375 2.153 1.953 28.500 1.568 0.549 9437 1451 284 5.625 2.294 1.807 29.500 1.609 0.521 9437 1451 285 9437 1451 286 AVERAGE VALUES OF THE FISSION CROSS SECTIONS COMPARED TO THE 9437 1451 287 ENDF/B-VI STANDARD EVALUATION(11) AND ALPHA VALUES COMPARED TO 9437 1451 288 SOME EXPERIMENTAL DATA ARE GIVEN IN THE FOLLOWING TABLE: 9437 1451 289 9437 1451 290 9437 1451 291 ENERGY CROSS SECTIONS(BARN) ALPHA 9437 1451 292 (KEV) (1) (2) (3) (4) (5) (6) (7) (8) 9437 1451 293 2- 3 3.284 3.304 2.894 3.31 0.881 1.000 1.108 1.028 9437 1451 294 3- 4 2.992 3.000 2.213 2.20 0.740 0.720 0.895 0.820 9437 1451 295 4- 5 2.394 2.383 2.073 2.07 0.866 0.870 0.821 0.837 9437 1451 296 5- 6 2.266 2.301 1.863 1.91 0.822 0.820 0.867 0.834 9437 1451 297 6- 7 2.006 2.008 1.677 1.63 0.836 0.790 0.816 0.793 9437 1451 298 7- 8 2.134 2.054 1.409 1.34 0.660 0.640 0.630 0.605 9437 1451 299 8- 9 2.207 2.216 1.245 1.23 0.564 0.540 0.575 0.530 9437 1451 300 9-10 1.867 1.864 1.136 1.05 0.608 0.550 0.617 0.569 9437 1451 301 9437 1451 302 1-10 2.628 2.622 2.014 2.06 0.767 0.752 0.806 0.768 9437 1451 303 10-20 1.762 1.764 0.876 0.85 0.497 0.480 0.466 0.498 9437 1451 304 20-30 1.597 1.595 0.606 0.58 0.379 0.350 0.373 0.388 9437 1451 305 9437 1451 306 (1) FISSION CROSS SECTION , PRESENT EVALUATION (0K) 9437 1451 307 (2) FISSION CROSS SECTION , ENDF/B-VI STANDARD(11) 9437 1451 308 (3) CAPTURE CROSS SECTION, PRESENT EVALUATION (293K) 9437 1451 309 (4) CAPTURE CROSS SECTION, GWIN ET AL.1976(4) 9437 1451 310 (5) ALPHA VALUE , PRESENT EVALUATION (293K) 9437 1451 311 (6) ALPHA VALUE FROM GWIN ET AL.1976(4) 9437 1451 312 (7) ALPHA VALUE FROM SOWERBY-KONSHIN EVALUATION 1971(16) 9437 1451 313 (8) AVERAGE ALPHA VALUE FROM EXPERIMENTAL DATA 9437 1451 314 9437 1451 315 ********************************************** 9437 1451 316 9437 1451 317 9437 1451 318 THE FISSION AND CAPTURE RESONANCE INTEGRALS AT 0K ARE COMPARED TO 9437 1451 319 ENDF/B-5 DATA IN THE FOLLOWING TABLE: 9437 1451 320 9437 1451 321 ********************************************************* 9437 1451 322 ENERGY RANGE(EV) FISSION(BARN) CAPTURE(BARN) 9437 1451 323 ********************************************************* 9437 1451 324 ENDF/B-5 PRESENT ENDF/B-5 PRESENT 9437 1451 325 0.5 - 5.0 86.02 85.71 32.31 28.65 9437 1451 326 5.0 - 10.0 26.03 25.08 20.14 19.06 9437 1451 327 10.0 - 50.0 100.25 96.87 78.66 77.19 9437 1451 328 50.0 - 100.0 40.32 40.47 27.23 25.93 9437 1451 329 100.0 - 301.0 19.98 19.68 19.52 17.95 9437 1451 330 301.0 -1000.0 10.15 10.05 8.54 8.35 9437 1451 331 ******************************************************** 9437 1451 332 0.5 -1000.0 282.76 277.85 186.30 177.13 9437 1451 333 ******************************************************** 9437 1451 334 9437 1451 335 BY ADDING THE ENDF/B-5 VALUE ABOVE 1 KEV ON OBTAIN FROM THE PRE- 9437 1451 336 SENT EVALUATION: 9437 1451 337 RI FISSION 297.22 BARNS 9437 1451 338 RI CAPTURE 184.93 BARNS 9437 1451 339 THE CORRESPONDING VALUES FROM ENDF/B-5 EVALUATION ARE: 9437 1451 340 RI FISSION 302.13 BARNS 9437 1451 341 RI CAPTURE 194.10 BARNS 9437 1451 342 9437 1451 343 ************************************************* 9437 1451 344 REFERENCES 9437 1451 345 ************************************************* 9437 1451 346 9437 1451 347 1-R.GWIN ET AL.,NUCL.SCI.ENG.,45,25(1971) 9437 1451 348 2-A.J.DERUYTER ET AL.,J.NCL.ENER.,26,293(1972) 9437 1451 349 3-J.BLONS, NUCL.SCI.ENG.,51,130(1973) 9437 1451 350 4-R.GWIN ET AL.,NUCL.SCI.ENG.,59,79(1976) 9437 1451 351 5-R.GWIN ET AL.,NUCL.SCI.ENG.,61,116(1976) 