IRDF2002 in metrology format cross sections (050112)              6000 0  0
 3.00600E+3 5.96340E+0          0          0         34         10 325 1451
 0.0        0.0                 0          0          0          6 325 1451
 1.00000E+0 2.00000E+7          0          0         10       2002 325 1451
 3.00000E+2 0.0                 1          0        237          3 325 1451
  3-Li-  6 LANL       EVAL-APR89 G.M.HALE, P.G.YOUNG               325 1451
                      DIST-Feb2004                                 325 1451
----IRDF-2002         MATERIAL  325                                325 1451
-----INCIDENT NEUTRON DATA                                         325 1451
------ENDF-6 FORMAT                                                325 1451
****************************************************************** 325 1451
  3-LI - 6 LANL       EVAL-APR89 G.M.HALE, P.G.YOUNG               325 1451
                      DIST-SEP91 REV1-JUL91            19910806    325 1451
----ENDF/B-VI         MATERIAL  325         REVISION 1             325 1451
****************************************************************** 325 1451
  3-LI - 6 LANL       EVAL-APR89 G.M.HALE, P.G.YOUNG               325 1451
                      DIST-FEB91                       910201      325 1451
----ENDF/B-VI         MATERIAL  325                                325 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 325 1451
DOSIMETRY                                                          325 1451
****************************************************************** 325 1451
                                                       891221      325 1451
----ENDF/B-VI         MATERIAL  325                                325 1451
                                                                   325 1451
****************************************************************** 325 1451
                                                                   325 1451
 ***************************************************************** 325 1451
 ***************************************************************** 325 1451
                                                                   325 1451
                  MOD1 OF ENDF/B-VI                                325 1451
                                                                   325 1451
     The following revisions were made for MOD1 of ENDF/B-VI:      325 1451
                                                                   325 1451
 1. MF=1,MT=451 - Comments were added regarding estimated          325 1451
    (expanded) covariance for the Standards Cross Sections.        325 1451
 2. MF=3,MT=53 - LF flag and Q-value corrected.                    325 1451
                                                                   325 1451
 ***************************************************************** 325 1451
****************************************************************** 325 1451
                                                                   325 1451
                        ENDF/VI EVALUATION                         325 1451
                     G. M. Hale and P. G. Young                    325 1451
                                                                   325 1451
 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE:                       325 1451
                                                                   325 1451
   1. Inclusion of the ENDF/B-VI standard (n,t) cross section      325 1451
      from the simultaneous standards analysis (ca85) over the     325 1451
      energy range thermal to 1 MeV.                               325 1451
   2. Replacement of all major cross sections and elastic angular  325 1451
      distributions at energies between 10^-5 eV and 3 MeV with    325 1451
      results from the R-matrix analysis performed in conjunction  325 1451
      with the simultaneous standards analysis.                    325 1451
   3. Revision of the elastic cross sections and angular distri-   325 1451
      butions at energies between 3 and 20 MeV to match recent     325 1451
      experimental data, resulting in a general decrease of the    325 1451
      elastic cross section in this energy range.                  325 1451
   4. Revision of the (n,n')d cross sections to account for        325 1451
      recent measurements, resulting in a general increase in      325 1451
      the total (n,n')d cross section that tends to offset the     325 1451
      decrease in the elastic cross section and maintain about     325 1451
      the same total cross section as before.                      325 1451
                                                                   325 1451
                                                                   325 1451
****************************************************************** 325 1451
                                                                   325 1451
                        STANDARDS COVARIANCES                      325 1451
                                                                   325 1451
 Phase 1 reviewers of the ENDF/B-VI standards cross sections have  325 1451
 expressed the concern that the uncertainties resulting from the   325 1451
 combination of R-matrix and simultaneous evaluations might have   325 1451
 led to uncertainties that are too small.  As a result, the        325 1451
 Standards Subcommittee produced (at the May, 1990 CSEWG meeting)  325 1451
 a set of expanded covariance estimates for the standard cross     325 1451
 section reactions.  These uncertainties are estimates such that   325 1451
 if a modern day experiment were performed on a given standard     325 1451
 cross section using the best techniques, approximately 2/3 of     325 1451
 the results should fall within these expanded uncertainties. The  325 1451
 expanded uncertainties for the Li-6(n,t) cross section are given  325 1451
 in the following table and are compared to values from the        325 1451
 combined output of the standards covariance analysis:             325 1451
                                                                   325 1451
    Energy Range    Estimated Uncertainty    Combined Analysis     325 1451
      (keV)              (percent)               (percent)         325 1451
                                                                   325 1451
 1.0E-08 - 0.1               0.3                     0.14          325 1451
     0.1 - 1.0               0.5                                   325 1451
     1.0 - 10.               0.7                     0.14          325 1451
     10. - 50.               0.9                                   325 1451
     50. - 90.               1.1                     0.25          325 1451
     90. - 150               1.5                                   325 1451
     150 - 450               2.0                     0.29          325 1451
     450 - 650               5.0                                   325 1451
     650 - 800               2.0                     0.36          325 1451
     800 - 1000              5.0                                   325 1451
                                                                   325 1451
 ****************************************************************  325 1451
                                                                   325 1451
                                                                   325 1451
 mf=2 --------- resonance parameters ----------------------------  325 1451
                                                                   325 1451
   mt=151  effective scattering radius = 2.31175e-13 cm.           325 1451
                                                                   325 1451
 mf=3 --------- smooth cross sections ---------------------------  325 1451
                                                                   325 1451
      the 2200 m/s cross sections are as follows:                  325 1451
          mt=1     sigma = 941.6928  barns                         325 1451
          mt=2     sigma = 0.67157   barns                         325 1451
          mt=102   sigma = 0.03850   barns                         325 1451
          mt=105   sigma = 940.9827  barns                         325 1451
                                                                   325 1451
   mt=105        (n,t) cross section                               325 1451
       below 3 mev, values are taken from the r-matrix analysis,   325 1451
          which includes (n,t) measurements from re78, la78, br77, 325 1451
          ov74, and ba75.  between 3 and 5 mev, the values are     325 1451
          based on ba75, and at higher energies are taken from the 325 1451
          evaluation of pe64, extended to 20 mev considering the   325 1451
          data of ke58.                                            325 1451
                                                                   325 1451
                                                                   325 1451
                                                                   325 1451
 mf=33----------cross section covariances------------------------- 325 1451
                   (to be added later)                             325 1451
                                                                   325 1451
  the relative covariances for mt=1,2, and 105 below 4 mev are     325 1451
  given in file 33. they are based on calculations using the co-   325 1451
  variances of the r-matrix parameters in first-order error        325 1451
  propogation.                                                     325 1451
   mt=1         total                                              325 1451
       relative covariances entered as nc-type sub-subsection,     325 1451
          implying that they are to be constructed from those for  325 1451
          mt=2 and 105.  they are not intended for use at energies 325 1451
          above 4 mev.                                             325 1451
   mt=2,105     elastic and (n,t)                                  325 1451
       relative covariances among these two cross sections are     325 1451
          entered explicitly as ni-type sub-subsections in the     325 1451
          lb=5 (direct) representation at energies below 4 mev.    325 1451
          although values for the 3.95-4.05 mev bin are repeated   325 1451
          in a 4-20 mev bin, the covariances are not intended for  325 1451
          use at energies above 4 mev.                             325 1451
                                                                   325 1451
                                                                   325 1451
--------------------- references --------------------------------  325 1451
                                                                   325 1451
ab70  U.Abbondanno, Nuo.Cim. A166,139(1970).                       325 1451
aj74  f.ajzenberg-selove and t.lauritsen, nucl. phys. a227,55      325 1451
      (1974).                                                      325 1451
ar64  a.h.armstrong, j.gammel, l.rosen, and g.m.frye, nucl. phys.  325 1451
      52,505 (1964).                                               325 1451
as63  v.j.ashby et al, phys. rev. 129,1771 (1963).                 325 1451
ba53  m.e.battat and f.l.ribe, phys.rev. 89,80 (1953).             325 1451
ba63  r.batchelor and j.h.towle, nucl. phys. 47,385 (1963).        325 1451
ba65  r.bass, c.bindhardt, and k.kruger, eandc(e)-57u (1965).      325 1451
ba75  c.m.bartle, proc. conf. on nuclear cross sections and        325 1451
      technology, vol.2,688 (1975),  and private communication     325 1451
      (1976).  see also nucl. phys. a330, 1 (1979).                325 1451
be75  besotosnyj et al., yk-19, 77 (1975).                         325 1451
br77  r.e.brown,g.g.ohlsen,r.f.haglund, and n.jarmie, phys. rev.   325 1451
      16c, 513 (1977).                                             325 1451
ca85  a.d.carlson,w.p.poenitz,g.m.hale, and r.w.peele, nuclear     325 1451
      data for basic and applied science (santa fe, n.m.), 1429    325 1451
      (1985).                                                      325 1451
co67  j.a.cookson and d.dandy, nucl. phys. a91,273 (1967).         325 1451
co82  h.conde,t.andersson,l.nilsson, and c.nordborg, nuclear data  325 1451
      for science and technology (antwerp, belgium), 447 (1982).   325 1451
de73  F.Demanins et al., infn/be-73 (1973).                        325 1451
dr82  m.drosg,d.m.drake,r.a.hardekopf, and g.m.hale, la-9129-ms    325 1451
      (1982).                                                      325 1451
dr85  m.drosg et al., santa fe conf.1, 145(1985).                  325 1451
fo71  d.g.foster and d.w.glasgow, phys. rev. c3,576 (1971).        325 1451
fr54  g.m.frye, phys. rev. 93,1086 (1954).                         325 1451
go72  c.a.goulding and p.stoler, eandc(us)-176u,161 (1972).        325 1451
ha75  j.a.harvey and n.w.hill, nuclear cross sections and          325 1451
      technology (washington, d.c.), 244 (1975).                   325 1451
ha84  g.m.hale, nuclear standard reference data (geel,belgium)     325 1451
      iaea tecdoc-335, 103 (1984). describes preliminary analysis. 325 1451
ho68  j.c.hopkins,d.m.drake, and h.conde, nucl. phys. a107,139     325 1451
      (1968), and j.c.hopkins, d.m.drake, and h.conde, la-3765     325 1451
      (1967).                                                      325 1451
ho79  h.h.hogue et al., n.s.&e. 69, 22 (1979).                     325 1451
ju73  e.t.jurney, lasl, private communication (1973).              325 1451
ke58  r.d.kern and w.e.kreger, phys. rev. 112, 926 (1958).         325 1451
ke79  j.d.kellie,g.p.lamaze, and r.b.schwartz, nuclear cross       325 1451
      sections for technology (knoxville, tn.), 48 (1979).         325 1451
kn77  h.h.knitter,c.budtz-jorgensen,m.mailly, and r.vogt, eur-     325 1451
      5726e (1977).                                                325 1451
kn79  h.d.knox,r.m.white, and r.o.lane, n.s.&e. 69, 223 (1979).    325 1451
kn83  h.h.knitter,c.budtz-jorgensen,d.l.smith, and d.marletta,     325 1451
      n.s.&e. 83, 229(1983).                                       325 1451
la61  r.o.lane,a.s.langsdorf,j.e.monahan, and a.j.elwyn, ann.      325 1451
      phys.12, 135 (1961).                                         325 1451
la78  g.p.lamaze,o.a.wasson,r.a.schrack, and a.d.carlson, n.s.&e.  325 1451
      68, (1978).                                                  325 1451
li80  p.w.lisowski et al., la-8342 (1980).                         325 1451
ma69  d.s.mather and l.f.paine, awre-o-47/69 (1969).               325 1451
me65  f.merchez,n.v.sen,v.regis, and r.bouchez, compt. rend. 260,  325 1451
      3922 (1965).                                                 325 1451
ov74  j.c.overley,r.m.sealock, and d.h.ehlers, nucl. phys. a221,   325 1451
      573 (1974).                                                  325 1451
pe64  e.d.pendlebury, awre-o-60/64 (1964).                         325 1451
pr69  g.presser et al., nuc.phys. a131, 679(1969).                 325 1451
re78  c.renner,j.a.harvey,n.w.hill,g.l.morgan, and k.pusk, bull.   325 1451
      am. phys. soc. 23, 526 (1978).                               325 1451
sa82  e.t.sadowski,h.knox,d.a.resler, and r.o.lane, bap 27,624(c5) 325 1451
      (1982).                                                      325 1451
sm77  a.b.smith,p.guenther,d.havel, and j.f.whalen, anl/ndm-29     325 1451
      (1977).                                                      325 1451
sm82  a.b.smith,p.t.guenther, and j.f.whalen, nucl. phys. a373,    325 1451
      305 (1982).                                                  325 1451
wo62  c.wong,j.d.anderson, and j.w.mcclure, nucl. phys. 33,680     325 1451
      (1962).                                                      325 1451
****************************************************************** 325 1451
                                                                   325 1451
                                                                   325 1451
                                                                   325 1451
                                                                   325 1451
****************************************************************** 325 1451
                                                                   325 1451
