MODIFIED version of "CEA JEF-2.x" (July 2005) 1 0 0 9.423900+4 2.369990+2 1 1 2 19437 1451 1 0.000000+0 0.000000+0 0 0 0 69437 1451 2 1.000000+0 0.000000+0 0 0 10 229437 1451 3 0.000000+0 0.000000+0 0 0 564 839437 1451 4 94-PU-239 CAD/ORNL EVAL-SEP90 FORT, DERRIEN AND DESAUSSURE 9437 1451 5 DIST-JAN92 REV2-JUL05 050707 9437 1451 6 ----JEF-2.x MATERIAL 9437 REVISION 2 9437 1451 7 -----INCIDENT NEUTRON DATA 9437 1451 8 ------ENDF-6 FORMAT 9437 1451 9 ******************************************************************9437 1451 10 ******************************************************************9437 1451 11 ** **9437 1451 12 ** The "CEA JEF-2.x" evaluation is described in JEF/DOC-807 **9437 1451 13 ** **9437 1451 14 ** Further modification(s) of the "CEA JEF-2.x" evaluation: **9437 1451 15 ** **9437 1451 16 ** 2005-07 E. Dupont (CEA) and Ch. Dean (Serco Group) **9437 1451 17 ** Unresolved Resonance Parameters (MF2,MT151,LRU=2) **9437 1451 18 ** Extension of average parameters up to EH = 30 keV **9437 1451 19 ** **9437 1451 20 ******************************************************************9437 1451 21 ******************************************************************9437 1451 22 * JEF-2 9437 1451 23 * NEW EVALUATION FOR JEF-2. FIRST VERSION DISTRIB. IN OCT.90 9437 1451 24 ******************************************************************9437 1451 25 * 1-NOV-90: MF=1, MT=452,455,AND 456 MODIFIED ABOVE 20 KEV. 9437 1451 26 * 3-JAN-91: ONE LEVEL SPACING CORRECTED AT 1.825 KEV (LINE 9437 1451 27 * NUMBER 3767) 9437 1451 28 * AMUN CHANGED TO 2.0 FOR L=1, J=1 UNRESOLVED 9437 1451 29 * RESONANCE PARAMETERS. 9437 1451 30 * 17-APR-91: ISOTROPIC ANGULAR DISTRIBUTIONS ADDED FOR 9437 1451 31 * MT=19,20,21 9437 1451 32 * FIRST UNRESOLVED RESONANCE ENERGY MODIFIED TO 9437 1451 33 * AGREE WITH UPPER LIMIT OF RESOLVED RESONANCES 9437 1451 34 * CROSS SECTIONS ENERGIES MODIFIED ACCORDINGLY 9437 1451 35 * 10-FEB-92: NEUTRON WIDTHS FOR L=1, J=1 UNRESOLVED RESONANCE 9437 1451 36 * PARAMETERS DIVIDED BY 2, FOLLOWING THE CORRECTION9437 1451 37 * OF AMUN ON 3-JAN-91 (SEE COMMENT ABOVE). 9437 1451 38 ******************************************************************9437 1451 39 9437 1451 40 STATUS OF THE EVALUATION - SEPTEMBER 1990 9437 1451 41 THIS EVALUATION IS A NEW ONE FROM 30 KEV TO 20 MEV 9437 1451 42 IT INCLUDES THE WORK BY DERRIEN AND DE SAUSSURE /14/ (RESONANCE 9437 1451 43 PARAMETERS UP TO 1 KEV) AND BY DERRIEN IN THE UNRESOLVED RANGE 9437 1451 44 FROM 1KEV TO 30 KEV (NOTE TECHNIQUE DRP/SPRC/LEPH) 9437 1451 45 9437 1451 46 *************************************************** 9437 1451 47 THERMAL REGION 9437 1451 48 *************************************************** 9437 1451 49 9437 1451 50 THE REICH-MOORE RESONANCE PARAMETERS HAVE BEEN OBTAINED FROM THE 9437 1451 51 ANALYSIS OF SEVERAL TRANSMISSION,FISSION,ABSORPTION AND CAPTURE 9437 1451 52 EXPERIMENTAL DATA(1,4,5,7,10).THE CROSS-SECTIONS AND THE WESCOTT 9437 1451 53 FACTORS OBTAINED FROM THESE PARAMETERS AT 0.0253 EV ARE GIVEN IN 9437 1451 54 THE FOLLOWING TABLE: 9437 1451 55 9437 1451 56 CALCULATED VALUES(BARN) PROPOSED 9437 1451 57 300K 0K STANDARD VALUES(BARN)(11) 9437 1451 58 **************************************************************** 9437 1451 59 FISSION 747.19 746.84 747.99+-1.87 9437 1451 60 CAPTURE 270.17 270.00 271.43+-2.14 9437 1451 61 SCATTERING 8.85 7.88+-0.97 9437 1451 62 ABSORPTION 1017.36 1016.84 1019.42+-4.00 9437 1451 63 TOTAL 1026.21 1027.30+-5.00 9437 1451 64 **************************************************************** 9437 1451 65 GF 1.0526 1.0563+-0.0022 9437 1451 66 GA 1.0768 1.0782+-0.0024 9437 1451 67 **************************************************************** 9437 1451 68 9437 1451 69 THE EXPERIMENTAL FISSION CROSS-SECTIONS WERE RENORMALIZED TO THE 9437 1451 70 VALUE OF 748.0 BARNS AT 0.0253 EV IN AGREEMENT WITH THE PROPOSED 9437 