9437 1451 352 6-W.WAGEMANS,ANNALS OF NUCL.ENER.7,9,495(1980) 9437 1451 353 7-R.GWIN ET AL.,NUCL.SCI.ENG.,88,37(1984) 9437 1451 354 8-L.W.WESTON ET AL.,NUCL.SCI.ENG.88,567(1984) 9437 1451 355 9-J.A.HARVEY ,PRIVATE COMMUNICATION(1985) 9437 1451 356 10-R.R.SPENCER ET AL.,NUCL.SCI.ENG.,96,318(1987) 9437 1451 357 11-A.CARLSON ET AL.,PRELIMINARY RESULTS OF THE ENDF/B-6 STANDARD 9437 1451 358 EVALUATION(SEPT 8 1987) 9437 1451 359 12-A.J.DERUYTER,J.NUCL.ENER.,26,293(1972) 9437 1451 360 13-N.M.LARSON ET AL.,ORNL/TM-7485,ORNL/TM-9179,ORNL/TM-9719/R1 9437 1451 361 14-H.DERRIEN AND G. DE SAUSSURE,ORNL-TM-10986(1988) 9437 1451 362 15-CH.LAGRANGE AND D.G.MADLAND,PHYS.REV.C,33,5(1986) 9437 1451 363 16-M.G.SOWERBY ET AL.,AT.ENERGY REV.,10,4,453,IAEA,VIENNA(1972) 9437 1451 364 17-H.DERRIEN,THESIS,ORSAY SERIE A,1172(1973) 9437 1451 365 9437 1451 366 *****************************************************************9437 1451 367 ENERGY REGION 0.03 TO 20 MEV 9437 1451 368 *****************************************************************9437 1451 369 9437 1451 370 Principal LANL evaluators: P.G.Young, R.E.MacFarlane, E.D.Arthur 9437 1451 371 9437 1451 372 The evaluation above 10 keV is based on a detailed theoretical 9437 1451 373 analysis utilizing the available experimental data. Coupled 9437 1451 374 channel optical model calculations with the ECIS code (Ra70) 9437 1451 375 were used to provide the total, elastic, and inelastic cross 9437 1451 376 sections to the first 7 members of the ground state rotational 9437 1451 377 band, as well as neutron elastic and inelastic angular distri- 9437 1451 378 butions to the rotational levels. The ECIS code was also 9437 1451 379 used to calculate neutron transmission coefficients. Hauser- 9437 1451 380 Feshbach statistical theory calculations were carried out with 9437 1451 381 the GNASH (Ar88, Yo77) and COMNUC (Du70) code systems, including 9437 1451 382 preequilibrium and fission. DWBA calculations were performed 9437 1451 383 with the DWUCK code (Ku70) for several vibrational levels, using 9437 1451 384 B(El) values inferred from (d,d') data on Pu238 and Pu240, as 9437 1451 385 well as Coulomb excitation measurements. A weak coupling 9437 1451 386 model (Pe69) was used to apply the Pu238 and Pu240 results to 9437 1451 387 states in Pu239. 9437 1451 388 This analysis is an extension of the calculations used for the 9437 1451 389 ENDF/B-V.2 evaluation, described in reference Ar82. 9437 1451 390 9437 1451 391 **********MF=1 Descriptive and Nubar Information*****************9437 1451 392 9437 1451 393 MT=452 Total Nubar. Sum of MT=455 and 456. 9437 1451 394 MT=455 Delayed Neutron Yields. England (En89). 9437 1451 395 MT=456 Prompt Neutron Yields. From 10-5 to 650 eV, based on the 9437 1451 396 evaluation of Fort (Fo88), after minor renormalization 9437 1451 397 for consistency with CSEWG standards. Fort's values 9437 1451 398 were multiplied by 1.000411 from 10-5 to 10 eV to 9437 1451 399 fall within 1/2 std. deviation of Pu239 thermal nubar 9437 1451 400 value, with the factor varying linearly to 1.00282 at 9437 1451 401 500 eV for consistency with Cf252 nubar. Fort's eval. 9437 1451 402 (renomalized) was used intact below 62.3 eV but higher 9437 1451 403 energy data were thinned with a thinning criterion of 9437 1451 404 0.036%, a factor of 5 less than the standard deviation 9437 1451 405 of the CSEWG standard value of Pu239 thermal nubar. 