92-U -235 JAEA+ EVAL-OCT09 O.Iwamoto,N.Otuka,S.Chiba,+ DIST-DEC21 20111206 ----JENDL-5 MATERIAL 9228 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 07-07 Calculation with CCONE code was performed. 07-09 Fission spectra up to 5 MeV were replaced with JENDL-3.3. 07-11 Fission cross section was revised with simultaneous evaluation. 07-12 Fission cross section was revised with new results of simultaneous evaluation. 08-01 Fission cross section was revised. New CCONE calculation was adopted. 08-02 Fission and capture cross sections, and nu-p were revised. CCONE calculation was made with revised parameters. Data were compiled as JENDL/AC-2008/1/. 09-08 (MF1,MT458) was evaluated. 09-10 nu-p and fission cross section were revised. 10-03 Covarinace data were given. 11-07 Covariance data in RRR were revised. 18-09 JENDL-5a (MF2/MT151) RRP from ENDF/B-VIII.0 (MF3/MT18) preliminary SOK result (MF3/MT19,20,21,38) renormalized 19-12 JENDL-5a2 (MF2/MT151) RPs of neg. and second resonances were modified (MF3/MT18) SOK result in 2019 (MF3/MT19,20,21,38) renormalized 20-03 JENDL-5a3 (MF2/MT151) AWR was changed to the same vale as others (MF5/MT18) ENDF/B-VIII.0 was adopted (En < 5 MeV) (MF31/MT452) AWR was changed to the same vale as others 20-07 (MT1/MF456) revised (MT3/MF18) IAEA standard was adopted (2.25-10 keV) (MF3/MT19,20,21,38) deleted 20-10 JENDL-5a4 (MF3/MT18) SOK(20201009) was adopted above 10 keV 21-06 JENDL-5b1 (MF1/MT456) revised below 700 keV (MF2/MT151) RRP from ENDF/B-VIII.0 (MF3/MT18) SOK(20210404) was adopted above 10 keV (MF5/MT18) new evaluation below 5 MeV 21-07 JENDL-5b2 (MF2/MT151) adjuisted RRP above 100 eV (MF3/MT1) sum of partial (MF3/MT2) adjusted (MF3/MT18) adjusted (MF5/MT102) adjusted 21-11 revised by O.Iwamoto (MF8/MT16-18,37,102) JENDL/AD-2017 adopted (MF8/MT4) added 21-11 above 20 MeV, JENDL-4.0/HE merged by O.Iwamoto 21-11 (MF6/MT5) recoil spectrum added by O.Iwamoto 21-12 (MF33/MT1,2,18,102) modified by O.Iwamoto ENDF/B-VIII.0 (En < 2.25 keV) 21-12 (MF33/MT18) modified by O.Iwamoto IAEA neutron data standards 2017 (2.25 keV < En < 10 keV) SOK(20210404,rectangle) (En > 10 keV) 21-12 (MF3/MT18) above 20 MeV SOK(20210404) adoped by O.Iwamoto 21-12 (MF1/MT452,456) by O.Iwamoto revise nu-p (En > 20 MeV) by a systematics (MF6/MT5) by O.Iwamoto modify multiplicity of ZAP=1 to compensate the revision of nu-p MF= 1 MT=452 Total number of neutrons per fission Sum of MT=455 and 456. MT=455 Delayed neutron data (same as JENDL-3.3/2/) Evaluated by using the least-squares method on the basis of the following experimental data in each energy region. Thermal region: Keepin/3/, Conant/4/, Synetos/5/, Reeder/6/, Borzakov/7/ 50 keV - 7 MeV: Keepin/3/, Maksyutenko/8/, Masters/9/, Krick/10/, Evans/11/, Cox/12/, Besant/13/, Gudkov/14/, Loaiza/15/ 14 - 15 MeV : Keepin/16/ Decay constants at the thermal energy were adopted from Keepin et al./17/ MT=456 Number of prompt neutrons per fission Below 500 eV, JENDL-3.3 was adopted, which was evaluated on the basis of experimental data of Gwin et al./18,19/ Above 500 eV, experimental data were anlyzed by the GMA code /20/ with Chiba-Smith approach/21/ for PPP minimization. Experimental data are renormalized with nu-p of CF-252 spontaneous fission (3.756+/-0.031) reported by Vorobyev et al./22/ if standards to derive original data were known. Experimental data sets are summarized below. r: re-normalized by nu-p(252Cf spon) of A.S.Vorobyev et