1451 71 STANDARD VALUE AND WITH THE UP-DATED ABSOLUTE VALUE OF DERUYTER 9437 1451 72 (12).THE OTHER EXPERIMENTAL ABSOLUTE VALUE IS THE 1025.0+-6.0 9437 1451 73 BARNS OBTAINED BY SPENCER(10) FOR THE TOTAL CROSS-SECTION.FITTING 9437 1451 74 THE RENORMALIZED EXPERIMENTAL FISSION,CAPTURE AND ABSORPTION DATA 9437 1451 75 AND SPENCER EXPERIMENTAL TRANSMISSION OVER THE ENERGY RANGE 0.02 9437 1451 76 EV TO 7.0 EV ONE OBTAINS,AT 0.0253 EV,270.17 BARNS FOR THE CAPTURE9437 1451 77 CROSS-SECTION AND 747.19 BARNS FOR THE FISSION CROSS-SECTION ,IN 9437 1451 78 VERY GOOD AGREEMENT WITH THE PROPOSED STANDARD VALUES. 9437 1451 79 9437 1451 80 THE SCATTERING CROSS-SECTION IS 0.97 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 300 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 CROSSSECTION.THIS DIFFERENCE IS OF THE SAME ORDER OF MAGNITUDE9437 1451 91 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 EXP CALC 9437 1451 104 GWIN71 FISS 631.41 843.71 9437 1451 105 GWIN76 FISS 631.41 838.39 9437 1451 106 GWIN84 FISS(*) 631.41 630.54(-0.13%) 837.18 836.79(-0.05%) 9437 1451 107 DERUYTER70 FISS 631.41 859.43 9437 1451 108 WAGEMANS80 FISS 631.41 862.56 9437 1451 109 GWIN71 CAPTURE 243.84 242.83(-0.42%) 524.75 517.67(-1.37%) 9437 1451 110 GWIN76 ABSORPT(*) 875.90 873.53(-0.27%) 1359.96 1355.67(-0.32%) 9437 1451 111 SPENCER84 TOT(*) 883.20 882.40(-0.09%) 1361.69 1367.46(+0.48%) 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 1 KEV.FOUR NEGATIVE ENERGY RESONANCES AND THREE FICTI- 9437 1451 136 TIOUS RESONANCES ABOVE 1 KEV ARE USED TO REPRESENT THE EFFECT OF 9437 1451 137 THE EXTERNAL REGION RESONANCES.A CONSTANT VALUE OF 9.46 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 WESTON DATA, WERE USED TO IDENTIFY THE 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) STANDARD 9437 1451 166 *********************************************************** 9437 1451 167 0.1-0.2 18.070 18.095(-3.1%) 18.93(+1.4%) 18.66+-0.13 9437 1451 168 0.2-0.3 17.371 17.441(-2.7%) 17.79(-0.5%) 17.88+-0.12 9437 1451 169 0.3-0.4 8.083 8.130(-3.6%) 8.91(+5.7%) 8.43+-0.06 9437 1451 170 0.4-0.5 9.387 9.337(-2.5%) 9.71(+1.5%) 9.57+-0.07 9437 1451 171 0.5-0.6 15.086 15.170(-2.6%) 15.51(-0.3%) 15.56+-0.11 9437 1451 172 0.6-0.7 4.159 4.192(-6.4%) 4.63(+3.8%) 4.46+-0.04 9437 1451 173 0.7-0.8 5.316 5.385(-4.5%) 5.94(+5.5%) 5.63+-0.04 9437 1451 174 0.8-0.9 4.726 4.765(-4.5%) 5.11(+2.6%) 4.98+-0.04 9437 1451 175 0.9-1.0 8.086 8.165(-1.7%) 8.57(+3.3%) 8.30+-0.07 9437 1451 176 *********************************************************** 9437 1451 177 0.1-1.0 10.033 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 0.4% 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) CALCULATED VALUES GWIN 1976 DATA 9437 1451 206 ******************************************************* 9437 1451 207 CAPT ABSORP ALPHA ABSORP ALPHA 9437 1451 208 7.3- 16.0 79.16 195.40 0.681 208.00 0.74(*) 9437 1451 209 16.0- 37.5 21.06 44.68 0.892 46.50 0.89(*) 9437 1451 210 37.5- 50.0 49.28 69.82 2.401 83.15 2.96(*) 9437 1451 211 50.0-100.0 34.16 91.12 0.600 92.84 0.63 9437 1451 212 100.0-200.0 15.69 33.76 0.868 33.66 0.87 9437 1451 213 200.0-300.0 16.01 33.38 0.922 34.69 0.94 9437 1451 214 300.0-400.0 9.65 17.68 1.194 18.31 1.16 9437 1451 215 400.0-500.0 3.88 13.27 0.414 13.56 0.44 9437 1451 216 500.0-600.0 10.18 25.27 0.675 26.54 0.72 9437 1451 217 600.0-700.0 6.30 10.46 1.515 11.57 1.54 9437 1451 218 700.0-800.0 4.65 9.97 0.875 10.52 0.97 9437 1451 219 800.0-900.0 3.51 8.24 0.743 9.30 0.82 9437 1451 220 900.0-999.9 4.96 12.99 0.613 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 6.2% LOWER 9437 1451 226 ABOVE 500 EV .THE CALCULATED ALPHA VALUES ARE ON AVERAGE 7.4% 9437 1451 227 LOWER ABOVE 500 EV. 9437 1451 228 9437 1451 229 UNRESOLVED RESONANCE REGION******************* 9437 1451 230 9437 1451 231 THE AVERAGE RESONANCE PRAMETERS ARE GIVEN IN THE ENERGY RANGE 