92-U -233 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,et al. DIST-DEC21 20111206 ----JENDL-5 MATERIAL 9222 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-07 Total and fission cross sections were modified. 06-10 Nu-p was modified. 07-06 Theoretical calculation was made with CCONE code. 07-08 Theoretical calculation was made with CCONE code. 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. 08-02 Fission cross section and nu-p were revised. CCONE calculation was made with revised parameters. 08-03 Interpolation of (5,18) was changed. Data were compiled as JENDL/AC-2008/1/. 09-04 MF01 was revised. 09-08 (MF1,MT458) was evaluated. 09-10 fission cross section was revised. 10-01 Data of prompt gamma rays due to fission were given. 10-03 Covariance data were given. 11-07 Covariance data in RRR were revised. 18-08 JENDL-5a (MF3/MT18) preliminary SOK result (MF3/MT19,20,21,38) renormalized 19-12 JENDL-5a2 (MF3/MT18) SOK-2019 result was adopted (MF3/MT19,20,21,38) renormalized 20-03 JENDL-5a3 (MF2/MT151) AWR was changed to the same vale as others (MF3/MT1) recalculated 20-10 JENDL-5a4 (MF3/MT18) SOK(20201009) was adopted (MF3/MT19,20,21,38) deleted 21-06 JENDL-5b1 (MF1/MT456) revised for 1-3 MeV (MF3/MT18) SOK(20210404) was adopted above 10 keV 21-11 revised by O.Iwamoto (MF8/MT4,16-18,37,102) added 21-12 (MF33/MT18) SOK(20210404,rectangle) adopted by O.Iwamoto MF= 1 General information MT=452 Number of Neutrons per fission Sum of MT=455 and 456. MT=455 Delayed neutron data At the thermal neutron energy, nu-d of 0.0067 was obtained from the data of Borzakov et al./2/, Conant et al./3/ and Keepin et al./4/. Nu-d above 10 keV was determined from experimental data measured by Krick and Evans/5,6/, Piksaykin et al./7/, and Masters et al./8/. Decay constants were taken from Ref./9/. MT=456 Number of prompt neutrons per fission Nu-p was determined from the data of Protopopov et al./10/, Smirenkin et al./11/, Flerov et al./12/, Hopkins et al./13/, Colvin et al./14,15/, Mather et al./16/, Walsh et al./17/, Hockenbury/18/, Nurpeisov et al./19,20/, Sergachev et al./21/, Nefedov et al./22/, and Gwin et al./23,24/ Nu-p of Cf-252 SF = 3.756+-0.031 /25/ was used. They were reproduced with two straight lines below and above about 1.5 MeV. 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/26/. 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 2000/27/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/28/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/29/ 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: 1.0-5 eV - 600 eV) Evaluation by Leal et al. for ENDF/B-VII.0 was adopted. See Appendix A1. Unresolved resonance parameters (600 eV - 30 keV) Parameters were determined with ASREP code/30/ so as to reproduce total, fission and capture cross sections. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 588.77 elastic 12.18 fission 531.34 775 capture 45.26 139 ------------------------------------------------------- (*) In the energy range from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections Cross sections above the resolved resonance region except for the total (MT=1), elastic scattering (MT=2) and fission cross sections (MT=18, 19, 20, 21, 38) were calculated with CCONE code/31/. The model parameters were determined by considering integral experimental data as well as measured cross-section data. In the CCONE calculation the CC OMP of