ENDF Reader

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endf_reader

ENDF Files¶

Library class¶

Below is an example of how to grab and graph cross section data from ENDF files using the Library class.

In [1]:
%matplotlib inline

import os
try:
    from urllib.request import urlretrieve
except ImportError:
    from urllib import urlretrieve
    
import matplotlib.pyplot as plt
import numpy as np
from IPython.display import HTML
from tabulate import tabulate

from pyne.endf import Library, Evaluation
/home/romano/.local/lib/python2.7/site-packages/pyne-0.5.0_rc1-py2.7.egg/pyne/rxdata.py:5: QAWarning: pyne.rxdata is not yet QA compliant.
  warn(__name__ + " is not yet QA compliant.", QAWarning)
/usr/local/lib/python2.7/dist-packages/IPython/kernel/__main__.py:14: QAWarning: pyne.rxname is not yet QA compliant.
/usr/local/lib/python2.7/dist-packages/IPython/kernel/__main__.py:14: QAWarning: pyne.endf is not yet QA compliant.
In [2]:
if not os.path.isfile("U235-VII.txt"):
    urlretrieve("http://t2.lanl.gov/nis/data/data/ENDFB-VII-neutron/U/235", "U235-VII.txt")
In [3]:
u235 = Library("U235-VII.txt")
xs_data = u235.get_xs(922350000, 16)[0]
In [4]:
fig = plt.figure()
Eints, sigmas = xs_data['e_int'], xs_data['xs']
plt.step(Eints, sigmas, where = "pre")
plt.suptitle(r'(n, 2n) Reaction in $^{235}$U')
plt.ylabel(r'$\sigma(E)$ (barns)')
plt.xlabel(r'$E_{int} (eV)$')
plt.xscale('log')
plt.yscale('log')
plt.savefig('u235_2n.eps')
In [5]:
if not os.path.isfile("U238-VII.txt"):
    urlretrieve("http://t2.lanl.gov/nis/data/data/ENDFB-VII-neutron/U/238", "U238-VII.txt")
In [6]:
u238 = Library("U238-VII.txt")
xs_data = u238.get_xs(922380000, 1)[0]
In [7]:
fig = plt.figure()
Eints, sigmas = xs_data['e_int'], xs_data['xs']
plt.step(Eints, sigmas, where = "pre")
plt.suptitle(r'Total Cross Section for $^{238}$U')
plt.ylabel(r'$\sigma(E)$ (barns)')
plt.xlabel(r'$E_{int} (eV)$')
plt.xlim(xmin = 10000)
plt.xscale('log')
plt.yscale('log')

Evaluation class¶

The pyne.endf.Evaluation class provides a facility for parsing data in an ENDF file. Parsing of all data other than covariances (MF=30+) is supported has been tested against the ENDF/B-VII.1 neutron, photoatomic, electroatomic, atomic relaxation, and photonuclear sublibraries. In this example, we will use the Evaluation class to look at typical data in the ENDF/B-VII.1 evaluation of U-235.

In [8]:
u235 = Evaluation("U235-VII.txt")
Reading MF=1, MT=451 Descriptive Data

By default, when an Evaluation is instantiated, only the descriptive data in MF=1, MT=451 is parsed. This allows us to get basic information about an evaluation without necessarily reading the whole thing. This useful data can be found in the info and target attributes.