9437 1451 232 1 KEV TO 30 KEV FOR 94 ENERGY POINTS.THEY WERE OBTAINED BY U- 9437 1451 233 -SING THE CADARACHE STATISTICAL CODE FISINGA TO FIT THE GROSS 9437 1451 234 STRUCTURE OF THE SACLAY EXPERIMENTAL TOTAL CROSS SECTIONS(17) 9437 1451 235 BELOW 4 KEV AND OF SELECTED EXPERIMENTAL FISSION CROSS SECTIONS9437 1451 236 NORMALIZED TO ENDF/B-VI STANDARD EVALUATION(11).ABOVE 4 KEV NO 9437 1451 237 HIGH RESOLUTION TOTAL CROSS SECTION DATA ARE AVAILABLE;AVERAGE 9437 1451 238 TOTAL CROSS SECTIONS WERE CALCULATED TO BE CONSISTENT WITH THE 9437 1451 239 STASTISTICAL PARAMATERS OBTAINED IN THE RESOLVED RESONANCE RE- 9437 1451 240 GION(14) AND WITH THE OPTICAL MODEL PARAMETERS OF LAGRANGE AND 9437 1451 241 MADLAND(15) OBTAINED BY FITTING THE EXPERIMENTAL DATA IN THE 9437 1451 242 HIGH ENERGY RANGE.A VALUE OF 9.46 F WAS USED FOR THE EFFECTIVE 9437 1451 243 RADIUS ,THE SAME AS THE ONE OBTAINED IN THE RESOLVED REGION 9437 1451 244 BY ANALYSING THE EXPERIMENTAL TRANSMISSION DATA.THE VALUES OB- 9437 1451 245 TAINED FOR ALPHA ARE CONSISTENT WITH THE 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 BY PROCESSING THE EVALUATED FILE 9437 1451 252 BY THEMIS,THE FRENCH VERSION OF NJOY,ARE GIVEN IN THE FOLLO- 9437 1451 253 -WING TABLE,'FISS' FOR THE FISSION VALUES AND 'CAPT' FOR THE 9437 1451 254 CAPTURE VALUES: 9437 1451 255 9437 1451 256 ENERGY(KEV) CROSS SECTIONS ENERGY(KEV) CROSS SECTIONS 9437 1451 257 (BARN) (BARN) 9437 1451 258 FISS CAPT FISS CAPT 9437 1451 259 1.025 7.064 3.592 5.875 2.233 1.763 9437 1451 260 1.075 4.548 5.058 6.125 2.474 1.584 9437 1451 261 1.125 7.286 3.909 6.375 1.944 1.768 9437 1451 262 1.175 4.916 3.162 6.625 1.881 1.667 9437 1451 263 1.225 4.010 2.580 6.875 1.747 1.681 9437 1451 264 1.275 5.538 5.627 7.125 1.804 1.578 9437 1451 265 1.325 7.558 2.525 7.375 2.420 1.305 9437 1451 266 1.375 6.904 2.404 7.625 2.470 1.248 9437 1451 267 1.425 3.686 3.725 7.875 1.799 1.520 9437 1451 268 1.475 4.369 2.769 8.125 2.186 1.304 9437 1451 269 1.525 2.844 3.684 8.375 2.417 1.180 9437 1451 270 1.575 2.200 3.652 8.625 2.038 1.305 9437 1451 271 1.625 4.626 3.875 8.875 2.292 1.164 9437 1451 272 1.675 3.455 3.543 9.125 1.819 1.314 9437 1451 273 1.725 2.961 3.956 9.375 1.642 1.313 9437 1451 274 1.775 3.897 4.326 9.625 2.133 1.054 9437 1451 275 1.825 5.995 4.163 9.875 1.778 1.165 9437 1451 276 1.875 4.195 2.632 10.250 1.794 1.151 9437 1451 277 1.925 2.226 3.125 10.750 1.981 1.059 9437 1451 278 1.975 1.856 3.948 11.250 1.843 1.071 9437 1451 279 2.050 1.879 3.137 11.750 1.812 1.042 9437 1451 280 2.150 3.119 3.315 12.250 1.900 0.968 9437 1451 281 2.250 2.691 2.800 12.750 1.864 0.947 9437 1451 282 2.350 3.436 3.331 13.250 1.857 0.917 9437 1451 283 2.450 4.280 2.456 13.750 1.715 0.942 9437 1451 284 2.550 2.725 2.754 14.250 1.492 0.948 9437 1451 285 2.650 3.103 3.425 14.750 1.797 0.854 9437 1451 286 2.750 4.169 2.010 15.250 1.883 0.797 9437 1451 287 2.850 4.126 2.077 15.750 1.697 0.843 9437 1451 288 2.950 3.362 3.710 16.250 1.801 0.782 9437 1451 289 3.050 3.017 1.998 16.750 1.628 0.824 9437 1451 290 3.150 4.896 1.934 17.250 1.498 0.819 9437 1451 291 3.250 3.954 2.277 17.750 1.862 0.701 9437 1451 292 3.350 1.710 2.166 18.250 1.711 0.736 9437 1451 293 3.450 2.198 2.572 18.750 1.632 0.748 9437 1451 294 3.550 2.214 1.885 19.250 1.738 0.694 9437 1451 295 3.650 2.394 2.948 19.750 1.743 0.677 9437 1451 296 3.750 3.067 1.624 20.500 1.672 0.679 9437 1451 297 3.850 3.556 2.122 21.500 1.646 0.661 9437 1451 298 3.950 2.930 2.397 22.500 1.472 0.697 9437 1451 299 4.125 2.114 2.270 23.500 1.632 0.619 9437 1451 300 4.375 2.509 2.129 24.500 1.636 0.596 9437 1451 301 4.625 2.774 1.717 25.500 1.547 0.607 9437 1451 302 4.875 1.980 2.186 26.500 1.627 0.562 9437 1451 303 5.125 2.406 1.916 27.500 1.544 0.572 9437 1451 304 5.375 2.153 1.953 28.500 1.568 0.549 9437 1451 305 5.625 2.294 1.807 29.500 1.609 0.521 9437 1451 306 9437 1451 307 AVERAGE