Soukhovitskii et al./32/ was modified so as to reproduce well the experimental data of total cross section measured by Poenitz et al./33,34/ and Guber et al./35/ Other parameters of CCONE calculation were adjusted to the fission cross section of JENDL-3.3 and the capture cross section measured by Hopkins and Diven/36/. The results of CC calculation for the elastic scattering was increased by 0.2 b above 1.5 MeV to improve integral benchmark tests. MT=1 Total cross section Experimental data measured after 1960 were anlyzed by the GMA code/37/ with the Chiba and Smith approach/38/ for PPP minimization. Experimental data sets are summarized below. -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 10047.095 2.26E+6 - 1.50E+7 D.G.Foster Jr.+ /39/ 10225.028 5.05E+5 - 7.96E+6 L.Green+ /40/ 10025.028 9.00E+5 - 9.89E+6 L.Green+ /40/ 10833.003 4.00E+4 - 2.09E+5 W.P.Poenitz+ /41/ 10833.002 5.80E+4 - 4.43E+6 W.P.Poenitz+ /41/ 10935.005 4.80E+4 - 4.81E+6 W.P.Poenitz+ /33/ 12323.002 3.40E+3 - 1.61E+6 D.C.Stupegia /42/ 12333.002 6.01E+2 - 8.81E+3 N.J.Pattenden+ /43/ 12853.053 1.82E+6 - 2.03E+7 W.P.Poenitz+ /34/ 13891.004 6.09E+2 - 6.85E+5 K.H.Guber+ /35/ -------------------------------------------------------------- MT=2 Elastic scattering cross section Calculated as total - non-elstic scattering cross sections MT=18 Fission cross section Below 10 keV, experimental data measured after 1960 were anlyzed by the GMA code/37/ with the Chiba and Smith approach/38/ for PPP minimization. Data were normalized to absolute cross section by adopting the JENDL-3.3 U-235(n,f) cross section if the data were given as the ratios to the U-235(n,f) cross section. Above 10 keV, experimental data measured after 1960 were analyzed by simultaneous fitting of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241 fission cross sections and their ratio by the SOK code/44/. Covariance matrix reported in Manabe et al./45/ was also considered in the analysis. Experimental data sets are summarized below. g: used in GMA analysis, s: used in SOK analysis -------------------------------------------------------------- Cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- gs 13890.002 6.00E+2 - 7.01E+5 K.H.Guber+ /46/ gs 40927.002 1.94E+6 V.I.Shpakov /47/ gs 12910.002 1.46E+7 K.R.Zasadny+ /48/ s 40911.003 1.47E+7 I.D.Alkhazov+ /49/ s 40610.002 4.40E+4 E.A.Zhagrov+ /50/ s 40587.002 2.45E+4 A.V.Murzin+ /51/ gs 10756.002 1.37E+5 - 8.05E+6 W.P.Poenitz /52/ gs 40547.003 1.48E+7 V.M.Adamov+ /53/ gs 32625.002 5.00E+5 - 1.00E+6 W.G.Yan+ /54/ s 20446.002 5.00E+3 - 3.00E+4 S.Nizamuddin+ /55/ g 20003.005 6.00E+2 - 3.00E+3 M.G.Cao+ /56/ gs 10056.002 6.00E+2 - 9.87E+3 D.W.Bergen /57/ s 30035.003 1.41E+7 R.H.Iyer+ /58/ s 10267.041 7.50E+3 - 8.50E+4 R.Gwin+ /59/ g 10056.002 6.41E+4 - 2.85E+6 D.W.Bergen /57/ g 12360.002 6.00E+2 - 9.78E+5 D.W.Bergen+ /60/ g 21463.002 4.00E+4 - 5.05E+5 P.H.White+ /61/ g 40650.002 2.80E+5 - 2.63E+6 G.N.Smirenkin+ /62/ g 12341.002 6.13E+2 - 9.60E+2 M.S.Moore+ /63/ -------------------------------------------------------------- Ratio to U-235(n,f) cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- gs 41455.002 5.77E+5 - 9.75E+6 O.A.Shcherbakov+ /64/ gs 41455.002 1.01E+7 - 1.94E+7 O.A.Shcherbakov+ /64/ gs 13134.004 1.47E+7 J.W.Meadows+ /65/ g 41432.003 2.00E+4 - 6.38E+6 D.L.Shpak /66/ gs 22014.003 4.90E+5 - 6.97E+6 K.Kanda+ /67/ gs 40607.002 6.42E+5 - 8.25E+5 D.L.Shpak+ /68/ gs 40474.002 