92-U -232 ORNL,LANL+ EVAL-FEB05 M.B.CHADWICK, P.G.YOUNG DIST- REV2- ----ENDF/B-VI MATERIAL 9219 REVISION 2 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT ****************************************************************** ****************************************************************** MOD 3A - Incorporate revised delayed neutron data fron W.B.Wilson [02-28-05] New DN 6-gp parameters from CINDER'90 summation calculations are available. The CINDER calculations are based on a new CINDER library in which beta-decay half-lives and beta-delayed neutron- emission probablities are obtained from the evaluated experimental data file NuBase2003 (G. Audi, O. Bersillon, J. Blanchot, and A. H. Wapstra, Nuclear Physics A729 (2003) p. 3-129), when avail- able there. When experimental data are not available there they are calculated in a model where allowed Gamow-Teller decays are treated in a microscopic quasi-particle random-phase approximation (QRPA) and the first forbidden decays are treated in the statis- tical gross theory (P. Moller, B. Pfeiffer, and K.-L. Kratz, Phys. Rev. C. 67 (2003) 055802). Further refinements are anticipated before the final release of ENDF-7. ****************************************************************** MOD 3 - 232U + n REVISION ABOVE THE RESONANCE REGION [9/2001] ENDF/B-VI EVALUATION ABOVE THE UNRESOLVED RESONANCE REGION M.B.Chadwick and P.G. Young (LANL) Above the unresolved resonance region, new evaluations were performed of the neutron total, elastic, (n,n'), (n,2n), (n,3n), (n,4n), (n,f), (n,nf), (n,2nf), (n,4nf), and (n,gamma) cross sections, as well as their associated angular and energy distributions. The energy range of the evaluation was increased to 30 MeV. To provide the new data, coupled channel optical model calculations were performed with the ECIS96 code for the lowest 3 members of the U232 ground state rotational band. Optical model parameters were obtained from an isospin-dependent analysis of all stable U isotopes. These calculations were used to provide elastic and inelastic cross sections and angular distributions, and to provide neutron transmission coefficients for nuclear reaction theory calculations with the GNASH (Yo96) Hauser-Feshbach statistical/fission/preequilibrium code and the COMNUC (Du70) Hauser-Feshbach statistical/fission code. These theory calculations were used to provide the MT=51-91 inelastic, (n,2n), (n,3n), and (n,4n) cross sections, angular distributions, and energy distributions. The compound nucleus inelastic calculations with GNASH and COMNUC were combined with direct cross sections and angular distributions from ECIS96 for MT=51-53. For MT=54-90, the direct cross section components were obtained from the ENDF/B-VI, MOD 5, evaluation of 238U + n reactions, which were obtained by using ECIS to extrapolate 14-MeV experimental data. The inelastic data above MT=66 are for combinations of (n,n') levels. Preequilibrium angular distribution shapes were used for MT=72- 90 at all energies, and for MT=54-71 above 6 MeV. Below 6 MeV, preequilibrium shapes were combined with compound nucleus angular distributions from COMNUC. The GNASH calculations were used directly for the correlated energy-angle spectra in MF=6. Prompt nubar (MT=456) was extrapolated to 30 MeV. Total and delayed nubar were also extended to 30 MeV. NOTE. The (n,5n) cross section becomes significant above about 25 MeV. Because there is no ENDF-6 MT number for this reaction, we have arbitrarily included this cross section in the (n,4n) reaction. Consequently, the (n,4n) cross section becomes unphysical above about 25 MeV. ------------------------------------------------------------------ MOD3 FILE DETAILS MF=1 FISSION NEUTRON MULTIPLICITIES, ENERGY RELEASE MT=452 Total neutrons per fission (nubar). Sum of MT=455 and MT=456 from 1.e-5 eV to 30 MeV. MT=455 Delayed neutron decay constants and multiplicities. Taken from the