JEFF-3.1.1 General purpose neutron file update release Jan. 2009 2311 0 0 0 9.123200+4 2.300450+2 1 1 2 19134 1451 1 0.000000+0 1.000000+0 0 0 0 69134 1451 2 1.000000+0 2.000000+7 1 0 10 319134 1451 3 0.000000+0 0.000000+0 0 0 211 259134 1451 4 91-Pa-232 ORNL,TIT EVAL-DEC99 R. Q. WRIGHT, N. TAKAGI 9134 1451 5 DIST-JAN09 20090105 9134 1451 6 ----JEFF-311 MATERIAL 9134 9134 1451 7 -----INCIDENT NEUTRON DATA 9134 1451 8 ------ENDF-6 FORMAT 9134 1451 9 *************************** JEFF-3.1.1 *************************9134 1451 10 ** **9134 1451 11 ** Original data taken from: JEFF-3.1 **9134 1451 12 ** **9134 1451 13 ******************************************************************9134 1451 14 ***************************** JEFF-3.1 *************************9134 1451 15 ** **9134 1451 16 ** Original data taken from: ENDF/B-VI.8 **9134 1451 17 ** **9134 1451 18 ******************************************************************9134 1451 19 9134 1451 20 **************************************************************** 9134 1451 21 9134 1451 22 ENDF/B-VI MOD 2 Revision, December 1999, R.Q. Wright (ORNL) 9134 1451 23 9134 1451 24 The JENDL-3.2 evaluation was adopted and updated as follows: 9134 1451 25 9134 1451 26 MF=1, 9134 1451 27 MT=456 NUMBER OF PROMPT NEUTRONS 9134 1451 28 Changed from LNU=1, polynomial representation to LNU=2, 9134 1451 29 tabulated representation. MT=452, 455, and 456 are now 9134 1451 30 represented as tabulated functions. 9134 1451 31 9134 1451 32 MF=2 RESONANCE PARAMETERS-- MLBW FORMALISM USED. 9134 1451 33 The resolved resonance parameter evaluation is based on 9134 1451 34 the measurements of Danon [Da1996]. Preliminary results were 9134 1451 35 given in the paper by Moore et al. [Mo1994]. 9134 1451 36 For the levels below 5 eV, the parameters from Table 1 9134 1451 37 of Moore et al. were used. The 2nd column of Table 1 of that 9134 1451 38 reference is actually g*Gamma-n0, not 2g*Gamma-n0 as 9134 1451 39 indicated. 9134 1451 40 For the 5 to 10 eV range, the parameters from Table 1 of 9134 1451 41 Danon are used. In this table, Column 2 should be 9134 1451 42 2g*Gamma-n0, not 2g*Gamma-n as indicated. In the current 9134 1451 43 evaluation the neutron widths for the 6.44 and 8.85 eV 9134 1451 44 resonances were both increased slightly. 9134 1451 45 The resulting parameters give a very good fit to the 9134 1451 46 fission cross section for the energy range below 10 eV. 9134 1451 47 9134 1451 48 2200-m/s cross sections and resonance integrals 9134 1451 49 2200 m/s value Res. Int. 9134 1451 50 Total 1762.19 b - 9134 1451 51 Elastic 32.75 b - 9134 1451 52 Fission 1517.34 b 868 b 9134 1451 53 Capture 212.10 b 146 b 9134 1451 54 9134 1451 55 MF=3 NEUTRON CROSS SECTIONS 9134 1451 56 MT=1 TOTAL CROSS SECTION 9134 1451 57 For the energy range 10 eV to 35 keV, the total cross 9134 1451 58 section is calculated as sum of MT = 2, 18 and 102, using 9134 1451 59 the revised values of MT = 2, 18, and 102. above 35 keV 9134 1451 60 The cross section is 133.35 barns at 10 eV. The revised 9134 1451 61 total is about the same as the JENDL-3.2 evaluation; it is 9134 1451 62 slightly higher below 4 eV and slightly lower between 4 eV 9134 1451 63 and 3 keV. 9134 1451 64 9134 1451 65 MT=2 ELASTIC SCATTERING CROSS SECTION 9134 1451 66 The elastic cross section is revised below 100 eV to join 9134 1451 67 smoothly with the resolved resonance range; it is unchanged 9134 1451 68 above 100 eV. The elastic cross section is 17.6 barns at 10 9134 1451 69 eV, and is slightly higher between 10 and 100 eV. 9134 1451 70 9134 1451 71 MT=18 FISSION CROSS SECTION 9134 1451 72 The fission cross section is revised below 35 keV to join 9134 1451 73 smoothly with the resolved resonance range; it is unchanged 9134 1451 74 above 35 keV. The cross section at 10 eV (100.0 barns) is 9134 1451 75 based on the average fission from 2 to 10 eV, as caluclated 9134 1451 76 from the resolved resonance parameters. 9134 1451 77 The average fission cross section = 108.4 barns for the 9134 1451 78 energy range 5.5 to 10 eV. For the energy range 100 ev to 9134 1451 79 30 keV, the revised cross section is very close to the 9134 1451 80 ENDF/B-VI 235U fission cross section. 9134 1451 81 9134 1451 82 MT=102 CAPTURE CROSS SECTION 9134 1451 83 The capture cross section is revised below 35 keV to join 9134 1451 84 smoothly with the resolved resonance range. It is unchanged 9134 1451 85 above 35 keV. 9134 1451 86 The cross section at 10 eV (15.75 barns) is based on the 9134 1451 87 average capture from 2 to 10 eV (resolved resonance range). 9134 1451 88 9134 1451 89 ---------------------------------------------------------------- 9134 1451 90 REFERENCES 9134 1451 91 9134 1451 92 Da1996 Y. Danon, M.S. Moore, P.E. Koehler et al. Nucl.Sci.Eng. 9134 1451 93 124, 482 (1996). 9134 1451 94 Mo1994 M.S. Moore, P.E. Koehler, P.E. Littleton et al., Int. 9134 1451 95 Conference on Nuclear Data for Science and Technology, 9134 1451 96 Gatlinburg, Tennessee, May 9-13, 1994 (American Nuclear 9134 1451 97 Society, 1994) Vol. 2, p. 1075. 9134 1451 98 9134 1451 99 **************************************************************** 9134 1451 100 9134 1451 101 JENDL-3.2 COMMENTS FOLLOW-- 9134 1451 102 9134 1451 103 HISTORY 9134 1451 104 88-08 NEW EVALUATION WAS MADE BY N. TAKAGI (TOKYO INSTITUTE OF 9134 1451 105 TECHNOLOGY, TIT) /1/. 9134 1451 106 94-06 JENDL-3.2. 9134 1451 107 NU-P, NU-D AND NU-TOTAL WERE MODIFIED. 9134 1451 108 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9134 1451 109 9134 1451 110 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 9134 1451 111 (1,452), (1,455), (1,456) 9134 1451 112 *********************************************************** 9134 1451 113 9134 1451 114 MF=1 GENERAL INFORMATION 9134 1451 115 MT=451 COMMENT AND DICTIONARY 9134 1451 116 MT=452 NUMBER OF NEUTRONS PER FISSION 9134 1451 117 SUM OF NU-P NAD NU-D. 9134 1451 118 MT=455 DELAYED NEUTRONS PER FISSION 9134 1451 119 AVERAGE VALUES OF SYSTEMATICS BY TUTTLE/2/, BENEDETTI ET 9134 1451 120 AL./3/ AND WALDO ET AL./4/ DECAY CONSTANTS WERE ADOPTED 9134 1451 121 FROM THE EVALUATION BY BRADY AND ENGLAND/5/. 