In [9]:
u235.info
Out[9]:
{u'author': ' Young,Chadwick,Talou,Madland,Leal',
 u'date': 'EVAL-SEP06',
 u'date_distribution': 'DIST-DEC06',
 u'date_entry': '        ',
 u'date_release': 'REV-      ',
 u'derived': False,
 u'description': ['***************************************************************** ',
  '                                                                  ',
  '                  ENDF/B-VII EVALUATION                           ',
  '                                                                  ',
  'P.G.Young, M.B.Chadwick, R.E.MacFarlane, W.B.Wilson, D.G.Madland, ',
  '                P.Talou, T. Kawano (LANL)                         ',
  '                          and                                     ',
  '   L. C. Leal, H. Derrien, N. M. Larson, R. Q. Wright (ORNL)      ',
  '                          and                                     ',
  '              D.A. Brown, J.Pruet (LLNL)                          ',
  '                                                                  ',
  '                ----- SUMMARY -----                               ',
  '                                                                  ',
  'Major features of the ENDF/B-VII evaluation are:                  ',
  '1. A new evaluation of the (n,f) cross section based on ENDF/B-   ',
  '   VII standard cross section analysis is incorporated.           ',
  '2. A new evaluation above thermal energy of prompt nubar          ',
  '   consistent with experimental data within uncertainties and     ',
  '   with fast critical benchmark measurements is included.         ',
  '3. New unresolved resonance parameter data are incorporated.      ',
  '4. A new analysis of the prompt fission neutron spectrum matrix   ',
  '   based on the Los Alamos model is used to calculate neutron     ',
  '   spectra at all energies. [The ENDF/B-VI results from the Los   ',
  '   Alamos model are used for the thermal spectrum.]               ',
  '5. Improved delayed neutron data are incorporated.                ',
  '6. New reaction theory calculations are utilized for (n,xn) and   ',
  '   other reactions.  Direct reaction cross sections and angular   ',
  '   distributions are extended to an excitation energy of 4 MeV.   ',
  '7. Improved fission energy release values are incorporated.       ',
  '                                                                  ',
  '             ----- DETAILED DESCRIPTION -----                     ',
  '                                                                  ',
  '>> MF=1 GENERAL INFORMATION                                       ',
  '                                                                  ',
  'MT=452: AVERAGE TOTAL NUMBER OF NEUTRONS PER FISSION (TOTAL       ',
  '        NUBAR)                                                    ',
  '   Sum of MT=455 and 456. The ENDF/B-VI values are preserved      ',
  'below 1 eV.                                                       ',
  '                                                                  ',
  'MT=455: DELAYED NEUTRON DATA                                      ',
  '   The decay constants and abundances are based on new delayed    ',
  "neutron 6-group parameters from CINDER'90 summation calculations. ",
  'The CINDER calculations (Wi05) utilize 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 (Au03), when available there. When experimental   ',
  'data are not available, the data 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 statistical gross theory      ',
  '(Mo03).                                                           ',
  '   The delayed nubar values were carried over from ENDF/B-VI,     ',
  'except between 1.0-5 eV and 50 keV where they were modified to    ',
  'agree with JENDL-3.3 values, following a suggestion from C.       ',
  'Lubitz (Lu06).                                                    ',
  '                                                                  ',
  'MT=456: AVERAGE NUMER OF PROMPT NEUTRONS RELEASED PER FISSION     ',
  '   1.e-5 eV - 10 keV:  Very similar to ENDF/B-VI. Minor changes   ',
  'from the ENDF/B-VI values were made to more closely approximate   ',
  'the energy dependence of the JENDL-3.3 evaluation and to          ',
  'accommodate our change in delayed nubar (above) while at the same ',
  'time keeping total nubar unchanged.                               ',
  '   10 keV - 20 MeV:  Generally based on results from the          ',
  'covariance analysis of experimental data performed for ENDF/B-VI, ',
  'after renormalization of the data above 6-8 MeV for consistency   ',
  'with the ENDF/B-VII standard 252Cf nubar value.                   ',
  '(Between 10 and 50 keV, small changes were also included to       ',
  'compensate for the delayed nubar change described above.) We      ',
  'attempted to follow the covariance data as well as possible but   ',
  'mainly to stay within uncertainties in the data and to keep good  ',
  'agreement with fast critical benchmarks. The structure in the     ',
  'Version VI covariance analysis around En=0.1-0.4 MeV, which was   ',
  'smoothed in the ENDF/B-VI evaluation, was restored in the Version ',
  'VII evaluation. Also, the evaluation was adjusted slightly        ',
  'between 1.0 and 2.5 MeV to better represent the covariance        ',
  'analysis. Above 2.8 MeV, nubar in ENDF/B-VII differs from ENDF/B- ',
  'VI by a factor of 1.0004.                                         ',
  '   Only limited experimental data has been obtained since the     ',
  'ENDF/B-VI analysis.  The new measurements include data by         ',
  'Khokhlov et al. (Kh94) and Boykov et al. (Bo90), but these        ',
  'measurements are consistent with our earlier covariance analysis  ',
  '(within experimental uncertainties) and are not expected to       ',
  'modify our results significantly.                                 ',
  '                                                                  ',
  'MT=458:  ENERGY RELEASE FROM FISSION                              ',
  '   Modifications were made to MT=458 based on a new analysis by   ',
  'Madland (Ma06). The average total fission product kinetic energy  ',
  '(EFR) and the average total prompt fission gamma-ray energy (EGP) ',
  'were taken from the Madland analysis. The average total prompt    ',
  'fission neutron kinetic energy (ENP) was obtained from NJOY,      ',
  'using the MF=5,MT=18 fission neutron spectra and prompt nubar     ',
  '(MT=1,MT=456) from the evaluation. (This value of total neutron   ',
  "energy is close to Madland's result.) The kinetic energy of       ",
  'delayed fission neutrons (END), the total energy from delayed     ',
  'gamma rays (EGD), the total energy released by delayed betas      ',
  '(EB), and the energy released by neutrinos (ENU) were carried     ',
  'over from the ENDF/B-VI evaluation. The total energy release per  ',
  'fission is: ET = EFR+ENP+END+EGP+EGD+EB+ENU, and the quantity ER  ',
  '(total energy less the neutrino energy, or pseudo-Q value) is     ',
  'simply ET-ENU. ER is also included as the fission Q-value in      ',
  'MF=3, MT=18,19,20,21,38.                                          ',
  '   The uncertainties in EFR, ENP, END, EGP, EGD, EB, ENU, ER, and ',
  'ET are carried over from ENDF/B-VI.                               ',
  '   There is currently no format for including energy dependence   ',
  'in the ENDF/B file.  These are included in the NJOY processing.   ',
  '                                                                  ',
  'MT=460: BETA-DELAYED FISSION GAMMA DATA                           ',
  '   See D.A. Brown (Br06). Evaluated by J. Pruet, based on Pruet   ',
  'et al. (Pr04).                                                    ',
  '                                                                  ',
  '              ----- REFERENCES (MF=1) -----                       ',
  'Au03 G. Audi, O. Bersillon, J. Blanchot, and A. H. Wapstra, Nucl. ',
  '     Phys. A729 (2003) p. 3-129.                                  ',
  'Bo90 G.S. Boykov, V.D. Dmitriev, G.A. Kudyaev, M.I. Svirin, G.N.  ',
  '     Smirenkin, Atomnaya Energiya 69, 23 (1990).                  ',
  'Br06 D.A. Brown, Lawrence Livermore National Laboratory report    ',
  '     UCRL-TR-223148 (2006).                                       ',
  "Kh94 Yu.A. Khoklhov, I.A. Ivanin, V.I. In'kov, Yu.I. Vinogradov,  ",
  '     L.D. Dailin, B.N. Polynov, Proc. Int. Conf. on Nucl. Data    ',
  '     for Science & Tech., Gatlinberg,1994, v.1, p.272 (1994).     ',
  'Lu06 C. Lubitz, personal communication, August, 2006.             ',
  'Ma06 D. G. Madland, Nucl. Phys. A772, 113 (2006).                 ',
  'Mo03 P. Moller, B. Pfeiffer, and K.-L. Kratz, Phys. Rev. C. 67    ',
  '     (2003) 055802.                                               ',
  'Pr04 J. Pruet, J. Hall, M.-A. Descalle, S.G. Prussin, Nucl. Inst. ',
  '     Meth. B 222, 403 (2004).                                     ',
  'Wi05 W. B. Wilson, personal communication, 2005.                  ',
  '                                                                  ',
  '>> MF=2 RESONANCE PARAMETERS                                      ',
  '                                                                  ',
  'MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS               ',
  '                                                                  ',
  ' ----- RESOLVED RESONANCE PARAMETERS (1.e-5 - 2250 eV) -----      ',
  '  The resolved resonance parameters are the same as those in      ',
  'ENDF/B-VI Release 8.  These parameters are from a new analysis    ',
  'for 235U by Leal et al. (Le97), using the multilevel R-matrix     ',
  'analysis code SAMMY (La96).  The energy range for resolved region ',
  '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 were obtained in the present evaluation,     ',
  'first, using the microscopic experimental data only, and second,  ',
  'including the integral data as well. (Note that Version VII       ',
  'thermal standards are not incorporated.) Our results are compared ',
  'to the SAMMY input in the following table:                        ',
  '                                                                  ',
  '       Parameter        SAMMY input     Fit to        Fit to      ',
  '                           value        diff data       diff      ',
  '                                        alone        & integ      ',
  '                                                        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       15.67      ',
  '       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 ',
  'obtained in this evaluation are compared with experimental data:  ',
  '                                                                  ',
  '       Experimental and calculated fission cross sections.        ',
  '       Cross sections were calculated with the code SAMMY.        ',
  '                                                                  ',
  '       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.  Cross ',
  '       sections were calculated with the code SAMMY.              ',
  '                                                                  ',
  '       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 meters (0.02 to 20 eV)                          ',
  '       (5) Fission and (6) capture data of de Saussure et al.     ',
  '           (De67) on the ORELA 25.2-meter flight path (0.01       ',
  '           to 2250 eV)                                            ',
  '       (7) Fission and (8) capture data of Perez et al. (Pe72)    ',
  '           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 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) Absorption and (13) 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)                                ',
  '                                                                  ',
  '----- UNRESOLVED RESONANCE PARAMETERS (2250 eV - 25 keV) -----    ',
  '   The unresolved resonance parameter evaluation was revised      ',
  'using a new ORNL analysis (10/1/03).                              ',
  '                                                                  ',
  '                 ----- REFERENCES (MF=2) -----                    ',
  'Ca93 A. Carlson, W. P. Poenitz, G. M. Hale, R. W. Peele, D. C.    ',
  '     Dodder, C. Y. Fu, and W. Manhart, The ENDF/B-6  Neutron      ',
  "     Cross Section Measurements Standards, NISTIR-5177, Natn'l    ",
  '     Institute of Standards and Technology, Gaithersburg, MD (May ',
  '     1993).                                                       ',
  'De67 G. de Saussure, R. Gwin, L. W. Weston, and R. W. Ingle,      ',
  '     Simultaneous Measurements of the Neutron Fission and Capture ',
  '     Cross Section for 235U for Incident Neutron Energy from 0.4  ',
  '     eV to 3 keV, ORNL/TM-1804, Martin Marietta Energy Systems,   ',
  '     Inc., Oak Ridge National Laboratory, Oak Ridge, TN (1967).   ',
  'Gw84 R. Gwin, R. R. Spencer, R. W. Ingle, J. H. Todd, and S.      ',
  '     W. Scoles, Nuc. Sci. Eng. 88, 37 (1984).                     ',
  'Gw96 R. Gwin, To be published in Nuclear Science Engineering.     ',
  'Ha79 J. Hardy, 235U Resonance Fission Integral and Alpha Based on ',
  '     Integral Measurements, ENDF-300, Sec. B.1, Brookhaven        ',
  '     National Laboratory, Upton, NY (1979).                       ',
  'Ha86 J. A. Harvey, N. W. Hill, F. G. Perey, G. L. Tweed, and      ',
  '     L. C. Leal, Proc. Int. Conf. On Nuclear Data for Science     ',
  '     and Technology, May 30-June 3, 1988, Mito, Japan.            ',
  'Le97 L. C. Leal, H. Derrien, N. M. Larson, R. Q. Wright, "R-      ',
  '     Matrix Analysis of 235U Neutron Transmission and Cross       ',
  '     Sections in the Energy Range 0 eV to 2.25 keV," ORNL/TM-     ',
  '     13516, Lockheed Martin Energy Research Corp., Oak Ridge      ',
  '     National Laboratory, Oak Ridge, TN (1997).                   ',
  'Mo92 M. C. Moxon, J. A. Harvey, and N. W. Hill, Private           ',
  '     communication ORNL (1992).                                   ',
  'Pe72 R. B. Perez, G. de Saussure, and E. G. Silver, Nucl. Sci.    ',
  '     Eng. 52, 46 (1973).                                          ',
  'Sc88 R. A. Schrack, "Measurement of the 235U(n,f) Reaction from   ',
  '     Thermal to 1 keV," Nuclear Data for Science and Technology,  ',
  '     p. 101, Mito, Japan (1988).                                  ',
  'Sp84 R. R. Spencer, J. A. Harvey, N. W. Hill, and L. Weston,      ',
  '     Nucl. Sci. Eng. 96, 318 (1987).                              ',
  'Wa87 J. A. Wartena, H. Weigmann, and C. Burkholz, Report  IAEA    ',
  '     Tecdoc 491, p.123 (1987).                                    ',
  'Wa88 C. Wagemans, P. Schillebeeckx, A. J. Deruyter, and R.        ',
  '     Barthelemy, "Subthermal fission Cross Section                ',
  '     measurements for 233U and 239Pu," Nuclear Data for           ',
  '     Science and Technology, p. 91, Mito, Japan (1988).           ',
  'We84 L. W. Weston and J. H. Todd, Nucl. Sci. Eng. 88, 567 (1984). ',
  'We90 H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach, and   ',
  '     J. A. Wartena, Proc. Intern.  Conf. on The Physics of        ',
  '     Reactors, Marseille 1990, Vol. P1, p. 133 (1990).            ',
  'We92 L. W. Weston and J. H. Todd, Nucl. Sci. Eng. 111, 415        ',
  '     (1992).                                                      ',