9437 1451 406 Above 650 eV, the evaluation is based on a covariance 9437 1451 407 analysis of all available exp. data from the CSISRS 9437 1451 408 data file at the NNDC, BNL, using the GLUCS analysis 9437 1451 409 code (He80). A smooth curve was passed through the 9437 1451 410 GLUCS results with structure removed. All the exp. 9437 1451 411 data were renormalized to ENDF/B-VI standards 9437 1451 412 prior to the covariance analysis. 9437 1451 413 9437 1451 414 **********MF=3 Smooth Cross Sections*****************************9437 1451 415 9437 1451 416 MT=1 Neutron Total Cross Section. 0.03 to 20 MeV, based on 9437 1451 417 coupled-channel optical calculations and the exp. data 9437 1451 418 of Po81, Sh78, Po83, Sc74, Fo71, Sm73, Na73, Pe60, Ca73,9437 1451 419 Li90. The exp. and theoretical results were combined 9437 1451 420 through a covariance analysis with the GLUCS code 9437 1451 421 system (He80). The covariance analysis results, which 9437 1451 422 agreed well with Derrien's (De89) unresolved resonance 9437 1451 423 analysis at 30 keV, were smoothly joined to those 9437 1451 424 results between 30 and 50 keV. 9437 1451 425 MT=2 0.030 to 20 MeV, based on subtraction of MT=4,16,17,18,37,9437 1451 426 and 102 from MT=1. 9437 1451 427 MT=4 Sum of MT=51-91 9437 1451 428 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 429 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 430 MT=18 The Pu239(n,f) cross section that resulted from the 9437 1451 431 simultaneous standards analysis for Vers.6 was 9437 1451 432 renormalized by a factor of 1.007 and used directly 9437 1451 433 from 50 keV to 1.25 MeV. Between 30 and 50 keV, the 9437 1451 434 evaluation was matched smoothly to the Derrien 9437 1451 435 unresolved resonance parameter analysis, requiring a 9437 1451 436 a reduction of about 4% near 30 keV. Above 1.25 MeV, 9437 1451 437 the evaluation is based on a new covariance analysis 9437 1451 438 of all available ratio and absolute Pu239(n,f) data, 9437 1451 439 including new measurements (Li88,Me88) that were not 9437 1451 440 available for the simul. standards analysis. The GLUCS 9437 1451 441 code was used for the covariance analysis, and a 9437 1451 442 smooth curve was drawn through the analysis results. 9437 1451 443 The new covariance analysis agrees reasonably with the 9437 1451 444 simultaneous standards results (both cross section and 9437 1451 445 standard deviation) at energies below the newer exp. 9437 1451 446 results. At higher energies, the effect of the new data 9437 1451 447 on the analysis is to raise the (n,f) cross section 9437 1451 448 somewhat at higher energies, particularly near 9 MeV 9437 1451 449 (about 4%) and above 15 MeV (few %). 9437 1451 450 MT=19 (n,f) first-chance fission cross section. 9437 1451 451 Ratio of first-chance to total fission obtained from 9437 1451 452 GNASH calculations. 9437 1451 453 MT=20 (n,nf) second-chance fission cross section. 9437 1451 454 Ratio of second-chance to total fission obtained from 9437 1451 455 GNASH calculations. 9437 1451 456 MT=21 (n,2nf) third-chance fission cross section. 9437 1451 457 Ratio of third-chance to total fission obtained from 9437 1451 458 GNASH calculations. 9437 1451 459 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 460 MT=38 (n,3nf) fourth-chance fission cross section. 9437 1451 461 Ratio of fourth-chance to total fission obtained from 9437 1451 462 GNASH calculations. 