9134 1451 122 MT=456 PROMPT NEUTRONS PER FISSION 9134 1451 123 BASED ON SYSTEMATICS BY MANERO AND KONSHIN/6/, AND BY 9134 1451 124 HOWERTON/7/. 9134 1451 125 9134 1451 126 MF=2 RESONANCE PARAMETERS 9134 1451 127 MT=151 RESONANCE PARAMETERS 9134 1451 128 NO RESONANCE PARAMETERS WERE GIVEN. 9134 1451 129 9134 1451 130 2200-M/S CROSS SECTIONS AND RESONANCE INTEGRALS 9134 1451 131 2200 M/S VALUE RES. INT. 9134 1451 132 TOTAL 1176.23 B - 9134 1451 133 ELASTIC 12.23 B - 9134 1451 134 FISSION 700.00 B 313 B 9134 1451 135 CAPTURE 464.00 B 309 B 9134 1451 136 9134 1451 137 MF=3 NEUTRON CROSS SECTIONS 9134 1451 138 MT=1 TOTAL CROSS SECTION 9134 1451 139 BELOW 1 EV, CALCULATED AS SUM OF MT'S = 2, 18 AND 102. 9134 1451 140 ABOVE 1 EV, OPTICAL MODEL CALCULATION WAS MADE WITH 9134 1451 141 CASTHY/8/. THE POTENTIAL PARAMETERS/9/ USED ARE AS 9134 1451 142 FOLLOWS, 9134 1451 143 V = 41.0 - 0.05*EN (MEV) 9134 1451 144 WS= 6.4 - 0.15*SQRT(EN) (MEV) 9134 1451 145 WV= 0 , VSO = 7.0 (MEV) 9134 1451 146 R = RSO = 1.31 , RS = 1.38 (FM) 9134 1451 147 A = ASO = 0.47 , B = 0.47 (FM) 9134 1451 148 9134 1451 149 MT=2 ELASTIC SCATTERING CROSS SECTION 9134 1451 150 BELOW 1 EV, ASSUMED TO BE THE SAME AS SHAPE ELASTIC 9134 1451 151 SCATTERING CROSS SECTION CALCULATED WITH THE OPTICAL 9134 1451 152 MODEL. ABOVE 1 EV, OPTICAL MODEL CALCULATION WAS 9134 1451 153 ADOPTED. 9134 1451 154 9134 1451 155 MT=4, 91 INELASTIC SCATTERING CROSS SECTIONS. 9134 1451 156 OPTICAL AND STATISTICAL MODEL CALCULATION WAS MADE WITH 9134 1451 157 CASTHY/8/. NO EXCITED LEVELS WERE RECOMMENDED IN 9134 1451 158 REF./10/. 9134 1451 159 NO ENERGY(KEV) SPIN-PARITY 9134 1451 160 G.S. 0.0 2 - 9134 1451 161 LEVELS ABOVE 50 KEV WERE ASSUMED TO BE OVERLAPPING. 9134 1451 162 THE LEVEL DENSITY PARAMETERS WERE TAKEN FROM REF./11/. 9134 1451 163 9134 1451 164 MT=16,17,37 (N,2N), (N,3N) AND (N,4N) REACTION CROSS SECTIONS 9134 1451 165 CALCULATED WITH EVAPORATION MODEL. 9134 1451 166 9134 1451 167 MT=18 FISSION CROSS SECTION 9134 1451 168 MEASURED THERMAL CROSS SECTION OF 700 BARNS WAS TAKEN FROM9134 1451 169 REF./12/, AND 1/V FORM WAS ASSUMED BELOW 1 EV. FOR 9134 1451 170 ENERGIES ABOVE 1 EV, THE SHAPE WAS ASSUMED TO BE THE SAME 9134 1451 171 AS U-233 FISSION CROSS SECTION AND NORMALIZED TO THE 9134 1451 172 SYSTEMATICS BY BEHRENS AND HOWERTON/13/. 9134 1451 173 9134 1451 174 MT=102 CAPTURE CROSS SECTION 9134 1451 175 MEASURED THERMAL CROSS SECTION OF 464 BARNS WAS TAKEN FROM9134 1451 176 REF./12/, AND 1/V FORM WAS ASSUMED BELOW 1 EV. THE CROSS 9134 1451 177 SECTION SHAPE NEAR 1 EV WAS ADJUSTED SO AS TO REPRODUCE 9134 1451 178 THE RESONANCE INTEGRAL/12/. ABOVE 1 EV, CALCULATED WITH 9134 1451 179 CASTHY. THE GAMMA-RAY STRENGTH FUNCTION WAS ESTIMATED FROM9134 1451 180 GAMMA-GAMMA = 0.040 EV AND LEVEL SPACING = 0.417 EV. 9134 1451 181 9134 1451 182 MT=251 MU-L 9134 1451 183 CALCULATED WITH CASTHY. 