  '                                                                  ',
  '>> MF=3 NEUTRON CROSS SECTIONS                                    ',
  '                                                                  ',
  '              ----- GENERAL INFORMATION -----                     ',
  '   The maximum energy of the evaluation remains at 20 MeV, the    ',
  'same as ENDF/B-VI.                                                ',
  '   The total (MT=1), fission (MT=18), and radiative capture       ',
  '(MT=102) cross sections are based mainly on experimental data,    ',
  'complimented by nuclear model calculations. Model parameters for  ',
  'the calculations were obtained by optimization with experimental  ',
  'data. The neutron total and (n,f) cross section revisions include ',
  'new experimental data that were not included in the ENDF/B-VI     ',
  '235U analysis.                                                    ',
  '   The present evaluation utilizes some data from the ENDF/B-VI   ',
  'evaluation, in particular, discrete inelastic scattering data for ',
  'levels below an excitation energy of 1.1 MeV. That is, the MT=51- ',
  '71 data in ENDF/B-VII correspond to the MT=51-82 data in ENDF/B-  ',
  "VI, or combinations thereof.  The discrete (n,n') cross sections  ",
  'and angular distributions in ENDF/B-VI are based on nuclear       ',
  'theory/model code calculations with the ECIS70 coupled-channels   ',
  'optical model code (Ra70) and with the GNASH (Ar88, Yo77) and     ',
  'COMNUC (Du71) Hauser-Feshbach codes, with model parameters        ',
  'optimized to experimental data. The GNASH calculations also       ',
  'include preequilibrium contributions. The coupled-channels        ',
  'optical model potential used is potential #3 in the International ',
  "Atomic Energy Agency's Reference Input Parameter Library (RIPL-2) ",
  'optical model parameter library (see Yo94).                       ',
  '   DWBA calculations were performed with the DWUCK code (Ku70)    ',
  'for several vibrational levels, using B(El) values inferred from  ',
  "(d,d') data on U234, U235, U238, as well as Coulomb excitation    ",
  'measurements.  A weak coupling model (Pe69) was used to apply the ',
  'U234 and U238 results to states in U235. A preliminary            ',
  'description of the ENDF/B-VI analysis was given at the Mito       ',
  'conference (Yo88).                                                ',
  '   The neutron total cross section (MT=1) below 25 keV is also    ',
  'the same as ENDF/B-VI.                                            ',
  '   An updated 235U analysis was performed with the ECIS94 (Ra94)  ',
  'and GNASH codes (Yo98) for ENDF/B-VII. This new analysis provide  ',
  'the basis for the ENDF/B-VII evaluation of the following data:    ',
  'MF=3,6 MT=16,17,37,91. Additionally, direct reaction cross        ',
  'sections and angular distributions, inferred from neutron         ',
  'spectrum measurements on 238U, are included for groups of states  ',
  'in the MT=72-90 data.                                             ',
  '                                                                  ',
  '           ----- DETAILS OF LANL REVISION -----                   ',
  '                                                                  ',
  'MT=1: TOTAL CROSS SECTION                                         ',
  '   2.25 - 25 keV:  Obtained from the ENDF/B-VI evaluation.        ',
  'Corresponds to the unresolved resonance region.                   ',
  '   25 keV - 20 MeV:  The previous ENDF/B-VI evaluation of the     ',
  'neutron total cross                                               ',
  'section in the MeV region resulted from a covariance analysis     ',
  'with the GLUCS code (He80) of the experimental data available at  ',
  'that time. Experimental data used include Fo71, Ve80, Bo71, Po81, ',
  'Gr73, Sc74, Po83, Pe60, Wh65, Ca73, Ga60, and Br58. The present   ',
  'revision is an extension of the previous results by incorporation ',
  'of the experimental data of Lisowski (Li90) into the GLUCS        ',
  'analysis.  The result of this analysis is a general lowering of   ',
  'the total cross section by a few tenths of a percent above 50     ',
  'keV.  The new result is lower than the previous one by 0.4% at 3  ',
  'MeV, is unchanged at 8 MeV, and is lowered by 0.5% at 14 MeV and  ',
  'by 1.3% (maximum change) at 20 MeV.                               ',
  '                                                                  ',
  'MT=2: ELASTIC SCATTERING CROSS SECTION                            ',
  '  The elastic cross section is obtained by subtracting the        ',
  'nonelastic cross section (MT=3) from the total cross section      ',
  '(MT=1).                                                           ',
  '                                                                  ',
  'MT=3: NONELASTIC CROSS SECTION                                    ',
  '   The (redundant) nonelastic cross section is the sum of the     ',
  'following reactions: MT=4,16,17,18,37,102.                        ',
  '                                                                  ',
  'MT=4: INELASTIC CROSS SECTION                                     ',
  '   Sum of MT=51-91.                                               ',
  '                                                                  ',
  'MT=16: (n,2n) CROSS SECTION                                       ',
  '   The revised (n,xn) cross sections (and energy-angle            ',
  'distributions) result from a revision of the GNASH nuclear model  ',
  'code (Yo98) analysis that corrects an error in the analysis used  ',
  'for the previous ENDF/B-VI evaluation.  The major change in the   ',
  'analysis is a correction of an inconsistent treatment of          ',
  'preequilibrium effects in the presence of fission that was        ',
  'present in previous versions of the GNASH code.  The changes to   ',
  'the cross sections are not large but are non-negligible.  For     ',
  'example, changes in the (n,2n) cross section are +6% near 8 MeV,- ',
  '5% at 12 MeV, +3.5% at 14 MeV, and +16% at 20 MeV.  The new GNASH ',
  'analysis also results in modified energy-angle distributions in   ',
  'MF = 6.                                                           ',
  '   The revised (n,2n) cross section is in good agreement with the ',
  'experimental data of Becker et al. (Be98), Frehaut et al. (Fr80), ',
  'and Mathur et al. (Ma69,Ma72).  The Becker data are based on new  ',
  'experimental data from a LANSCE-GEANIE experiment. The new        ',
  'results are deduced from a combination of measured partial gamma- ',
  'ray cross sections and enhanced Hauser-Feshbach reaction          ',
  'modeling.                                                         ',
  '                                                                  ',
  'MT=17,37: (n,xn) CROSS SECTIONS                                   ',
  '   The (n,xn) cross sections are based on an analysis with the    ',
  'GNASH nuclear model code (Yo98) that corrects an error in the     ',
  'calculations used for the previous ENDF/B-VI evaluation.  See     ',
  'comments under MT=16 above.                                       ',
  '                                                                  ',
  'MT=18: FISSION NEUTRON CROSS SECTION                              ',
  "   2.25-25 keV:  Incorporated the cross section from Leal's       ",
  'analysis in the unresolved resonance region, after                ',
  'renormalization of the fission cross section (-2%) to agree with  ',
  'the average of the ENDF/B-VII 235U(n,f) standard cross section in ',
  'this energy range.                                                ',
  '   25 keV - 20 MeV:  The 235U(n,f) Version VII standard cross     ',
  'section (Pr05) was incorporated into the 235U evaluation with     ',
  'minimal smoothing.  The original standard energy grid is included ',
  'as a subset of a larger grid.  The expansion to the denser grid   ',
  'was accomplished using a spline fit to a log-log file of the      ',
  'standard data.                                                    ',
  '   The Q-value was changed from 193.72 MeV to 193.4834 MeV to     ',
  'maintain consistency with the MF=1,MT=458 data.                   ',
  '                                                                  ',
  'MT=19,20,21,38: MULTI-CHANCE FISSION CROSS SECTIONS               ',
  '  The ratios of the multi-chance fission cross sections to the    ',
  'total fission cross section were obtained from GNASH              ',
  'calculations.  The evaluated multi-chance fission cross sections  ',
  'were then obtained by multiplying the MT=18 total fission cross   ',
  'section by the ratios.                                            ',
  '   The Q-value was changed from 193.72 MeV to 193.4834 MeV to     ',
  'maintain consistency with the MF=1,MT=458 data.                   ',
  '                                                                  ',
  'MT=51-71 DISCRETE INELASTIC LEVEL CROSS SECTIONS                  ',
  '   The cross sections for MT=51-71 are obtained from the ENDF/B-  ',
  'VI levels data or combinations thereof. The correspondance        ',
  'between the level data in ENDF/B-VII and ENDF/B-VI is as follows: ',
  '        ENDF/B-VII       ENDF/B-VI                                ',
  '         MT=51-56        MT=51-56                                 ',
  '         MT=57           MT57 + MT58                              ',
  '         MT=58           MT=59                                    ',
  '         MT=59           MT60 + MT61 +MT62                        ',
  '         MT=60           MT=63                                    ',
  '         MT=61           MT64 + MT65                              ',
  '         MT=62           MT66 + MT67                              ',
  '         MT=63           MT68 + MT69                              ',
  '         MT=64           MT70 + MT71 + MT72                       ',
  '         MT=65           MT73 + MT74                              ',
  '         MT=66           MT75 + MT76                              ',
  '         MT=67           MT=77                                    ',
  '         MT=68           MT=78                                    ',
  '         MT=69           MT=79                                    ',
  '         MT=70           MT=80                                    ',
  '         MT=71           MT=81 + MT82                             ',
  '   The levels MT=53,56,58,60 are the 9/2-, 11/2-, 13/2-, 15/2-    ',
  'members of the 7/2- ground-state rotational band.  The cross      ',
  'sections for these levels contain direct contributions calculated ',
  'with the ECIS70 code (Ra70).  The levels defined by MT=51-66 all  ',
  'contain compound nucleus contributions calculated with the COMNUC ',
  'code (Du71).  The level cross sections corresponding to MT=67-71  ',
  'are based on distorted wave Born approximation calculations with  ',
  'the DWUCK code (Ku70) for 2+ and 3- vibrational states.           ',
  '                                                                  ',
  'MT=72-90: DISCRETE INELASTIC LEVEL CROSS SECTIONS (DIRECT         ',
  '          REACTIONS TO GROUPS OF STATES)                          ',
  "   The cross sections for MT=72-90 are based upon Baba's neutron  ",
  'emission spectra measurements (Ba89) at 14 MeV for n + 238U       ',
  'reactions.  To accomplish this, we adopted in the present         ',
  'evaluation the MT=72-90 data from the ENDF/B-VII 238U evaluation  ',
  '(MAT=9237).  To free up the necessary MT slots in the present     ',
  'evaluation, it was necessary to combine data for some of the      ',
  'weaker levels, as tabulated above.                                ',
  '   The 238U results are based on DWBA calculations run with the   ',
  'ECIS94 code (Ra94), assuming a set of 2+ or 3- (mainly)           ',
  'vibrational states.  Deformation parameters were determined by    ',
  'matching the 14-MeV Baba data. The tabulated cross sections are   ',
  'actually sums of contributions to several states.  The DWBA       ',
  'calculations were used to extrapolate the 14-MeV cross sections   ',
  'to lower and higher energies, and to obtain the MF=4 angular      ',
  'distributions for each assumed state. The spins, parities, and    ',
  'deformation parameters used in the calculations are given in the  ',
  'table below. These results affect the evaluation in the           ',
  'excitation energy range Ex=1.17-3.91 MeV.                         ',
  '                                                                  ',
  '  MT    Ex (MeV)    J   Pi    Beta                                ',
  '      0.00000000   0.0  +1  0.0000E+00                            ',
  '  72  1.17000000   3.0  -1  3.8087E-02                            ',
  '  73  1.25000000   2.0  +1  3.0175E-02                            ',
  '  74  1.44000000   3.0  -1  5.6001E-02                            ',
  '  75  1.59000000   3.0  -1  3.8111E-02                            ',
  '  76  1.75000000   3.0  -1  3.9460E-02                            ',
  '  77  1.85000000   3.0  -1  3.5265E-02                            ',
  '  78  1.95000000   3.0  -1  4.0750E-02                            ',
  '  79  2.15000000   3.0  -1  4.7400E-02                            ',
  '  80  2.30000000   3.0  -1  5.3002E-02                            ',
  '  81  2.39000000   4.0  +1  8.8154E-03                            ',
  '  82  2.49000000   2.0  +1  2.5122E-02                            ',
  '  83  2.94000000   2.0  +1  2.7150E-02                            ',
  '  84  3.18900000   2.0  +1  2.5287E-02                            ',
  '  85  3.38800000   2.0  +1  2.5070E-02                            ',
  '  86  3.53800000   2.0  +1  1.5390E-02                            ',
  '  87  3.63700000   2.0  +1  1.6125E-02                            ',
  '  88  3.73700000   2.0  +1  1.6472E-02                            ',
  '  89  3.83700000   2.0  +1  1.4293E-02                            ',
  '  90  3.90900000   2.0  +1  1.5091E-02                            ',
  '                                                                  ',
  'MT=91: INELASTIC CONTINUUM NEUTRON CROSS SECTION                  ',
  '   Based on the GNASH Hauser-Feshbach statistical/preequilibrium  ',
  'calculations, described above. Note that MT=91 thresholds at 0.5  ',
  'MeV.  Therefore, discrete states with excitation energies above   ',
  '0.5 MeV (MT=67-90) lie in the MT=91 continuum region.             ',
  '                                                                  ',
  'MT=102: NEUTRON RADIATIVE CAPTURE CROSS SECTION                   ',
  '   2.25 TO 1000 keV:  The radiative capture cross section at      ',
  'these energies is based on alpha measurements (See ANL-83-4       ',
  'supplement).                                                      ',
  '   1-20 MeV:  Based on re-normalized COMNUC/GNASH calculations,   ',
  'with a semi-direct component added above a few MeV.               ',
  '                                                                  ',
  '             ----- REFERENCES (MF=3) -----                        ',
  'Ar88 E.D. Arthur, LA-UR-88-382 (1988).                            ',
  'Ba89 M. Baba, H. Wakabayashi, N. Itoh, K. Maeda, and N. Hirakawa, ',
  '     "Measurements of Prompt Fission Neutron Spectra and Double-  ',