al. -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- r 12326.004 2.80E+5 - 1.45E+7 J.C.Hopkins+ /23/ r 12870.004 1.70E+7 - 1.96E+7 R.E.Howe /24/ r 13101.003 5.00E+2 - 9.00E+6 R.Gwin+ /25/ r 20427.002 2.25E+5 - 1.36E+6 F.Kaeppeler+ /26/ 21696.004 2.50E+6 - 1.41E+7 I.Johnstone /27/ r 21785.003 1.14E+6 - 1.47E+7 J.Frehaut+ /28/ r 40262.002 8.60E+5 - 5.35E+6 M.V.Savin+ /29/ r 40493.002 1.98E+5 - 9.85E+5 M.V.Savin+ /30/ 40785.002 1.43E+7 Ju.A.Vasilev+ /31/ -------------------------------------------------------------- In the energy region from 1 keV to 1 MeV, GMA results were increased by 0.2%. MT=458 Components of energy release due to fission Total energy and prompt energy were calculated from mass balance using JENDL-4 fission yields data and mass excess evaluation. Mass excess values were from Audi's 2009 evaluation/32/. Delayed energy values were calculated from the energy release for infinite irradiation using JENDL FP Decay Data File 2000 and JENDL-4 yields data. For delayed neutron energy, as the JENDL FP Decay Data File 2000/33/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/34/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/35/ when they were provided. When the fractions were not given by Sher, averaged fractions were used. MF= 2 Resonance parameters MT=151 Resolved resonance parameters (RM: 1.0E-5 - 500 eV) Adopted are parameters for Reich-Moore formula evaluated by Leal et al./36/ In the present file, the upper boundary of resolved resonance region is set to 500 eV. See Appendix A-1 Unresolved resonance parameters (500 eV - 30 keV) Unresolved resonance parameters were determined with ASREP code/37/ so as to reproduce the cross sections. The parameters are used only for calculation of self-shielding factors. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 698.90 elastic 15.12 fission 585.08 274 capture 98.71 139 ------------------------------------------------------- (*) In the energy range from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections Between 500 eV and 2.25 keV: Cross sections were calculated with resonance parameters of JENDL-3.3 which were taken from ENDF/B-VI.5/36/ and broadened with a resolution function of R(E)=0.03*E. The capture cross section was multiplied by ratios of average capture cross sections of JENDL-3.2 and JENDL-3.3 to lower the cross sections to those of JENDL-3.2. Above 2.25 keV: Cross sections except for the total (M=1), elastic scattering (MT=2), fission (MT=18, 19, 20, 21, 38) and capture cross sections were calculated with CCONE code/38/. The model parameters were determined by considering integral experimental data as well as measured cross-section data. MT= 1 Total cross section In the energy range from 2.25 to 500 keV, cross section was calculated with CCONE code/38/. Above 500 keV, the cross section was determined by spline fitting to experimental data of Schwartz et al./39/, Poenitz et al./40,41/, Harvey et al./42/, Cabe and Cance/43/, and Uttley et al./44/ These experimental data were used also for adjustment of the OMP of Soukhovitskii et al./45/ MT=2 Elastic scattering cross section Calculated as total - non-elstic scattering cross section MT=16 (n,2n) cross section Calculated with CCONE code. The experimental data of Frehaut et al./46/ were considered to determine the model parameters of CCONE calculation. MT=18 Fission cross sections From 2.25 keV to 10 keV, JENDL-3.3/2/ was adopted. The data of JENDL-3.3 were