MOD 2, described below. Extrapolated to 30 MeV. MT=456 Prompt Neutron Multiplicities. ). Used MOD 2 evaluation (below). Linear extrapolation to 30 MeV. MF=2 RESONANCE PARAMETERS MT=151 Used resolved and unresolved resonance parameters from the MOD 2 evaluation (see below). MF=3 SMOOTH CROSS SECTIONS MT=1 Neutron Total Cross Section. From 2 keV to 30 MeV, based on ECIS96 coupled-channels calculations. MT=2 Neutron Elastic Cross Section. From 2 keV to 30 MeV, obtained by subtracting MT=3 from MT=1. MT=3 Neutron Nonelastic Cross Section. From 2 keV to 30 MeV, sum of the cross sections from all nonelastic reactions. MT=16 (n,2n) Cross Section. Threshold to 30 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations. MT=17 (n,3n) Cross Section. Threshold to 20 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations. MT=18 Fission Cross Section (Total). From 1 keV to 7.5 MeV, based on measured cross sections of Fursov et al. [Fu86], as formatted in MOD 2 evaluation (below). Above 7.5 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations MT=19 First-Chance Fission (n,f) Cross Section. From 1 keV to 30 MeV, obtained by multiplying MT=18 cross section by ratio of first chance to total fission cross section from GNASH calculation. MT=20 Second-Chance Fission (n,nf) Cross Section. From 1 keV to 30 MeV, obtained by multiplying MT=18 cross section by ratio of second chance to total fission cross section from GNASH calculation. MT=21 Third-Chance Fission (n,2nf) Cross Section. From 1 keV to 30 MeV, obtained by multiplying MT=18 cross section by ratio of third chance to total fission cross section from GNASH calculation. MT=37 (n,4n) Cross Section. Threshold to 20 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations. MT=38 Fourth-Chance Fission (n,3nf) Cross Section. From 1 keV to 30 MeV, obtained by multiplying MT=18 cross section by ratio of fourth chance to total fission cross section from GNASH calculation. MT=54-90 Discrete inelastic, threshold to 30 MeV, combination of Hauser-Feshbach and direct cross sections from 238U analysis. MT=91 Continuum Inelastic Cross Section. Threshold to 30 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations MT=102 Radiative Capture Cross Section. 1 keV to 30 MeV, based on GNASH Hauser-Feshbach/preequilibrium calculations. MF=4 ANGULAR DISTRIBUTIONS MT=2 Elastic Scattering. Below 10 MeV, Legendre polynomial coefficients obtained by combining COMNUC compound elastic with ECIS96 shape elastic angular distributions. Above 10 MeV, used tabulations of ECIS96 shape elastic angular distributions. MT=51-90 Discrete (n,n'). Obtained by combining compound nucleus and direct reaction angular distributions. For the direct reaction contributions, ECIS96 coupled-channel shapes were used for MT=51-53, and preequilibrium shapes were used for MT=54-90. MF=5 ENERGY DISTRIBUTIONS MF=18,19,20,21,38 Fission Neutron Energy Distributions. Taken from the MOD 1 evaluation (see below). MF=455 Delayed Neutron Energy Distributions. Taken from the MOD 1 evaluation (see below). MF=6 CORRELATED ENERGY-ANGLE DISTRIBUTIONS MT=16 (n,2n) Distributions. GNASH Hauser-Feshbach /preequilibrium calculations. Neutron distributions only. MT=17 (n,3n) Distributions. GNASH Hauser-Feshbach /preequilibrium calculations. Neutron distributions only. MT=37 (n,3n) Distributions. GNASH Hauser-Feshbach /preequilibrium calculations. Neutron distributions only. MT=91 (n,n') Continuum Distributions. GNASH Hauser-Feshbach /preequilibrium calculations. Neutron distributions only. MF=8 RADIOACTIVE DECAY AND FISSION PRODUCT YIELD DATA MT=16,17,102 Taken from the MOD 1 evaluation (see below). MF=12 PHOTON-PRODUCTION MULTIPLICITIES MT=102 Radiative Capture. Based on GNASH Hauser-Feshbach /preequilibrium calculations. MF=13 PHOTON-PRODUCTION CROSS SECTIONS MT=3 Nonelastic Reactions. Taken directly from the ENDF/B- VI, MOD 5 238U +n evaluation. MF=14 PHOTON ANGULAR