  '     Differential Neutron Inelastic Scattering Cross Sections for ',
  '     238-U and 232-Th," IAEA Int. Nucl. Data report INDC(JPN)-129 ',
  '     (1989).                                                      ',
  'Be98 J. A. Becker et al., personal communication from W. Younes,  ',
  '     1998.                                                        ',
  'Bo72 K. Boeckhoff et al., J.Nuc.En.26,91(1972).                   ',
  'Br58 A. Bratenahl et al., Phys.Rev.110,927(1958).                 ',
  'Ca73 J. Cabe et al., CEA-R-4524 (1973).                           ',
  'Du71 C.L.Dunford, "Compound Nucleus Reaction Analysis Programs    ',
  '     COMNUC and CASCADE," Atomics International North American    ',
  '     Rockwell report AI-AEC-12931 (document no. TI-707-130-013)   ',
  '     (February, 1971).                                            ',
  'Fo71 D. Foster & D. Glasgow, Phys.Rev.C3,576(1971).               ',
  'Fr80 J. Frehaut et al., Nucl.Sci.Eng. 74, 29 (1980); J. Frehaut   ',
  '     et al., Brookhaven National Lab. report BNL-NCS-512457       ',
  '(1980)                                                            ',
  '     p.399.                                                       ',
  'Ga60 L. A. Galloway, III, Case Institute, TID-11005 (1960) p.19.  ',
  'Gr73 L. Green et al., USNDC-9 (1973) p.170                        ',
  'He80 D. Hetrick & C.Y. Fu, ORNL/TM-7341 (1980).                   ',
  'Ku70 P.D. Kunz, DWUCK: A Distorted-Wave Born Approximation        ',
  '     Program, unpublished report.                                 ',
  'Li90 P.W. Lisowski, personal communication of a measurement       ',
  '     done at WNR in 1985 (1990).                                  ',
  'Ma69 D. Mather, report AWRE-O-47 (1969).                          ',
  'Ma72 D. Mather et al., report AWRE-O-72 (1972).                   ',
  'Pe60 J.Peterson et al., Phys.Rev.120, 521(1960).                  ',
  'Pe69 R.J.Peterson, Ann.Phys. 53, 40 (1069).                       ',
  'Po81 W.Poenitz et al., Nuc.Sci.Eng.78, 333(1981).                 ',
  'Po83 W.Poenitz et al., ANL-NDM-80, 1983.                          ',
  'Pr05 V.G. Pronyaev, S.A.Badikov, A.D. Carlson, Z. Chen, E.V. Gai, ',
  '     G.M. Hale, F.-J. Hambsch, H.M. Hofmann, T. Kawano, N.M.      ',
  '     Larson, S.-Y. Oh, D.L. Smith, S. Tagesen, and H. Vonach,     ',
  '     personal communication (2005); see also: "An International   ',
  '     Evaluation of the Neutron Cross Section Standards," to be    ',
  '     published as an IAEA Technical report (2006).                ',
  'Ra70 J.Raynal,IAEA SMR-9/8 (1970).                                ',
  'Ra94 J. Raynal, "Notes on ECIS94," Centre d\'Etudes Nucleaires     ',
  '     (Saclay) report CEA-N-2772 (1994).                           ',
  'Sc74 R.Schwartz et al., Nuc.Sci.Engr.54,322(1974).                ',
  'Wh65 W.Whalen et al., ANL-7110 (1965) p.15.                       ',
  'Yo77 P.G.Young & E.D.Arthur, LA-6947 (1977).                      ',
  'Yo88 P.G.Young & E.D.Arthur, Nuc.Data for Sci.& Tech., Mito       ',
  '      Conference (1988) p.603.                                    ',
  'Yo94 P. G. Young, "Experience at Los Alamos with Use of the       ',
  '     Optical Model for Applied Nuclear Data Calculations," Los    ',
  '     Alamos National Laboratory report LA-UR-94-3104 (1994).      ',
  'Yo98 P. G. Young, E. D. Arthur, and M. B. Chadwick,               ',
  '     "Comprehensive Nuclear Model Calculations: Theory and Use of ',
  '     the GNASH Code," Proc. Workshop on 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.                           ',
  '                                                                  ',
  '>> MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY PARTICLES              ',
  '                                                                  ',
  'MT=2: NEUTRON ELASTIC SCATTERING ANGULAR DISTRIBUTIONS            ',
  '   1.e-5 eV -  20.0 MeV:  Elastic scattering angular distribution ',
  'based on ECIS70 (Ra70,Yo94) coupled-channels calculations, with a ',
  'compound elastic component from COMNUC included below 6 MeV       ',
  '(Du71). These data are taken from ENDF/B-VI.                      ',
  '                                                                  ',
  'MT=51,52,54,55,57,59,61-66: DISCRETE INELASTIC NEUTRON ANGULAR    ',
  '                            DISTRIBUTIONS FOR COMPOUND NUCLEUS    ',
  '                            REACTIONS                             ',
  '   Threshold to 6.0 MeV:  Angular distributions obtained using    ',
  'compound-nucleus reaction theory calculations with width          ',
  'fluctuation corrections in the COMNUC code (Du71). [Note that     ',
  'MT=65 contains a direct reaction component from DWUCK             ',
  'calculations and extends to 20 MeV.] These data are taken from    ',
  'ENDF/B-VI. For the combined levels (see MF=3, MT=51-71 above),    ',
  'the angular distributions for the first MT number were used in    ',
  'the level combinations.                                           ',
  '                                                                  ',
  'MT=53,56,58,60: DISCRETE INELASTIC NEUTRON ANGULAR DISTRIBUTIONS  ',
  '                FOR THE GROUND-STATE ROTATIONAL BAND              ',
  '   Threshold to 20.0 MeV:  Angular distributions determined from  ',
  'coupled-channel optical model calculations with ECIS70 (Ra70),    ',
  'plus compound-nucleus contributions from COMNUC calculations.     ',
  'These are the 9/2-, 11/2-, 13/2-, 15/2- members of the 7/2-       ',
  'ground-state rotational band. These data are taken from ENDF/B-   ',
  'VI.                                                               ',
  '                                                                  ',
  'MT=67-90: DISCRETE INELASTIC NEUTRON ANGULAR DISTRIBUTIONS FOR    ',
  '          DIRECT REACTIONS                                        ',
  '   Threshold to 20.0 MeV:  Angular distributions determined for   ',
  'MT=67-71 with distorted wave Born approximation calculations      ',
  'using DWUCK code (Ku70) and for MT=72-90 with the ECIS94 code     ',
  '(Ra94).                                                           ',
  '                                                                  ',
  '            ----- REFERENCES (MF=4) -----                         ',
  'Du71 C.L.Dunford, "Compound Nucleus Reaction Analysis Programs    ',
  '     COMNUC and CASCADE," Atomics International North American    ',
  '     Rockwell report AI-AEC-12931 (document no. TI-707-130-013)   ',
  '     (February, 1971).                                            ',
  'Ku70 P.D. Kunz, DWUCK: A Distorted-Wave Born Approximation        ',
  '     Program, unpublished report.                                 ',
  'Ra70 J.Raynal,IAEA SMR-9/8 (1970).                                ',
  'Ra94 J. Raynal, "Notes on ECIS94," Centre d\'Etudes Nucleaires     ',
  '     (Saclay) report CEA-N-2772 (1994).                           ',
  'Yo94 P. G. Young, "Experience at Los Alamos with Use of the       ',
  '     Optical Model for Applied Nuclear Data Calculations," Los    ',
  '     Alamos National Laboratory report LA-UR-94-3104 (1994).      ',
  '                                                                  ',
  '>> MF=5 ENERGY DISTRIBUTIONS OF SECONDARY PARTICLES               ',
  '                                                                  ',
  'MT=18 PROMPT FISSION NEUTRON SPECTRUM MATRIX                      ',
  '   Prompt fission neutron spectrum matrix for the n + 235U system ',
  'was calculated (Ma03) using the Los Alamos (Madland-Nix) model    ',
  '(MN82) in its exact formulation with energy-dependent compound    ',
  'nucleus formation cross sections for the inverse processes. The   ',
  'matrix includes first-, second-, and third-chance fission         ',
  'components and also includes the neutrons evaporated prior to     ',
  'fission in second- and third-chance fission. The tabulated        ',
  'distribution law (LF=1) is used.                                  ',
  '   The matrix is calculated for 19 incident neutron energies.     ',
  'These are:                                                        ',
  '     0.0, 0.5, 1.0, 1.5, 2.0, 2.5, 3.0, 4.0, 5.0, 6.0, 7.0, 8.0,  ',
  '     9.0, 10.0, 11.0, 12.0, 13.0, 14.0, 15.0 MeV.                 ',
  '   The 20-MeV spectrum is simply a duplication of the 15-MeV      ',
  'spectrum.                                                         ',
  '   Note that the thermal spectrum is identical to that of ENDF/B- ',
  'VI. This is due to the fact that the two most recent differential ',
  'measurements of the thermal spectrum and the most accepted set of ',
  'integral cross section measurements in the thermal spectrum       ',
  'constitute three mutually incompatible experimental sets. This    ',
  'incompatibility has yet to be resolved.                           ',
  '                                                                  ',
  'MT=455 DELAYED NEUTRON EMISSION SPECTRA FROM FISSION              ',
  '   The abundances are based on the new delayed-neutron            ',
  'calculations (Wi05) described under MF=1, MT=455. However, the    ',
  'earlier six-group spectra from ENDF/B-VI are carried over to      ',
  'ENDF/B-VII because new multigroup spectra have not yet been       ',
  'calculated with the new library.                                  ',
  '                                                                  ',
  '           ----- REFERENCES (MF=5) -----                          ',
  'Ma03 D. G. Madland, private communication, 15 May 2003.           ',
  'MN82 D. G. Madland and J. R. Nix, Nucl. Sci. Eng. 81 (1982)       ',
  '     213-271.                                                     ',
  'Wi05 W. B. Wilson, personal communication, 2005.                  ',
  '                                                                  ',
  '>> MF=6 PRODUCT ENERGY-ANGULAR DISTRIBUTIONS                      ',
  '                                                                  ',
  'MT=16,17,37: (n,xn) CONTINUUM DISTRIBUTIONS                       ',
  '   Based on the GNASH Hauser-Feshbach statistical/preequilibrium  ',
  'calculations described above. Updated Kalbach-Mann systematics    ',
  'are used for specifying neutron distributions (Ka88). Only        ',
  'neutron distributions are given.                                  ',
  '                                                                  ',
  "MT=91: (n,n') CONTINUUM DISTRIBUTIONS                             ",
  '   Based on the GNASH Hauser-Feshbach statistical/preequilibrium  ',
  'calculations described above. Updated Kalbach-Mann systematics    ',
  'are used for specifying neutron distributions (Ka88). Only        ',
  'neutron distributions are given.                                  ',
  '                                                                  ',
  '          ----- REFERENCES (MF=6) -----                           ',
  'Ka88 C. Kalbach, "Systematics of Continuum Angular                ',
  '     Distributions: Extensions to Higher Energies," Phys.Rev.C    ',
  '     37, 2350 (1988); see also C. Kalbach and F. M. Mann,         ',
  '     "Phenomenology of Continuum Angular Distributions. I.        ',
  '     Systematics and Parameterization," Phys.Rev.C 23, 112        ',
  '     (1981).                                                      ',
  '                                                                  ',
  '>> MF=12 PHOTON PRODUCTION MULTIPLICITIES AND TRANSITION          ',
  '         PROBABILITIES                                            ',
  '                                                                  ',
  'MT=4,18,102: PHOTON MULTIPLICITIES FROM INELASTIC SCATTERING,     ',
  '             FISSION, AND RADIATIVE CAPTURE                       ',
  '   1.e-5 eV - 1.09 MeV:  Based on the ENDF/B-VI evaluation by     ',
  'Hunter and Stewart (Hu72).                                        ',
  '  The yield for MT=18 is changed from 7.17 to 7.0437 in order to  ',
  ' maintain consistency with the MF=1,MT=458 data.                  ',
  '                                                                  ',
  '>> MF=13 PHOTON PRODUCTION CROSS SECTIONS                         ',
  '                                                                  ',
  'MT=3: PHOTON PRODUCTION CROSS SECTIONS FROM NONELASTIC REACTIONS  ',
  '   1.09 - 20.0 MeV:  Based on the experimental data in references ',
  'Bu71, Dr74, Dr78, and Ne66. The evaluation is the same as the     ',
  'ENDF/B-VI evaluation [see reference St76].                        ',
  '                                                                  ',
  '>> MF=14 PHOTON ANGULAR DISTRIBUTIONS                             ',
  '                                                                  ',
  'MT=3,4,18,102: PHOTON ANGULAR DISTRIBUTIONS FROM NONELASTIC       ',
  '               REACTIONS                                          ',
  '   Same as the ENDF/B-VI evaluation [see reference St76].         ',
  '                                                                  ',
  '>> MF=15 CONTINUOUS PHOTON ENERGY SPECTRA                         ',
  '                                                                  ',
  'MT=3: PHOTON ENERGY SPECTRA FROM NONELASTIC REACTIONS             ',
  '   Based on experimental data in references Dr74 and Dr78.        ',
  '                                                                  ',
  'MT=18,102: PHOTON ENERGY SPECTRA FROM FISSION AND RADIATIVE       ',
  '           CAPTURE                                                ',
  '   Same as the ENDF/B-VI evaluation [see reference St76].         ',
  '                                                                  ',
  '            ----- REFERENCES (MF=12-15) -----                     ',
  'Bu71 P. S. Buchanan, D. O. Nellis and W. E. Tucker,"A Compilation ',
  '     of Cross Sections and Angular Distributions of Gamma Rays    ',
  '     Produced by by Neutron Bombardment of Various Nuclei," ORO-  ',
  '     2791-32(Feb.1971)                                            ',
  'Dr74 D. M. Drake, Nucl. Sci. Eng. 55, 427 (1974).                 ',
  'Dr78 D. M. Drake, E. D. Arthur and M. G. Silbert, Nucl. Sci. Eng. ',
  '     65, 49 (1978).                                               ',
  'Hu72 R. E. Hunter and L. Stewart,"Evaluated Neutron-Induced       ',
  '     Gamma-Ray Production for Pu-239 and Pu-240," LA-4901(1972).  ',
  'Ne66 D. O. Nellis and I. L. Morgan ORO-2791-17(June 1966)         ',
  'St76 L. Stewart, H. Alter and R. Hunter, ENDF-201 (1976)          ',
  '                                                                  ',
  '************************* C O N T E N T S *********************** '],
 u'energy_max': 20000000.0,
 u'format': 6,
 u'identifier': ['----ENDF/B-VII        MATERIAL 9228         REVISION -            ',
  '-----INCIDENT NEUTRON DATA                                        ',
  '------ENDF-6 FORMAT                                               '],
 u'laboratory': 'ORNL,LANL,+',
 u'library': (u'ENDF/B', 7, 0),
 u'modification': 7,
 u'reference': '                     ',
 u'sublibrary': 10}
In [10]:
u235.target
Out[10]:
{u'ZA': 92235,
 u'excitation_energy': 0.0,
 u'fissionable': True,
 u'isomeric_state': 0,
 u'mass': 233.0248,
 u'stable': False,
 u'state': 0,
 u'temperature': 0.0,
 u'zsymam': ' 92-U -235 '}