based on the experimental data of Weston and Todd/47/. Above 10 keV, experimental data measured after 1980 were analyzed by simultaneous fitting of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241 fission cross section and its ratio by the SOK code /48/. -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 41112.002 1.88E+6 - 2.37E+6 V.A.Kalinin+ /49/ 22304.006 2.60E+6 - 1.47E+7 K.Merla+ /50/ 22304.002 4.45E+6 - 1.88E+7 K.Merla+ /50/ 12924.002 1.07E+6 - 5.99E+6 R.G.Johnson+ /51/ 40969.011 6.24E+5 - 7.85E+5 N.N.Buleeva+ /52/ 22091.002 1.35E+7 - 1.49E+7 T.Iwasaki+ /53/ 30721.002 1.42E+7 J.W.Li+ /54/ 12877.004 5.05E+3 - 2.05E+5 L.W.Weston+ /47/ 10987.002 3.10E+5 - 2.82E+6 A.D.Carlson+ /55/ 30634.002 1.47E+7 J.W.Li+ /56/ 12826.002 1.46E+7 M.Mahdavi+ /57/ 10971.002 1.41E+7 O.A.Wasson+ /58/ 21620.002 2.50E+6 - 4.45E+6 M.Cance+ /59/ 21777.002 5.40E+3 - 8.25E+4 F.Corvi+ /60/ 10950.002 2.45E+5 - 1.20E+6 O.A.Wasson+ /61/ 40601.002 6.00E+3 - 4.50E+4 A.A.Bergman+ /62/ -----.--- 3.03E+6 - 2.96E+7 A.D.Carlson+ /63/ -------------------------------------------------------------- The obtained cross section from 1 to 4 MeV and from 7 to 8 MeV was slightly modified for JENDL-4. The fission cross section of JENDL-3.3 were used to determine the parameters in the CCONE calculation. MT=19, 20, 21, 38 Multi-chance fission cross sections Calculated with CCONE code, and renormalized to the total fission cross section (MT=18). MT=102 Capture cross section From 2.25 keV to 1 MeV, experimental data measured after 1970 were analyzed by the GMA code/20/ with the Chiba and Smith approach/21/ for PPP minimization/64/. All experimental data are given in the form of alpha-value (=ratios to the U-235(n,f) cross section), which were normalized to absolute cross section by the JENDL-3.3 U-235(n,f) cross section. Data points of Kononov et al./65/ below 20 keV were not considered because their systematic error was very large in this energy region. Experimental data sets are summarized below. -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 12424.003 3.47E+3 - 2.56E+4 J.B.Czirr+ /66/ 20158.002 8.50E+3 - 5.50E+4 R.E.Bandl+ /67/ 20880.002 1.06E+4 - 1.96E+5 H.Beer+ /68/ 20880.003 1.04E+5 - 3.07E+5 H.Beer+ /68/ 20880.004 1.70E+4 - 6.40E+4 H.Beer+ /68/ 20880.005 3.79E+5 - 4.81E+5 H.Beer+ /68/ 21777.004 2.50E+3 - 8.25E+4 F.Corvi+ /60/ 40412.002 2.04E+4 - 1.10E+6 V.N.Kononov+ /65/ 40502.002 3.49E+3 - 8.89E+3 Ju.V.Ryabov /69/ 40581.002 2.50E+3 - 4.50E+4 G.V.Muradyan+ /70/ 40609.004 2.45E+4 V.P.Vertebnyy+ /71/ 12409.003 2.00E+5 - 6.00E+5 G.de Saussure+ /72/ 12407.002 1.23E+4 - 6.90E+5 L.W.Weston+ /73/ 12331.005 3.00E+4 - 1.00E+6 J.C.Hopkins+ /74/ 12416.002 1.78E+5 - 1.01E+6 B.C.Diven+ /75/ -------------------------------------------------------------- The results of the GMA were used to determine the parameters in the CCONE calculation. Above 1.3 MeV, results of the CCONE calculation were adopted. The present results were slightly modified by considering integral data. MF= 4 Angular distributions of secondary neutrons MT=2 Elastic scattering Calculated with CCONE code. MT=18 Fission Isotropic distributions in the laboratory system were assumed. MF=5 Energy Distributions of Secondary Neutrons MT=18 Prompt fission neutron spectra Below 5 MeV, spectra given in JENDL-3.3/2/ were adopted. Relevant part of JENDL-3.3 comments: *DISTRIBUTIONS WERE CALCULATED WITH A MODIFIED MADLAND-NIX MODEL WITH CONSIDERATION FOR MULTIMODAL NATURE