9134 1451 184 9134 1451 185 MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 9134 1451 186 MT=2,91 CALCULATED WITH OPTICAL MODEL. 9134 1451 187 MT=16,17,18,37 ISOTROPIC IN THE LAB SYSTEM. 9134 1451 188 9134 1451 189 MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS 9134 1451 190 MT=16,17,37,91 EVAPORATION SPECTRA 9134 1451 191 OBTAINED FROM LEVEL DENSITY PARAMETERS. 9134 1451 192 9134 1451 193 MT=18 MAXWELLIAN FISSION SPECTRUM. 9134 1451 194 TEMPERATURE WAS ESTIMATED FROM Z**2/A DEPENDENCE/14/. 9134 1451 195 9134 1451 196 REFERENCES 9134 1451 197 1) TAKAGI N. ET AL.: J. NUCL. SCI. TECHNOL., 27, 853 (1990). 9134 1451 198 2) TUTTLE R.J.: INDC(NDS)-107/G+SPECIAL, P.29 (1979), 9134 1451 199 3) BENEDETTI G. ET AL.: NUCL. SCI. ENG., 80, 379 (1982). 9134 1451 200 4) WALDO R. ET AL.: PHYS. REV., C23, 1113 (1981). 9134 1451 201 5) BRADY M.C. AND ENGLAND T.R.: NUCL. SCI. ENG., 103, 129(1989). 9134 1451 202 6) MANERO F. AND KONSHIN V.A.: AT. ENERGY REV.,10, 637 (1972). 9134 1451 203 7) HOWERTON R.J.: NUCL. SCI. ENG., 62, 438 (1977). 9134 1451 204 8) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991). 9134 1451 205 9) OHSAWA T., OHTA M.: J. NUCL. SCI. TECHNOL., 18, 408 (1981). 9134 1451 206 10) SCHMORAX M.R.: NUCL. DATA SHEETS, 36, 367 (1982). 9134 1451 207 11) GILBERT A., CAMERON A.G.W.: CAN. J. PHYS., 43, 1446 (1965). 9134 1451 208 12) MUGHABGHAB S.F.: "NEUTRON CROSS SECTIONS, VOL.1, NEUTRON 9134 1451 209 RESONANCE PARAMETERS AND THERMAL CROSS SECTIONS , PART B, 9134 1451 210 Z=61-100", ACADEMIC PRESS (1984). 9134 1451 211 13) BEHRENS J.W., HOWERTON R.J: NUCL. SCI. ENG., 65, 464, (1978). 9134 1451 212 14) SMITH A.B. ET AL.: ANL/NDM-50 (1979). 9134 1451 213 9134 1451 214 ************************ C O N T E N T S *********************** 9134 1451 215 1 451 240 19134 1451 216 1 452 5 19134 1451 217 1 455 7 19134 1451 218 1 456 5 19134 1451 219 2 151 18 19134 1451 220 3 1 22 19134 1451 221 3 2 20 19134 1451 222 3 4 15 19134 1451 223 3 16 9 19134 1451 224 3 17 6 19134 1451 225 3 18 19 19134 1451 226 3 37 4 19134 1451 227 3 91 15 19134 1451 228 3 102 20 19134 1451 229 4 2 240 19134 1451 230 4 16 10 19134 1451 231 4 17 10 19134 1451 232 4 18 10 19134 1451 233 4 37 10 19134 1451 234 4 91 10 19134 1451 235 5 16 17 19134 1451 236 5 17 22 19134 1451 237 5 18 7 19134 1451 238 5 37 25 19134 1451 239 5 91 10 19134 1451 240 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1452 1 0.000000+0 0.000000+0 0 0 1 59134 1452 2 5 2 9134 1452 3 1.000000-5 2.285800+0 2.530000-2 2.285800+0 5.000000+6 2.845800+09134 1452 4 6.000000+6 2.952800+0 2.000000+7 4.520800+0 9134 1452 5 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1455 1 0.000000+0 0.000000+0 0 0 6 09134 1455 2 1.290000-2 3.470000-2 1.150000-1 2.856000-1 