 ***************** Program LINEAR (VERSION 2002-1) *************** 325 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value      325 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent      325 1451
 ***************** Program SIGMA1 (VERSION 2002-1) *************** 325 1451
 Data Doppler Broadened to 300.000000 Kelvin                       325 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value      325 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent      325 1451
 ***************** Program FIXUP (Version 2002-1) **************** 325 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes  325 1451
 Corrected Thresholds-----------------------------------------Yes  325 1451
 Extended Cross Sections to 20 MeV----------------------------No   325 1451
 Allow Cross Section Deletion---------------------------------No   325 1451
 Allow Cross Section Reconstruction---------------------------No   325 1451
 Make All Cross Sections Non-Negative-------------------------Yes  325 1451
 Delete Energies Not in Ascending Order-----------------------Yes  325 1451
 Deleted Duplicate Points-------------------------------------Yes  325 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes  325 1451
 Check for Legal MF/MT Numbers--------------------------------Yes  325 1451
 Allow Creation of Missing Sections---------------------------No   325 1451
 Allow Insertion of Energy Points-----------------------------No   325 1451
 Create Uniform Energy Grid-----------------------------------No   325 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes  325 1451
 ***************** Program GROUPIE (VERSION 2002-1) ************** 325 1451
 Unshielded Group Averages Using  640 Groups                       325 1451
 Weighting Spectrum: Flat (Constant) Spectrum                      325 1451
                                1        451        244          2 325 1451
                                3        105        217          1 325 1451
                               33        105         33          1 325 1451
                                                                   325 1  0
                                                                   325 0  0
 3.00600E+3 5.96340E+0          0          0          0          0 325 3105
 4.78380E+6 4.78380E+6          0          0          1        641 325 3105
        641          1                                             325 3105
 .000100000 14794.9461 .000105000 14449.9534 .000110000 14122.2061 325 3105
 .000115000 13820.1942 .000120000 13467.1458 .000127500 13075.4530 325 3105
 .000135000 12717.5721 .000142500 12388.6847 .000150000 12032.7367 325 3105
 .000160000 11661.9312 .000170000 11324.5439 .000180000 11014.8459 325 3105
 .000190000 10726.6978 .000200000 10463.6234 .000210000 10216.1135 325 3105
 .000220000 9986.21324 .000230000 9772.62292 .000240000 9521.91859 325 3105
 .000255000 9246.40272 .000270000 9033.79728 .000280000 8797.12696 325 3105
 .000300000 8508.50325 .000320000 8247.16668 .000340000 8007.82467 325 3105
 .000360000 7787.37574 .000380000 7586.13528 .000400000 7376.33068 325 3105
 .000425000 7161.64277 .000450000 6965.50583 .000475000 6785.66231 325 3105
 .000500000 6616.52235 .000525000 6461.94757 .000550000 6315.93048 325 3105
 .000575000 6180.13284 .000600000 6040.94642 .000630000 5897.75751 325 3105
 .000660000 5766.73697 .000690000 5641.05202 .000720000 5507.32050 325 3105
 .000760000 5363.93249 .000800000 5230.80234 .000840000 5108.78635 325 3105
 .000880000 4992.81623 .000920000 4886.50890 .000960000 4784.70637 325 3105
 .001000000 4679.19964 .001050000 4569.04414 .001100000 4465.76191 325 3105
 .001150000 4370.59678 .001200000 4258.33842 .001275000 4134.96155 325 3105
 .001350000 4022.16216 .001425000 3916.88925 .001500000 3805.04153 325 3105
 .001600000 3688.10030 .001700000 3581.34573 .001800000 3482.60854 325 3105
 .001900000 3392.40441 .002000000 3308.77765 .002100000 3230.31157 325 3105
 .002200000 3158.44480 .002300000 3089.72099 .002400000 3011.30697 325 3105
 .002550000 2924.17510 .002700000 2856.02126 .002800000 2781.96182 325 3105
 .003000000 2690.80549 .003200000 2607.85391 .003400000 2531.98785 325 3105
 .003600000 2462.79638 .003800000 2398.91608 .004000000 2332.33692 325 3105
 .004250000 2264.72689 .004500000 2203.02081 .004750000 2145.35559 325 3105
 .005000000 2092.44576 .005250000 2043.40654 .005500000 1997.05482 325 3105
 .005750000 1954.57647 .006000000 1909.96714 .006300000 1865.38391 325 3105
 .006600000 1823.18871 .006900000 1784.04467 .007200000 1741.48065 325 3105
 .007600000 1695.96373 .008000000 1654.34447 .008400000 1615.27928 325 3105
 .008800000 1578.91854 .009200000 1545.13798 .009600000 1513.00017 325 3105
 .010000000 1479.73644 .010500000 1444.57603 .011000000 1412.42313 325 3105
 .011500000 1381.75051 .012000000 1346.59615 .012750000 1307.68584 325 3105
 .013500000 1271.64181 .014250000 1238.56317 .015000000 1203.27730 325 3105
 .016000001 1166.23862 .017000001 1132.28052 .017999999 1101.28202 325 3105
 .018999999 1072.79317 .020000000 1046.06875 .021000000 1021.61496 325 3105
 .022000000 998.572690 .023000000 977.099923 .024000000 951.867476 325 3105
 .025500000 924.650334 .027000001 903.085714 .028000001 879.664218 325 3105
 .029999999 850.841313 .032000002 824.584971 .034000002 800.612528 325 3105
 .035999998 778.732778 .037999999 758.549586 .039999999 737.487875 325 3105
 .042500000 716.110259 .045000002 696.571444 .047499999 678.366928 325 3105
 .050000001 661.623345 .052499998 646.130877 .055000000 631.464013 325 3105
 .057500001 618.035173 .059999999 603.916290 .063000001 589.802419 325 3105
 .066000000 576.497152 .068999998 564.095301 .071999997 550.648650 325 3105
 .075999998 536.246480 .079999998 523.089245 .083999999 510.719324 325 3105
 .088000000 499.211911 .092000000 488.553833 .096000001 478.374129 325 3105
 .100000001 467.864268 .104999997 456.740352 .109999999 446.578283 325 3105
 .115000002 436.871661 .119999997 425.753899 .127499998 413.473559 325 3105
 .135000005 402.085828 .142499998 391.626836 .150000006 380.471804 325 3105
 .159999996 368.744950 .170000002 358.015375 .180000007 348.198886 325 3105
 .189999998 339.184378 .200000003 330.720268 .209999993 322.982244 325 3105
 .219999999 315.682189 .230000004 308.885854 .239999995 301.048031 325 3105
 .254999995 292.298232 .270000011 285.625887 .280000001 278.184188 325 3105
 .300000012 269.039086 .319999993 260.740769 .340000004 253.189079 325 3105
 .360000014 246.269336 .379999995 239.820982 .400000006 233.210767 325 3105
 .425000012 226.472440 .449999988 220.243253 .474999994 214.511408 325 3105
 .500000000 209.256293 .524999976 204.282890 .550000012 199.729453 325 3105
 .574999988 195.403926 .600000024 190.991063 .629999995 186.517846 325 3105
 .660000026 182.279486 .689999998 178.400407 .720000029 174.092531 325 3105
 .759999990 169.580921 .800000012 165.402550 .839999974 161.482057 325 3105
 .879999995 157.896838 .920000017 154.457639 .959999979 151.313264 325 3105
 1.00000000 147.924471 1.04999995 144.457738 1.10000002 141.215388 325 3105
 1.14999998 138.163257 1.20000005 134.661002 1.27499998 130.732247 325 3105
 1.35000002 127.134297 1.42499995 123.849352 1.50000000 120.293925 325 3105
 1.60000002 116.585459 1.70000005 113.209752 1.79999995 110.125223 325 3105
 1.89999998 107.244386 2.00000000 104.595170 2.09999990 102.145940 325 3105
 2.20000005 99.8337043 2.29999995 97.7171215 2.40000010 95.2040034 325 3105
 2.54999995 92.4334553 2.70000005 90.3148276 2.79999995 87.9381155 325 3105
 3.00000000 85.0523601 3.20000005 82.4380221 3.40000010 80.0560785 325 3105
 3.59999990 77.8616733 3.79999995 75.8730582 4.00000000 73.7205775 325 3105
 4.25000000 71.6078328 4.50000000 69.6686317 4.75000000 67.8455059 325 3105
 5.00000000 66.1838277 5.25000000 64.5853545 5.50000000 63.1383990 325 3105
 5.75000000 61.7747034 6.00000000 60.3789510 6.30000019 58.9664576 325 3105
 6.59999990 57.6264409 6.90000010 56.4004006 7.19999981 55.0410801 325 3105
 7.59999990 53.6130860 8.00000000 52.2937668 8.39999962 51.0497077 325 3105
 8.80000019 49.9143943 9.19999981 48.8331393 9.60000038 47.8550339 325 3105
 10.0000000 46.7674076 10.5000000 45.6690583 11.0000000 44.6357610 325 3105
 11.5000000 43.7022554 12.0000000 42.5635629 12.7500000 41.3292052 325 3105
 13.5000000 40.1934306 14.2500000 39.1440984 15.0000000 38.0253790 325 3105
 16.0000000 36.8529448 17.0000000 35.7847221 18.0000000 34.8158482 325 3105
 19.0000000 33.8933511 20.0000000 33.0523746 21.0000000 32.2802396 325 3105
 22.0000000 31.5498964 23.0000000 30.8900598 24.0000000 30.0863941 325 3105
 25.5000000 29.2029000 27.0000000 28.5498764 28.0000000 27.7953130 325 3105
 30.0000000 26.8770847 32.0000000 26.0473656 34.0000000 25.2913699 325 3105
 36.0000000 24.6097325 38.0000000 23.9570474 40.0000000 23.2925112 325 3105
 42.5000000 22.6200978 45.0000000 21.9989139 47.5000000 21.4210590 325 3105
 50.0000000 20.8953058 52.5000000 20.3989799 55.0000000 19.9413886 325 3105
 57.5000000 19.5133607 60.0000000 19.0727461 63.0000000 18.6328468 325 3105
 66.0000000 18.2080695 69.0000000 17.8128281 72.0000000 17.3948470 325 3105
 76.0000000 16.9347672 80.0000000 16.5172078 84.0000000 16.1339473 325 3105
 88.0000000 15.7710512 92.0000000 15.4221555 96.0000000 15.1053270 325 3105
 100.000000 14.7700328 105.000000 14.4288661 110.000000 14.0965950 325 3105
 115.000000 13.8008879 120.000000 13.4398837 127.500000 13.0466643 325 3105
 135.000000 12.6979150 142.500000 12.3611838 150.000000 12.0094109 325 3105
 160.000000 11.6403331 170.000000 11.2996687 180.000000 10.9872033 325 3105
 190.000000 10.7071240 200.000000 10.4367810 210.000000 10.1921719 325 3105
 220.000000 9.96131944 230.000000 9.74939037 240.000000 9.49846206 325 3105
 255.000000 9.22265603 270.000000 9.01502091 280.000000 8.77535127 325 3105
 300.000000 8.48444616 320.000000 8.22571754 340.000000 7.98405981 325 3105
 360.000000 7.76663440 380.000000 7.56467016 400.000000 7.35376806 325 3105
 425.000000 7.13932704 450.000000 6.94443603 475.000000 6.76435003 325 3105
 500.000000 6.59661622 525.000000 6.44315602 550.000000 6.29397110 325 3105
 575.000000 6.15893198 600.000000 6.02324074 630.000000 5.87760877 325 3105
 660.000000 5.74686517 690.000000 5.62296263 720.000000 5.48674802 325 3105
 760.000000 5.34606717 800.000000 5.21088095 840.000000 5.08749818 325 3105
 880.000000 4.97416821 920.000000 4.86865092 960.000000 4.76557776 325 3105
 1000.00000 4.66214100 1050.00000 4.55068283 1100.00000 4.44663325 325 3105
 1150.00000 4.35342467 1200.00000 4.23972217 1275.00000 4.11709895 325 3105
 1350.00000 4.00425444 1425.00000 3.89861606 1500.00000 3.78734478 325 3105
 1600.00000 3.67248621 1700.00000 3.56544145 1800.00000 3.46853432 325 3105
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 10000000.0 .037800215 10100000.0 .037400218 10200000.0 .037000222 325 3105
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 10600000.0 .035400235 10700000.0 .035000238 10800000.0 .034600241 325 3105
 10900000.0 .034200244 11000000.0 .033900196 11100000.0 .033700097 325 3105
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 19900000.0 .015090401 20000000.0 0.0                              325 3105
                                                                   325 3  0
                                                                   325 0  0
 3.00600E+3 5.96340E+0          0          0          0          1 32533105
 0.000000+0 0.000000+0          0        105          0          2 32533105
 0.000000+0 0.000000+0          1          5         36          8 32533105
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                                                                   32533  0
                                                                   325 0  0
                                                                     0 0  0
 5.00000E+3 1.07191E+1         -1          0         34          0 500 1451
 0.0        0.0                 0          0          0          6 500 1451
 1.00000E+0 2.00000E+7          0          0         10       2002 500 1451
 3.00000E+2 0.0                 1          0         75          2 500 1451
  5-B -  0  NDS       IAEA-JAN04                                   500 1451
                      DIST-Feb2004                                 500 1451
----IRDF-2002                                                      500 1451
-----INCIDENT NEUTRON DATA                                         500 1451
------ENDF-6 FORMAT                                                500 1451
 ***************************************************************** 500 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 500 1451
DOSIMETRY     Assembled at NDS from the ENDF/B-VI evaluations for  500 1451
              5-B -10 and 5-B -11. The files were processed        500 1451
              through the 2002 pre-processing codes LINEAR,RECENT, 500 1451
              SIGMA1 and FIXUP to produce pointwise files prior to 500 1451
              input to MIXER.                                      500 1451
                                                                   500 1451
MIXER INPUT FILE                                                   500 1451
                                                                   500 1451
Boron   DENSITY  2.34 G/CC. CONSTITUENTS:  10-19.8 11-80.2         500 1451
FIXUP.out                                                          500 1451
MIXER.OUT                                                          500 1451
    5000      500 3  1            (ID FOR COMBINED REACTION)       500 1451
    5010             1 0.429086                                    500 1451
    5011             1 1.910914                                    500 1451
                                 (BLANK LINE TERMINATES LIST)      500 1451
****************************************************************** 500 1451
                                                                   500 1451
  5-B - 10 LANL       EVAL-NOV89 G.M.HALE, P.G.YOUNG               500 1451
                      DIST-SEP91 REV1-JUL91            19910806    500 1451
----ENDF/B-VI         MATERIAL  525         REVISION 1             500 1451
 ***************************************************************** 500 1451
  5-B - 11 LANL       EVAL-MAY89 P.G.YOUNG                         500 1451
                      DIST-SEP 1 REV1-                 20010926    500 1451
----ENDF/B-VI         MATERIAL  528                                500 1451
 ***************** Program MIXER (VERSION 2002-1) **************** 500 1451
Boron   DENSITY  2.34 G/CC. CONSTITUENTS:  10-19.8                 500 1451
 11-80.2                                                           500 1451
 ----------------------------------------                          500 1451
 Composition                                                       500 1451
 ----------------------------------------                          500 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                          500 1451
 ----------------------------------------                          500 1451
    5010  3    1  .198003533   .429086000                          500 1451
    5011  3    1  .801996467   1.91091400                          500 1451
 ----------------------------------------                          500 1451
    5000  3    1  1.00000000   2.34000000                          500 1451
 ----------------------------------------                          500 1451
 Composition                                                       500 1451
 ----------------------------------------                          500 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                          500 1451
 ----------------------------------------                          500 1451
    5010  3  101  .198003533   .429086000                          500 1451
    5011  3  101  .801996467   1.91091400                          500 1451
 ----------------------------------------                          500 1451
    5000  3  101  1.00000000   2.34000000                          500 1451
 ----------------------------------------                          500 1451
 Composition                                                       500 1451
 ----------------------------------------                          500 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                          500 1451
 ----------------------------------------                          500 1451
    5010  3  102  .198003533   .429086000                          500 1451
    5011  3  102  .801996467   1.91091400                          500 1451
 ----------------------------------------                          500 1451
    5000  3  102  1.00000000   2.34000000                          500 1451