To look at cross sections, secondary energy and angle distributions, and resonance data, we need to parse the rest of the data in the file, which can be done through the Evaluation.read(...) method.

In [11]:
u235.read()
Reading MF=1, MT=452 Total Neutrons per Fission
Reading MF=1, MT=455 Delayed Neutron Data
Reading MF=1, MT=456 Prompt Neutrons per Fission
Reading MF=1, MT=458 Energy Release Due to Fission
Reading MF=1, MT=460 Delayed Photon Data
Reading MF=2, MT=151 Resonance Parameters
Reading MF=3, MT=1 (z,total)
Reading MF=3, MT=2 (z,elastic)
Reading MF=3, MT=3 (z,nonelastic)
Reading MF=3, MT=4 (z,n)
Reading MF=3, MT=16 (z,2n)
Reading MF=3, MT=17 (z,3n)
Reading MF=3, MT=18 (z,fission)
Reading MF=3, MT=19 (z,f)
Reading MF=3, MT=20 (z,nf)
Reading MF=3, MT=21 (z,2nf)
Reading MF=3, MT=37 (z,4n)
Reading MF=3, MT=38 (z,3nf)
Reading MF=3, MT=51 (z,n1)
Reading MF=3, MT=52 (z,n2)
Reading MF=3, MT=53 (z,n3)
Reading MF=3, MT=54 (z,n4)
Reading MF=3, MT=55 (z,n5)
Reading MF=3, MT=56 (z,n6)
Reading MF=3, MT=57 (z,n7)
Reading MF=3, MT=58 (z,n8)
Reading MF=3, MT=59 (z,n9)
Reading MF=3, MT=60 (z,n10)
Reading MF=3, MT=61 (z,n11)
Reading MF=3, MT=62 (z,n12)
Reading MF=3, MT=63 (z,n13)
Reading MF=3, MT=64 (z,n14)
Reading MF=3, MT=65 (z,n15)
Reading MF=3, MT=66 (z,n16)
Reading MF=3, MT=67 (z,n17)
Reading MF=3, MT=68 (z,n18)
Reading MF=3, MT=69 (z,n19)
Reading MF=3, MT=70 (z,n20)
Reading MF=3, MT=71 (z,n21)
Reading MF=3, MT=72 (z,n22)
Reading MF=3, MT=73 (z,n23)
Reading MF=3, MT=74 (z,n24)
Reading MF=3, MT=75 (z,n25)
Reading MF=3, MT=76 (z,n26)
Reading MF=3, MT=77 (z,n27)
Reading MF=3, MT=78 (z,n28)
Reading MF=3, MT=79 (z,n29)
Reading MF=3, MT=80 (z,n30)
Reading MF=3, MT=81 (z,n31)
Reading MF=3, MT=82 (z,n32)
Reading MF=3, MT=83 (z,n33)
Reading MF=3, MT=84 (z,n34)
Reading MF=3, MT=85 (z,n35)
Reading MF=3, MT=86 (z,n36)
Reading MF=3, MT=87 (z,n37)
Reading MF=3, MT=88 (z,n38)
Reading MF=3, MT=89 (z,n39)
Reading MF=3, MT=90 (z,n40)
Reading MF=3, MT=91 (z,nc)
Reading MF=3, MT=102 (z,gamma)
Reading MF=4, MT=2 (z,elastic)
Reading MF=4, MT=18 (z,fission)
Reading MF=4, MT=51 (z,n1)
Reading MF=4, MT=52 (z,n2)
Reading MF=4, MT=53 (z,n3)
Reading MF=4, MT=54 (z,n4)
Reading MF=4, MT=55 (z,n5)
Reading MF=4, MT=56 (z,n6)
Reading MF=4, MT=57 (z,n7)
Reading MF=4, MT=58 (z,n8)
Reading MF=4, MT=59 (z,n9)
Reading MF=4, MT=60 (z,n10)
Reading MF=4, MT=61 (z,n11)
Reading MF=4, MT=62 (z,n12)
Reading MF=4, MT=63 (z,n13)
Reading MF=4, MT=64 (z,n14)
Reading MF=4, MT=65 (z,n15)
Reading MF=4, MT=66 (z,n16)
Reading MF=4, MT=67 (z,n17)
Reading MF=4, MT=68 (z,n18)
Reading MF=4, MT=69 (z,n19)
Reading MF=4, MT=70 (z,n20)
Reading MF=4, MT=71 (z,n21)
Reading MF=4, MT=72 (z,n22)
Reading MF=4, MT=73 (z,n23)
Reading MF=4, MT=74 (z,n24)
Reading MF=4, MT=75 (z,n25)
Reading MF=4, MT=76 (z,n26)
Reading MF=4, MT=77 (z,n27)
Reading MF=4, MT=78 (z,n28)
Reading MF=4, MT=79 (z,n29)
Reading MF=4, MT=80 (z,n30)
Reading MF=4, MT=81 (z,n31)
Reading MF=4, MT=82 (z,n32)
Reading MF=4, MT=83 (z,n33)
Reading MF=4, MT=84 (z,n34)
Reading MF=4, MT=85 (z,n35)
Reading MF=4, MT=86 (z,n36)
Reading MF=4, MT=87 (z,n37)
Reading MF=4, MT=88 (z,n38)
Reading MF=4, MT=89 (z,n39)
Reading MF=4, MT=90 (z,n40)
Reading MF=5, MT=18 (z,fission)
Reading MF=5, MT=455 Delayed Neutron Data
Reading MF=6, MT=16 (z,2n)
Reading MF=6, MT=17 (z,3n)
Reading MF=6, MT=37 (z,4n)
Reading MF=6, MT=91 (z,nc)
Reading MF=12, MT=4 (z,n)
Reading MF=12, MT=18 (z,fission)
Reading MF=12, MT=102 (z,gamma)
Reading MF=12, MT=460 Delayed Photon Data
Reading MF=13, MT=3 (z,nonelastic)
Reading MF=14, MT=3 (z,nonelastic)
Reading MF=14, MT=4 (z,n)
Reading MF=14, MT=18 (z,fission)
Reading MF=14, MT=102 (z,gamma)
Reading MF=14, MT=460 Delayed Photon Data
Reading MF=15, MT=3 (z,nonelastic)
Reading MF=15, MT=18 (z,fission)
Reading MF=15, MT=102 (z,gamma)