6.706000-1 2.311100+09134 1455 3 0.000000+0 0.000000+0 0 0 1 59134 1455 4 5 2 9134 1455 5 1.000000-5 1.576000-2 2.530000-2 1.576000-2 5.000000+6 1.576000-29134 1455 6 6.000000+6 1.083000-2 2.000000+7 1.083000-2 9134 1455 7 0.000000+0 0.000000+0 0 0 0 09134 1 099999 9.123200+4 2.300450+2 0 2 0 09134 1456 1 0.000000+0 0.000000+0 0 0 1 59134 1456 2 5 2 9134 1456 3 1.000000-5 2.270040+0 2.530000-2 2.270040+0 5.000000+6 2.830040+09134 1456 4 6.000000+6 2.941970+0 2.000000+7 4.509970+0 9134 1456 5 0.000000+0 0.000000+0 0 0 0 09134 1 099999 0.000000+0 0.000000+0 0 0 0 09134 0 0 0 9.123200+4 2.300450+2 0 0 1 09134 2151 1 9.123200+4 1.000000+0 0 1 1 09134 2151 2 1.000000-5 1.000000+1 1 2 0 09134 2151 3 2.000000+0 9.890000-1 0 0 1 09134 2151 4 2.300450+2 0.000000+0 0 0 78 139134 2151 5 -5.000000+0 2.500000+0 4.906800-1 5.068000-2 4.000000-2 4.000000-19134 2151 6 3.300000-1 1.500000+0 2.310962-1 9.620000-5 4.000000-2 1.910000-19134 2151 7 6.900000-1 2.500000+0 3.400415-1 4.150000-5 4.000000-2 3.000000-19134 2151 8 1.400000+0 1.500000+0 3.400560-1 5.600000-5 4.000000-2 3.000000-19134 2151 9 2.730000+0 1.500000+0 2.176600-1 6.600000-4 4.000000-2 1.770000-19134 2151 10 3.050000+0 2.500000+0 2.101570-1 1.570000-4 4.000000-2 1.700000-19134 2151 11 4.150000+0 2.500000+0 3.753200-1 3.200000-4 4.000000-2 3.350000-19134 2151 12 6.440000+0 1.500000+0 3.086300-1 6.300000-4 4.000000-2 2.680000-19134 2151 13 7.450000+0 2.500000+0 5.261960-1 1.960000-4 4.000000-2 4.860000-19134 2151 14 8.410000+0 2.500000+0 5.192780-1 2.780000-4 4.000000-2 4.790000-19134 2151 15 8.850000+0 2.500000+0 4.537650-1 7.650000-4 4.000000-2 4.130000-19134 2151 16 1.100000+1 1.500000+0 3.613270-1 1.327000-3 4.000000-2 3.200000-19134 2151 17 1.185000+1 2.500000+0 4.408600-1 8.600000-4 4.000000-2 4.000000-19134 2151 18 0.000000+0 0.000000+0 0 0 0 09134 2 099999 0.000000+0 0.000000+0 0 0 0 09134 0 0 0 9.123200+4 2.300450+2 0 0 0 09134 3 1 1 0.000000+0 0.000000+0 0 0 2 569134 3 1 2 6 2 56 5 9134 3 1 3 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.590600-1 0.000000+09134 3 1 4 1.000000+0 0.000000+0 1.000000+1 0.000000+0 1.000000+1 1.333500+29134 3 1 5 3.162280+1 7.912700+1 1.000000+2 4.912960+1 3.162280+2 3.413684+19134 3 1 6 1.000000+3 2.544520+1 3.162280+3 2.038895+1 1.000000+4 1.708060+19134 3 1 7 3.000000+4 1.476440+1 3.250000+4 1.445023+1 3.500000+4 1.415950+19134 3 1 8 4.000000+4 1.390960+1 5.000000+4 1.349490+1 5.021730+4 1.348670+19134 3 1 9 7.000000+4 1.285810+1 1.000000+5 1.214320+1 1.100000+5 1.194180+19134 3 1 10 1.200000+5 1.175310+1 1.500000+5 1.124790+1 1.800000+5 1.081060+19134 3 1 11 2.000000+5 1.054780+1 2.500000+5 9.968470+0 3.000000+5 9.476870+09134 3 1 12 4.000000+5 8.686870+0 5.000000+5 8.091510+0 6.000000+5 7.644420+09134 3 1 13 7.000000+5 7.315260+0 8.000000+5 7.081530+0 9.000000+5 6.925250+09134 3 1 14 1.000000+6 6.831530+0 2.000000+6 7.310760+0 3.000000+6 7.819010+09134 3 1 15 4.000000+6 7.814340+0 5.000000+6 