9437 1451 463 MT=51-55,57 Thres. to 20 MeV, coupled-channel optical model 9437 1451 464 calculations (3/2+ to 13/2+ members of the K=1/2 ground 9437 1451 465 state rotational band) using the ECIS code. Compound 9437 1451 466 nucleus contributions, obtained from COMNUC calcula- 9437 1451 467 tions, are also included. 9437 1451 468 MT=56,58-69,71,72,74-77 Threshold to 6.0 MeV, Compound 9437 1451 469 nucleus reaction theory calculation with width fluctua- 9437 1451 470 tions, using the COMNUC code. 9437 1451 471 MT=70,73,78-81 Threshold to 20 MeV, distorted wave Born 9437 1451 472 approximation calculations with the DWUCK code for 9437 1451 473 l=2 and l=3 vibrational states. The l=2 states are 9437 1451 474 MT=78,80 and the l=3 states are MT=70,73,79, and 81. 9437 1451 475 Compound nucleus contributions were included in 9437 1451 476 the data for MT = 70 and 73. 9437 1451 477 MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 478 Note that the MT=78-81 vibrational states lie in the 9437 1451 479 MT=91 continuum region. 9437 1451 480 MT=102 0.030-20 MeV, obtained using the values of alpha (ratio 9437 1451 481 of capture to fission cross sections) from ENDF/B-V.2, 9437 1451 482 together with MT=18 from the present evaluation. Between9437 1451 483 40 and 100 MeV, structure in the cross section was 9437 1451 484 smoothed out. Below 50 keV, the results were smoothly 9437 1451 485 joined to the unresolved res. result at 30 keV, 9437 1451 486 requiring an increase of 0.2% at 30 keV. 9437 1451 487 9437 1451 488 **********MF=4 Neutron Angular Distributions******************** 9437 1451 489 9437 1451 490 MT=2 Elastic scattering angular distribution based on ECIS 9437 1451 491 coupled-channel calculations, with a compound elastic 9437 1451 492 component from COMNUC included below 6 MeV. 9437 1451 493 MT=51-55,57 Thres. to 20 MeV, Coupled-channel optical model 9437 1451 494 calculations plus compound-nucleus contributions. 9437 1451 495 MT=56,58-69,71,72,74-77 Threshold to 6.0 MeV, Compound 9437 1451 496 nucleus reaction theory calculation with width fluctua- 9437 1451 497 tions, using the COMNUC code. 9437 1451 498 MT=70,73,78-81 Thres. to 20 MeV, Distorted wave Born approx. 9437 1451 499 imation calculations with DWUCK code. Compound-nucleus 9437 1451 500 contributions are included for MT=70 and 73. 9437 1451 501 9437 1451 502 ************MF=5 Neutron Energy Distributions******************* 9437 1451 503 9437 1451 504 MF=18 Composite neutron energy distributions from fission. 9437 1451 505 Based on calculations by D.Madland (Ar84) using Madland-9437 1451 506 Nix formalism. The calculations include the first-, 9437 1451 507 second-, and third-chance fission neutron components. 9437 1451 508 These data are the same as were used for Revision 2 of 9437 1451 509 ENDF/B-V. Parameters for the calculation were adjusted 9437 1451 510 to give the same average fission neutron energy at 9437 1451 511 thermal as ENDF/B-V.0. Tabulated data (LF=1) used. 9437 1451 512 MT=455 Tal England (En89). 9437 1451 513 9437 1451 514 ************MF=6 Correlated Energy-Angle Distributions********** 9437 1451 515 9437 1451 516 MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 517 Updated Kalbach-Mann systematics used for specifying 9437 1451 518 neutron distributions (Ka87). Only neutrons given. 9437 1451 519 MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 520 Updated Kalbach-Mann systematics used for specifying 9437 1451 521 neutron distributions (Ka87). Only neutrons given. 9437 1451 522 MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 523 Updated Kalbach-Mann systematics used for specifying 9437 1451 524 neutron distributions (Ka87). Only neutrons given. 