Most of the data that is parsed resides in the reactions attribute, which is a dictionary that is keyed by the MT value.

In [12]:
elastic = u235.reactions[2]
print('Elastic scattering has the following attributes:')
for attr in elastic.__dict__:
    if elastic.__dict__[attr]:
        print('  ' + attr)
Elastic scattering has the following attributes:
  files
  angular_distribution
  mt
  xs

Now with our reaction we can look at the cross section and any other associated data. The cross section elastic.xs is a Tab1 object whose (x,y) pairs can be accessed from the x and y attributes. The first ten values of the cross section are:

In [13]:
zip(elastic.xs.x[:10], elastic.xs.y[:10])
Out[13]:
[(1.0000000000000001e-05, 0.0),
 (0.0253, 0.0),
 (77.129580000000004, 0.0),
 (2250.0, 0.0),
 (2250.0, 11.204420000000001),
 (2300.0, 12.13951),
 (2500.0, 11.891019999999999),
 (2650.0, 11.528930000000001),
 (2900.0, 11.56729),
 (3000.0, 11.59604)]

Since resonances haven't been reconstructed, everything below the unresolved resonance range at 2250 keV is zero. Above that energy, we can use elastic.xs like a function to get a value at a particular energy. For example, to get the elastic cross section at 1 MeV:

In [14]:
elastic.xs(1.0e6)
Out[14]:
3.6496849999999998

We can also take a look at the angular distribution for elastic scattering.