OF THE FISSION PROCESS/76,77/. THE COMPOUND NUCLEUS FORMATION CROSS SEC- TIONS FOR FISSION FRAGMENTS WERE CALCULATED USING BECCHETTI- GREENLEES POTENTIAL/78/. THE IGNATYUK FORMULA/79/ WERE USED TO GENERATE THE LEVEL DENSITY PARAMETERS. UP TO 3rd-CHANCE-FISSION WERE CONSIDERED AT HIGH INCIDENT NEUTRON ENERGIES. PARAMETERS ADOPTED FOR THERMAL-NEUTRON FISSION: (S1: STANDARD-1, S2: STANDARD-2, SL: SUPERLONG MODES) TOTAL AVERAGE FRAGMENT KINETIC ENERGY = 187 MEV FOR S1 = 167 MEV FOR S2 = 157 MEV FOR SL AVERAGE ENERGY RELEASE = 194.49 MEV FOR S1 = 184.86 MEV FOR S2 = 190.95 MEV FOR SL AVERAGE MASS NUMBER OF LIGHT FF = 96 AVERAGE MASS NUMBER OF HEAVY FF = 140 LEVEL DENSITY OF THE LIGHT FF = 10.31(S2), 11.43(S1) LEVEL DENSITY OF THE HEAVY FF = 8.89(S1), 13.25(S2) MODE BRANCHING RATIO = 0.18342(S1), 0.81589(S2), 0.00069(SL) NOTE THAT THE PARAMETERS VARY WITH THE INCIDENT ENERGY WITHIN THE INDICATED RANGE. Above 5.5 MeV, calculated with CCONE code/38/. MT=455 Delayed neutron spectra (same as JENDL-3.3) Taken from Brady and England/80/. Group abundace parameters were adjusted so as to reproduce total delayed neutron emission rate measured by Keepin/17/, Piksaikin/81/ and East/82/. MF= 6 Energy-angle distributions Calculated with CCONE code. Distributions from fission (MT=18) are not included. MF=12 Photon Production Multiplicities (option 1) MT=18 Fission (same as JENDL-3.3) The thermal neutron-induced fission gamma spectrum measured by Verbinski et al./83/ was adopted. MF=14 Photon Angular Distributions MT=18 (same as JENDL-3.3) Isotropic distributions were assumed. MF=15 Continuous Photon Energy Spectra MT=18 (same as JENDL-3.3) Experimental data by Verbinski et al./83/ were adopted. MF=31 Covariances of average number of neutrons per fission MT=452 Number of neutrons per fission Combination of covariances for MT=455 and MT=456. MT=455 (Same as JENDL-3.3/2/) MT=456 Below 500 eV, the covariance of JENDL-3.3 was adopted. Above 500 eV, it was obtained by fitting to the experimental data described above. The error of nu-p was multiplied by a factor of 2.0. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/85/ and the covariances of model parameters used in the theoretical calculations. For the following cross sections, covariances were determined by different methods. MT= 1, 2 Total and elastic scattering cross sections Below 500 eV, covariance matrices was calculated from those of resonance parameters/84/. In the energy range from 200 to 500 eV, uncertainty of 2% was assumed for the total cross section, and 4% for the elastic scattering. From 0.5 to 2.25 keV, uncertainty of 5% was assumed. Above 2.25 keV, covariance of the CCONE calculation was adopted. MT=18 Fission cross section Below 500 eV, covariance matrices was calculated from those of resonance parameters/84/. In the energy range from 200 to 500 eV, uncertainty of 2% was assumed. From 500 eV to 9 keV, uncertainty was assumed to be 5%. Above 9 keV, covariance matrix was obtained by simultaneous evaluation among the fission cross sections of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241(See MF=3, MT=18, and /1/). Since the variances are very small, they were adopted by multiplying a factor of 2. MT=102 Capture cross section Below 500 eV, covariance matrices was calculated from those of resonance parameters/84/. In the energy range from 200 to 500 eV, uncertainty of 3% was assumed. From 500 eV to 2.25 keV, uncertainty was assumed to