7.464290+0 5.585000+6 6.968820+09134 3 1 16 6.000000+6 6.993430+0 7.000000+6 6.544650+0 8.000000+6 6.138000+09134 3 1 17 9.000000+6 5.817650+0 1.000000+7 5.601610+0 1.100000+7 5.491100+09134 3 1 18 1.200000+7 5.480780+0 1.242870+7 5.501695+0 1.300000+7 5.545290+09134 3 1 19 1.400000+7 5.660120+0 1.500000+7 5.811850+0 1.600000+7 5.980510+09134 3 1 20 1.700000+7 6.141030+0 1.800000+7 6.265730+0 1.824700+7 6.285382+09134 3 1 21 1.900000+7 6.346220+0 2.000000+7 6.405530+0 9134 3 1 22 0.000000+0 0.000000+0 0 0 0 09134 3 099999 9.123200+4 2.300450+2 0 0 0 09134 3 2 1 0.000000+0 0.000000+0 0 0 2 499134 3 2 2 4 2 49 5 9134 3 2 3 1.000000-5 0.000000+0 2.530000-2 0.000000+0 1.000000+1 0.000000+09134 3 2 4 1.000000+1 1.760000+1 3.162280+1 1.588000+1 1.000000+2 1.431960+19134 3 2 5 1.000000+3 1.379520+1 1.000000+4 1.254460+1 3.000000+4 1.162840+19134 3 2 6 3.250000+4 1.153840+1 3.500000+4 1.145510+1 4.000000+4 1.131820+19134 3 2 7 5.000000+4 1.106610+1 5.021730+4 1.106040+1 7.000000+4 1.057820+19134 3 2 8 1.000000+5 9.954650+0 1.100000+5 9.768500+0 1.200000+5 9.589910+09134 3 2 9 1.500000+5 9.092570+0 1.800000+5 8.643270+0 2.000000+5 8.366490+09134 3 2 10 2.500000+5 7.743780+0 3.000000+5 7.222580+0 4.000000+5 6.398080+09134 3 2 11 5.000000+5 5.788480+0 6.000000+5 5.328500+0 7.000000+5 4.983510+09134 3 2 12 8.000000+5 4.728490+0 9.000000+5 4.542310+0 1.000000+6 4.412280+09134 3 2 13 2.000000+6 4.519400+0 3.000000+6 5.028580+0 4.000000+6 5.076990+09134 3 2 14 5.000000+6 4.767700+0 6.000000+6 4.308940+0 7.000000+6 3.856530+09134 3 2 15 8.000000+6 3.482500+0 9.000000+6 3.202950+0 1.000000+7 3.024650+09134 3 2 16 1.100000+7 2.946740+0 1.200000+7 2.958650+0 1.300000+7 3.040290+09134 3 2 17 1.400000+7 3.166930+0 1.500000+7 3.317030+0 1.600000+7 3.476580+09134 3 2 18 1.700000+7 3.637010+0 1.800000+7 3.788450+0 1.900000+7 3.917850+09134 3 2 19 2.000000+7 4.019540+0 9134 3 2 20 0.000000+0 0.000000+0 0 0 0 09134 3 099999 9.123200+4 2.300450+2 0 0 0 09134 3 4 1 0.000000+0-4.999990+4 0 0 1 369134 3 4 2 36 3 9134 3 4 3 5.021730+4 0.000000+0 7.000000+4 5.375490-2 1.000000+5 1.715890-19134 3 4 4 1.100000+5 2.143010-1 1.200000+5 2.578450-1 1.500000+5 3.881160-19134 3 4 5 1.800000+5 5.055980-1 2.000000+5 5.718650-1 2.500000+5 6.892510-19134 3 4 6 3.000000+5 7.672190-1 4.000000+5 8.992360-1 5.000000+5 1.001640+09134 3 4 7 6.000000+5 1.073600+0 7.000000+5 1.124250+0 8.000000+5 1.179130+09134 3 4 8 9.000000+5 1.197670+0 1.000000+6 1.225040+0 2.000000+6 1.707470+09134 3 4 9 3.000000+6 1.752350+0 4.000000+6 1.775090+0 5.000000+6 1.758010+09134 3 4 10 6.000000+6 1.342940+0 7.000000+6 3.454040-1 8.000000+6 8.610000-29134 3 4 11 9.000000+6 2.479920-2 1.000000+7 9.764760-3 1.100000+7 6.162670-39134 3 4 12 1.200000+7 5.326760-3 1.300000+7 5.012310-3 1.400000+7 4.997520-39134 3 4 13 1.500000+7 4.973620-3 1.600000+7 4.966510-3 1.700000+7 4.939580-39134 3 4 14 1.800000+7 4.922910-3 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