VALUES OF THE FISSION CROSS SECTIONS COMPARED TO THE 9437 1451 308 ENDF/B-VI STANDARD EVALUATION(11) AND ALPHA VALUES COMPARED TO 9437 1451 309 SOME EXPERIMENTAL DATA ARE GIVEN IN THE FOLLOWING TABLE: 9437 1451 310 9437 1451 311 ENERGY CROSS SECTIONS(BARN) ALPHA 9437 1451 312 (KEV) (1) (2) (3) (4) (5) (6) (7) (8) 9437 1451 313 1- 2 4.507 4.467 3.612 3.79 0.801 0.840 0.927 0.892 9437 1451 314 2- 3 3.289 3.304 2.901 3.31 0.882 1.000 1.108 1.028 9437 1451 315 3- 4 2.994 3.000 2.192 2.20 0.732 0.720 0.895 0.820 9437 1451 316 4- 5 2.344 2.383 2.076 2.07 0.886 0.870 0.821 0.837 9437 1451 317 5- 6 2.272 2.301 1.860 1.91 0.819 0.820 0.867 0.834 9437 1451 318 6- 7 2.012 2.008 1.675 1.63 0.833 0.790 0.816 0.793 9437 1451 319 7- 8 2.123 2.054 1.413 1.34 0.665 0.640 0.630 0.605 9437 1451 320 8- 9 2.233 2.216 1.238 1.23 0.554 0.540 0.575 0.530 9437 1451 321 9-10 1.843 1.864 1.212 1.05 0.658 0.550 0.617 0.569 9437 1451 322 1-10 2.624 2.622 2.020 2.06 0.770 0.752 0.806 0.768 9437 1451 323 10-20 1.762 1.764 0.876 0.85 0.497 0.480 0.466 0.498 9437 1451 324 20-30 1.595 1.595 0.606 0.58 0.380 0.350 0.373 0.388 9437 1451 325 9437 1451 326 (1) FISSION CROSS SECTION , PRESENT EVALUATION 9437 1451 327 (2) FISSION CROSS SECTION , ENDF/B-VI STANDARD(11) 9437 1451 328 (3) CAPTURE CROSS SECTION, PRESENT EVALUATION 9437 1451 329 (4) CAPTURE CROSS SECTION, GWIN ET AL.1976(4) 9437 1451 330 (5) ALPHA VALUE , PRESENT EVALUATION 9437 1451 331 (6) ALPHA VALUE FROM GWIN ET AL.1976(4) 9437 1451 332 (7) ALPHA VALUE FROM SOWERBY-KONSHIN EVALUATION 1971(16) 9437 1451 333 (8) AVERAGE ALPHA VALUE FROM EXPERIMENTAL DATA 9437 1451 334 9437 1451 335 ********************************************** 9437 1451 336 9437 1451 337 THE FISSION AND CAPTURE RESONANCE INTEGRALS ARE COMPARED TO ENDF/ 9437 1451 338 B-5 DATA IN THE FOLLOWING TABLE: 9437 1451 339 9437 1451 340 ********************************************************* 9437 1451 341 ENERGY RANGE(EV) FISSION(BARN) CAPTURE(BARN) 9437 1451 342 ********************************************************* 9437 1451 343 ENDF/B-5 PRESENT ENDF/B-5 PRESENT 9437 1451 344 0.5 - 5.0 86.02 85.62 32.31 27.91 9437 1451 345 5.0 - 10.0 26.03 25.19 20.14 19.06 9437 1451 346 10.0 - 50.0 100.25 96.82 78.66 77.41 9437 1451 347 50.0 - 100.0 40.32 40.41 27.23 25.54 9437 1451 348 100.0 - 301.0 19.98 19.66 19.52 17.91 9437 1451 349 301.0 -1000.0 10.15 10.03 8.54 8.16 9437 1451 350 ******************************************************** 9437 1451 351 0.5 -1000.0 282.76 277.73 186.30 175.99 9437 1451 352 ******************************************************** 9437 1451 353 9437 1451 354 BY ADDING THE ENDF/B-5 VALUE ABOVE 1 KEV ON OBTAIN FROM THE PRE- 9437 1451 355 SENT EVALUATION: 9437 1451 356 RI FISSION 297.10 BARNS 9437 1451 357 RI CAPTURE 183.79 BARNS 9437 1451 358 THE CORRESPONDING VALUES FROM ENDF/B-5 EVALUATION ARE: 9437 1451 359 RI FISSION 302.13 BARNS 9437 1451 360 RI CAPTURE 194.10 BARNS 9437 1451 361 9437 1451 362 ******************************************************************9437 1451 363 MF=1 9437 1451 364 ******************************************************************9437 1451 365 MT=452 NU PROMPT 9437 1451 366 ************************************************* 9437 1451 367 THE FLUCTUATIONS OBSERVED IN THE RESOLVED RANGE HAVE BEEN ANALYZED9437 1451 368 IN TERMS OF SPIN AND (N, F) EFFECTS.PREDOMINENT IN THE EPITHERMAL9437 1451 369 RANGE ,THE SPIN EFFECT PROGRESSIVELY DECREASES AND THE REDUCTION 9437 1451 370 OF NUP IS DUE TO THE (N,GF) EFFECT ONLY.SINCE THIS EFFECT IS LIKE9437 1451 371 LY EXISTING ON WIDE RANGE THE EXISTENCE OF NUP REDUCTION IS SUSPEC9437 1451 372 TED AT HIGHER ENERGIES,IE IN THE UNRESOLVED RANGE AND PROBABLY IN 9437 1451 373 THE RANGE OF TENS OR HUNDREDS KEV.A FORMALISM HAS BEEN ESTABLISHED9437 1451 374 TO DESCRIBE NUP AS A FUNCTION OF ENERGY BETWEEN 10-5 EV AND 4 KEV.9437 1451 375 AT HIGHER ENERGIES THE EVALUATED DATA HAVE BEEN OBTAINED FROM A 9437 1451 376 RENORMALIZATION OF THE EXPERIMENTEL DATA AND THE BARYCENTRE METHOD9437 1451 377 .MORE DETAILS WILL BE FOUND IN REF/18/.AN OVERALL 0.35% ACCURACY 9437 1451 378 IS ASSIGNED.IN THE REVISED JEF2.1 VERSION MODIFICATIONS HAVE BEEN 9437 1451 379 BROUGHT.THEY CONCERN: 9437 1451 380 1/ A RENORMALIZATION IN THE RANGE 4 KEV_20 MEV BY A FACTOR 1.002429437 1451 381 =3.7661/3.757 TO ACCOUNT FOR A REDEFINITION OF THE PROMPT NEUTRON 9437 1451 382 YIELD OF 252 CF/19/. 