be 10%. In the energy region from 2.25 keV to 1 MeV, capture cross section was obtained with GMA code/64/. Its covariance matrix was obtained simultaneously. Above 1 MeV, covariance of the CCONE calculation was adopted. MF=34 Covariances for Angular Distributions MT=2 Elastic scattering Covariances were given only to P1 components. MF=35 Covariances for Energy Distributions MT=18 Fission spectra Below 5 MeV, based on the covarinaces given in JENDL-3.3. Above 5 MeV, estimated with CCONE and KALMAN codes. ***************************************************************** Calculation with CCONE code ***************************************************************** Models and parameters used in the CCONE/38/ calculation 1) Coupled channel optical model Levels in the rotational band were included. Optical model potential and coupled levels are shown in Table 1. 2) Two-component exciton model/86/ * Global parametrization of Koning-Duijvestijn/87/ was used. * Gamma emission channel/88/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/89/ was included. * Neutron, gamma and fission decay channel were included. * Transmission coefficients of neutrons were taken from coupled channel calculation in Table 1. * The level scheme of the target is shown in Table 2. * Level density formula of constant temperature and Fermi-gas model were used with shell energy correction and collective enhancement factor. Parameters are shown in Table 3. * Fission channel: Double humped fission barriers were assumed. Fission barrier penetrabilities were calculated with Hill-Wheler formula/90/. Fission barrier parameters were shown in Table 4. Transition state model was used and continuum levels are assumed above the saddles. The level density parameters for inner and outer saddles are shown in Tables 5 and 6, respectively. * Gamma-ray strength function of Kopecky et al/91/,/92/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,3,6,9 (see Table 2) * optical potential parameters /45/ Volume: V_0 = 49.8613 MeV lambda_HF = 0.01004 1/MeV C_viso = 15.9 MeV A_v = 12.04 MeV B_v = 81.36 MeV E_a = 385 MeV r_v = 1.2568 fm a_v = 0.61 fm Surface: W_0 = 17.1117 MeV B_s = 11.19 MeV C_s = 0.01361 1/MeV C_wiso = 23.5 MeV r_s = 1.1803 fm a_s = 0.61 fm Spin-orbit: V_so = 5.75 MeV lambda_so = 0.005 1/MeV W_so = -3.1 MeV B_so = 160 MeV r_so = 1.1214 fm a_so = 0.59 fm Coulomb: C_coul = 1.3 r_c = 1.2452 fm a_c = 0.545 fm Deformation: beta_2 = 0.201044 beta_4 = 0.11 beta_6 = 0.0015 * Calculated strength function S0= 0.93e-4 S1= 2.11e-4 R'= 9.50 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of U-235 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 7/2 - * 1 0.00008 1/2 + 2 0.01304 3/2 + 3 0.04621 9/2 - * 4 0.05171 5/2 + 5 0.08174 7/2 + 6 0.10304 11/2 - * 7 0.12930 5/2 + 8 0.15047 9/2 + 9 0.17071 13/2 - * 10 0.17139 7/2 + 11 0.19712 11/2 + 12 0.22542 9/2 + 13 0.24913 15/2 - 14 0.25900 7/2 + 15 0.29114 11/2 + 16 0.29467 13/2 + 17 0.33285 5/2 + 18 0.33852 17/2 - 19 0.35730 15/2 + 20 0.36707 7/2 + 21 0.36900 13/2 + 22 0.39322 3/2 + 23 0.41478 9/2 + 24 0.42675 5/2 + 25 0.43860 19/2 - 26 0.44572 7/2 + 27 0.45450 15/2 + 28 0.47130 11/2 + 29 0.47430 7/2 + 30 0.48500 17/2 + 31 0.50992 9/2 + 32 0.53240 13/2 + 33 0.53323 9/2 + 34 0.55040 21/2 - 35 0.56800 19/2 + 36 0.58780 11/2 + 37 0.60808 11/2 + 38 0.61640 15/2 + 39 0.63317 5/2 - 40 0.63781 3/2 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-236 18.6157 1.5623 2.7551 0.3856 -0.1537 