 ----------------------------------------                          500 1451
 Composition                                                       500 1451
 ----------------------------------------                          500 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                          500 1451
 ----------------------------------------                          500 1451
    5010  3  107  .198003533   .429086000                          500 1451
    5011  3  107  .801996467   1.91091400                          500 1451
 ----------------------------------------                          500 1451
    5000  3  107  1.00000000   2.34000000                          500 1451
 ----------------------------------------                          500 1451
 ***************************************************************** 500 1451
 ***************** Program GROUPIE (VERSION 2002-1) ************** 500 1451
 Unshielded Group Averages Using  640 Groups                       500 1451
 Weighting Spectrum: Flat (Constant) Spectrum                      500 1451
                                1        451         81          0 500 1451
                                3          1        217          0 500 1451
                                                                   500 1  0
                                                                   500 0  0
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 1.00000000 123.780844 1.04999995 120.979502 1.10000002 118.361080 500 3  1
 1.14999998 115.897618 1.20000005 113.067735 1.27499998 109.891331 500 3  1
 1.35000002 106.981880 1.42499995 104.329054 1.50000000 101.458040 500 3  1
 1.60000002 98.4619583 1.70000005 95.7344614 1.79999995 93.2450433 500 3  1
 1.89999998 90.9191035 2.00000000 88.7743706 2.09999990 86.7961282 500 3  1
 2.20000005 84.9314533 2.29999995 83.2225104 2.40000010 81.1922432 500 3  1
 2.54999995 78.9523968 2.70000005 77.2405994 2.79999995 75.3210344 500 3  1
 3.00000000 72.9906486 3.20000005 70.8785626 3.40000010 68.9541277 500 3  1
 3.59999990 67.1812474 3.79999995 65.5745264 4.00000000 63.8352665 500 3  1
 4.25000000 62.1282650 4.50000000 60.5624597 4.75000000 59.0903402 500 3  1
 5.00000000 57.7484882 5.25000000 56.4578220 5.50000000 55.2898433 500 3  1
 5.75000000 54.1886310 6.00000000 53.0612548 6.30000019 51.9351776 500 3  1
 6.59999990 50.8628833 6.90000010 49.8318790 7.19999981 48.7534904 500 3  1
 7.59999990 47.6206168 8.00000000 46.5237547 8.39999962 45.5355385 500 3  1
 8.80000019 44.6037394 9.19999981 43.7383185 9.60000038 42.9248150 500 3  1
 10.0000000 42.0606968 10.5000000 41.1776911 11.0000000 40.3423886 500 3  1
 11.5000000 39.5880260 12.0000000 38.6679061 12.7500000 37.6703547 500 3  1
 13.5000000 36.7520020 14.2500000 35.9019256 15.0000000 34.9993163 500 3  1
 16.0000000 34.0655369 17.0000000 33.1869710 18.0000000 32.4040102 500 3  1
 19.0000000 31.6588069 20.0000000 30.9789236 21.0000000 30.3549965 500 3  1
 22.0000000 29.7635975 23.0000000 29.2193144 24.0000000 28.5750189 500 3  1
 25.5000000 27.8695586 27.0000000 27.3402256 28.0000000 26.7310982 500 3  1
 30.0000000 25.9887737 32.0000000 25.3180434 34.0000000 24.7074621 500 3  1
 36.0000000 24.1574705 38.0000000 23.6306474 40.0000000 23.0936576 500 3  1
 42.5000000 22.5505658 45.0000000 22.0486452 47.5000000 21.5814268 500 3  1
 50.0000000 21.1567832 52.5000000 20.7559690 55.0000000 20.3863362 500 3  1
 57.5000000 20.0408924 60.0000000 19.6855835 63.0000000 19.3302654 500 3  1
 66.0000000 18.9869251 69.0000000 18.6677337 72.0000000 18.3306065 500 3  1
 76.0000000 17.9594100 80.0000000 17.6222249 84.0000000 17.3127579 500 3  1
 88.0000000 17.0197600 92.0000000 16.7380090 96.0000000 16.4820209 500 3  1
 100.000000 16.2112453 105.000000 15.9359587 110.000000 15.6678476 500 3  1
 115.000000 15.4291974 120.000000 15.1378186 127.500000 14.8203232 500 3  1
 135.000000 14.5397363 142.500000 14.2691231 150.000000 13.9867706 500 3  1
 160.000000 13.6818332 170.000000 13.4109171 180.000000 13.1555221 500 3  1
 190.000000 12.9290606 200.000000 12.7104858 210.000000 12.5127684 500 3  1
 220.000000 12.3261774 230.000000 12.1548099 240.000000 11.9518957 500 3  1
 255.000000 11.7288684 270.000000 11.5609853 280.000000 11.3672011 500 3  1
 300.000000 11.1319023 320.000000 10.9228323 340.000000 10.7276606 500 3  1
 360.000000 10.5517956 380.000000 10.3883506 400.000000 10.2176820 500 3  1
 425.000000 10.0440874 450.000000 9.88629658 475.000000 9.74048627 500 3  1
 500.000000 9.60470812 525.000000 9.48052301 550.000000 9.35992088 500 3  1
 575.000000 9.25110798 600.000000 9.14177387 630.000000 9.02441847 500 3  1
 660.000000 8.91883084 690.000000 8.81865059 720.000000 8.70850763 500 3  1
 760.000000 8.59473669 800.000000 8.48608332 840.000000 8.38851795 500 3  1
 880.000000 8.29316123 920.000000 8.20644559 960.000000 8.12315904 500 3  1
 1000.00000 8.03946134 1050.00000 7.94925379 1100.00000 7.86504477 500 3  1
 1150.00000 7.78964206 1200.00000 7.69766506 1275.00000 7.59849181 500 3  1
 1350.00000 7.50749127 1425.00000 7.42287540 1500.00000 7.33376315 500 3  1
 1600.00000 7.23670734 1700.00000 7.15187245 1800.00000 7.07332652 500 3  1
 1900.00000 6.99994868 2000.00000 6.93257038 2100.00000 6.86828554 500 3  1
 2200.00000 6.81015996 2300.00000 6.75548978 2400.00000 6.69166669 500 3  1
 2550.00000 6.62125864 2700.00000 6.56668506 2800.00000 6.50669835 500 3  1
 3000.00000 6.43357144 3200.00000 6.36681847 3400.00000 6.30451939 500 3  1
 3600.00000 6.24910370 3800.00000 6.19848492 4000.00000 6.14452902 500 3  1
 4250.00000 6.08944901 4500.00000 6.03907865 4750.00000 5.99352313 500 3  1
 5000.00000 5.95106929 5250.00000 5.91098294 5500.00000 5.87402293 500 3  1
 5750.00000 5.83987590 6000.00000 5.80436403 6300.00000 5.76947723 500 3  1
 6600.00000 5.73549550 6900.00000 5.70427986 7200.00000 5.67022453 500 3  1
 7600.00000 5.63332483 8000.00000 5.59949985 8400.00000 5.56855642 500 3  1
 8800.00000 5.53922932 9200.00000 5.51319652 9600.00000 5.48895931 500 3  1
 10000.0000 5.46369329 10500.0000 5.43846691 11000.0000 5.41544923 500 3  1
 11500.0000 5.39315352 12000.0000 5.36803123 12750.0000 5.34047638 500 3  1
 13500.0000 5.31546685 14250.0000 5.29293566 15000.0000 5.26954641 500 3  1
 16000.0000 5.24479834 17000.0000 5.22305385 18000.0000 5.20353444 500 3  1
 19000.0000 7.12541739 20000.0000 7.24241608 21000.0000 5.30190192 500 3  1
 22000.0000 5.27208174 23000.0000 5.24361992 24000.0000 5.20954070 500 3  1
 25500.0000 5.17078401 27000.0000 5.13980809 28000.0000 5.10455490 500 3  1
 30000.0000 5.06574581 32000.0000 5.03067657 34000.0000 4.99761670 500 3  1
 36000.0000 4.96584401 38000.0000 4.93518887 40000.0000 4.90246721 500 3  1
 42500.0000 4.86690317 45000.0000 4.83340578 47500.0000 4.80129341 500 3  1
 50000.0000 4.77057581 52500.0000 4.74064233 55000.0000 4.71163987 500 3  1
 57500.0000 4.69070842 60000.0000 4.67139378 63000.0000 4.65111479 500 3  1
 66000.0000 4.63211792 69000.0000 4.61365727 72000.0000 4.59310668 500 3  1
 76000.0000 4.57056718 80000.0000 4.54847495 84000.0000 4.52740384 500 3  1
 88000.0000 4.50728513 92000.0000 4.48720889 96000.0000 4.46740112 500 3  1
 100000.000 4.44870019 105000.000 4.43069561 110000.000 4.41328427 500 3  1
 115000.000 4.39598139 120000.000 4.37454597 127500.000 4.34637401 500 3  1
 135000.000 4.31858511 142500.000 4.29123051 150000.000 4.25933798 500 3  1
 160000.000 4.22542717 170000.000 4.19635793 180000.000 4.17003849 500 3  1
 190000.000 4.15300908 200000.000 4.14638603 210000.000 4.14356882 500 3  1
 220000.000 4.13121488 230000.000 4.11053499 240000.000 4.06861928 500 3  1
 255000.000 4.00069574 270000.000 3.94037507 280000.000 3.86665886 500 3  1
 300000.000 3.77145733 320000.000 3.72772652 340000.000 3.74655040 500 3  1
 360000.000 3.77211330 380000.000 4.02057150 400000.000 5.36633122 500 3  1
 425000.000 6.04627264 450000.000 4.26821710 475000.000 3.56523272 500 3  1
 500000.000 3.31205000 525000.000 3.16060395 550000.000 3.02711998 500 3  1
 575000.000 2.90538691 600000.000 2.78646325 630000.000 2.68123077 500 3  1
 660000.000 2.55864635 690000.000 2.44576134 720000.000 2.32608590 500 3  1
 760000.000 2.18668294 800000.000 2.07697838 840000.000 2.00464709 500 3  1
 880000.000 1.91487082 920000.000 1.85988545 960000.000 1.85966237 500 3  1
 1000000.00 1.99844282 1100000.00 2.59010874 1200000.00 4.01108567 500 3  1
 1300000.00 3.29526407 1400000.00 2.58285930 1500000.00 2.34079584 500 3  1
 1600000.00 2.26392860 1700000.00 2.49474450 1800000.00 2.24504623 500 3  1
 1900000.00 2.03376151 2000000.00 1.91019507 2100000.00 1.84747066 500 3  1
 2200000.00 1.83318987 2300000.00 2.12063722 2400000.00 2.49494552 500 3  1
 2500000.00 2.50719296 2600000.00 2.13667038 2700000.00 2.01931759 500 3  1
 2800000.00 1.99504862 2900000.00 1.89499661 3000000.00 1.76919974 500 3  1
 3100000.00 1.68098607 3200000.00 1.59990897 3300000.00 1.56423816 500 3  1
 3400000.00 1.59961572 3500000.00 1.66588162 3600000.00 1.61988151 500 3  1
 3700000.00 1.56748149 3800000.00 1.53208693 3900000.00 1.49023547 500 3  1
 4000000.00 1.41791554 4100000.00 1.39714830 4200000.00 1.41892165 500 3  1
 4300000.00 1.42692084 4400000.00 1.46503309 4500000.00 1.62307184 500 3  1
 4600000.00 1.72925520 4700000.00 1.75089854 4800000.00 1.74667762 500 3  1
 4900000.00 1.64537827 5000000.00 1.64952549 5100000.00 1.69619115 500 3  1
 5200000.00 1.71851055 5300000.00 1.78666562 5400000.00 1.44672473 500 3  1
 5500000.00 1.44124459 5600000.00 1.72178195 5700000.00 1.63618106 500 3  1
 5800000.00 1.72992253 5900000.00 1.64376607 6000000.00 1.57266729 500 3  1
 6100000.00 1.68638830 6200000.00 1.74083966 6300000.00 1.62271447 500 3  1
 6400000.00 1.55876905 6500000.00 1.53342879 6600000.00 1.54450370 500 3  1
 6700000.00 1.55372630 6800000.00 1.53101850 6900000.00 1.55950292 500 3  1
 7000000.00 1.54318696 7100000.00 1.51561331 7200000.00 1.50573037 500 3  1
 7300000.00 1.49005228 7400000.00 1.50076964 7500000.00 1.48213986 500 3  1
 7600000.00 1.47902476 7700000.00 1.53547522 7800000.00 1.64734023 500 3  1
 7900000.00 1.57205020 8000000.00 1.51171522 8100000.00 1.46456050 500 3  1
 8200000.00 1.42233878 8300000.00 1.39253049 8400000.00 1.37306198 500 3  1
 8500000.00 1.37095860 8600000.00 1.41436323 8700000.00 1.42332974 500 3  1
 8800000.00 1.39454074 8900000.00 1.37760768 9000000.00 1.37347945 500 3  1
 9100000.00 1.36667184 9200000.00 1.36164867 9300000.00 1.36455817 500 3  1
 9400000.00 1.36882641 9500000.00 1.40362699 9600000.00 1.49427005 500 3  1
 9700000.00 1.56920927 9800000.00 1.44474503 9900000.00 1.42598883 500 3  1
 10000000.0 1.42652300 10100000.0 1.45975500 10200000.0 1.43764007 500 3  1
 10300000.0 1.39331617 10400000.0 1.36668711 10500000.0 1.37095629 500 3  1
 10600000.0 1.39204230 10700000.0 1.41662913 10800000.0 1.44704417 500 3  1
 10900000.0 1.46829442 11000000.0 1.46436689 11100000.0 1.44554160 500 3  1
 11200000.0 1.43183455 11300000.0 1.41973046 11400000.0 1.40928706 500 3  1
 11500000.0 1.40245608 11600000.0 1.40090111 11700000.0 1.40411213 500 3  1
 11800000.0 1.40939096 11900000.0 1.41500850 12000000.0 1.41899236 500 3  1
 12100000.0 1.41927267 12200000.0 1.42002358 12300000.0 1.42055289 500 3  1
 12400000.0 1.42114992 12500000.0 1.42179096 12600000.0 1.42196161 500 3  1
 12700000.0 1.42229522 12800000.0 1.42235987 12900000.0 1.42263263 500 3  1
 13000000.0 1.42405226 13100000.0 1.42427635 13200000.0 1.42446247 500 3  1
 13300000.0 1.42459770 13400000.0 1.42479286 13500000.0 1.42496373 500 3  1
 13600000.0 1.42513722 13700000.0 1.42529220 13800000.0 1.42573183 500 3  1
 13900000.0 1.42636889 14000000.0 1.42653165 14100000.0 1.42620005 500 3  1
 14200000.0 1.42604859 14300000.0 1.42607695 14400000.0 1.42620803 500 3  1
 14500000.0 1.42645059 14600000.0 1.42640948 14700000.0 1.42608295 500 3  1
 14800000.0 1.42604421 14900000.0 1.42627763 15000000.0 1.42640597 500 3  1
 15100000.0 1.42642142 15200000.0 1.42620980 15300000.0 1.42577383 500 3  1
 15400000.0 1.42552047 15500000.0 1.42545867 15600000.0 1.42546625 500 3  1
 15700000.0 1.42552674 15800000.0 1.42540050 15900000.0 1.42507008 500 3  1
 16000000.0 1.42454069 16100000.0 1.42380209 16200000.0 1.42319399 500 3  1
 16300000.0 1.42273229 16400000.0 1.42230819 16500000.0 1.42192518 500 3  1
 16600000.0 1.42134787 16700000.0 1.42055453 16800000.0 1.41979771 500 3  1
 16900000.0 1.41908886 17000000.0 1.41825331 17100000.0 1.41730518 500 3  1
 17200000.0 1.41640109 17300000.0 1.41553330 17400000.0 1.41458477 500 3  1
 17500000.0 1.41355380 17600000.0 1.41256255 17700000.0 1.41161949 500 3  1
 17800000.0 1.41074469 17900000.0 1.40994157 18000000.0 1.40904997 500 3  1
 18100000.0 1.40808103 18200000.0 1.40715562 18300000.0 1.40626368 500 3  1
 18400000.0 1.40541552 18500000.0 1.40460736 18600000.0 1.40372433 500 3  1
 18700000.0 1.40276326 18800000.0 1.40182117 18900000.0 1.40089943 500 3  1
 19000000.0 1.39988249 19100000.0 1.39876550 19200000.0 1.39769893 500 3  1
 19300000.0 1.39669575 19400000.0 1.39567395 19500000.0 1.39465125 500 3  1
 19600000.0 1.39354180 19700000.0 1.39233869 19800000.0 1.39113602 500 3  1
 19900000.0 1.38993562 20000000.0 0.0                              500 3  1
                                                                   500 3  0
                                                                   500 0  0
                                                                     0 0  0
 5.01000E+3 9.92692E+0          0          0         34         10 525 1451
 0.0        0.0                 0          0          0          6 525 1451
 1.00000E+0 2.00000E+7          0          0         10       2002 525 1451
 3.00000E+2 0.0                 1          0        256          3 525 1451
  5-B - 10 LANL       EVAL-NOV89 G.M.HALE, P.G.YOUNG               525 1451
                      DIST-Feb2004                                 525 1451
----IRDF-2002         MATERIAL  525                                525 1451
-----INCIDENT NEUTRON DATA                                         525 1451
------ENDF-6 FORMAT                                                525 1451
****************************************************************** 525 1451
  5-B - 10 LANL       EVAL-NOV89 G.M.HALE, P.G.YOUNG               525 1451
                      DIST-SEP91 REV1-JUL91            19910806    525 1451
----ENDF/B-VI         MATERIAL  525         REVISION 1             525 1451
****************************************************************** 525 1451
  5-B - 10 LANL       EVAL-NOV89 G.M.HALE, P.G.YOUNG               525 1451
                      DIST-FEB91                       910201      525 1451
----IRDF-90           MATERIAL  525                                525 1451
                                                                   525 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 525 1451
DOSIMETRY                                                          525 1451
****************************************************************** 525 1451
                                                                   525 1451
                  MOD1 OF ENDF/B-VI                                525 1451
                                                                   525 1451
     The following revisions were made for MOD1 of ENDF/B-VI:      525 1451
                                                                   525 1451
 1. MF=1,MT=451 - Comments were added regarding estimated          525 1451
    (expanded) covariance for the Standards Cross Sections.        525 1451
 2. MF=3,MT=55,57,62,64,65,68,70,71,73,74,76-81,83,84 - LR flags   525 1451
    and Q-values corrected.                                        525 1451
                                                                   525 1451
 ***************************************************************** 525 1451
****************************************************************** 525 1451
                                                                   525 1451
                      ENDF/VI EVALUATION                           525 1451
                 G. M. Hale and P. G. Young                        525 1451
                                                                   525 1451
 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE:                       525 1451
                                                                   525 1451
   1. Inclusion of the ENDF/B-VI standard (n,alpha) and (n,alpha1) 525 1451
      results from the simultaneous standards analysis (Ca85) over 525 1451
      the standard energy range thermal to 100 keV.                525 1451
   2. Replacement of all major cross sections and elastic angular  525 1451
      distributions from 10-5 eV to 1 MeV with results from the    525 1451