9437 1451 525 MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc. 9437 1451 526 Updated Kalbach-Mann systematics used for specifying 9437 1451 527 neutron distributions (Ka87). Only neutrons given. 9437 1451 528 9437 1451 529 ************MF=12,13,14,15 Photon-Production Data*************** 9437 1451 530 9437 1451 531 All photon-production data were carried over from ENDF/B-V.2, 9437 1451 532 MAT=1399. Data are given for MF=12, MT=4,18,102; MF=13, MT=3; 9437 1451 533 MF=14, MT=3,4,18,102; and, MF=15, MT=3,4,18,102. 9437 1451 534 9437 1451 535 ******************References************************************ 9437 1451 536 9437 1451 537 Ar84 E.Arthur et al., Nuc.Sci.Eng.88,56(1984). 9437 1451 538 Ar88 E.D.Arthur, LA-UR-88-382 (1988). 9437 1451 539 Ca73 J.Cabe et al., CEA-R-4524 (1973). 9437 1451 540 Ca85 A.Carlson et al., Nuc.Data for Basic & Applied Science, 9437 1451 541 Santa Fe, NM (1985) p.1429. 9437 1451 542 De89 H.Derrien & G.de Saussure, ORNL-10986 (1989). 9437 1451 543 Du70 C.L.Dunford, AI-AEC-12931 (1970). 9437 1451 544 En89 T.R.England et al,LA11151-MS(1988),LA-11534-T(1989); 9437 1451 545 LAUR-88-4118 to be published in NSE(1989). 9437 1451 546 Fo71 D.Foster & D.Glasgow, Phys.Rev.C3,576(1971). 9437 1451 547 Fo88 E.Fort et al., Nuc.Sci.Eng.99,375(1988). 9437 1451 548 Fr86 J.Frehaut, NEANDC(E) 238/L (1986). 9437 1451 549 He80 D.Hetrick & C.Y.Fu, ORNL/TM-7341 (1980). 9437 1451 550 Ka87 C.Kalbach, LA-UR-87-4139 (1987) to be pub.in Phys.Rev.C!. 9437 1451 551 Ku70 P.D.Kunz, DWUCK: A Distorted-Wave Born Approximation 9437 1451 552 Program, unpublished report. 9437 1451 553 Li88 P.Lisowski et al., Nuc.Data for Sci.& Tech.,Mito Conf.,p97.9437 1451 554 Li90 P.Lisowski, Pers.Communication of WNR data taken in 1985. 9437 1451 555 Me88 J.Meadows, Ann.Nucl.Energy 15, 421 (1988). 9437 1451 556 Na73 K.Nadolny et al., USNDC-9 (1973)p.170 9437 1451 557 Pe60 J.Peterson et al., Phys.Rev.120, 521(1960). 9437 1451 558 Pe69 R.J.Peterson, Ann.Phys. 53, 40 (1069). 9437 1451 559 Po81 W.Poenitz et al., Nuc.Sci.Eng.78, 333(1981). 9437 1451 560 Po83 W.Poenitz et al., ANL-NDM-80, 1983. 9437 1451 561 Ra70 J.Raynal,IAEA SMR-9/8 (1970). 9437 1451 562 Sc74 R.Schwartz et al., Nuc.Sci.Engr.54,322(1974). 9437 1451 563 Sh78 R.Shamu et al., personal communication, 1978. 9437 1451 564 Sm73 A.Smith et al., J.Nuc.En.27,317(1973). 9437 1451 565 Yo77 P.G.Young & E.D.Arthur, LA-6947 (1977). 9437 1451 566 Yo88 P.G.Young & E.D.Arthur, Nuc.Data for Sci.& Tech., Mito 9437 1451 567 Conference (1988) p.603. 9437 1451 568 9437 1451 569 **************************************************************** 9437 1451 570 1 451 573 19437 1451 571 2 151 1179 19437 1451 572 3 18 104 19437 1451 573 9437 1 0 574 9437 0 0 575 9.423900+4 2.369990+2 0 0 1 09437 2151 576 9.423900+4 1.000000+0 0 1 3 09437 2151 577 1.000000-5 1.000000+3 1 3 0 09437 2151 578 5.000000-1 9.410000-1 0 0 1 19437 2151 579 2.369990+2 0.000000+0 0 0 2418 4039437 2151 580 -1.500200+2 1.000000+0 4.289400-1 4.451600-2 1.917200-1 0.000000+09437 2151 581 -1.546700+1 1.000000+0 1.354600-4 2.685500-2-2.552800-6 0.000000+09437 2151 582 -6.908700+0 0.000000+0 1.242500-2 2.600200-1-9.416900-1 2.962300-19437 2151 583 -2.194400-1 0.000000+0 3.065100-5 2.590600-3-1.614300-3-5.824600-19437 2151 584 2.956243-1 1.000000+0 7.992700-5 3.930300-2 5.737700-2 0.000000+09437 2151 585 7.815800+0 1.000000+0 7.920400-4 3.775400-2-4.474800-2 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