3.8909 U-235 18.4419 0.7828 2.6265 0.3721 -0.7434 2.8828 U-234 18.4800 1.5689 2.5578 0.3899 -0.1502 3.9076 U-233 18.4122 0.7861 2.4694 0.3817 -0.8188 2.9881 U-232 18.3442 1.5757 2.6095 0.3885 -0.1133 3.8795 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- U-236 6.201 1.040 5.417 0.550 U-235 5.700 0.400 5.600 0.300 U-234 6.050 1.040 5.400 0.600 U-233 5.970 0.800 5.450 0.520 U-232 5.800 1.040 5.100 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-236 20.9712 1.8226 2.6000 0.3244 -0.5720 3.8226 U-235 20.3497 0.9133 2.6000 0.3515 -1.8195 3.2133 U-234 20.2753 1.8304 2.6000 0.3522 -0.9023 4.1304 U-233 20.2008 0.9172 2.6000 0.3312 -1.4942 2.9172 U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-236 21.5182 1.8226 0.1400 0.3654 0.0498 3.9226 U-235 20.3497 0.9133 0.1000 0.4065 -1.2054 3.4133 U-234 20.2753 1.8304 0.0600 0.3939 -0.1191 4.1304 U-233 20.2008 0.9172 0.0200 0.3732 -0.7817 2.9172 U-232 20.1263 1.8383 -0.0200 0.3745 0.1401 3.8383 -------------------------------------------------------- Table 7. Gamma-ray strength function for U-236 -------------------------------------------------------- K0 = 1.500 E0 = 4.500 (MeV) * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 303.18 (mb) ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb) * M1: ER = 6.63 (MeV) EG = 4.00 (MeV) SIG = 2.69 (mb) * E2: ER = 10.19 (MeV) EG = 3.28 (MeV) SIG = 6.51 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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C33, 818 (1986). 87) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 88) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 89) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 90) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 91) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 92) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). ***************************************************************** Appendix A-1 from ENDF/B-VI.8 ***************************************************************** File 2 MT=151 Resonance parameters, from a new analysis by Leal et al. [LE97], using the multilevel R-matrix analysis code SAMMY [LA96]. Energy range for U235 is 0 to 2.25 keV. For the first time, integral data were fitted during the analysis process: Thermal cross sections (fission, capture, and elastic), Westcott g-factors (fission and absorption) are from the ENDF/B-6 standards [CA93], and the K1 value is from Hardy [HA79]. Thermal parameters obtained in the present evaluation, first using the microscopic experimental data only, and second including the integral data as well, are compared to the SAMMY input in the following Tabld: Parameter SAMMY input Fit to Fit to diff. value diff data & integ. alone data ----------- ----------------- -------- ------------ Fission 584.25 +/- 1.11 582.28 584.88 Capture 98.96 +/- 0.74 99.18 98.66 Scattering 15.46 +/- 1.06 15.44 15.12 Westcott gf 0.9771 +/- 0.0008 0.9743 0.9764 Westcott ga 0.9790 +/- 0.0008 0.9774 0.9785 Westcott gg 0.9956 0.9910 K1(barn) 722.70 +/- 3.90 717.48 722.43 The final adjustment of nu by SAMMY to the recommended K1 value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission and absorption cross sections calculated from the final resonance parameters. In the following Tables, the fission and capture cross sections calculated in this evaluation with the code SAMMY are compared with experimental data. Experimental and calculated total cross sections. Energy Range