      R-matrix analysis performed in conjunction with the          525 1451
      simultaneous standards analysis.                             525 1451
   3. Replaced the total cross section 1-20 MeV with results       525 1451
      from a covariance analysis of available data.                525 1451
   4. Revised elastic and inelastic cross sections for low-lying   525 1451
      levels incorporating new elastic, inelastic, and (n,xgamma)  525 1451
      experimental data. We attempted to better reconcile the      525 1451
      inelastic and gamma ray data.                                525 1451
   5. Refit all elastic angular distributions from 1-20 MeV with   525 1451
      Legendre expansions and incorporated results from new        525 1451
      measurements.                                                525 1451
   6. Fit inelastic neutron angular distributions for first 5      525 1451
      excited states of B10 with Legendre expansions.              525 1451
   7. Incorporated new (n,t2alpha) cross section data into MT113   525 1451
      and adjusted (n,alpha) cross sections above standard region  525 1451
      for better consistency with data as well as other cross      525 1451
      sections (esp. total and elastic) determined by data.        525 1451
                                                                   525 1451
 *****Note that covariance data will be added at a later date.     525 1451
                                                                   525 1451
****************************************************************** 525 1451
                                                                   525 1451
                        STANDARDS COVARIANCES                      525 1451
                                                                   525 1451
 Phase 1 reviewers of the ENDF/B-VI standards cross sections have  525 1451
 expressed the concern that the uncertainties resulting from the   525 1451
 combination of R-matrix and simultaneous evaluations might have   525 1451
 led to uncertainties that are too small.  As a result, the        525 1451
 Standards Subcommittee produced (at the May, 1990 CSEWG meeting)  525 1451
 a set of expanded covariance estimates for the standard cross     525 1451
 section reactions.  These uncertainties are estimates such that   525 1451
 if a modern day experiment were performed on a given standard     525 1451
 cross section using the best techniques, approximately 2/3 of     525 1451
 the results should fall within these expanded uncertainties. The  525 1451
 expanded uncertainties for the B-10(n,alpha0) and B-10(n,alpha1)  525 1451
 cross sections are given in the following tables and are          525 1451
 compared to values from the combined output of the standards      525 1451
 covariance analysis:                                              525 1451
                                                                   525 1451
                                                                   525 1451
                     B-10(n,alpha0) Cross Section                  525 1451
                                                                   525 1451
    Energy Range    Estimated Uncertainty    Combined Analysis     525 1451
      (keV)              (percent)               (percent)         525 1451
                                                                   525 1451
 1.0E-08 - 0.1               0.5                     0.21          525 1451
     0.1 - 5.0               1.5                                   525 1451
     5.0 - 30.               3.0                     0.38          525 1451
     30. - 90.               5.0                                   525 1451
     90. - 150               10.0                    0.86          525 1451
     150 - 200               12.0                                  525 1451
     200 - 250               15.0                    0.79          525 1451
                                                                   525 1451
                                                                   525 1451
                                                                   525 1451
                     B-10(n,alpha1) Cross Section                  525 1451
                                                                   525 1451
    Energy Range    Estimated Uncertainty    Combined Analysis     525 1451
      (keV)              (percent)               (percent)         525 1451
                                                                   525 1451
 1.0E-08 - 0.1               0.2                    0.16           525 1451
     0.1 - 5.0               0.4                                   525 1451
     5.0 - 30.               0.6                    0.20           525 1451
     30. - 90.               1.0                                   525 1451
     90. - 150               1.5                    0.48           525 1451
     150 - 200               2.0                                   525 1451
     200 - 250               2.5                    0.62           525 1451
                                                                   525 1451
 ****************************************************************  525 1451
                                                                   525 1451
 mf=2 --------- resonance parameters ----------------------------  525 1451
                                                                   525 1451
   mt=151  effective scattering radius = 4.129038-13 cm            525 1451
                                                                   525 1451
 mf=3 --------- smooth cross sections ---------------------------  525 1451
                                                                   525 1451
      the 2200 m/s cross sections are as follows,                  525 1451
          mt=1    sigma = 3842.146  barns                          525 1451
          mt=2    sigma = 2.142435  barns                          525 1451
          mt=102  sigma = 0.5       barns                          525 1451
          mt=103  sigma = 0.000566  barns                          525 1451
          mt=107  sigma = 3839.496  barns                          525 1451
          mt=113  sigma = 0.0069993 barns                          525 1451
          mt=600  sigma = 0.000566  barns                          525 1451
          mt=800  sigma = 241.2677  barns                          525 1451
          mt=801  sigma = 3598.228  barns                          525 1451
                                                                   525 1451
   mt=1        total cross section                                 525 1451
      0 to 1 mev, calculated from r-matrix parameters obtained     525 1451
          from simultaneous standards analysis (ca85) used to      525 1451
          obtain the endf/b-vi standard cross sections.            525 1451
      1 to 20 mev, covariance analysis of measurements of di67,    525 1451
          ts62,fo61,co52,au79, and co54, constrained to match      525 1451
          r-matrix fit at 1 mev. glucs covariance analysis code    525 1451
          (he80) was used in the calculations.                     525 1451
                                                                   525 1451
                                                                   525 1451
   mt=107   (n,alpha) cross section                                525 1451
      0 to 20 mev, sum of mt=800,801.                              525 1451
                                                                   525 1451
   mt=800     (n,alpha0) cross section                             525 1451
      0 to 1 mev, calculated from the r-matrix parameters          525 1451
          described for mt=1. experimental (n,alpha0) data input   525 1451
          to the fit were those of ma68 and da61. in addition, the 525 1451
          angular distributions of va72 for the inverse reaction   525 1451
          were included in the analysis.                           525 1451
      1 to 20 mev, based on da61 measurements, with smooth extra-  525 1451
          polation from 8 to 20 mev using 14-MeV data of an69. The 525 1451
          da61 data above approximately 2 mev were renormalized    525 1451
          by a factor of approximately 1.4. Note that some of the  525 1451
          structure seen in da61 was expanded to give consistent   525 1451
          nonelastic, elastic, and total cross sections when       525 1451
          compared with experimental data.                         525 1451
                                                                   525 1451
   mt=801     (n,alpha1) cross section                             525 1451
      0 to 1 mev, calculated from the r-matrix parameters          525 1451
          described for mt=1. experimental (n,alpha1) data in-     525 1451
          cluded in the fit are those of sc76. in addition, the    525 1451
          absolute differential cross-section measurements of      525 1451
          se76 were included in the analysis.                      525 1451
      1 to 20 mev, smooth curve through measurements of da61 and   525 1451
          ne70, with smooth extrapolation from 15 to 20 mev. the   525 1451
          da61 data above approximately 2 mev were renormalized    525 1451
          by a factor of approximately 1.4. Note that some of the  525 1451
          structure seen in da61 was expanded to give consistent   525 1451
          nonelastic, elastic, and total cross sections when       525 1451
          compared with experimental data.                         525 1451
                                                                   525 1451
                                                                   525 1451
                                                                   525 1451
 ----------------------- references -----------------------------  525 1451
                                                                   525 1451
 aj75   f. ajzenberg-selove, nucl. phys. a248,6 (1975)             525 1451
 aj88   f. ajzenberg-selove, nucl. phys. a490,1 (1988)             525 1451
 an69   B. Antolkovic, Nuc.Phys.A139, 10 (1969).                   525 1451
 as70   a. asami and m.c. moxon, j.nucl.energy 24,85 (1970)        525 1451
 au79   g.auchampaugh et al., nucl.sci.eng.69,30(1979)             525 1451
 ba60   r.bardes and g.e. owen, phys.rev.120,1369 (1960)           525 1451
 be56   r.l. becker and h.h. barschall, phys.rev.102,1384 (1956)   525 1451
 bo51   c.k.bockelman et al., phys.rev. 84,69 (1951)               525 1451
 bo69   d.bogart and l.l.nichols, nucl.phys.a125,463 (1969)        525 1451
 ca85   a.carlson et al., nuc.data for basic & applied science,    525 1451
        santa fe, nm (1985) p.1429.                                525 1451
 co52   j.h.coon et al., phys.rev. 88,562 (1952)                   525 1451
 co54   c.f.cook and t.w. bonner,phys.rev. 94,651 (1954)           525 1451
 co67   s.a. cox and f.r. pontet, j.nucl.energy 21,271 (1967)      525 1451
 co69   j.a. cookson and j.g.locke,nucl.phys.a146,417(1970)        525 1451
 co73   m.s. coates et al., priv. comm. to l.stewart (1973)        525 1451
 da56   r.b.day,phys.rev.102,767 (1956)                            525 1451
 da60   r.b. day and m.walt,phys.rev.117,1330 (1960)               525 1451
 da61   e.a. davis et al., nucl.phys.27,448 (1961)                 525 1451
 di67   k.m. diment, aere-r-5224 (1967)                            525 1451
 di88   j.k.dickens, proc.conf. on nuc.data for sci.& tech.,mito,  525 1451
        japan (1988) p.213.                                        525 1451
 fo61   d.m. fossan et al., phys.rev. 123,209 (1961)               525 1451
 fr56   g.m. frye and j.h. gammel,phys.rev. 103,328 (1956)         525 1451
 gl82   s.glendinning, nuc.sci.eng.80,256(1982)                    525 1451
 ha73   s.l.hausladen, thesis, ohio univ. coo-1717-5 (1973)        525 1451
 he80   d.hetrick & c.y.fu, ornl/tm-7341 (1980).                   525 1451
 hy69   m.hyakutake, eandc(j)-13 (1969) p.29                       525 1451
 ho69   j.c. hopkins, priv. comm. lasl (1969)                      525 1451
 ir67   d.c.irving, ornl-tm-1872 (1967)                            525 1451
 ka87   R.Kavanagh & R.Marcley, Phys.Rev.C36, 1194 (1987).         525 1451
 la71   r.o. lane et al., phys.rev.c4,380 (1971)                   525 1451
 ma68   r.l.macklin and j.h.gibbons,phys.rev.165,1147 (1968)       525 1451
 mo66   f.p.mooring et al.,nucl.phys.82,16 (1966)                  525 1451
 ne54   n.g.nereson,la-1655 (1954)                                 525 1451
 ne70   d.o.nellis et al., phys.rev. c1,847 (1970)                 525 1451
 po70   d.porter et al., awre o 45/70 (1970)                       525 1451
 qa85   S.Qaim et al., Santa Fe Conf. (1985)p.97.                  525 1451
 qa88   S.Qaim et al., Mito Conf. (1988) p.225.                    525 1451
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        (1976)                                                     525 1451
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 ****************************************************************  525 1451
                                                                   525 1451
                                                                   525 1451
                                                                   525 1451
                                                                   525 1451
 ***************** Program LINEAR (VERSION 2002-1) *************** 525 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value      525 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent      525 1451
 ***************** Program SIGMA1 (VERSION 2002-1) *************** 525 1451
 Data Doppler Broadened to 300.000000 Kelvin                       525 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value      525 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent      525 1451
 ***************** Program FIXUP (Version 2002-1) **************** 525 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes  525 1451
 Corrected Thresholds-----------------------------------------Yes  525 1451
 Extended Cross Sections to 20 MeV----------------------------No   525 1451
 Allow Cross Section Deletion---------------------------------No   525 1451
 Allow Cross Section Reconstruction---------------------------No   525 1451
 Make All Cross Sections Non-Negative-------------------------Yes  525 1451
 Delete Energies Not in Ascending Order-----------------------Yes  525 1451
 Deleted Duplicate Points-------------------------------------Yes  525 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes  525 1451
 Check for Legal MF/MT Numbers--------------------------------Yes  525 1451
 Allow Creation of Missing Sections---------------------------No   525 1451
 Allow Insertion of Energy Points-----------------------------No   525 1451
 Create Uniform Energy Grid-----------------------------------No   525 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes  525 1451
 ****************************************************************  525 1451
 ***************** Program GROUPIE (VERSION 2002-1) ************** 525 1451
 Unshielded Group Averages Using  640 Groups                       525 1451
 Weighting Spectrum: Flat (Constant) Spectrum                      525 1451
                                1        451        263          2 525 1451
                                3        107        217          1 525 1451
                               33        107         26          1 525 1451
                                                                   525 1  0
                                                                   525 0  0
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 12100000.0 .041277362 12200000.0 .041285476 12300000.0 .041293590 525 3107
 12400000.0 .041301704 12500000.0 .041374170 12600000.0 .041510989 525 3107
 12700000.0 .041647808 12800000.0 .041784627 12900000.0 .041921446 525 3107
 13000000.0 .042109005 13100000.0 .042347306 13200000.0 .042585606 525 3107
 13300000.0 .042823906 13400000.0 .043062207 13500000.0 .043300507 525 3107
 13600000.0 .043538808 13700000.0 .043777108 13800000.0 .044015408 525 3107
 13900000.0 .044253709 14000000.0 .044426535 14100000.0 .044533888 525 3107
 14200000.0 .044641241 14300000.0 .044748594 14400000.0 .044855947 525 3107
 14500000.0 .044877859 14600000.0 .044814332 14700000.0 .044750804 525 3107
 14800000.0 .044687276 14900000.0 .044623749 15000000.0 .044450191 525 3107