9437 1451 383 2/ THE BUMP IN NUP CURVE OBSERVED BY GWIN /20/ BETWEEN 20 KEV AND 9437 1451 384 50 KEV HAS BEEN INTRODUCED ACCORDING TO THE FOLLOWING ARGUMENT: 9437 1451 385 WITH THE EXCEPTION OF THIS BUMP GWIN'S DATA OF THIS SET ARE IN 9437 1451 386 PERFECT AGREEMENT WITH CALCULATIONS PERFORMED WITH THE FORMALISM 9437 1451 387 USED AT LOWER ENERGY AND USING AVERAGE PARAMETERS DEDUCED FROM 9437 1451 388 THE RESOLVED RANGE.OF COURSE ,A PHYSICS EXPLANATION HAS TO BE 9437 1451 389 FOUND ,BUT IT IS NOT UNDERSTOOD WHY ONLY THIS POINT WOULD BE 9437 1451 390 ERRONEOUS AND ALL THE OTHER ONES CORRECT. 9437 1451 391 ***************************************** 9437 1451 392 MT=455 NU DELAYED 9437 1451 393 ***************************************** 9437 1451 394 THE TOTAL DELAYED NEUTRON YIELD NUD HAS BEEN CALCULATED WITH A 9437 1451 395 NEW FORMALISM DUE TO LENDL /21/ WHICH ACCOUNTS FOR THE COMPETITION9437 1451 396 WITH PROMPT NEUTRON EMISSION .THIS MODEL IS BASED ON A ZP FORMA 9437 1451 397 LISM (NUP DEPENDENT) THAT IS CONSISTENT WITH THE ZP MODEL BY WAHL.9437 1451 398 THE CONSISTENCY IN THE THERMAL RANGE HAS BEEN CHECKED.THE EVALUA 9437 1451 399 TED NUD VALUES ARE CONSISTENT WITH SUMMATION CALCULATIONS IN THE 9437 1451 400 THERMAL ,FAST AND 14 MEV RANGES/22/. 9437 1451 401 9437 1451 402 ******************************************************************9437 1451 403 MF=3 9437 1451 404 ******************************************************************9437 1451 405 9437 1451 406 THE CROSS SECTIONS HAVE BEEN CALCULATED WITH THE STATISTICAL- 9437 1451 407 OPTICAL MODEL FORMALISM FROM 1 KEV TO 5.6 MEV USING THE FISINGA 9437 1451 408 CODE . IN THE HIGHER ENERGY RANGE , THE PRECOMPOUND COMPONENT 9437 1451 409 IS TAKEN INTO ACCOUNT IN THE SI4N CODE. 9437 1451 410 ******************************************************** 9437 1451 411 MT=1,2,4,16,17,51.....,91 NEUTRON CHANNEL CROSS-SECTIONS 9437 1451 412 ******************************************************** 9437 1451 413 THE NEUTRON CHANNELS ARE DESCRIBED BY COUPLED CHANNEL CALCULATIONS9437 1451 414 (COUPLING BASIS:1/2+,3/2+,5/2+,7/2+,9/2+) USING A PARAMETRIZATION 9437 1451 415 RESULTING FROM SEMIMICROSCOPIC CALCULATIONS BY LAGRANGE,MADLAND 9437 1451 416 AND GIROD/23/. SUCH A PARAMETRIZATION HAS BEEN PREFERED TO PHENO- 9437 1451 417 MENOLOGICAL PARAMETRIZATIONS OF EQUIVALENT QUALITY CONCERNING THE 9437 1451 418 FITTING OF THE CROSS-SECTIONS BECAUSE OF A BETTER DESCRIPTION OF 9437 1451 419 THE NEUTRON ANGULAR DISTRIBUTIONS OF IMPORTANCE FOR THE TRANSPORT 9437 1451 420 CROSS SECTION CALCULATION. NEVERTHELESS , THE SCATTERING 9437 1451 421 RADIUS TOO SMALL A VALUE (9,15 FM) USED IN HANDLING THE SO CALLED9437 1451 422 "SPRT" METHOD IS NOT CONSISTENT WITH THE VALUE DEDUCED FROM THE 9437 1451 423 RESOLVED RESONANCE PARAMETER ANALYSIS INTRODUCING A SMALL DISCON- 9437 1451 424 TINUITY IN THE SCATTERING CROSS SECTION AT 30 KEV THAT HAS BEEN 9437 1451 425 REMOVED BY AVERAGING. 