Calculated Schrack Weston Weston (eV) (b.eV) (b.eV) (b.eV) (b.eV) --------------- ---------- ------- ------ ------ 0.5 - 20.0 910.4 929.9 20.0 - 60.0 1867.8 1882.8 1869.9 60.0 - 100.0 954.0 968.0 954.2 100.0 - 200.0 2032.7 2092.7 2089.5 2073.9 200.0 - 300.0 2062.2 2007.0 2060.0 2054.6 300.0 - 400.0 1280.8 1321.6 1297.1 1292.9 400.0 - 500.0 1333.2 1391.5 1351.8 1347.9 500.0 - 600.0 1489.2 1467.9 1499.2 1494.3 600.0 - 700.0 1126.6 1156.4 1134.1 1132.6 700.0 - 800.0 1088.7 1085.8 1093.3 1075.7 800.0 - 900.0 797.6 784.0 813.0 804.9 900.0 - 1000.0 724.4 723.9 738.2 721.4 1000.0 - 2000.0 7036.1 7054.2 Experimental and calculated capture cross sections. Energy Range Calculated De Saussure Perez (eV) (b.eV) (b.eV) (b.eV) --------------- ---------- ----------- ------ 0.5 - 20.0 653.5 647 20.0 - 60.0 1066.1 1084 1057 60.0 - 100.0 490.2 477 504 100.0 - 200.0 1158.8 1148 1138 200.0 - 300.0 907.8 904 940 300.0 - 400.0 660.2 658 642 400.0 - 500.0 495.9 506 478 500.0 - 600.0 533.3 506 562 600.0 - 700.0 494.8 481 449 700.0 - 800.0 490.1 513 475 800.0 - 900.0 439.8 444 397 900.0 - 1000.0 504.2 542 482 1000.0 - 1100.0 509.6 522 463 1100.0 - 1200.0 413.7 395 332 1200.0 - 1300.0 340.4 372 267 1300.0 - 1400.0 304.1 304 225 1400.0 - 1500.0 355.7 301 254 --------------- ---------- ----------- ------ 20.0 - 1500.0 9164.7 9046 8665 The fission and capture resonance integral calculated from the present evaluation are 276.04 b and 140.49 b, respectively, giving a capture-to-fission ratio (alpha value) of 0.509 in excellent agreement with the value obtained from integral measurements. The following energy-differential data were included in the analysis: (1) Transmission data of Harvey et al. [HA86] on the ORELA 18-meter flight path, with sample thickness of 0.03269 atoms/barn, cooled to 77 K (0.4 to 68 eV). (2) Transmission data of Harvey et al. [HA86] on the ORELA 80-meter flight path, with sample thickness of 0.00233 atoms/barn, cooled to 77 K (4 to 2250 eV). (3) Transmission data of Harvey et al. [HA86] on the ORELA 80-meter flight path, with sample thickness of 0.03269 atoms/barn, cooled to 77 K (4 to 2250 eV). (4) Fission data of Schrack [SC88] on the RPI Linac at 8.4 meter flight path (0.02 to 20 eV). (5,6) Fission and capture data of de Saussure et al. [DE67] on the ORELA 25.2-meter flight path (0.01 to 2250 eV). (7,8) Fission and capture data of Perez et al. [PE73] on the ORELA 39-meter flight path (0.01 to 100 eV). (9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter flight path (0.01 to 20 eV). (10) Transmission data of Spencer et al. [SP84] on the ORELA ORELA 18-meter flight path, sample thickness of 0.001468 atom/barn (0.01 to 1.0 eV). (11) Fission data of Wagemans et al. [WA88] on the Geel 18- meter flight path (0.001 to 1.0 eV) (12,13) Absorption and fission data of Gwin [GW96] at ORELA (0.01 to 4.0 eV). (14) Fission data of Weston and Todd [WE84] on the ORELA 18.9-meter flight path (14 to 2250 eV). (15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to 1.0 eV). (16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to 0.15 eV). (17) Fission data of Weston and Todd [WE92] on the ORELA 86.5-meter flight path (100 to 2000 eV). (18) Fission yield data of Moxon et al. [MO92] at ORELA (0.01 to 50.0 eV). ---------------------------------------------------------------- REFERENCES FOR RESOLVED RESONANCE REGION [CA93] A. Carlson, W.P. Poenitz, G.M. 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