 15100000.0 .044166602 15200000.0 .043883014 15300000.0 .043599426 525 3107
 15400000.0 .043315837 15500000.0 .043024909 15600000.0 .042726641 525 3107
 15700000.0 .042428372 15800000.0 .042130104 15900000.0 .041831836 525 3107
 16000000.0 .041533568 16100000.0 .041235299 16200000.0 .040937031 525 3107
 16300000.0 .040638763 16400000.0 .040340494 16500000.0 .040042226 525 3107
 16600000.0 .039743958 16700000.0 .039445690 16800000.0 .039147421 525 3107
 16900000.0 .038849153 17000000.0 .038554716 17100000.0 .038264109 525 3107
 17200000.0 .037973502 17300000.0 .037682895 17400000.0 .037392289 525 3107
 17500000.0 .037101682 17600000.0 .036811075 17700000.0 .036520469 525 3107
 17800000.0 .036229862 17900000.0 .035939255 18000000.0 .035648648 525 3107
 18100000.0 .035358042 18200000.0 .035067435 18300000.0 .034776828 525 3107
 18400000.0 .034486221 18500000.0 .034195615 18600000.0 .033905008 525 3107
 18700000.0 .033614401 18800000.0 .033323794 18900000.0 .033033188 525 3107
 19000000.0 .032742581 19100000.0 .032451974 19200000.0 .032161368 525 3107
 19300000.0 .031870761 19400000.0 .031580154 19500000.0 .031289547 525 3107
 19600000.0 .030998941 19700000.0 .030708334 19800000.0 .030417727 525 3107
 19900000.0 .030127120 20000000.0 0.0                              525 3107
                                                                   525 3  0
                                                                   525 0  0
 5.01000E+3 9.92692E+0          0          0          0          1 52533107
 0.000000+0 0.000000+0          0        107          0          2 52533107
 0.000000+0 0.000000+0          1          5         36          8 52533107
 1.000000-5 1.010000+6 3.000000+6 5.000000+6 7.000000+6 1.100000+7 52533107
 1.500000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 52533107
 0.000000+0 0.000000+0 0.000000+0 2.250000-2 7.000000-3 2.000000-3 52533107
 0.000000+0 0.000000+0 0.000000+0 3.240000-2 1.000000-3 3.000000-4 52533107
 0.000000+0 0.000000+0 1.690000-2 6.000000-3 2.000000-3 0.000000+0 52533107
 2.040000-2 7.000000-3 2.000000-3 1.280000-2 4.000000-3 1.690000-2 52533107
 0.000000+0 0.000000+0          1          5         91         13 52533107
 1.000000-5 1.000000+0 1.000000+2 9.000001+2 9.000000+3 3.750000+4 52533107
 9.000000+4 1.350000+5 2.050000+5 2.750000+5 4.125000+5 6.750000+5 52533107
 1.010000+6 2.490741-6 2.484915-6 2.438099-6 2.368409-6 2.230656-6 52533107
 1.982202-6 1.807934-6 1.568326-6 1.552257-6 1.555066-6 1.157951-6 52533107
 1.015610-6 2.484192-6 2.461351-6 2.413927-6 2.243903-6 2.000165-6 52533107
 1.927335-6 1.633085-6 1.490809-6 1.647004-6 1.200942-6 1.128964-6 52533107
 2.552806-6 2.624028-6 2.350393-6 2.167907-6 2.540208-6 1.939686-6 52533107
 1.230224-6 2.113235-6 1.424748-6 1.662929-6 2.916571-6 2.927420-6 52533107
 3.121191-6 3.670394-6 2.363825-6 1.299338-6 2.929283-6 1.881633-6 52533107
 2.253798-6 4.641171-6 6.294455-6 5.607427-6 2.930722-6 2.917079-6 52533107
 4.089449-6 2.707325-6 2.287332-6 1.083199-5 1.198476-5 7.899932-6 52533107
 5.246804-6 7.085706-6 4.943829-6 5.243390-6 2.019061-5 1.881386-5 52533107
 8.922377-6 9.271985-6 7.711829-6 9.574556-6 2.834902-5 2.180783-5 52533107
 7.471345-6 8.265557-6 6.775191-6 3.334345-5 1.719968-5 8.391731-6 52533107
 6.705476-6 4.365900-5 1.552481-5 1.322365-5 4.232958-5 2.020522-5 52533107
 1.222438-4                                                        52533107
                                                                   52533  0
                                                                   525 0  0
                                                                     0 0  0
 9.01900E+3 1.88352E+1          0          0         34         10 925 1451
 0.0        0.0                 0          0          0          6 925 1451
 1.00000E+0 3.00001E+7          0          0         10       2002 925 1451
 3.00000E+2 0.0                 1          0        164          3 925 1451
  9-F - 19 FEI        EVAL-OCT01 K.I.Zolotarev                     925 1451
                      DIST-Feb2004                                 925 1451
----IRDF-2002         MATERIAL  925                                925 1451
-----INCIDENT NEUTRON DATA                                         925 1451
------ENDF-6 FORMAT                                                925 1451
*****************************************************************  925 1451
  9-F -19  FEI        EVAL-OCT01 K.I.Zolotarev                     925 1451
                      DIST-JAN02                       20020125    925 1451
----BROND-2           MATERIAL  925                                925 1451
-----INCIDENT NEUTRON DATA                                         925 1451
------ENDF-6 FORMAT                                                925 1451
------Russian Reactor Dosimetry File RRDF-2002                     925 1451
*****************************************************************  925 1451
Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko            925 1451
*****************************************************************  925 1451
MF= 3                                                              925 1451
MT= 16 -(n,2n) cross section data                                  925 1451
--------------------------------------                             925 1451
    Excitation function for the  F-19(n,2n)F-18  reaction  in the  925 1451
energy region from threshold to 30 MeV was evaluated  by means of  925 1451
statistical analysis of experimental cross section data [1-17] and 925 1451
data from GNASH [18,19] calculation.                               925 1451
    Experimental data [2-3],[5],[9-13],[15-17]  were renormalized  925 1451
to  the new standards for monitor reactions cross sections.  Data  925 1451
of Mc Crary et al. [1], Rayburn et al.[2], Bormann et al.[4], and  925 1451
Shiokawa et al. [7]  were renormalized  to the results of precise  925 1451
measurements Vonach et al. [9] and Ikeda et al. [15] in the 14 MeV 925 1451
region. Data of Chatterjee et al. [8] were corrected to their pre- 925 1451
vious measurements at the 14.8 MeV point [6]. Results of measure-  925 1451
ments of Nagel [5] were taken only from SUBENT 005, EXFOR 20198.   925 1451
    Experimental cross section data from ref. [20-31] were rejec-  925 1451
ted due to their big discrepancy with the main bulk of experimen-  925 1451
al data [1-17] and data from theoretical model calculation.        925 1451
    The final procedure of evaluation  F-19(n,2n)F-18  excitation  925 1451
function from threshold to 30 MeV has been carried out within the  925 1451
framework of generalized least squares method.  Rational function  925 1451
was used  as  model function [32].  Calculations was performed by  925 1451
means of  Pade-2 code [33].                                        925 1451
     U-235 thermal fission [34] and  Cf-252 spontaneous  fission   925 1451
neutron spectra [35] averaged cross-sections calculated from the   925 1451
evaluated F-19(n,2n)F-18 excitation function are the following:    925 1451
                                                                   925 1451
 ----------------------------------------------------------------- 925 1451
    TYPE OF SPECTRUM   I <CS>,mb (calc.) I   <CS>, mb (measured)   925 1451
 ----------------------I-----------------I------------------------ 925 1451
 U-235 neutron fission I   0.007299      I 0.007200+-0.001000 [36] 925 1451
                       I                 I 0.006509+-0.000300 [37] 925 1451
                       I                 I 0.008653+-0.000464 [38] 925 1451
 ----------------------I-----------------I------------------------ 925 1451
 CF-252 spont. fission I   0.01615       I 0.01628+-0.00054   [39] 925 1451
                       I                 I 0.01612+-0.00054   [40] 925 1451
 ----------------------------------------------------------------- 925 1451
                                                                   925 1451
MF=33                                                              925 1451
MT= 16 -(n,2n) cross section cov. matrix                           925 1451
---------------------------------------------                      925 1451
    Uncertainties in the evaluated  excitation function  for the   925 1451
reaction  F-19(n,2n)F-18 are given in the form of relative covari- 925 1451
ance matrix  for the 26-neutron energy groups (LB=5).  Covariance  925 1451
matrix  of  uncertainties  was  calculated  simultaneously  with   925 1451
recommended cross section data by means of PADE-2 code.            925 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix   925 1451
given in the 33-file are the following:                            925 1451
                                                                   925 1451
    3.67622E-08    3.74071E-08    3.79051E-08    3.91111E-08       925 1451
    4.11094E-08    4.32321E-08    4.68199E-08    4.98738E-08       925 1451
    5.55426E-08    6.01322E-08    7.00100E-08    7.85485E-08       925 1451
    9.57150E-08    1.06820E-07    1.28694E-07    1.47868E-07       925 1451
    2.58858E-07    6.32301E-07    3.80384E-06    5.45023E-04       925 1451
    1.36448E-03    3.28506E-03    8.26671E-03    1.39754E-02       925 1451
    1.58981E-02    6.37025E-02                                     925 1451
                                                                   925 1451
References :                                                       925 1451
 1. J.H.Mc Crary,I.L.Morgan  Bull. American Phys. Soc., v.5,       925 1451
    p.246, April 1960 ;                                            925 1451
    J.H.Mc Crary,I.L.Morgan  Report AFSWC-TR-60-30, 1960           925 1451
 2. L.A.Rayburn  Proc. of Conf. on Direct Interactions and Nuclear 925 1451
    Reaction Mechanisms, Padua, 3-8 September 1962,  Gordon and    925 1451
    Breach, New York, 1963, p.322. ;                               925 1451
    L.A.Rayburn  Bull. Am. Phys. Soc., v.7, p.335, April 1962      925 1451
 3. M.Cevolani, S.Petralia  Nuovo Cimento, v.26, p.1328, Dec.1962  925 1451
 4. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965          925 1451
 5. W.Nagel  EXFOR 20198.005, 1966                                 925 1451
 6. A.Chatterjee et al. Progress Report BARC-305, p.30, Nov. 1967  925 1451
 7. T.Shiokawa et al. J. Inorg. Nucl. Chem., v.30, p.1, Jan. 1968  925 1451
 8. A.Chatterjee e.a. Proc.of Symposium Nucl.Phys. and Solid State 925 1451
    Phys., Roorke, India, 1969, v.2, p.117, December 1969          925 1451
 9. H.K.Vonach et al.  Proc. of 2-nd Conference on Nuclear Cross-  925 1451
    Sections and Technology, Washington D.C., 4-7 March 1968, v.2, 925 1451
    p.885                                                          925 1451
10. R.C.Barrall et al.  Report AFWL-TR-68-134, March 1969          925 1451
11. R.Mogharrab, H.Neuert Atomkernenergie, v.19, p.107, April 1972 925 1451
12. J.C.Robertson et al.  J. Nucl. Energ., v.27, p.531, Aug. 1973  925 1451
13. R.A.Sigg  Dissert. Abstr., sect. B, v.37, p.2237, Nov. 1976    925 1451
14. T.B.Ryves et al.  J. of Physics, pt.G, v.4, n.11, p.1783, 1978 925 1451
15. Y.Ikeda et al.  Report JAERI-1312, 1988                        925 1451
16. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990     925 1451
17. C.L.Hartmann, P.M.DeLuca  Nucl. Sci. Eng., v.109, p.319, 1991  925 1451
18. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear     925 1451
    Model Code for Calculation of Cross Section and Emission       925 1451
    Spectra. Report LA-6947, Los Alamos, 1977                      925 1451
19. E.L.Trykov, G.Ya.Tertychnyi  Private communication, IPPE,      925 1451
    Obninsk, May 1999                                              925 1451
20. E.B.Paul, R.L.Clarke  Canadian J. Phys., v.31, p.267, 1953     925 1451
21. V.J.Ashby e.a.  Phys. Rev., v.111, p.616, 1958                 925 1451
22. L.A.Rayburn  Phys. Rev., v.122, p.168, 1961                    925 1451
23. O.D.Brill et al.  Doklady Akademii Nauk, v.136,n.1,p.55, 1961; 925 1451
    O.D.Brill et al.  Soviet Physics-Doklady, v.6, p.24, 1961      925 1451
24. J.Picard, C.F.Williamson  Nucl. Phys., v.63, p.673, April 1965 925 1451
25. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965           925 1451
26. J.Csikai  Report EANDC-50 S, paper 102, July 1965              925 1451
27. A.Pasquarelli  Nucl. Phys. A, v.93, p.218, March 1967          925 1451
28. M.Bormann, I.Riehle  Zeitschrift f.Physik, v.207, p.64,        925 1451
    October 1967                                                   925 1451
29. H.O.Menlove et al.  Phys. Rev., v.163, p.1308, 1967            925 1451
30. D.Crumpton  J.Inorg. Nucl. Chem., v.31, p.3727, December 1969  925 1451
31. J.Araminowicz, J.Dresler  Report INR-1464, p.14, May 1973      925 1451
32. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali-  925 1451
    st's Meeting on Evaluation and Processing of Covariance Data,  925 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105   925 1451
33. S.A.Badikov et.al.  Preprint FEI-1686, Obninsk, 1985           925 1451
34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI  925 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989               925 1451
35. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987         925 1451
36. F.Nasyrov, B.D.Sciborskij  Atomnaja Energija, v.25, no.5,      925 1451
    p.437, November 1968                                           925 1451
37. M.Najzer, J.Rant  Report IAEA-208, v.2, p.247, 1978            925 1451
38. W.Mannhart   Progress Report INDC(Ger)-045, pp.40-43, 1999     925 1451
39. W.Mannhart  Handbook on Nuclear Activation Data. IAEA Tech-    925 1451
    nical Report series No.273, p.413, 1987                        925 1451
40. W.Mannhart  Validation of Differential Cross Sections with     925 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna,  925 1451
    September 2002                                                 925 1451
*****************************************************************  925 1451
 File 2 added to the pointwise file containing only the effective  925 1451
 scattering radius with no resonance parameters given.             925 1451
 Taken from ENDF/B-VI                                              925 1451
                                                                   925 1451
*****************************************************************  925 1451
 ***************** Program LINEAR (VERSION 2002-1) *************** 925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value      925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent      925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) *************** 925 1451
 Data Doppler Broadened to 300.000000 Kelvin                       925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value      925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent      925 1451
 ***************** Program FIXUP (Version 2002-1) **************** 925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes  925 1451
 Corrected Thresholds-----------------------------------------Yes  925 1451
 Extended Cross Sections to 20 MeV----------------------------No   925 1451
 Allow Cross Section Deletion---------------------------------No   925 1451
 Allow Cross Section Reconstruction---------------------------No   925 1451
 Make All Cross Sections Non-Negative-------------------------Yes  925 1451
 Delete Energies Not in Ascending Order-----------------------Yes  925 1451
 Deleted Duplicate Points-------------------------------------Yes  925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes  925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes  925 1451
 Allow Creation of Missing Sections---------------------------No   925 1451
 Allow Insertion of Energy Points-----------------------------No   925 1451
 Create Uniform Energy Grid-----------------------------------No   925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes  925 1451
 ***************** Program GROUPIE (VERSION 2002-1) ************** 925 1451
 Unshielded Group Averages Using  640 Groups                       925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                      925 1451
                                1        451        171          1 925 1451
                                3         16         34          1 925 1451
                               33         16         71          1 925 1451
                                                                   925 1  0
                                                                   925 0  0