9437 1451 426 THE N,XN CROSS SECTIONS HAVE BEEN CALCULATED BY LOOKING AT THE CON9437 1451 427 SISTENCY OF THE NEUTRON EMISSION PROBABILITY OF A GIVEN NUCLEUS 9437 1451 428 INVOLVED IN THE NEUTRON CASCADE WITH THE EXPERIMENTAL DATA OF 9437 1451 429 ITS RELATED NEUTRON CROSS SECTIONS.DIFFICULTIES HAVE BEEN ENCOUNTE9437 1451 430 RED TO REPRODUCE THE N,2N CROSS SECTION DATA BY FREHAUT/24/.VERY 9437 1451 431 LOW VALUES ARE OBSERVED IN A 2 MEV RANGE ABOVE THE THRESHOLD.THEY 9437 1451 432 ARE CONFIRMED BY INTEGRAL DATA OBTAINED IN FUEL IRRADIATION IN 9437 1451 433 PHENIX POWER PLANT.EXPLANATIONS ABOUT THIS SINGULAR CROSS SECTION 9437 1451 434 BEHAVIOUR ARE STILL TO BE FOUND.THE FIT INTO THE DATA HAS BEEN 9437 1451 435 OBTAINED BY ADJUSTING THE THEORETICAL FIRST CHANCE FISSION CROSS 9437 1451 436 IN THIS 2 MEV RANGE ABOVE THE N,2N THRESHOLD.IN THE CALCULATIONS 9437 1451 437 ABOVE 5 MEV PREEQUILIBRIUM EFFECTS ARE TAKEN IN ACCOUNT BY A 9437 1451 438 CORRECTIVE TERM BASED ON THE EXCITON MODEL BY GRIFFITH. 9437 1451 439 THE DISCRETE INELASTIC LEVELS CROSS SECTIONS ARE SLIGHTLY UNDER 9437 1451 440 ESTIMATED WHILE THE TOTAL INELASTIC CROSS-SECTION DEDUCED BY 9437 1451 441 SMITH AND GUENTHER/25/FROM "PSEUDO ELASTIC SCATTERING" CROSS- 9437 1451 442 SECTION MEASUREMENTS IS CORRECTLY REPRESENTED. 9437 1451 443 ********************************************* 9437 1451 444 MT=18,19,20,21,38 FISSION CROSS SECTIONS 9437 1451 445 ********************************************* 9437 1451 446 THE EVALUATION OF THE FISSION CROSS-SECTION PROVED VERY DIFFICULT.9437 1451 447 VERY LARGE DISCREPANCIES EXIST IN BOTH RELATIVE AND ABSOLUTE DATA.9437 1451 448 THE EXPERIMENTAL EVALUATION OBTAINED BY KANDA AND NAKAJIMA BY A 9437 1451 449 GENERALIZED LEAST SQUARES METHOD HAS BEEN FINALLY RETAINED AS A 9437 1451 450 STARTING POINT. ACTUALLY,THE EVALUATED DATA ARE THEORETICAL DERI- 9437 1451 451 VED BY FITTING INTO THE EXPERIMENTAL EVALUATION. FOR JEF2.0 THE 9437 1451 452 FIT WAS EYEGUIDED ACCOUNTING FOR 1984 WESTON'S DATA WHICH WERE NOT9437 1451 453 INCLUDED IN KANDA'S DATA BASE. A VALIDATION ON JEZEBEL,JEZEBEL-PU 9437 1451 454 FLATTOP-PU /26/REVEALED,AMONG OTHERS,AN UNDERESTIMATION OF THE 9437 1451 455 FISSION CROSS-SECTION .THE VERSION JEF2.1 HAS BEEN OBTAINED BY 9437 1451 456 ADJUSTING BY A STATISTICAL METHOD (BASED ON A THE MAXIMUM LIKELY- 9437 1451 457 HOUD FUNCTION) THE NUCLEAR MODEL PARAMETERS ON KANDA'S DATA AND 9437 1451 458 DISCARDING WESTON'S RESULTS.THE ADJUSTMENT HAS BEEN PERFORMED IN 9437 1451 459 REQUIRED CONDITIONS OF LINEARITY AND THE BASIC PARAMETERS HAVE 9437 1451 460 BEEN MODIFIED WITHIN THE ERROR BARS. THE ADVANTAGE OF SUCH AN 9437 1451 461 METHOD WHICH HAS STRONG PHYSICAL BASIS IS TO PRODUCE THE COVARI- 9437 1451 462 ANCE MATRIX C(P) FOR THE MODEL PARAMETERS P SO THAT THE COVARIANCE9437 1451 463 MATRICES OF THE CROSS-SECTIONS CAN BE PRODUCED ONCE THE SENSITIVI-9437 1451 464 TIES OF THE CROSS-SECTIONS TO THE PARAMETERS ARE KNOWN: C =SC(P)S+9437 1451 465 .THE COVARIANCE MATRICES FOR FISSION AND CAPTURE CROSS SECTIONS 9437 1451 466 HAVE BEEN CALCULATED THAT WAY /27/. 9437 1451 467 .COMPARED TO JEF2.0 VERSION THE FISSION CROSS-SECTION HAS BEEN 9437 1451 468 RESCALED BY A FACTOR VARYING BETWEEN 1.01 AND 1.017. 9437 1451 469 9437 1451 470 ************************************************************* 9437 1451 471 MF = 4 ANGULAR DISTRIBUTION (TAKEN FROM BROND UP TO 3MEV) 9437 1451 472 ************************************************************* 9437 1451 473 9437 1451 474 THE ELASTIC AND INELASTIC NEUTRON SCATTERING ANGULAR 9437 1451 475 DISTRIBUTIONS FROM 100 KEV TO 3 MEV ARE OBTAINED BY THE 9437 1451 476 COUPLED CHANNELS CALCULATIONS WITH THE ISOTROPIC PART 9437 1451 477 CALCULATED FROM THE COMPOUND PROCESS ADDED AND BY THE 9437 1451 478 ANALYSIS OF THE EXPERIMENTAL DATA AVAILABLE.ABOVE 3 MEV THE 9437 1451 479 COMPOUND COMPONENT IS NEGLIGBLE AND THE DISTRIBUTIONS ARE 9437 1451 480 THOSE OF THE DIRECT INTERACTION AS CALCULATED BY LAGRANGE. 9437 1451 481 9437 1451 482 ************************************************ 9437 1451 483 MF=5 ENERGY DISTRIBUTIONS OF SECOND. NEUTRONS 9437 1451 484 ************************************************ 9437 1451 485 9437 1451 486 MT=16,17,37,91 EVAPORATION SPECTRUM. 