In [15]:
esad = elastic.angular_distribution
print(esad)
<pyne.endf.AngularDistribution object at 0x7ff929d76450>
In [16]:
# Elastic scattering angular distribution at 100 keV
E = esad.energy[5]
pdf = esad.probability[5]

theta = np.linspace(0., 2*np.pi, 1000)
mu = np.cos(theta)

plt.subplot(111, polar=True)
plt.plot(theta, pdf(mu))
Out[16]:
[<matplotlib.lines.Line2D at 0x7ff929a05610>]

Ah, but elastic scattering is a simple reaction you say. What if I want information about something more complicated like fission! In the special case of fission, there is the normal reaction data as well as a special attribute on the Evaluation class called fission:

In [17]:
print(u235.reactions[18])
print(u235.fission.keys())
<ENDF Reaction: MT=18, (z,fission)>
[u'energy_release', u'yield_independent', u'yield_cumulative', u'nu', u'delayed_photon']

We can look at the neutrons released per fission:

In [18]:
E = np.logspace(-5, 6)
plt.semilogx(E, u235.fission['nu']['total'](E))
plt.xlabel('Energy (eV)')
plt.ylabel('Neutrons per fission')
Out[18]:
<matplotlib.text.Text at 0x7ff929a11d90>

The components of energy release from fission are also available to us:

In [19]:
for component, coefficients in u235.fission['energy_release'].items():
    if component != 'order':
        print('{}: {} +/- {} MeV'.format(component, coefficients[0,0], coefficients[1,0]))
fission_products: 169130000.0 +/- 490000.0 MeV
delayed_betas: 6500000.0 +/- 50000.0 MeV
neutrinos: 8750000.0 +/- 70000.0 MeV
total_less_neutrinos: 193483400.0 +/- 150000.0 MeV
prompt_neutrons: 4916000.0 +/- 70000.0 MeV
delayed_neutrons: 7400.0 +/- 1110.0 MeV
total: 202233400.0 +/- 130000.0 MeV
prompt_gammas: 6600000.0 +/- 500000.0 MeV
delayed_gammas: 6330000.0 +/- 50000.0 MeV

To look at the fission energy distribution, we must use the normal reaction data:

In [20]:
# Get prompt fission neutron spectra
fission = u235.reactions[18]
pfns = fission.energy_distribution[0]

# Plot the distribution for the lowest incoming energy
plt.semilogx(pfns.pdf[0].x, pfns.pdf[0].y)
plt.xlabel('Energy (eV)')
plt.ylabel('Probability / eV')
plt.title('Neutron spectrum at E={} eV'.format(pfns.energy[0]))
Out[20]:
<matplotlib.text.Text at 0x7ff9297deb90>

Finally, let's take a look at resolved resonance data, which can be found in the resonances dictionary.

In [21]:
rrr = u235.resonances['resolved']
print(rrr)

# Show all (l,J) combinations
print(rrr.resonances.keys())
<pyne.endf.ReichMoore object at 0x7ff929dc7850>
[(0, 3.0), (0, 4.0)]
In [22]:
# Set up headers for table
headers = ['Energy', 'Neutron width', 'Capture width', 'FissionA width', 'FissionB width']

# Get resonance data for l=0, J=3
l = 0
J = 3.0

# Create table data
data = [[r.energy, r.width_neutron, r.width_gamma, r.width_fissionA, r.width_fissionB]
        for r in rrr.resonances[l,J]]