 9.01900E+3 1.88352E+1          0          0          0          0 925 3 16
-1.04322E+7-1.04322E+7          0          0          1         92 925 3 16
         92          1                                             925 3 16
 10900000.0 2.02752E-6 11000000.0 5.63229E-5 11100000.0 .000137655 925 3 16
 11200000.0 .000292275 11300000.0 .000548721 11400000.0 .000917543 925 3 16
 11500000.0 .001402145 11600000.0 .002002095 11700000.0 .002714490 925 3 16
 11800000.0 .003534705 11900000.0 .004456885 12000000.0 .005474310 925 3 16
 12100000.0 .006579685 12200000.0 .007765400 12300000.0 .009023740 925 3 16
 12400000.0 .010347050 12500000.0 .011727850 12600000.0 .013159000 925 3 16
 12700000.0 .014633700 12800000.0 .016152650 12900000.0 .017695750 925 3 16
 13000000.0 .019265950 13100000.0 .020863250 13200000.0 .022460550 925 3 16
 13300000.0 .024071763 13400000.0 .025696888 13500000.0 .027322013 925 3 16
 13600000.0 .028947138 13700000.0 .030572263 13800000.0 .032197388 925 3 16
 13900000.0 .033822513 14000000.0 .035447638 14100000.0 .037049625 925 3 16
 14200000.0 .038628475 14300000.0 .040207325 14400000.0 .041786175 925 3 16
 14500000.0 .043331888 14600000.0 .044844463 14700000.0 .046357038 925 3 16
 14800000.0 .047869613 14900000.0 .049339675 15000000.0 .050767225 925 3 16
 15100000.0 .052194775 15200000.0 .053622325 15300000.0 .054999450 925 3 16
 15400000.0 .056326150 15500000.0 .057652850 15600000.0 .058979550 925 3 16
 15700000.0 .060248850 15800000.0 .061460750 15900000.0 .062672650 925 3 16
 16000000.0 .063884550 16100000.0 .065033000 16200000.0 .066118000 925 3 16
 16300000.0 .067203000 16400000.0 .068288000 16500000.0 .069304550 925 3 16
 16600000.0 .070252650 16700000.0 .071200750 16800000.0 .072148850 925 3 16
 16900000.0 .073024838 17000000.0 .073828713 17100000.0 .074632588 925 3 16
 17200000.0 .075436463 17300000.0 .076166188 17400000.0 .076821763 925 3 16
 17500000.0 .077477338 17600000.0 .078132913 17700000.0 .078714000 925 3 16
 17800000.0 .079220600 17900000.0 .079727200 18000000.0 .080233800 925 3 16
 18100000.0 .080667413 18200000.0 .081028038 18300000.0 .081388663 925 3 16
 18400000.0 .081749288 18500000.0 .082040025 18600000.0 .082260875 925 3 16
 18700000.0 .082481725 18800000.0 .082702575 18900000.0 .082858125 925 3 16
 19000000.0 .082948375 19100000.0 .083038625 19200000.0 .083128875 925 3 16
 19300000.0 .083180900 19400000.0 .083152770 19500000.0 .083082710 925 3 16
 19600000.0 .083012650 19700000.0 .082942590 19800000.0 .082872530 925 3 16
 19900000.0 .082740370 20000000.0 0.0                              925 3 16
                                                                   925 3  0
                                                                   925 0  0
 9.01900E+3 1.88352E+1          0          0          0          1 92533 16
 0.000000+0 0.000000+0          0         16          0          1 92533 16
 0.000000+0 0.000000+0          1          5        406         28 92533 16
 1.000000-5 1.090000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7 92533 16
 1.400000+7 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7 92533 16
 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7 92533 16
 2.000000+7 2.100000+7 2.200000+7 2.300000+7 2.400000+7 2.500000+7 92533 16
 2.700000+7 2.800000+7 2.900000+7 3.000000+7 0.000000+0 0.000000+0 92533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16
 0.000000+0 7.844560-3 4.898350-3 3.027060-3 1.929170-3 1.321910-3 92533 16
 9.926111-4 8.097050-4 6.990610-4 6.218640-4 5.596050-4 5.048120-4 92533 16
 4.555230-4 4.121780-4 3.758470-4 3.473540-4 3.269110-4 3.140740-4 92533 16
 3.073570-4 3.120960-4 3.231580-4 3.310920-4 3.298620-4 3.041000-4 92533 16
 2.525070-4 2.039030-4 1.464150-4 3.648070-3 2.360540-3 1.449110-3 92533 16
 9.104930-4 6.319380-4 5.059420-4 4.569550-4 4.388680-4 4.266530-4 92533 16
 4.088340-4 3.819840-4 3.470900-4 3.071470-4 2.656870-4 2.259100-4 92533 16
 1.902720-4 1.486420-4 1.141580-4 1.016320-4 1.039070-4 1.134800-4 92533 16
 1.281520-4 1.350840-4 1.314670-4 1.212070-4 1.826270-3 1.354210-3 92533 16
 9.782671-4 7.059900-4 5.267220-4 4.207220-4 3.665780-4 3.449020-4 92533 16
 3.398810-4 3.397380-4 3.365240-4 3.256910-4 3.054490-4 2.760780-4 92533 16
 2.392420-4 1.754770-4 8.900070-5 1.611170-5-3.205140-5-5.099350-5 92533 16
-2.365820-5 5.330250-5 1.291950-4 2.188860-4 1.233180-3 1.050650-3 92533 16
 8.482810-4 6.625510-4 5.129570-4 4.047920-4 3.347390-4 2.952800-4 92533 16
 2.776250-4 2.733860-4 2.755370-4 2.787900-4 2.796490-4 2.762050-4 92533 16
 2.614270-4 2.288330-4 1.889520-4 1.513570-4 1.231830-4 1.085890-4 92533 16
 1.236560-4 1.520640-4 1.922260-4 1.030240-3 9.309550-4 7.910340-4 92533 16
 6.422140-4 5.056870-4 3.931520-4 3.093600-4 2.543630-4 2.252570-4 92533 16
 2.174790-4 2.257800-4 2.449250-4 2.701550-4 3.105180-4 3.558030-4 92533 16
 3.811660-4 3.836510-4 3.655150-4 3.091160-4 2.316670-4 1.739440-4 92533 16
 1.144030-4 9.301140-4 8.611210-4 7.506000-4 6.229440-4 4.974770-4 92533 16
 3.876710-4 3.015590-4 2.426370-4 2.107830-4 2.032840-4 2.157690-4 92533 16
 2.430530-4 3.003160-4 3.817500-4 4.465680-4 4.803090-4 4.784630-4 92533 16
 4.115300-4 2.889300-4 1.821310-4 6.207860-5 8.648120-4 8.135850-4 92533 16
 7.261930-4 6.210140-4 5.137590-4 4.162000-4 3.360620-4 2.773020-4 92533 16
 2.407400-4 2.248090-4 2.263440-4 2.532360-4 3.094430-4 3.658040-4 92533 16
 4.035410-4 4.133410-4 3.692840-4 2.697090-4 1.753330-4 6.488820-5 92533 16
 8.276580-4 7.994630-4 7.403910-4 6.622230-4 5.756870-4 4.895590-4 92533 16
 4.102830-4 3.420170-4 2.868660-4 2.452630-4 2.075860-4 1.890180-4 92533 16
 1.937990-4 2.070150-4 2.180840-4 2.167570-4 1.920930-4 1.606910-4 92533 16
 1.194350-4 8.367870-4 8.381600-4 8.077220-4 7.513760-4 6.760250-4 92533 16
 5.886680-4 4.958200-4 4.031020-4 3.150170-4 2.001340-4 8.278800-5 92533 16
 1.074360-5-2.078540-5-2.009100-5 2.343720-5 9.478700-5 1.508320-4 92533 16
 2.094530-4 9.030970-4 9.293760-4 9.163470-4 8.669560-4 7.869960-4 92533 16
 6.839560-4 5.660230-4 4.412210-4 2.582200-4 4.586960-5-1.068070-4 92533 16
-1.912730-4-2.107930-4-1.358110-4 2.138250-5 1.612100-4 3.183170-4 92533 16
 1.012600-3 1.049310-3 1.037800-3 9.812610-4 8.866920-4 7.633760-4 92533 16
 6.215440-4 3.970180-4 1.170730-4-1.006010-4-2.340860-4-2.800390-4 92533 16
-1.974460-4 1.093280-5 2.077500-4 4.350440-4 1.136340-3 1.170610-3 92533 16
 1.152160-3 1.086070-3 9.810220-4 8.476920-4 6.195350-4 3.158060-4 92533 16
 6.329550-5-1.052330-4-1.794050-4-1.175480-4 9.273780-5 3.048660-4 92533 16
 5.562250-4 1.254720-3 1.286290-3 1.267530-3 1.204830-3 1.107390-3 92533 16
 9.184080-4 6.431870-4 3.946850-4 2.121830-4 1.124980-4 1.261560-4 92533 16
 2.843660-4 4.638940-4 6.851270-4 1.376450-3 1.420770-3 1.422120-3 92533 16
 1.386320-3 1.278440-3 1.085040-3 8.822720-4 7.106800-4 5.921990-4 92533 16
 5.353610-4 5.914810-4 6.921020-4 8.298950-4 1.539830-3 1.622460-3 92533 16
 1.669850-3 1.679580-3 1.616300-3 1.498950-3 1.363950-3 1.236560-3 92533 16
 1.094900-3 1.009270-3 9.922760-4 1.001160-3 1.798240-3 1.944450-3 92533 16
 2.099870-3 2.205700-3 2.207720-3 2.133310-3 2.009010-3 1.777510-3 92533 16
 1.524740-3 1.362900-3 1.209860-3 2.198710-3 2.517890-3 2.820120-3 92533 16
 2.967440-3 2.974120-3 2.866070-3 2.549010-3 2.119470-3 1.798830-3 92533 16
 1.465550-3 3.093180-3 3.712880-3 4.106620-3 4.263970-3 4.206430-3 92533 16
 3.790030-3 3.114020-3 2.560510-3 1.957740-3 4.741660-3 5.470710-3 92533 16
 5.854090-3 5.903210-3 5.430360-3 4.514070-3 3.710880-3 2.809610-3 92533 16
 6.497010-3 7.107000-3 7.299680-3 6.881850-3 5.882950-3 4.952680-3 92533 16
 3.884000-3 7.920590-3 8.280630-3 8.031330-3 7.128890-3 6.218540-3 92533 16
 5.143690-3 8.820440-3 8.840330-3 8.205440-3 7.459920-3 6.540140-3 92533 16
 9.408330-3 9.443150-3 9.180490-3 8.759390-3 1.037760-2 1.080020-2 92533 16
 1.110710-2 1.176100-2 1.264710-2 1.416170-2                       92533 16
                                                                   92533  0
                                                                   925 0  0
                                                                     0 0  0
 1.10230E+4 2.27923E+1          0          0         34         101125 1451
 0.0        0.0                 0          0          0          61125 1451
 1.00000E+0 2.00000E+7          0          0         10       20021125 1451
 3.00000E+2 0.0                 1          0        214          51125 1451
 11-Na- 23 ORNL,TOH   EVAL-DEC77 D. C. LARSON,S.IWASAKI           1125 1451
                      DIST-Feb2004                                1125 1451
----IRDF-2002         MATERIAL 1125                               1125 1451
-----INCIDENT NEUTRON DATA                                        1125 1451
------ENDF-6 FORMAT                                               1125 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1125 1451
DOSIMETRY                                                         1125 1451
******************************************************************1125 1451
 11-NA- 23 ORNL       EVAL-DEC77 D. C. LARSON                     1125 1451
                      DIST-SEP 1 REV1-JUL91            20010926   1125 1451
----ENDF/B-VI         MATERIAL 1125         REVISION 1            1125 1451
                                                                  1125 1451
 **************************************************************** 1125 1451
 11-NA- 23 TOH        EVAL-AUG96 S.IWASAKI                        1125 1451
                      DIST-JUL98                                  1125 1451
----JENDL/D-99        MATERIAL 1125                               1125 1451
-----INCIDENT NEUTRON DATA                                        1125 1451
------ENDF-6 FORMAT                                               1125 1451
 <DOSIMETRY CROSS SECTION>                                        1125 1451
      NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y)                    1125 1451
 **************************************************************** 1125 1451
 ================================================================ 1125 1451
 **************  Start of ENDF/B-VI Bibliography  **************  1125 1451
 ================================================================ 1125 1451
 ENDF/B-VI MOD 2 Revision, September 2000, S.C.Frankle, R.C.Reedy,1125 1451
                 P.G.Young (LANL)                                 1125 1451
                                                                  1125 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   1125 1451
 MT 102) has been updated using new experimental data.            1125 1451
 The previous evaluation (MOD 1: 61 gamma rays) is replaced with  1125 1451
 292 discrete gamma rays based on thermal neutron experimental    1125 1451
 data. Thermal spectrum assumed to 20 MeV.                        1125 1451
 The Q-value for radiative capture was also updated in File 3.    1125 1451
                                                                  1125 1451
 Gas production added as MF=3, MT=203,207                         1125 1451
                                                                  1125 1451
 Details of these changes are described in Frankle et al. [Fr01]. 1125 1451
                                                                  1125 1451
****************************************************************  1125 1451
                                                                  1125 1451
 ENDF/B-VI MOD 1 Revision, July 1991, V. McLane (NNDC)            1125 1451
                                                                  1125 1451
 REVISION 1 CORRECTION                                            1125 1451
                                                                  1125 1451
 Corrected number of energy ranges in File 32, MT = 151 to be 1.  1125 1451
                                                                  1125 1451
****************************************************************  1125 1451
                                                                  1125 1451
 ENDF/B-VI MOD 0, January 1990 (NNDC)                             1125 1451
                                                                  1125 1451
 ENDF/B-V Material 1311 converted to ENDF-6 format by NNDC.       1125 1451
                                                                  1125 1451
****************************************************************  1125 1451
                                                                  1125 1451
 ENDF/B-V Evaluation, December 1977, D. C. Larson (ORNL)          1125 1451
                                                                  1125 1451
MF=2                                                              1125 1451
   Resonance parameters are used from 600 eV to 500 keV.  The     1125 1451
 thermal capture cross section is given as 528 mb, in agreement   1125 1451
 with the experimental results of 526.9+-4.5 mb [RY71], as well   1125 1451
 as earlier results summarized in [RY71].  Using resonance        1125 1451
 parameters E=2.81 keV, GN=376 eV, and GG=0.353 eV for the large  1125 1451
 2.81 keV resonance gives the correct thermal capture value using 1125 1451
 a Breit-Wigner shape, and this form is used to calculate the     1125 1451
 capture crposs section from 1.0E-5 to 600 eV.  The capture       1125 1451
 width GG=0.353 eV is consistent with a recent result by Wilson   1125 1451
 [WI77] where they find 0.24<GG<0.4 eV with a 90% confidence      1125 1451
 level, and with a recent preliminary result of Macklin [MA76]    1125 1451
 where he found GG=0.385+-0.04 eV.  The GN and ER values are      1125 1451
 taken from Seltzer [SE74].  From 600 eV to 500 keV, the capture  1125 1451
 cross section is generated from resonance areas reported by      1125 1451
 Musgrove [MU77], and neutron widths and spins obtained from      1125 1451
 a multilevel analysis of the transmission data of Larson [LA76]. 1125 1451
                                                                  1125 1451
MF=3                                                              1125 1451
   MT=1  Resonance parameters + data of Larson [LA76].            1125 1451
   MT=2  MT=1 - MT=3                                              1125 1451
   MT=102  As discussed above, the shape of the capture cross     1125 1451
     section from 1.E-5 eV to 600 eV was taken from a single      1125 1451
     level Breit-Wigner form from 600 eV to 500 keV, File 3 gives 1125 1451
     no contribution.  From 500 keV to 20 MeV the capture file is 1125 1451
     the same as the ENDF/B-IV file by Pitterle and Paik [PI72].  1125 1451
MF=33                                                             1125 1451
   Uncertainty files for the capture cross section are estimated  1125 1451
 from the literature, and the spread in the various data sets.    1125 1451
                                                                  1125 1451
***************************************************************** 1125 1451
                                                                  1125 1451
REFERENCES                                                        1125 1451
                                                                  1125 1451
[Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos       1125 1451
   National Laboratory Report, LA-13812 (2001).                   1125 1451
[HE75] Hermsdorf et al., Report INDC(GDR)-2/L                     1125 1451
[KI76] W.E. Kinney and F.G. Perey, private communication          1125 1451
[LA76] D.C. Larson, J. Harvey and Hill, Oak Ridge National        1125 1451
   report ORNL-TM-5614 (1976)                                     1125 1451
[LA77] D.C. Larson and G.L. Morgan, to be published               1125 1451
[MA76] R.L. Macklin, private communication (1976)                 1125 1451
[MU77] Musgrove, Allen, and R.L. Macklin, preprint (1977)         1125 1451
[PI72] Pitterle and Paik, report WARD 3045T4B-2, Appendix A (1972)1125 1451
[RY71] T.B. Ryves, 71 Canterbury Conference p. 139                1125 1451
[SE74] J. Seltzer and F.W.K. Firk, Nucl.Sci.Eng. 53, 415 (1974)   1125 1451
[WI77] Wilson, Jackson, and Thomas, Nucl.Sci.Eng. 63, 55 (1977)   1125 1451
 ================================================================ 1125 1451
 ******** Processing history of ENDF/B-VI (n,g) component ******* 1125 1451
 ================================================================ 1125 1451