9437 1451 487 MT=455 TAKEN FROM T.ENGLAND (EN89) 9437 1451 488 9437 1451 489 MT=18 COMPOSITE NEUTRON ENERGY DISTRIBUTIONS FROM FISSION. 9437 1451 490 19 BASED ON CALCULATIONS BY D.MADLAND (AR84) USING MADLAND-9437 1451 491 20 NIX FORMALISM. THE CALCULATIONS INCLUDE THE FIRST-, 9437 1451 492 21 SECOND-, AND THIRD-CHANCE FISSION NEUTRON COMPONENTS. 9437 1451 493 38 THESE DATA ARE THE SAME AS WERE USED FOR REVISION 2 OF 9437 1451 494 ENDF/B-V. PARAMETERS FOR THE CALCULATION WERE ADJUSTED 9437 1451 495 TO GIVE THE SAME AVERAGE FISSION NEUTRON ENERGY AT 9437 1451 496 THERMAL AS ENDF/B-V.0. TABULATED DATA (LF=1) USED. 9437 1451 497 9437 1451 498 **************************************** 9437 1451 499 MF=12 PHOTON PRODUCTION MULTIPLICITIES 9437 1451 500 **************************************** 9437 1451 501 9437 1451 502 MT=18 : THE EXPERIMENTAL DATA FOR THIS NUCLEUS ARE VERY SCARCE9437 1451 503 AND ARE LIMITED TO THE THERMAL ENERGY.ON THE CONTRARY OF , 9437 1451 504 TOTAL GAMMMA ENERGY (EGT PROPORTIONNAL TO NUP),THERE IS NO EXPERI 9437 1451 505 MENTAL EVIDENCE OF THE INCIDENT ENERGY DEPENDENCE OF THE TOTAL 9437 1451 506 NUMBER OF EMITTED QUANTA .BUT SUCH A DEPENDENCE CAN BE DERI 9437 1451 507 VED FROM THE EXPRESSION =2*/S2 IF ONE KNOWS THE ENERGY 9437 1451 508 DEPENDENCE OF S2(VARIANCE OF THE DISTRIBUTION).FREHAUT ET AL HAVE 9437 1451 509 FOUND A LINEAR LAW FOR S2 AND NUP,ESTABLISHING IN THAT WAY THE 9437 1451 510 ENERGY DEPENDENCE OF S2 /28/.A VERY SMALL ENERGY DEPENDENCE HAS 9437 1451 511 BEEN FOUND FOR , THAT HAS BEEN NEGLECTED IN THE PRESENT 9437 1451 512 EVALUATION/30/.IT IS GIVEN =6.497 . 9437 1451 513 HOFFMAN AND HOFFMAN /29/ HAVE QUOTED RELATIONSHIPS FOR AND 9437 1451 514 .THESE ONES HAVE BEEN MODIFIED TO ACCOUNT FOR THE ANISOTROPY9437 1451 515 OF THE QUANTA EMISSION AND TO FIT INTO 252 CF RESULTS BY FREHAUT.9437 1451 516 FINALLY THE LAWS USED IN THE EVALUATION ARE: 9437 1451 517 =0.02772 A +0.0891 9437 1451 518 =0.106 A-18.943 WHERE A STANDS FOR THE ATOM MASS OF THE 9437 1451 519 FISSIONNING NUCLEUS. 9437 1451 520 FROM THE THERMAL NUP VALUE ONE OBTAINS FOR 240 PU FISSIONNING 9437 1451 521 =0.815 NUP+ 4.4 THE CONSTANT 4.4 BEING IN ACCORDANCE WITH 9437 1451 522 BOTH STATISTICAL MODEL PREDICTIONS AND EXPERIMENTAL EVIDENCE. 9437 1451 523 FINNALY ,ON THE WHOLE ENERGY RANGE,THE TOTAL GAMMA ENERGY RELEASED9437 1451 524 IN FISSION RESULTS FROM THE COMPETITION BETWEEN FIRST,SECOND,THIRD9437 1451 525 CHANCE FISSION.IN ADDITION A PREFISSION COMPONENT HAS BEEN ADDED 9437 1451 526 THAT HAS BEEN CALCULATED WITH THE SAME PARAMETER VALUES AS FOR 9437 1451 527 FISSION CROSS-SECTION AND NUP SO THAT A COMPLETE CONSISTENCY IS 9437 1451 528 PRESERVED. 9437 1451 529 9437 1451 530 ************************************************* 9437 1451 531 REFERENCES 9437 1451 532 ************************************************* 9437 1451 533 9437 1451 534 1-R.GWIN ET AL.,NUCL.SCI.ENG.,45,25(1971) 9437 1451 535 2-A.J.DERUYTER ET AL.,J.NCL.ENER.,26,293(1972) 9437 1451 536 3-J.BLONS, NUCL.SCI.ENG.,51,130(1973) 9437 1451 537 4-R.GWIN ET AL.,NUCL.SCI.ENG.,59,79(1976) 9437 1451 538 5-R.GWIN ET AL.,NUCL.SCI.ENG.,61,116(1976) 9437 1451 539 6-W.WAGEMANS,ANNALS OF NUCL.ENER.7,9,495(1980) 9437 1451 540 7-R.GWIN ET AL.,NUCL.SCI.ENG.,88,37(1984) 9437 1451 541 8-L.W.WESTON ET AL.,NUCL.SCI.ENG.88,567(1984) 9437 1451 542 9-J.A.HARVEY ,PRIVATE COMMUNICATION(1985) 9437 1451 543 10-R.R.SPENCER ET AL.,NUCL.SCI.ENG.,96,318(1987) 9437 1451 544 11-A.CARLSON ET AL.,PRELIMINARY RESULTS OF THE ENDF/B-6 STANDARD 9437 1451 545 EVALUATION(SEPT 8 1987) 9437 1451 546 12-A.J.DERUYTER,J.NUCL.ENER.,26,293(1972) 9437 1451 547 13-N.M.LARSON ET AL.,ORNL/TM-7485,ORNL/TM-9179,ORNL/TM-9719/R1 9437 1451 548 14-H.DERRIEN AND G. DE SAUSSURE,ORNL-TM- (1988) 9437 1451 549 15-CH.LAGRANGE AND D.G.MADLAND,PHYS.REV.C,33,5(1986) 9437 1451 550 16-M.G.SOWERBY ET AL.,AT.ENERGY REV.,10,4,453,IAEA,VIENNA(1972) 9437 1451 551 17-H.DERRIEN,THESIS,ORSAY SERIE A,1172(1973) 9437 1451 552 18-E.FORT ET AL,NSE99,375-389(1988). 