DISTRIBUTIONS MT=3 Nonelastic Reaction Photons. Assumed isotropic. MT=18 Fission Photons. Assumed isotropic MT=102 Radiative Capture Photons. Assumed isotropic. MF=15 PHOTON ENERGY DISTRIBUTIONS MT=3 Nonelastic Reactions. Taken directly from the ENDF/B-VI, MOD 5 238U +n evaluation. MT=18 Fission Photons. Taken directly from the ENDF/B-VI, MOD 5 238U +n evaluation. MT=102 Radiative Capture Photons. Based on GNASH Hauser- Feshbach /preequilibrium calculations. ---------------------------------------------------------------- REFERENCES [Fu86] B.I. Fursov, E.Yu. Baranov et al., At.En. 61,383 (1986); see Sov.J.At.En. 61, 963 (1987) for English translation. [Du70] C.L. Dunford, Atomics Int. report AI-AEC-12931 (1970) [Ra96] J. Raynal, personal communication through A. Koning, ECN Petten (1996). [Yo96] P.G.Young, E.D.Arthur, and M.B.Chadwick, "Comprehensive Nuclear Model Calculations: Theory and Use of the GNASH Code," Proc. Workshop NUCLEAR REACTION DATA AND NUCLEAR REACTORS, ICTP, Trieste, Italy, 15 April - 17 May 1996 [Ed: A. Gandini and G. Reffo, World Scientific Publ. Co., Singapore (1998)] p. 227-404. ****************************************************************** ****************************************************************** ENDF/B-VI MOD 2 Revision, October 1999, R.Q. Wright (ORNL) The JENDL-3.2 evaluation was adopted and updated as follows: The following sections are taken from ENDF/B-V. MT=452 Total number of neutrons per fission MT=455 Number of delayed neutrons MT=456 Number of prompt neutrons MF=2, MT=151 RESONANCE PARAMETERS Resolved resonance parameters (1.0E-5 to 194 eV) Reich-Moore parameters from Mughabghab [Mu84] Unresolved resonance range (194 to 2,000 eV) D0 = 4.60 eV from Mughabghab [Mu84] S0 = 1.20-4, capture width = 0.04 eV Scattering radius = 9.80 fm Calculated 2200-m/s cross sections and resonance integrals 2200-m/s (barns) Res. Int. Total 162.77 Elastic 10.82 Fission 76.76 383.8 barns Capture 75.19 181.3 barns MF=3 NEUTRON CROSS SECTIONS Modified above 2.0 keV as follows: MT=1 TOTAL Same as JENDL-3.2 except additional points were added by interpolation. MT=2 ELASTIC SCATTERING Elastic = total - nonelastic MT=18 FISSION 2 to 175 keV, fission taken from ENDF/B-V. 175 keV to 7 MeV, evaluation is based on the measured data of Fursov [2]. Only a small amount of smoothing was done-- cross section is taken from EXFOR "40919 file". REFERENCES [Fu87] B.I. Fursov, E.Yu. Baranov et al., At.En. 61,383 (1986); see Sov.J.At.En. 61, 963 (1987) for English translation. [Mu84] S.F. Mughabghab, Neutron Cross Sections, Vol. 1, Part B, (Academic Press, 1984). ************************ C O N T E N T S ************************* 1 451 373 3 1 452 5 3 1 455 7 3 1 456 5 3 2 151 63 2 3 1 154 3 3 2 154 3 3 3 154 3 3 4 88 3 3 16 28 3 3 17 20 3 3 18 128 3 3 19 128 3 3 20 99 3 3 21 51 3 3 37 10 3 3 38 32 3 3 51 88 3 3 52 85 3 3 53 81 3 3 54 77 3 3 55 76 3 3 56 74 3 3 57 73 3 3 58 72 3 3 59 70 3 3 60 69 3 3 61 68 3 3 62 67 3 3 63 65 3 3 64 64 3 3 65 63 3 3 66 62 3 3 67 61 3 3 68 60 3 3 69 60 3 3 70 59 3 3 71 58 3 3 72 57 3 3 73 56 3 3 74 52 3 3 75 48 3 3 76 47 3 3 77 46 3 3 78 45 3 3 79 44 3 3 80 43 3 3 81 42 3 3 82 41 3 3 83 40 3 3 84 38 3 3 85 37 3 3 86 37 3 3 87 36 3 3 88 35 3 3 89 35 3 3 90 34 3 3 91 62 3 3 102 82 3 4 2 647 3 4 18 2 3 4 19 2 3 4 20 2 3 4 21 2 3 4 38 2 3 4 51 183 3 4 52 181 3 4 53 177 3 4 54 104 3 4 55 99 3 4 56 97 3 4 57 95 3 4 58 91 3 4 59 98 3 4 60 92 3 4 61 91 3 4 62 91 3 4 63 93 3 4 64 91 3 4 65 91 3 4 66 91 3 4 67 89 3 4 68 91 3 4 69 89 3 4 70 89 3 4 71 89 3 4 72 87 3 4 73 72 3 4 74 70 3 4 75 68 3 4 76 66 3 4 77 66 3 4 78 64 3 4 79 64 3 4 80 64 3 4 81 64 3 4 82 62 3 4 83 62 3 4 84 60 3 4 85 60 3 4 86 58 3 4 87 58 3 4 88 58 3 4 89 58 3 4 90 58 3 5 18 9 3 5 19 9 3 5 20 8 3 5 21 8 3 5 38 7 3 6 16 1221 3 6 17 567 3 6 37 165 3 6 91 2918 3 12 18 4 3 12 102 20 3 13 3 40 3 14 3 1 3 14 18 1 3 14 102 1 3 15 3 299 3 15 18 32 3 15 102 2719 3