 ************************ C O N T E N T S *********************** 1125 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1125 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1125 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1125 1451
 ***************** Program RECENT (VERSION 2002-1) ***************1125 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1125 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     1125 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1125 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1125 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1125 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1125 1451
 ***************** Program FIXUP (Version 2002-1) ****************1125 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1125 1451
 Corrected Thresholds-----------------------------------------Yes 1125 1451
 Extended Cross Sections to 20 MeV----------------------------No  1125 1451
 Allow Cross Section Deletion---------------------------------No  1125 1451
 Allow Cross Section Reconstruction---------------------------No  1125 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1125 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1125 1451
 Deleted Duplicate Points-------------------------------------Yes 1125 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1125 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1125 1451
 Allow Creation of Missing Sections---------------------------No  1125 1451
 Allow Insertion of Energy Points-----------------------------No  1125 1451
 Create Uniform Energy Grid-----------------------------------No  1125 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1125 1451
***************************************************************** 1125 1451
 ================================================================ 1125 1451
 *************  End of ENDF/B-VI Bibliography (n,g) ************* 1125 1451
 ================================================================ 1125 1451
 ================================================================ 1125 1451
  **********  Start of JENDL/D-99 (n,2n) Bibliography  *********  1125 1451
 ================================================================ 1125 1451
 11-NA- 23 TOH        EVAL-AUG96 S.IWASAKI                        1125 1451
                      DIST-JUL98                                  1125 1451
----JENDL/D-99        MATERIAL 1125                               1125 1451
-----INCIDENT NEUTRON DATA                                        1125 1451
------ENDF-6 FORMAT                                               1125 1451
HISTORY                                                           1125 1451
96-08  (N,2N) CROSS SECTIONS WERE EVALUATED BY S. IWASAKI (TOH).  1125 1451
                                                                  1125 1451
            ===== POINT-WISE DATA FILE =====                      1125 1451
                                                                  1125 1451
 <DOSIMETRY CROSS SECTION>                                        1125 1451
      NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y)                    1125 1451
                                                                  1125 1451
MF=1  GENERAL INFORMATION                                         1125 1451
 MT=451  DESCRIPTIVE DATA AND DICTIONARY                          1125 1451
                                                                  1125 1451
MF=3  NEUTRON CROSS SECTIONS                                      1125 1451
 MT=16  (N,2N) CROSS SECTION                                      1125 1451
     EVALUATED USING SPLINE FITTING METHOD /6/TO A SET OF         1125 1451
     SELECTED EXPERIMENTAL DATA /9-14/ TAKEN FROM NESTOR-2/8/.    1125 1451
     MAINLY BASED ON THE EXPERIMENTAL DATA OF ADAMSKI/6/.         1125 1451
                                                                  1125 1451
MF=33  COVARIANCES OF NEUTRON CROSS SECTIONS                      1125 1451
 MT=16                                                            1125 1451
    GENERATED USING SPLINE FITTING METHOD /6/                     1125 1451
                                                                  1125 1451
REFERENCES                                                        1125 1451
 6) S. IWASAKI, NUCLEAR DATA FOR SCINECE AND TECHNOLOGY,          1125 1451
    GATLINBURG,TENNESSEE, MAY 9 13, 1994. PROCEEDINGS, VOL 2, 614,1125 1451
    ANS, INC., (1994).                                            1125 1451
 8) T. NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED.     1125 1451
 9) H.O. MENLOVE, ET AL. PHYS. REV. 163,1308 (1967).              1125 1451
10) L. ADAMSKI, ET AL. ANN. NUCL. ENERGY, 7, (7), 397, (1980).    1125 1451
11) Y. IKEDA, ET AL., JAERI-1312 (1988).                          1125 1451
12) LU HAN-LIN , ET AL. INDC(CRP)-16 (1989).                      1125 1451
13) B.STROHMAIER, ET AL. ANN. NUCL. ENERG., 20,8 , 533-545 (1991).1125 1451
14) M. SAKUMA, S. IWASAKI, H. SHIMADA, N. ODANO, K. SUDA,         1125 1451
    J.R. DUMAIS,AND K. SUGIYAMA, JAERI-M 92-027, P. 278 (1992).   1125 1451
 ================================================================ 1125 1451
 ******* Processing history of JENDL/D99 (n,2n) component ******* 1125 1451
 ================================================================ 1125 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1125 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1125 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1125 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1125 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1125 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1125 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1125 1451
 ***************** Program FIXUP (Version 2002-1) ****************1125 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1125 1451
 Corrected Thresholds-----------------------------------------Yes 1125 1451
 Extended Cross Sections to 20 MeV----------------------------No  1125 1451
 Allow Cross Section Deletion---------------------------------No  1125 1451
 Allow Cross Section Reconstruction---------------------------No  1125 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1125 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1125 1451
 Deleted Duplicate Points-------------------------------------Yes 1125 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1125 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1125 1451
 Allow Creation of Missing Sections---------------------------No  1125 1451
 Allow Insertion of Energy Points-----------------------------No  1125 1451
 Create Uniform Energy Grid-----------------------------------No  1125 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1125 1451
 ================================================================ 1125 1451
 ***********   End of JENDL/D-99 (n,2n) Bibliography  *********** 1125 1451
 ================================================================ 1125 1451
                                                                  1125 1451
                                                                  1125 1451
                                                                  1125 1451
                                                                  1125 1451
***************************************************************** 1125 1451
 ************************ C O N T E N T S *********************** 1125 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************1125 1451
 Unshielded Group Averages Using  640 Groups                      1125 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     1125 1451
                                1        451        223          21125 1451
                                3         16         27          11125 1451
                                3        102        217          21125 1451
                               33         16         32          11125 1451
                               33        102          5          01125 1451
                                                                  1125 1  0
                                                                  1125 0  0
 1.10230E+4 2.27923E+1          0          0          0          01125 3 16
-1.24184E+7-1.24184E+7          0          0          1         721125 3 16
         72          1                                            1125 3 16
 12900000.0 .000146800 13000000.0 .001190000 13100000.0 .0019700001125 3 16
 13200000.0 .002750000 13300000.0 .003530000 13400000.0 .0043100001125 3 16
 13500000.0 .005940000 13600000.0 .008420000 13700000.0 .0109000001125 3 16
 13800000.0 .013380000 13900000.0 .015860000 14000000.0 .0187000001125 3 16
 14100000.0 .021900000 14200000.0 .025100000 14300000.0 .0283000001125 3 16
 14400000.0 .031500000 14500000.0 .034400000 14600000.0 .0370000001125 3 16
 14700000.0 .039600000 14800000.0 .042200000 14900000.0 .0448000001125 3 16
 15000000.0 .047060000 15100000.0 .048980000 15200000.0 .0509000001125 3 16
 15300000.0 .052820000 15400000.0 .054740000 15500000.0 .0565000001125 3 16
 15600000.0 .058100000 15700000.0 .059700000 15800000.0 .0613000001125 3 16
 15900000.0 .062900000 16000000.0 .064470000 16100000.0 .0660100001125 3 16
 16200000.0 .067550000 16300000.0 .069090000 16400000.0 .0706300001125 3 16
 16500000.0 .072250000 16600000.0 .073950000 16700000.0 .0756500001125 3 16
 16800000.0 .077350000 16900000.0 .079050000 17000000.0 .0807799901125 3 16
 17100000.0 .082539970 17200000.0 .084299950 17300000.0 .0860599301125 3 16
 17400000.0 .087819910 17500000.0 .089519910 17600000.0 .0911599301125 3 16
 17700000.0 .092799950 17800000.0 .094439970 17900000.0 .0960799901125 3 16
 18000000.0 .097550000 18100000.0 .098850000 18200000.0 .1001500001125 3 16
 18300000.0 .101450000 18400000.0 .102750000 18500000.0 .1038000001125 3 16
 18600000.0 .104600000 18700000.0 .105400000 18800000.0 .1062000001125 3 16
 18900000.0 .107000000 19000000.0 .107550000 19100000.0 .1078500001125 3 16
 19200000.0 .108150000 19300000.0 .108450000 19400000.0 .1087500001125 3 16
 19500000.0 .109010000 19600000.0 .109230000 19700000.0 .1094500001125 3 16
 19800000.0 .109670000 19900000.0 .109890000 20000000.0 0.0       1125 3 16
                                                                  1125 3  0
 1.10230E+4 2.27922E+1          0          0          0          01125 3102
 6.95944E+6 6.95944E+6          0          0          1        6411125 3102
        641          1                                            1125 3102
 .000100000 8.30164087 .000105000 8.10805815 .000110000 7.924152501125 3102
 .000115000 7.75468967 .000120000 7.55659008 .000127500 7.336806781125 3102
 .000135000 7.13599486 .000142500 6.95145136 .000150000 6.751723291125 3102
 .000160000 6.54365768 .000170000 6.35434403 .000180000 6.180565381125 3102
 .000190000 6.01887822 .000200000 5.87126599 .000210000 5.732385051125 3102
 .000220000 5.60338070 .000230000 5.48352961 .000240000 5.342854671125 3102
 .000255000 5.18826207 .000270000 5.06896675 .000280000 4.936165701125 3102
 .000300000 4.77421396 .000320000 4.62757432 .000340000 4.493275061125 3102
 .000360000 4.36957608 .000380000 4.25665906 .000400000 4.138934231125 3102
 .000425000 4.01846711 .000450000 3.90841334 .000475000 3.807498011125 3102
 .000500000 3.71258783 .000525000 3.62585567 .000550000 3.543924191125 3102
 .000575000 3.46772455 .000600000 3.38962227 .000630000 3.309273501125 3102
 .000660000 3.23575512 .000690000 3.16523166 .000720000 3.090195551125 3102
 .000760000 3.00974077 .000800000 2.93504070 .000840000 2.866576321125 3102
 .000880000 2.80150387 .000920000 2.74185233 .000960000 2.684728051125 3102
 .001000000 2.62552423 .001050000 2.56371297 .001100000 2.505759491125 3102
 .001150000 2.45235729 .001200000 2.38936346 .001275000 2.320138341125 3102
 .001350000 2.25684272 .001425000 2.19776833 .001500000 2.135007841125 3102
 .001600000 2.06939155 .001700000 2.00949098 .001800000 1.954088431125 3102
 .001900000 1.90347414 .002000000 1.85654851 .002100000 1.812517101125 3102
 .002200000 1.77218922 .002300000 1.73362626 .002400000 1.689631451125 3102
 .002550000 1.64073850 .002700000 1.60249139 .002800000 1.560939111125 3102
 .003000000 1.50978745 .003200000 1.46324013 .003400000 1.420675721125 3102
 .003600000 1.38184811 .003800000 1.34600268 .004000000 1.308644841125 3102
 .004250000 1.27070585 .004500000 1.23607802 .004750000 1.203720151125 3102
 .005000000 1.17403839 .005250000 1.14652160 .005500000 1.120508881125 3102
 .005750000 1.09667581 .006000000 1.07164890 .006300000 1.046636701125 3102
 .006600000 1.02296441 .006900000 1.00100411 .007200000 .9771248781125 3102
 .007600000 .951588884 .008000000 .928240224 .008400000 .9063256531125 3102
 .008800000 .885929825 .009200000 .866981181 .009600000 .8489539411125 3102
 .010000000 .830295238 .010500000 .810572277 .011000000 .7925341741125 3102
 .011500000 .775328941 .012000000 .755618159 .012750000 .7337950341125 3102
 .013500000 .713577898 .014250000 .695027981 .015000000 .6752294951125 3102
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 11500.0000 8.72839E-5 12000.0000 7.58623E-5 12750.0000 6.49898E-51125 3102
 13500.0000 5.64971E-5 14250.0000 4.97604E-5 15000.0000 4.35907E-51125 3102
 16000.0000 3.80848E-5 17000.0000 3.38451E-5 18000.0000 3.05404E-51125 3102
 19000.0000 2.79591E-5 20000.0000 2.59203E-5 21000.0000 2.43219E-51125 3102
 22000.0000 2.30785E-5 23000.0000 2.21424E-5 24000.0000 2.13466E-51125 3102
 25500.0000 2.08431E-5 27000.0000 2.07704E-5 28000.0000 2.11847E-51125 3102
 30000.0000 2.27455E-5 32000.0000 2.80248E-5 34000.0000 .0205060161125 3102
 36000.0000 4.50860E-5 38000.0000 3.92173E-5 40000.0000 5.06367E-51125 3102
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 50000.0000 .002357099 52500.0000 .011060977 55000.0000 .0008522611125 3102
 57500.0000 .000242611 60000.0000 .000113112 63000.0000 6.47568E-51125 3102
 66000.0000 4.36530E-5 69000.0000 3.25442E-5 72000.0000 2.52201E-51125 3102
 76000.0000 2.02903E-5 80000.0000 1.73241E-5 84000.0000 1.54575E-51125 3102
 88000.0000 1.42691E-5 92000.0000 1.35572E-5 96000.0000 1.32289E-51125 3102
 100000.000 1.33801E-5 105000.000 1.48151E-5 110000.000 2.62443E-51125 3102
 115000.000 .010369566 120000.000 2.32799E-5 127500.000 1.61793E-51125 3102
 135000.000 3.05032E-5 142500.000 .002577990 150000.000 2.34513E-51125 3102
 160000.000 3.18537E-5 170000.000 5.13344E-5 180000.000 .0002606791125 3102
 190000.000 .002419398 200000.000 .001549094 210000.000 .0005946621125 3102
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 300000.000 .000905494 320000.000 5.43255E-5 340000.000 9.43173E-51125 3102
 360000.000 .000281610 380000.000 .000966779 400000.000 .0003782121125 3102
 425000.000 .000764072 450000.000 .000354808 475000.000 .0001314051125 3102
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 575000.000 .000340963 600000.000 .000349825 630000.000 .0003499871125 3102
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 7000000.00 .000167589 7100000.00 .000167843 7200000.00 .0001680971125 3102
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 7600000.00 .000169112 7700000.00 .000169366 7800000.00 .0001696191125 3102
 7900000.00 .000169873 8000000.00 .000170155 8100000.00 .0001704651125 3102
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 16900000.0 .000236010 17000000.0 .000236318 17100000.0 .0002366261125 3102
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 17500000.0 .000237859 17600000.0 .000238167 17700000.0 .0002384761125 3102
 17800000.0 .000238784 17900000.0 .000239092 18000000.0 .0002394001125 3102
 18100000.0 .000239709 18200000.0 .000240017 18300000.0 .0002403251125 3102
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 18700000.0 .000241558 18800000.0 .000241866 18900000.0 .0002421741125 3102
 19000000.0 .000242483 19100000.0 .000242791 19200000.0 .0002430991125 3102
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 19600000.0 .000244332 19700000.0 .000244640 19800000.0 .0002449491125 3102
 19900000.0 .000245257 20000000.0 0.0                             1125 3102
                                                                  1125 3  0
                                                                  1125 0  0
 1.10230E+4 2.27923E+1          0          0          0          1112533 16
 0.00000+ 0 0.00000+ 0          0         16          0          1112533 16
 0.00000+ 0 0.00000+ 0          1          5        171         18112533 16
 1.00000- 5 1.29000+ 7 1.29816+ 7 1.32500+ 7 1.37500+ 7 1.42500+ 7112533 16
 1.47500+ 7 1.52500+ 7 1.57500+ 7 1.62500+ 7 1.67500+ 7 1.72500+ 7112533 16
 1.77500+ 7 1.82500+ 7 1.87500+ 7 1.92500+ 7 1.97500+ 7 2.00000+ 7112533 16
 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0112533 16