2.40E+4 - 7.40E+6 B.I.Fursov+ /69/ g 40474.002 1.27E+5 - 7.00E+6 B.I.Fursov+ /69/ s 40361.003 1.50E+4 - 1.94E+6 D.L.Shpak+ /70/ gs 10236.002 1.42E+5 - 9.37E+6 J.W.Meadows /71/ g 10562.003 8.50E+2 - 1.95E+7 G.W.Carlson+ /72/ gs 20363.002 5.20E+3 - 1.01E+6 E.Pfletschinger+ /73/ g 10084.003 6.60E+2 - 2.40E+4 W.K.Lehto+ /74/ g 40309.003 4.85E+5 - 2.51E+6 V.G.Nesterov+ /75/ g 40027.004 3.30E+5 - 2.58E+6 G.N.Smirenkin+ /76/ s 22282.003 1.35E+7 - 1.49E+7 F.Manabe+ /45/ -------------------------------------------------------------- The obtained cross section in the energy range from 1 to 4 MeV and from 7 to 8 MeV was slightly modified for JENDL-4. MT=19, 20, 21, 38 Multi-chance fission cross sections Calculated with CCONE code, and renormalized to the total fission cross section (MT=18). 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 spectrum Below 5 MeV, data of JENDL-3.3/77/ were adopted. Comment of JENDL-3.3: * Distributions were calculated with the modified Madland-Nix model/78,79/. The compound nucleus formation cross sections for fission fragments (FF) were calculated using Becchetti-Greenlees potential/80/. Up to 4th-chance-fission were considered at high incident neuttron energies. The Ignatyuk formula/81/ were used to generate the level density parameters. Parameters adopted: Total average FF kinetic energy = 172.311-0.0212*E(MeV) Average energy release = 188.438 MeV Average mass number of light FF = 95 Average mass number of heavy FF = 139 Level density of the light FF = 9.999- 10.094 Level density of the heavy FF = 11.89 - 12.20 Note that the parameters vary with the incident energy within the indicated range. Above 5.5 MeV, the distributions were calculated with CCONE code/31/. MT=455 Delayed neutron spectrum. Summation calculation made by Brady and England/9/ was adopted. MF= 6 Energy-angle distributions Calculated with CCONE code. Distributions from fission (MT=18) are not included. MF=12 Photon production multiplicities MT=18 Fission Calculated from the total energy released by the prompt gamma-rays due to fission given in MF=1/MT=458 and the average energy of gamma-rays. MF=14 Photon angular distributions MT=18 Fission Isotoropic distributions were assumed. MF=15 Continuous photon energy spectra MT=18 Fission Experimental data measured by Verbinski et al./82/ for U-235 thermal fission 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 Nu-d at 0.0253eV was estimated as 0.00067+-0.0002 from experimental data/2,3,4/. Error of 3% was adopted. Above 100 eV error of 8% was assumed. MT=456 Covariance of obtained by fitting a stlight line to experimental data (See MF1,MT456). MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/84/ 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 600 eV, covariance matrices was calculated from those of resonance parameters/83/. Above 600 keV, Covariance matrix was obtained with CCONE and KALMAN codes/84/. MT=18 Fission cross section Below 600 eV, covariance matrices was calculated from those of resonance parameters/83/. 600 eV - 9 keV, covariances were obtained by GMA code. 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 600 eV, covariance matrices was calculated from those of resonance parameters/83/. Above 600 keV, Covariance matrix was obtained with CCONE and KALMAN codes/84/. 