9437 1451 553 19-E.J.AXTON,NUCL.SCI.TECHNOL.5,609 (1984). 9437 1451 554 20-R.GWIN ET AL,ORNL/TM-6246(1978), ORNL/TM-7148(1980) 9437 1451 555 21-A.LENDEL ET AL,ATOMNAYA ENERGIYA VOL 61 ,N3,PP215-216,(1986) 99437 1451 556 22-E.FORT ET AL,JEF/DOC 282 (1989). 9437 1451 557 23-C.LAGRANGE,D.MADLAND,M.GIROD,PHYS REW VOL33,5 MAY 1986. 9437 1451 558 24-J.FREHAUT ET AL PROCEEDINGS CONF SANTA FE P1561(1985) 9437 1451 559 25-A.B.SMITH AND P.GUENTHER ANL/NDM-63 (1982). 9437 1451 560 26-E.FORT,C.MOUNIE UNPUBLISHED (1990) 9437 1451 561 27-E.FORT,Y.KANDA,Y.NAKAJIMA.TO BE ISSUED AS A JEF/DOC REPORT. 9437 1451 562 28-J.FREHAUT,PROC CONS MEETING MITO(MAY 1988) INDC(NDS)-220 P81 9437 1451 563 29-D.C.HOFFMAN AND M.H.HOFFMAN ANN.REV.NUCL.SCI 24 (1974) 151 9437 1451 564 30.E.FORT,P.LONG,SPRC/LEPH 88/232 REPORT,SEE ALSO JEF/DOC REPORT 9437 1451 565 9437 1451 566 *************************************************** 9437 1451 567 9437 1451 568 1 451 651 29437 1451 569 1 452 777 09437 1451 570 1 455 779 09437 1451 571 1 456 777 09437 1451 572 2 151 887 29437 1451 573 3 1 102 09437 1451 574 3 2 98 09437 1451 575 3 4 101 09437 1451 576 3 16 26 09437 1451 577 3 17 14 09437 1451 578 3 18 97 09437 1451 579 3 19 97 09437 1451 580 3 20 28 09437 1451 581 3 21 26 09437 1451 582 3 37 6 09437 1451 583 3 38 14 09437 1451 584 3 51 101 09437 1451 585 3 52 92 09437 1451 586 3 53 89 09437 1451 587 3 54 83 09437 1451 588 3 55 68 09437 1451 589 3 56 63 09437 1451 590 3 57 62 09437 1451 591 3 58 61 09437 1451 592 3 59 59 09437 1451 593 3 60 58 09437 1451 594 3 61 58 09437 1451 595 3 62 56 09437 1451 596 3 63 55 09437 1451 597 3 64 55 09437 1451 598 3 65 54 09437 1451 599 3 66 53 09437 1451 600 3 67 52 09437 1451 601 3 68 52 09437 1451 602 3 69 51 09437 1451 603 3 70 50 09437 1451 604 3 71 48 09437 1451 605 3 91 60 09437 1451 606 3 102 85 09437 1451 607 3 104 19 09437 1451 608 3 105 20 09437 1451 609 3 106 20 09437 1451 610 4 2 78 09437 1451 611 4 16 2 09437 1451 612 4 17 2 09437 1451 613 4 18 2 09437 1451 614 4 19 2 09437 1451 615 4 20 2 09437 1451 616 4 21 2 09437 1451 617 4 37 2 09437 1451 618 4 38 2 09437 1451 619 4 51 76 09437 1451 620 4 52 74 09437 1451 621 4 53 72 09437 1451 622 4 54 86 09437 1451 623 4 55 2 09437 1451 624 4 56 2 09437 1451 625 4 57 2 09437 1451 626 4 58 2 09437 1451 627 4 59 2 09437 1451 628 4 60 2 09437 1451 629 4 61 2 09437 1451 630 4 62 2 09437 1451 631 4 63 2 09437 1451 632 4 64 2 09437 1451 633 4 65 2 09437 1451 634 4 66 2 09437 1451 635 4 67 2 09437 1451 636 4 68 2 09437 1451 637 4 69 2 09437 1451 638 4 70 2 09437 1451 639 4 71 2 09437 1451 640 4 91 2 09437 1451 641 5 16 17 09437 1451 642 5 17 22 09437 1451 643 5 18 1640 09437 1451 644 5 19 9 09437 1451 645 5 20 8 09437 1451 646 5 21 8 09437 1451 647 5 37 25 09437 1451 648 5 38 13 09437 1451 649 5 91 10 09437 1451 650 5 455 107 09437 1451 651 9437 1 0 652 9.423900+4 2.369990+2 0 2 0 09437 1452 653 0.000000+0 0.000000+0 0 0 1 23219437 1452 654 2321 2 9437 1452 655 1.000000-5 2.881240+0 1.796900-1 2.863279+0 2.890600-1 2.860490+09437 1452 656 1.136700+0 2.872952+0 2.757800+0 2.881390+0 4.093800+0 2.883551+09437 1452 657 5.648400+0 2.880384+0 7.000000+0 2.868778+0 7.791220+0 2.856864+09437 1452 658 7.800000+0 2.856874+0 8.200000+0 2.854115+0 8.500000+0 2.857651+09437 1452 659 8.800000+0 2.870571+0 9.200000+0 2.891561+0 9.400000+0 2.894689+09437 1452 660 9.800000+0 2.887377+0 1.043250+1 2.872683+0 1.150000+1 2.860411+09437 1452 661 1.178200+1 2.850828+0 1.178840+1 2.845887+0 1.201240+1 2.832718+09437 1452 662 1.210000+1 2.817229+0 1.230000+1 2.887516+0 1.240000+1 2.890595+09437 1452 663 1.310000+1 2.879029+0 1.399220+1 2.865700+0 1.427830+1 2.860600+09437 1452 664 1.444090+1 2.863698+0 1.450000+1 2.872434+0 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