# Render table
HTML(tabulate(data, headers=headers, tablefmt='html'))
Out[22]:
Energy Neutron width Capture width FissionA width FissionB width
-2038.3 0.019703 0.033792 -0.046652 -0.10088
-1812.1 0.0008574 0.037445 0.73617 -0.74187
-1586.2 0.0082845 0.034439 0.15365 -0.099186
-1357.5 0.050787 0.038506 -0.16914 -0.38622
-515.88 2.9884 0.03803 -0.81285 -0.81805
-74.766 0.38375 0.052085 -0.8644 -0.78652
-3.4928 8.539e-08 0.037791 -0.0068844 0.012977
-1.5043 8.5333e-08 0.037828 -0.0070397 0.011686
-0.56098 0.00029974 0.020855 0.095644 -0.011839
0.273793 4.2486e-06 0.046203 0.11771 0.00034848
2.03488 9.0045e-06 0.038045 -0.0096908 0.00037599
3.14289 2.3824e-05 0.038151 -0.031653 0.070502
3.8696 4.4501e-07 0.038904 0.13761 -0.0018067
6.16459 5.3266e-05 0.039758 -0.083615 0.10512
6.98885 1.5637e-06 0.040067 0.04406 0.071915
7.65909 2.0241e-06 0.039758 0.087785 0.04138
8.92592 8.295e-05 0.050484 -0.18998 0.10429
9.69634 3.5647e-05 0.039758 -0.003997 -0.22075
10.7111 7.2781e-06 0.039758 -0.0061735 -0.32271
12.39 0.0013712 0.040985 0.0173 -0.016389
13.598 0.00025477 0.040275 -0.0029924 -0.61789
14.1026 0.00020831 0.024255 -0.3445 0.14588
14.5531 0.00012853 0.046234 0.00087499 0.021571
15.2437 1.6697e-06 0.039857 0.18426 0.089568
17.5436 3.1317e-06 0.039753 0.048819 0.11383
18.0198 0.00035689 0.036368 1.0451e-06 -0.14981
18.492 1.3953e-05 0.040275 -0.63043 -0.1101
19.2283 0.00012108 0.042988 7.3124e-07 0.020707
20.1843 2.4933e-05 0.039857 0.027639 0.006801
20.932 1.5659e-05 0.039857 0.1348 0.059383
23.5912 0.00084254 0.04344 0.14677 -0.059554
24.2381 0.00041703 0.04974 -3.599e-05 -0.030401
25.5826 0.0012718 0.034386 -0.29131 0.33268
26.4641 0.00038439 0.039107 -0.22817 0.21599
27.1577 5.6624e-05 0.042795 -0.00025949 -0.054566
28.0186 3.7423e-06 0.039857 0.026191 0.032917
28.3705 0.00021981 0.044599 0.0080966 0.11572
29 5.2933e-06 0.039857 0.076317 0.23007
30.5915 0.00019587 0.041321 -0.011296 0.068178
32.0934 0.00081623 0.039591 -0.01916 0.010666
33.65 4.2152e-05 0.039857 0.036606 0.020347
34.5334 0.00043534 0.040859 0.011941 0.23936
34.8724 0.0017278 0.043851 -0.11628 0.02733
35.1716 0.0033819 0.041613 -0.076564 0.0028464
39.1285 3.9723e-05 0.044026 -0.0042766 0.0061865
39.8849 0.00035718 0.04265 0.0055795 -0.15973
41.535 0.00058982 0.041406 -0.045422 0.18643
41.8613 0.0014248 0.038191 -0.012106 -0.00059062
43.382 0.0007244 0.039765 -0.0041727 -0.013435
44.9335 0.00090145 0.041955 0.22338 -0.28673
46.9 8.1762e-05 0.039857 -0.1682 0.086147
48.3131 0.0010506 0.039202 0.18303 -0.00067141
48.7699 0.00090899 0.03994 0.0011572 -0.081059
50.094 0.00031195 0.036633 -0.021936 -0.0019242
50.4437 0.00096846 0.037045 0.042698 -0.0048907
51.1893 0.001369 0.041559 -0.10814 0.02774
52.2184 0.0031472 0.03653 -0.00026835 0.4237
53.4167 0.00057522 0.038436 -0.068472 -0.020913
54.9529 0.00066566 0.04059 -0.045892 0.020739
56.0303 0.00099273 0.038545 -0.13202 0.012265
57.682 0.00069665 0.041852 -0.14898 0.066486
58.0735 0.0017224 0.043263 0.05717 0.0022823
60.1737 0.0015618 0.041508 0.048639 0.25604
61.075 0.00040318 0.040461 -0.091954 0.00046587
61.7812 1.9374e-05 0.039857 0.0084772 0.17909
62.9689 3.1299e-06 0.042065 -0.45997 0.15533
63.643 0.0014389 0.043887 0.22748 0.7878
65.7752 0.00041467 0.038434 -0.013545 -0.016141
66.3305 7.5462e-05 0.040054 0.025122 0.14356
68.5923 8.8457e-05 0.041265 0.01718 -0.042938
69.6122 3.6018e-05 0.039861 0.056897 0.064609
70.2426 0.00045752 0.038354 -0.32232 -0.17636
70.4115 0.0014451 0.035698 0.28903 -6.8226e-06
70.7819 0.0022402 0.03791 -0.047156 0.12665
72.9251 0.00026466 0.040269 -3.773e-05 0.32275
74.4992 0.0014038 0.03865 -0.10777 0.054648
75.5603 0.0019775 0.055291 0.22111 0.071048
76.6829 3.4863e-05 0.039857 0.038207 0.034888
78.0839 0.0012847 0.044481 0.0015725 0.097842
80.3727 0.00094156 0.040298 0.1316 -0.00037581
81.4173 0.0011969 0.041417 0.00026211 -0.1174
82.426 0.00010664 0.041928 0.001809 0.049279
83.8556 0.0021151 0.042115 0.24387 -0.41409
84.5754 0.0002844 0.041161 -0.070914 0.07513
85.0531 0.0020853 0.041558 -0.014241 0.52568
86.7933 0.00053152 0.045098 0.00036126 -0.11453
87.5879 0.00058973 0.045062 -0.1933 0.069271
88.6979 0.002771 0.041289 -0.00081324 0.39522
89.7965 0.000128 0.039857 0.14792 0.07316
91.2525 0.0031395 0.040657 0.13037 0.15588
92.3511 7.0943e-05 0.039857 0.099071 0.088603
93.2226 0.00037718 0.043554 0.067649 -0.064681
94.7372 0.00051932 0.040858 -0.001474 0.043855
96.1418 0.00015908 0.039857 0.06763 0.084556
96.3482 0.0010144 0.041825 0.14669 0.43332
98.0853 0.0027879 0.040739 0.17133 0.026076
99.472 0.00041474 0.039404 0.04088 0.18134
101.255 5.6329e-07 0.041736 -0.21272 0.08824
101.815 0.00040812 0.041038 -0.067301 0.0002608
103.446 0.0001842 0.042204 0.19348 -0.051836
104.144 0.00028909 0.043382 -0.093562 0.10715
105.14 0.0026637 0.041799 -0.13877 0.044306
106.872 0.00040767 0.042197 0.31616 -0.21387
107.622 0.0047407 0.03585 -0.013362 0.00081017
109.045 1.1996e-08 0.041736 0.25065 0.13525
110.07 0.0013347 0.045588 -0.0089666 -0.30914
111.15 9.4973e-05 0.042271 0.1692 -0.11205
113.496 0.0016946 0.044383 -0.17782 -0.0042807
114.581 0.00017995 0.042036 -0.28578 1.3477
115.93 0.0022697 0.032442 0.00093706 -0.24458
117.642 0.00085932 0.042079 -0.86976 -0.33615
118.539 1.5221e-07 0.042141 -0.074636 0.24059
120.403 8.2405e-06 0.042162 0.026277 -0.081337
121.034 0.00063598 0.041939 0.33821 1.209
123.449 0.00048571 0.041169 0.043188 0.14327
123.755 0.00024751 0.04146 -0.39767 -0.0022612
125.234 0.00030917 0.041939 0.0060538 -0.061421
126.444 0.0035935 0.038837 0.10947 -0.13624
128.122 4.1875e-07 0.042141 -0.070884 -0.12297
129.536 0.00077359 0.042169 2.1659e-05 0.10724
131.384 0.0014388 0.040082 -0.42592 -0.0069382
132.283 0.0013274 0.04051 0.18721 0.056842
133.039 0.0010536 0.040795 0.12353 -0.070781
135.442 0.0050753 0.03469 0.047434 0.25815
135.805 4.3589e-06 0.042141 -0.45494 0.037587
139.126 0.00062902 0.049025 -0.00067134 0.029446
140.096 0.0013055 0.043245 -0.39004 0.0012005
141.819 0.0019195 0.045175 0.0021788 -0.06863
142.774 0.00038928 0.042151 -0.20068 0.28913
143.759 2.3462e-06 0.042141 0.086693 -0.019684
145.547 0.005322 0.044518 -0.00041795 -0.12686
146.987 2.4796e-06 0.042141 0.055245 -0.17432
147.327 0.0031476 0.040858 0.049243 -0.00027471
148.888 0.0021134 0.045226 -0.00046788 0.25518
149.269 0.00223 0.048649 0.10446 -0.17952
151.975 3.5621e-05 0.042228 -0.31763 0.33403
152.18 5.2297e-05 0.041894 -0.094372 0.026354
154.207 1.2817e-05 0.042277 -0.0054913 -0.0041703
155.289 0.00073623 0.04058 -0.068406 0.10856
156.169 0.0018308 0.032838 0.016941 0.002266
157.453 0.00090737 0.039242 0.01237 -0.0004948
158.508 0.0013254 0.038929 0.091898 0.069027
159.91 0.00080413 0.041008 -0.00064136 -0.000248
160.937 0.0075251 0.039461 -0.012377 0.00046038
161.774 4.6832e-05 0.042046 0.26723 0.12763
162.546 0.0018119 0.041271 -0.017126 -0.49988
166.259 0.0032934 0.039238 0.004909 0.31485
168.525 0.00038823 0.044061 0.0058684 0.0072023
169.394 0.0025306 0.042249 -0.024411 -0.17623
171.298 1.9404e-05 0.042083 0.067415 0.020764
171.728 0.0016202 0.050028 0.00042192 -0.063568
174.534 0.0046666 0.041825 0.23246 -0.001944
175.013 0.0013058 0.041781 -0.025537 -0.083466
175.912 8.6635e-05 0.04154 -0.06244 -0.00067126
177.536 0.0094439 0.042334 0.097858 -0.0030726
181.917 0.0023983 0.041995 0.13776 -0.2834
182.234 2.4305e-05 0.041998 -0.00033479 0.21591
182.279 0.0031532 0.048768 0.036026 -0.0009633
184.064 0.0010419 0.040429 0.0037021 -0.00080528
187.498 6.5726e-05 0.042086 0.040797 0.051298
187.81 0.00024896 0.042119 0.045622 -0.37825
188.865 0.00023489 0.043605 -0.0047533 0.00079472
189.501 0.0062153 0.03941 0.00061455 -0.03436
189.959 0.00029676 0.042624 0.032211 0.033276
192.337 0.010459 0.039421 -0.0059137 0.057524
193.076 3.0007e-06 0.042337 -0.40586 0.05478
193.427 6.7945e-05 0.042322 -0.025818 0.099755
196.043 0.00024529 0.042249 -0.18028 0.092694
197.157 0.0012277 0.046474 -0.02239 0.0073256
198.057 0.00055558 0.036025 -0.0040186 0.56528
198.832 0.0063807 0.030467 0.43231 -0.3613
200.571 0.00041985 0.042303 0.36571 -0.014424
203.07 0.00060354 0.04199 -0.071046 -0.022543
204.078 0.00086923 0.042994 0.028424 -0.05326
205.846 1.4323e-06 0.042141 -0.0094031 0.044546
208.357 0.00014367 0.042204 0.39473 0.00012514
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1921.32 0.0017676 0.039355 0.026059 0.019758
1924.67 0.00085985 0.045086 0.022971 -0.019812
1925.41 0.0021686 0.033951 0.031107 -0.025473
1926.62 0.0017163 0.034589 -0.024371 -0.02553
1928.81 0.0022531 0.042879 -0.016692 -0.019253
1930.78 0.023123 0.043084 0.075743 -0.081915
1934.19 0.0036593 0.034731 -0.028408 -0.03023
1935.29 0.012196 0.039575 0.049349 0.040138
1940.62 0.063246 0.064179 -0.13277 -0.021257
1941.19 0.00098359 0.045426 0.03017 -0.028276
1942.03 0.0043762 0.042748 -0.022186 0.021838
1945.24 0.024349 0.065254 0.02674 -0.025416
1947.91 0.016094 0.046102 -0.017254 -0.018765
1951.84 0.015315 0.047084 -0.024015 0.020842
1954.77 0.020039 0.048647 0.037573 0.0053673
1956.52 0.0044752 0.042926 0.0099142 -0.011447
1958.86 0.0012247 0.043192 -0.023341 -0.020494
1962.79 0.015261 0.052201 -0.13749 -0.034376
1967.21 0.0020805 0.046556 0.036492 -0.016928
1967.66 0.0010009 0.043864 0.019856 0.019114
1967.88 0.0036424 0.037227 -0.045879 0.026901
1971.02 0.0010565 0.039114 0.019693 -0.024305
1972.4 0.020009 0.068738 -0.018433 -0.015537
1976.6 0.0047396 0.042068 0.01271 -0.013549
1979.43 0.0010491 0.044184 -0.020947 -0.021164
1979.61 0.038195 0.037345 -0.27807 0.027119
1980.9 0.0015935 0.039991 0.027448 0.026496
1982.98 0.0077433 0.035771 -0.045142 -0.040527
1986.02 0.021988 0.033686 -0.046444 0.038636
1988.14 0.0020278 0.044554 -0.020297 0.021667
1988.56 0.0008954 0.045746 0.013701 -0.016255
1992.38 0.0060825 0.031092 0.0087928 0.01414
1994.53 0.0016925 0.04391 0.020833 -0.02262
1994.69 0.01261 0.031472 0.058614 -0.062237
1997.72 0.0018178 0.037261 0.034494 0.029043
1999.77 0.015427 0.034439 0.10063 0.039873
2000.05 0.0093308 0.0382 -0.051836 0.031842
2001.91 0.0021772 0.0382 0.030623 -0.03249
2002.58 0.0028282 0.0382 -0.005494 -0.016653
2002.91 0.018593 0.0382 -0.080057 0.10933
2004.92 0.00079571 0.0382 -0.02816 0.027478
2004.99 0.0011632 0.0382 0.026001 -0.02639
2006.56 0.0073174 0.0382 -0.035474 0.033374
2006.86 0.0038023 0.0382 0.023025 0.024506
2008.56 0.0066252 0.0382 -0.016243 -0.01819
2009.02 0.0022069 0.0382 0.021413 -0.022542
2010.35 0.005045 0.0382 -0.018759 0.021879
2011.56 0.005008 0.0382 -0.019649 0.022736
2012.77 0.0030409 0.0382 0.022573 0.025041
2013.63 0.0078751 0.0382 -0.01633 0.018803
2014.51 0.0011722 0.0382 0.023317 0.024986
2015.9 0.0011927 0.0382 -0.022873 0.024978
2016.28 0.005969 0.0382 -0.016827 0.019321
2017.77 0.0034393 0.0382 -0.021675 -0.024138
2018.22 0.0017224 0.0382 0.023265 -0.025221
2019.43 0.0026096 0.0382 0.021819 0.024688
2020.17 0.0046935 0.0382 0.020639 0.023451
2022.12 0.0019683 0.0382 -0.037242 0.027369
2022.25 0.0019326 0.0382 0.026208 0.028802
2024.3 0.001318 0.0382 -0.027753 -0.028687
2024.53 0.0032214 0.0382 -0.025627 0.025957
2025.03 0.0015565 0.0382 0.029508 -0.028103
2027.11 0.0019852 0.0382 -0.02165 -0.019886
2027.51 0.0064149 0.0382 -0.020663 0.019382
2029.04 0.001326 0.0382 0.024642 -0.023045
2029.41 0.00028889 0.0382 0.023906 0.023382
2030.88 0.00096993 0.0382 -0.025392 0.025712
2031.43 0.016121 0.0382 0.020069 -0.020338
2032.02 0.0011096 0.0382 0.02421 -0.024347
2034.29 0.001244 0.0382 0.026198 0.02586
2034.49 0.0085195 0.0382 0.016225 0.02004
2036.22 0.0013885 0.0382 0.02334 -0.023299
2036.46 0.0019842 0.0382 -0.023318 -0.023671
2037.65 0.0023487 0.0382 0.019267 -0.019444
2038.15 0.0075417 0.0382 0.0062329 -0.0063068
2040.73 0.012007 0.0382 0.017437 -0.019328
2041.67 0.0091782 0.0382 0.059714 0.058032
2042.73 0.0060076 0.0382 0.033256 0.036136
2043.87 0.0020114 0.0382 0.030676 -0.031808
2044.41 0.0064811 0.0382 -0.039584 0.039812
2045.96 0.0093723 0.0382 0.027179 -0.026893
2046.24 0.0010687 0.0382 -0.025641 0.025019
2047.89 0.00011285 0.0382 0.023767 0.023653
2049.03 0.001944 0.0382 -0.023811 -0.022776
2050.13 0.060423 0.0382 0.015403 0.013952
2050.81 0.003822 0.0382 -0.023187 -0.022164
2051.64 0.0016307 0.0382 0.023505 -0.023453
2053.02 0.040412 0.0382 -0.023078 0.029624
2054.65 0.014743 0.0382 -0.070943 0.076694
2054.74 0.0049157 0.0382 -0.029032 -0.03632
2056.31 0.0033487 0.0382 -0.058304 0.054341
2057.19 0.0032036 0.0382 0.046601 -0.049197
2057.76 0.0031836 0.0382 0.047269 -0.048257
2059.28 0.0041641 0.0382 -0.058436 0.057093
2060.53 0.0037644 0.0382 0.058568 0.052003
2061.12 0.0010697 0.0382 0.032298 0.030139
2062.14 0.0012681 0.0382 0.033795 0.030815
2063.11 0.0038404 0.0382 -0.05579 -0.043569
2064.33 0.01229 0.0382 0.078497 -0.048556
2065.72 0.0016651 0.0382 0.03676 -0.031725
2066 0.00079202 0.0382 -0.02761 0.026106
2067.27 0.007504 0.0382 -0.084841 0.049111
2068.36 0.0029839 0.0382 -0.032408 -0.031631
2069.02 0.0093839 0.0382 -0.025575 0.027508
2070.74 0.0055192 0.0382 0.025682 -0.034067
2071.25 0.0061451 0.0382 -0.023893 0.028644
2072.14 0.0092522 0.0382 0.030762 0.04244
2073.13 0.0022044 0.0382 -0.027281 0.033293
2074.23 0.00529 0.0382 0.029989 0.036534
2075.49 0.003388 0.0382 -0.027462 -0.030867
2077.03 0.0014002 0.0382 0.025571 0.026651
2077.76 0.0014714 0.0382 0.025852 -0.025967
2079.06 0.0024216 0.0382 -0.028452 -0.02776
2079.45 0.010051 0.0382 -0.023015 0.0202
2081.06 0.0049252 0.0382 0.042379 0.033751
2082.21 0.0060868 0.0382 0.0546 -0.036552
2082.91 0.0027417 0.0382 0.038074 0.026997
2083.86 0.0027964 0.0382 0.036652 0.023988
2084.99 0.018178 0.0382 -0.072512 0.018462
2085.97 0.0080975 0.0382 0.030989 0.014353
2087.22 0.015802 0.0382 -0.011348 -0.0048974
2087.77 0.0014975 0.0382 -0.021728 0.016844
2089.38 0.0013714 0.0382 0.022502 -0.018821
2090.65 0.014512 0.0382 -0.0055073 0.018507
2091.92 0.0036453 0.0382 0.047412 -0.035745
2092.21 0.011139 0.0382 -0.065353 0.122
2094.05 0.0059466 0.0382 0.10119 -0.057725
2094.46 0.0034543 0.0382 -0.071971 -0.051108
2095.44 0.0098637 0.0382 0.089655 0.043551
2096.64 0.0029249 0.0382 -0.052779 0.039103
2097.35 0.0015254 0.0382 -0.037591 -0.030728
2098.33 0.0042627 0.0382 -0.042595 0.028782
2099.47 0.0040152 0.0382 0.045515 -0.029843
2100.69 0.0038623 0.0382 -0.045312 0.040644
2101.83 0.0049606 0.0382 0.024381 0.022096
2103.45 0.0014785 0.0382 0.018104 0.018604
2103.5 0.0032082 0.0382 0.0099513 0.010092
2104.51 0.0029119 0.0382 -0.010767 -0.011429
2106.43 0.0093105 0.0382 0.0044212 0.0049668
2107.32 0.0023977 0.0382 0.018627 0.023447
2108.2 0.0037676 0.0382 -0.021832 -0.029411
2109.58 0.0087474 0.0382 0.032199 0.04307
2110.26 0.0020655 0.0382 -0.027852 -0.036066
2111.4 0.0028828 0.0382 -0.030153 -0.037809
2112.38 0.0049362 0.0382 0.02655 0.028283
2114.01 0.0020537 0.0382 -0.024603 -0.027202
2114.91 0.0018978 0.0382 0.025941 0.028496
2115.56 0.0028707 0.0382 0.032169 0.032865
2116.91 0.002534 0.0382 -0.028496 0.031954
2117.93 0.0047935 0.0382 0.037138 -0.03797
2118.71 0.0024955 0.0382 0.043544 0.044937
2119.35 0.0026014 0.0382 -0.048692 -0.046918
2120.94 0.0077472 0.0382 0.076551 0.061356
2121.98 0.003841 0.0382 0.060828 0.043035
2123.62 0.001916 0.0382 0.045918 -0.028861
2124.86 0.0077755 0.0382 0.030891 0.14181
2125.26 0.0041374 0.0382 -0.065434 -0.021773
2126.96 0.0019805 0.0382 -0.033677 0.025371
2127.32 0.0030501 0.0382 0.037586 -0.024885
2128.04 0.0042421 0.0382 0.052565 0.027774
2128.88 0.015393 0.0382 -0.029342 0.017799
2130.8 0.0065609 0.0382 0.042646 0.025445
2131.91 0.0073317 0.0382 -0.048228 0.029229
2132.72 0.0059409 0.0382 0.048985 0.066784
2133.97 0.0097019 0.0382 -0.12449 -0.045495
2134.77 0.0061109 0.0382 0.080003 -0.050121
2136.37 0.001427 0.0382 0.03353 -0.028596
2136.95 0.0076603 0.0382 -0.079768 0.044936
2138.02 0.0049646 0.0382 -0.04745 0.031393
2139.45 0.0048861 0.0382 -0.047506 0.065158
2140.65 0.0064117 0.0382 0.027502 0.018212
2141.71 0.012174 0.0382 -0.0094597 -0.006696
2142.87 0.0054702 0.0382 -0.019874 -0.013997
2144.01 0.0021488 0.0382 -0.02481 0.019635
2144.94 0.020402 0.0382 0.03558 0.015126
2145.95 0.0053819 0.0382 -0.033675 0.053928
2147.26 0.0039317 0.0382 -0.034278 -0.024414
2147.94 0.0074338 0.0382 -0.053689 0.02661
2149.06 0.009294 0.0382 0.042862 0.026795
2149.73 0.003625 0.0382 0.028205 -0.023949
2151.7 0.0041053 0.0382 0.030522 -0.023172
2152.8 0.009687 0.0382 -0.031405 0.057177
2153.8 0.0073233 0.0382 0.019609 -0.013634
2154.6 0.0039918 0.0382 0.019652 0.015272
2155.9 0.0059924 0.0382 0.013534 0.010462
2157 0.002269 0.0382 -0.02354 -0.020174
2158 0.00034811 0.0382 0.024713 0.023554
2159.4 0.0010644 0.0382 -0.027347 0.025085
2160.6 0.0082938 0.0382 0.049474 -0.034828
2161 0.0045599 0.0382 0.050711 0.037967
2162.5 0.0035674 0.0382 -0.043762 0.033615
2163.3 0.0084567 0.0382 0.053734 -0.031264
2164.8 0.016221 0.0382 0.030989 0.018305
2165.6 0.00092404 0.0382 -0.030387 -0.027426
2167.2 0.0014228 0.0382 0.031321 0.027767
2168.1 0.011164 0.0382 0.022173 -0.016109
2169.3 0.0022244 0.0382 0.029071 0.026836
2170.6 0.0027818 0.0382 0.025783 0.023851
2171.6 0.0036867 0.0382 -0.021947 -0.019493
2172.8 0.0039547 0.0382 0.0053325 -0.0049496
2173.7 0.0038338 0.0382 -0.0066797 -0.0059751
2174.9 0.0014211 0.0382 0.019637 0.017671
2176.1 0.00074753 0.0382 0.022554 -0.020593
2176.9 0.00083845 0.0382 0.023076 -0.020935
2178.3 0.0035617 0.0382 0.025758 0.021192
2179.1 0.0099385 0.0382 0.023604 -0.016383
2180.2 0.0088554 0.0382 -0.019818 -0.025215
2181.6 0.0091186 0.0382 0.063721 -0.047902
2182.3 0.0071399 0.0382 0.080962 -0.061477
2183.7 0.013718 0.0382 -0.040012 0.077713
2184.6 0.004173 0.0382 -0.04049 0.058405
2185.8 0.0040215 0.0382 0.035469 -0.028475
2187.7 0.0043409 0.0382 0.031631 -0.027096
2188.5 0.0043844 0.0382 -0.020531 0.025169
2189.4 0.012839 0.0382 -0.0090885 0.011822
2190.8 0.0027126 0.0382 -0.019592 -0.022255
2191.5 0.0031348 0.0382 -0.019382 0.021922
2193.2 0.014422 0.0382 0.014543 -0.010165
2193.6 0.0015244 0.0382 0.024775 -0.021324
2195.2 0.0013518 0.0382 0.024546 0.026885
2196.3 0.012099 0.0382 0.014183 -0.0082439
2197.1 0.0021905 0.0382 0.023298 0.025911
2198.7 0.0019632 0.0382 0.024374 0.024214
2199.2 0.0075346 0.0382 -0.010453 0.015822
2201.03 0.0053488 0.0382 -0.017223 0.021833
2201.48 0.0038585 0.0382 -0.023766 0.027926
2202.8 0.016707 0.0382 0.050669 0.029595
2204.42 0.0039979 0.0382 0.047888 -0.041653
2205.49 0.004326 0.0382 0.049637 0.043641
2206.16 0.0051826 0.0382 -0.036247 -0.039213
2207.32 0.0063124 0.0382 0.063116 -0.040227
2207.85 0.0015927 0.0382 -0.029583 -0.025505
2210.06 0.0025675 0.0382 0.035959 -0.025533
2210.17 0.0059857 0.0382 -0.0275 0.050875
2213.34 0.014411 0.0382 -0.086403 0.059736
2213.59 0.0085422 0.0382 0.040704 -0.032422
2216.24 0.0046507 0.0382 0.095386 -0.053696
2217.45 0.013259 0.0382 -0.070226 0.23877
2218.77 0.0020703 0.0382 0.052781 0.047614
2220.41 0.0045478 0.0382 0.069529 0.065413
2220.5 0.002708 0.0382 -0.03108 0.044514
2221.06 0.0052006 0.0382 -0.02721 -0.028573
2222.21 0.0055284 0.0382 -0.020963 -0.022048
2224.5 0.0063866 0.0382 -0.017455 0.063241
2225.89 0.0063003 0.0382 -0.037934 0.025096
2227.91 0.013086 0.0382 -0.033956 0.025442
2229.38 0.0034165 0.0382 -0.026144 -0.031188
2229.4 0.006321 0.0382 -0.026163 0.027536
2231.36 0.011273 0.0382 0.029341 -0.027157
2231.46 0.0018232 0.0382 -0.027217 -0.026068
2234.38 0.0032873 0.0382 0.026969 -0.039809
2234.38 0.0021682 0.0382 0.023174 0.027126
2234.49 0.0029577 0.0382 0.029143 0.028069
2236.97 0.0016565 0.0382 -0.033234 -0.035833
2237.08 0.0059469 0.0382 0.032795 -0.045803
2239.34 0.012216 0.0382 0.065732 -0.17723
2239.64 0.0053098 0.0382 -0.045341 0.040048
2241.35 0.0036008 0.0382 0.046788 0.029937
2241.41 0.0047699 0.0382 -0.074981 -0.10328
2244.19 0.002697 0.0382 -0.05473 -0.059622
2244.2 0.0015052 0.0382 -0.03882 0.026443
2245.96 0.003476 0.0382 -0.09574 0.06483
2246.14 0.0031453 0.0382 0.057812 -0.0839
2250.3 0.022405 0.068425 -0.41171 -0.12273
2254.2 0.025518 0.094863 0.026743 0.042032
2256.2 0.01423 0.049379 0.025013 0.03631
2283.8 7.159 0.099886 0.87653 0.46888
2630.4 7.8534 0.045164 0.7068 0.53647
3330.8 12.057 0.047228 0.47442 0.57129
4500.9 6.1439 0.033681 0.28662 0.36414