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/31/ 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/85/ * Global parametrization of Koning-Duijvestijn/86/ was used. * Gamma emission channel/87/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/88/ 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/89/. 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/90/,/91/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,2,3 (see Table 2) * optical potential parameters /32/ Volume: V_0 = 50.1895 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 = 1.25 fm a_v = 0.63 fm Surface: W_0 = 16.2027 MeV B_s = 11.19 MeV C_s = 0.01361 1/MeV C_wiso = 23.5 MeV r_s = 1.1803 fm a_s = 0.65 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 = 0.205125 beta_4 = 0.076 beta_6 = 0.0015 * Calculated strength function S0= 0.92e-4 S1= 2.11e-4 R'= 9.64 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of U-233 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 5/2 + * 1 0.04035 7/2 + * 2 0.09216 9/2 + * 3 0.15523 11/2 + * 4 0.19700 7/2 + 5 0.22947 13/2 + 6 0.29881 5/2 - 7 0.30194 13/2 - 8 0.31190 3/2 + 9 0.31460 15/2 + 10 0.32083 7/2 - 11 0.33042 7/2 + 12 0.34048 5/2 + 13 0.35379 9/2 - 14 0.38043 7/2 + 15 0.39756 11/2 - 16 0.39850 1/2 + 17 0.41117 17/2 + 18 0.41576 3/2 + 19 0.42500 17/2 + 20 0.43200 9/2 + 21 0.45611 5/2 + 22 0.49700 11/2 + 23 0.50362 7/2 - 24 0.51755 19/2 + 25 0.52200 15/2 - 26 0.54654 5/2 + 27 0.56160 9/2 - 28 0.56700 5/2 - 29 0.57200 1/2 - 30 0.57500 11/2 - ------------------- *) 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-234 18.4623 1.5689 2.5578 0.3903 -0.1513 3.9089 U-233 18.3945 0.7861 2.4694 0.3820 -0.8201 2.9898 U-232 18.3266 1.5757 2.6095 0.3888 -0.1142 3.8806 U-231 18.2588 0.7895 2.6793 0.3781 -0.7806 2.9485 U-230 18.1909 1.5825 2.6739 0.3937 -0.1509 3.9421 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- U-234 6.000 1.040 5.400 0.600 U-233 5.600 0.800 5.400 0.520 U-232 5.500 1.040 5.000 0.600 U-231 6.000 0.800 5.600 0.520 U-230 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-234 21.7235 1.8304 2.6000 0.3321 -0.7507 4.0304 U-233 20.2008 0.9172 2.6000 0.3667 -2.0299 3.4172 U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383 U-231 20.0518 0.9211 2.6000 0.3325 -1.4903 2.9211 U-230 19.9772 1.8463 2.6000 0.3332 -0.5651 3.8463 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- U-234 21.7235 1.8304 0.0600 0.3643 0.0626 3.9304 U-233 19.2990 0.9172 0.0200 0.4202 -1.2096 3.4172 U-232 18.3293 1.8383 -0.0200 0.4264 -0.2156 4.2383 U-231 20.0518 0.9211 -0.0600 0.3758 -0.7765 2.9211 U-230 19.9772 1.8463 -0.1000 0.3771 0.1493 3.8463 -------------------------------------------------------- Table 7. Gamma-ray strength function for U-234 -------------------------------------------------------- K0 = 1.600 E0 = 4.500 (MeV) * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 299.61 (mb) ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb) * M1: ER = 6.65 (MeV) EG = 4.00 (MeV) SIG = 2.81 (mb) * E2: ER = 10.22 (MeV) EG = 3.30 (MeV) SIG = 6.52 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. 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Rev. C41, 1941 (1990). 91) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). Appendix A1: Resolved resonance parameters (ENDF/B-VII.0) ========================================================= MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETER EVALUATIONS L. Leal, H. Derrien, Guber, N. Larson, R. Wright, and R.Spencer (ORNL) January, 2003 The resonance parameter evaluation was done by Leal, Derrien, Guber, Larson, Wright and Spencer [LE01] using the multilevel R-matrix analysis code SAMMY [LA96]. The resolved resonance evaluation were performed in the energy range from 0 to 600 eV. The unresolved resonance region evaluation covers the energy range from 600 eV to 40 keV. The evaluation included high resolution transmission data (GU00), fission cross section data [GU98] measured at the Oak Ridge Electron Linear Accelerator (ORELA), in addition to other experimental data. Integral data were also included in the evaluation. Integral quantities and thermal values calculated with the U-233 resonance parameter are shown in the Table below. Also shown are the results of calculations using U-233 ENDF/B-VI evaluation and Axton standard values: Quantity ENDF/B-VI Axton Present Eval -------- --------- ----- ------------ Fission 531.14 +/- 1.33 530.70 +/- 1.34 530.70 Capture 45.51 +/- 0.68 45.52 +/- 0.70 45.22 Scattering 12.13 +/- 0.66 12.19 +/- 0.67 12.18 Westcott ga 0.9996 +/- 0.0011 0.9995 +/- 0.0011 1.00325 Westcott gf 0.9955 +/- 0.0014 0.9955 +/- 0.0014 1.00045 K1 (barn) 742.60 +/- 2.40 742.25 +/- 0.0040 746.0 The following Table shows the average values of the fission and capture cross sections of the present evaluation compared to the previous ENDF-B6 evaluation in the energy range thermal to 600 eV. Energy Range Fission Capture (eV) Present ENDF/B-VI.5 Present ENDF/B-VI.5 -------------- -------- ----------- ------ ----------- 0.001 -0.020 980.19 971.62 82.03 83.27 0.020 -0.050 462.03 460.02 39.84 40.21 0.050 -0.400 201.88 202.62 20.86 20.36 0.400 - 1.00 127.29 126.91 11.93 9.95 1.0 - 2.10 389.14 378.56 66.35 67.37 2.10 - 2.75 206.76 198.02 111.45 112.00 2.75 - 3.00 49.84 50.46 7.91 7.48 3.00 - 15.0 104.25 101.26 17.66 17.65 15.0 - 30.0 94.72 91.80 13.51 13.27 30.0 - 50.0 40.72 38.85 5.66 5.46 50.0 - 75.0 41.24 41.21 5.69 4.61 75.0 - 100 36.92 33.72 8.94 4.35 100 - 125 38.24 29.94 6.10 3.88 125 - 150 21.11 22.10 3.72 3.54 150 - 200 20.99 21.34 3.06 3.18 200 - 300 23.10 19.87 3.51 2.72 300 - 400 18.28 16.66 2.45 2.33 400 - 500 11.06 13.17 1.39 2.08 500 - 600 13.52 13.40 2.00 1.90 ----------------------------------------------------------------- The fission and capture resonance integral calculated from the present evaluation are 776.64 b and 139.66 b, respectively, which compare to 760 +/ 17 b and 137 +/- 6 b reported by Mughabghab.[MU85] ----- REFERENCES (MF=2) ----- GU98 K. H. Guber et al., Nuc. Sci. Eng. 135, 1(2000). GU00 K. H. Guber et al., to be published in the Nuc. Sci. Eng. KA85 K. Kanda et al, Measurement of Fast Neutron Induced Fission Cross Sections of 232-Th, 233-U, and 234-U Relative to 235-U, Nuclear Data for Basic and Applied Science, Vol. 1, Santa Fe, New Mexico (May 1985). LA98 N. M. Lasrson, Updated User Guide for SAMMY: Multilevel R- Matrix Fits to Neutron Data Using Bayes' Equations, ORNL/TM-9179/R4 (December 1998). See also ORNL/TM-9170/R5. LE01 L. C. Leal, H. Derrien, J. A. Harvey, K. H. Guber, N. M. Larson and R. R. Spencer, R-Matrix Resonance Analysis and Statistical Properties of the Resonance Parameters of U-233 in the Neutron Energy Range from Thermal to 600 eV, ORNL/TM-2000/372, March 2001. LE96 L. C. Leal and R. Q. Wright, Assessment of the Available 233-U Cross Section Evaluations in the Calculation of Critical Benchmark Experiments, ORNL/TM-13313R (Sept 1996). MU85 S. F. Mughabghab, Neutron Cross Sections, Vol. I. Part B (1985).