 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0112533 16
 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 0.00000+ 0 1.00000+ 0112533 16
 1.06250- 1 4.73600- 2 3.12000- 3 2.70000- 3 2.42000- 3 1.30000- 3112533 16
 1.28000- 3 5.40000- 4 3.80000- 4 1.70000- 4 1.70000- 4 2.10000- 4112533 16
 1.80000- 4-2.70000- 4 0.00000+ 0 1.56250- 2 7.59999- 3 5.02500- 4112533 16
 6.07500- 4 4.12500- 4 3.25000- 4 2.00000- 4 2.25000- 4 2.37500- 5112533 16
 6.37500- 5 2.12500- 5 2.62500- 4 2.25000- 5 3.37500- 5 4.50000- 5112533 16
 4.09599- 3 3.57120- 4 4.66560- 4 3.87200- 4 2.49600- 4 2.04800- 4112533 16
 1.15200- 4 1.21600- 4 1.08800- 4 1.08800- 4 1.34400- 4 1.15200- 4112533 16
 1.72800- 4 2.30400- 5 3.60000- 5 5.18400- 5 5.08200- 5 4.68000- 5112533 16
 2.88000- 5 2.16000- 5 1.14000- 5 1.02000- 5 1.02000- 5 1.26000- 5112533 16
 1.08000- 5 1.62000- 5-2.16000- 5 8.10000- 5 9.70200- 5 8.30700- 5112533 16
 7.34400- 5 4.86000- 5 3.42000- 5 1.53000- 5 2.29500- 5 1.89000- 5112533 16
 2.26800- 5-2.43000- 5-3.24000- 5 1.21000- 4 1.38710- 4 1.28480- 4112533 16
 1.00980- 4 6.27000- 5 3.74000- 5 1.87000- 5 2.31000- 5 1.98000- 5112533 16
-2.97000- 5-3.96000- 5 1.69000- 4 1.99680- 4 1.73160- 4 1.25970- 4112533 16
 6.63000- 5 4.42000- 5 2.73000- 5 2.34000- 5 3.51000- 5-4.68000- 5112533 16
 2.56000- 4 2.76480- 4 2.03680- 4 1.41440- 4 8.16000- 5 6.72000- 5112533 16
 2.88000- 5 4.32000- 5 5.76000- 5 3.24000- 4 3.31740- 4 2.17260- 4112533 16
 1.04040- 4 1.13400- 4 6.48000- 5 4.86000- 5 6.48000- 5 3.61000- 4112533 16
 3.16540- 4 2.29330- 4 1.79550- 4 1.02600- 4 1.02600- 4 6.84000- 5112533 16
 2.89000- 4 2.83220- 4 2.64180- 4 1.65240- 4 1.37700- 4 1.22400- 4112533 16
 2.89000- 4 3.53430- 4 2.23380- 4 2.57040- 4 1.83600- 4 4.41000- 4112533 16
 3.70440- 4 4.87620- 4 4.00680- 4 3.24000- 4 4.81140- 4 4.66560- 4112533 16
 7.29000- 4 9.42840- 4 1.29600- 3                                 112533 16
                                                                  112533  0
 1.10230E+4 2.27922E+1          0          0          0          1112533102
 0.000000+0 0.000000+0          0        102          0          1112533102
 0.000000+0 0.000000+0          0          1         12          6112533102
 1.000000-5 4.000000-4 5.000000+1 2.500000-3 6.000000+2 1.000000-2112533102
 5.000000+5 4.000000-2 5.000000+6 6.250000-2 2.000000+7 0.000000+0112533102
                                                                  112533  0
                                                                  1125 0  0
                                                                     0 0  0
 1.20240E+4 2.37790E+1          0          0         34         101225 1451
 0.0        0.0                 0          0          0          61225 1451
 1.00000E+0 2.10000E+7          0          0         10       20021225 1451
 3.00000E+2 0.0                 1          0         40          31225 1451
 12-Mg- 24 IRK-VIENNA EVAL-APR90                                  1225 1451
                      DIST-Feb2004                                1225 1451
----IRDF-2002         MATERIAL 1225                               1225 1451
-----INCIDENT NEUTRON DATA                                        1225 1451
------ENDF-6 FORMAT                                               1225 1451
 *****************************************************************1225 1451
 12-MG- 24 IRK-VIENNA EVAL-APR90                                  1225 1451
                      DIST-JUN90                                  1225 1451
    IRK-EVAL.NLIB 25  1225                                        1225 1451
 *****************************************************************1225 1451
  The Q values and threshold energies were updated prior to pro-  1225 1451
  cessing through the codes to comply with the values obtained    1225 1451
  using the NNDC calculation program which is based on the 1995   1225 1451
  Update to the Atomic mass Evaluation.                           1225 1451
*****************************************************************11225 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1225 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1225 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1225 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1225 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1225 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1225 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1225 1451
 ***************** Program FIXUP (Version 2002-1) ****************1225 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1225 1451
 Corrected Thresholds-----------------------------------------Yes 1225 1451
 Extended Cross Sections to 20 MeV----------------------------No  1225 1451
 Allow Cross Section Deletion---------------------------------No  1225 1451
 Allow Cross Section Reconstruction---------------------------No  1225 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1225 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1225 1451
 Deleted Duplicate Points-------------------------------------Yes 1225 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1225 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1225 1451
 Allow Creation of Missing Sections---------------------------No  1225 1451
 Allow Insertion of Energy Points-----------------------------No  1225 1451
 Create Uniform Energy Grid-----------------------------------No  1225 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1225 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************1225 1451
 Unshielded Group Averages Using  640 Groups                      1225 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     1225 1451
                                1        451         47          11225 1451
                                3        103         54          11225 1451
                               33        103        191          11225 1451
                                                                  1225 1  0
                                                                  1225 0  0
 1.20240E+4 2.37790E+1          0          0          0          01225 3103
-4.73343E+6-4.73343E+6          0          0          1        1521225 3103
        152          1                                            1225 3103
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 8.153000-5 1.136000-4 9.074000-5 7.833000-5 6.488000-5 1.916000-4122533103
 2.484000-4 0.000000+0 1.364000-4 2.742000-5 5.675000-6 4.078000-6122533103
 2.047000-6 3.265000-6 6.286000-6 7.306000-6 3.756000-5 3.566000-5122533103
 3.666000-5 3.988000-5 3.708000-5 6.163000-5 8.552000-5 1.954000-4122533103
 2.193000-4 4.042000-5 3.101000-5 3.731000-5 4.210000-5 2.190000-5122533103
 6.633000-6 2.502000-5 3.546000-5 2.793000-5 2.402000-5 1.868000-5122533103
 5.986000-5 9.835000-5 0.000000+0 2.990000-5 1.234000-5 1.606000-5122533103
 1.677000-5 8.438000-6 2.638000-6 2.727000-6 4.223000-6 3.176000-6122533103
 5.216000-6 5.098000-6 1.078000-5 1.432000-5 7.229000-5 2.146000-5122533103
 1.217000-5 1.304000-5 7.310000-6 1.286000-6 1.058000-5 1.034000-5122533103
 1.053000-5 9.525000-6 8.492000-6 1.116000-5 1.378000-5 0.000000+0122533103
 2.728000-5 1.661000-5 8.260000-6 1.834000-6 1.729000-6 2.294000-6122533103
 1.904000-6 3.962000-6 4.040000-6 9.790000-6 1.247000-5 6.928000-5122533103
 3.051000-5 8.964000-6 1.622000-6 5.904000-6 8.294000-6 5.994000-6122533103
 8.828000-6 7.746000-6 2.092000-5 3.809000-5 9.510000-5 3.034000-5122533103
 1.586000-5 1.302000-5 1.255000-5 1.343000-5 1.401000-5 1.257000-5122533103
 2.020000-5 2.765000-5 6.808000-5 1.409000-4 1.474000-5 2.281000-5122533103
 1.493000-5 1.676000-5 1.412000-5 2.123000-5 1.232000-5 2.577000-5122533103
 4.629000-4 1.240000-4 1.180000-4 1.074000-4 9.052000-5 1.470000-4122533103
 1.828000-4 0.000000+0 4.242000-4 1.303000-4 1.194000-4 9.884000-5122533103
 2.045000-4 2.627000-4 0.000000+0 4.462000-4 1.101000-4 9.018000-5122533103
 1.636000-4 2.041000-4 0.000000+0 3.819000-4 1.030000-4 1.966000-4122533103
 2.921000-4 6.338000-4 3.281000-4 1.682000-4 2.465000-4 5.976000-4122533103
 1.025000-3 7.385000-4 1.611000-3 2.111000-3 3.391000-3 2.720000-2122533103
                                                                  122533  0
                                                                  1225 0  0
                                                                     0 0  0
 1.30270E+4 2.67497E+1          0          0         34         101325 1451
 0.0        0.0                 0          0          0          61325 1451
 1.00000E+0 2.00000E+7          0          0         10       20021325 1451
 3.00000E+2 0.0                 1          0        325          51325 1451
 13-Al- 27 FEI/IRK    EVAL-APR90 K.I.Zolotarev, M.Wagner et al.   1325 1451
                      DIST-Feb2004                                1325 1451
----IRDF-2002         MATERIAL 1325                               1325 1451
-----INCIDENT NEUTRON DATA                                        1325 1451
------ENDF-6 FORMAT                                               1325 1451
 *****************************************************************1325 1451
 13-AL- 27 FEI        EVAL-May03 K.I.Zolotarev                    1325 1451
                      DIST-Sep03                                  1325 1451
----BROND-3           MATERIAL 1325                               1325 1451
-----INCIDENT NEUTRON DATA                                        1325 1451
------ENDF-6 FORMAT                                               1325 1451
***************************************************************** 1325 1451
------Russian Reactor Dosimetry File RRDF-2002                    1325 1451
***************************************************************** 1325 1451
Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai        1325 1451
***************************************************************** 1325 1451
MF=3                                                              1325 1451
MT=103 - (n,p) cross section                                      1325 1451
 *****************************************************************1325 1451
 13-AL- 27 IRK-VIENNA EVAL-APR90                                  1325 1451
                      DIST-JUN90                                  1325 1451
    IRK-EVAL.NLIB 25  1325                                        1325 1451
MF=3                                                              1325 1451
MT=107 - (n,a) cross section                                      1325 1451
******************************************************************1325 1451
***************************************************************** 1325 1451
 **********   Start of (N,P) bibliographical component ********** 1325 1451
***************************************************************** 1325 1451
------Russian Reactor Dosimetry File RRDF-2002                    1325 1451
***************************************************************** 1325 1451
Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai        1325 1451
***************************************************************** 1325 1451
MF=3                                                              1325 1451
MT=103 - (n,p) cross section                                      1325 1451
-------------------------------------                             1325 1451
    Microscopic experimental data  [1-76]  were  analyzed in  the 1325 1451
process of preparation  of input data base  for the evaluation of 1325 1451
cross sections  and  their uncertainty  for the  Al-27(n,p)Mg-27  1325 1451
reaction.  During this procedure all experimental data  if it was 1325 1451
possible were corrected to the new recommended cross section data 1325 1451
for monitor reactions used in the measurements and to the new re- 1325 1451
commended decay data from ref. [77].                              1325 1451
    Excitation function for the  Al-27(n,p)Mg-27 reaction in the  1325 1451
energy region from threshold to 23.0 MeV was evaluated by  means  1325 1451
of statistical analysis of experimental cross section data [1-58].1325 1451
    Special correction was done with experimental data [3], [8],  1325 1451
[12], [13], [15], [17], [27], [28], [47] and [58].                1325 1451
    Experimental data of Hudson and Morgan [3], Gabbard et al.[8],1325 1451
Ferguson and Albergotti [15] , Cuzzocrea et al. [17] , Csikai and 1325 1451
Chimoe et al. [47]  were renormalized  to the results  of precise 1325 1451
absolute measurements of Ikeda et al.[53] in the overlapping ener-1325 1451
gy ranges. Correction factors for the experimental data [3], [8], 1325 1451
[15], [17]  were Fc=0.87555, Fc=1.41467, Fc=1.60840, Fc=0.89474,  1325 1451
respectively.  Data  of independent  measurements  of Csikai  and 1325 1451
Chimoe et al. [47] were multiplied to the coefficients Fc=0.89370 1325 1451
and Fc=0.92320, respectively.                                     1325 1451
    Cross sections for the  Al-27(n,p)Mg-27 reaction  measured by 1325 1451
Bass et al. [12] in the neutron energy range 6.00 - 9.00 MeV with 1325 1451
25 keV step were recalculated by averaging original  experimental 1325 1451
data over 100 keV energy intervals.                               1325 1451
    Experimental data of Calvi et al. [13] and Shimizu et al.[58] 1325 1451
were corrected  to the results  of Smith and Meadows measurements 1325 1451
with Li-7(p,n)Be-7 neutron source [27]  in the overlapping energy 1325 1451
intervals. Correction factors were Fc=0.96506 and Fc=1.47768, res-1325 1451
pectively. For the experimental data [58]  value Fc=1.47768 is a  1325 1451
total correction factor.  On the first step of correction data of 1325 1451
Shimizu et al. [58]  were renormalized to the new evaluated cross 1325 1451
sections for the monitor reaction In-115(n,n')In-115m [78].       1325 1451
    Data of  Smith and Meadows [27] measured  with using neutrons 1325 1451
from  D(d,n)He3 reaction were renormalized to the results of this 1325 1451
experiment obtained with  Li7(p,n)Be7 neutron source in the over- 1325 1451
lapping interval 5.398 - 5.870 MeV.  D(d,n)He3 data in the energy 1325 1451
range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 .   1325 1451
    Data given in ref. [28] by Mostafa  were renormalized  to the 1325 1451
absolute cross section value for Al-27(n,p)Mg-27  reaction evalu- 1325 1451
ated at 7.1 MeV with taking into account  experimental data [12], 1325 1451
[14] and [27].                                                    1325 1451
    Experimental data [5], [8], [12] and [44] were used partially.1325 1451
Data of Mani et al. [5]  were taken into account  only  for the 9 1325 1451
neutron energies: 11.92, 13.82, 14.75, 16.20, 16.63, 17.83, 18.13,1325 1451
20.15 and 21.72 MeV. Renormalized experimental data of Gabbard et 1325 1451
al. [8] were used  only  for the 8 neutron energies: 12.90, 13.10,1325 1451
13.70, 14.40, 14.90, 15.45, 16.80 and 17.40 MeV.  Information ob- 1325 1451
tained from Bass et al. experiment [12] was taken into account in 1325 1451
the evaluation in the energy region 6.0 - 8.6 MeV.  In the energy 1325 1451
interval 8.6 - 9.0 MeV  this experiment gives significantly over- 1325 1451
estimated cross sections in a comparison with corrected experimen-1325 1451
tal data [24], [27] and new experimental data [55]. Cross section 1325 1451
measured by Bradley et al. [44] for the neutron energy 11 MeV was 1325 1451
not input in the data base for evaluation  due to a very big dis- 1325 1451
crepancy with new experimental data of Csikai et al. [55].        1325 1451
    Experimental cross section data [59-76] were rejected  due to 1325 1451
their discrepancy with the main bulk of experimental data [1-58]. 1325 1451
    In the rejected experiments [59-70] and [75-76] cross section 1325 1451
values were measured  only  in a one energy point in the interval 1325 1451
14 - 15 MeV.                                                      1325 1451
    Statistical analysis of input cross section data  was carried 1325 1451
out by means of PADE-2 code [79].  Rational function  was used as 1325 1451
the model function [80].                                          1325 1451
    Evaluated excitation function  for the reaction Al27(n,p)Mg27 1325 1451
was  tested with  using integral experimental  data  [81-83]  for 1325 1451
U-235 thermal fission neutron spectrum and evaluated integral ex- 1325 1451
perimental data [83] for Cf-252 spontaneous fission neutron spec- 1325 1451
trum.  Calculated  and  measured average cross section values for 1325 1451
U-235 thermal fission neutron spectrum [84]  and  Cf-252  sponta- 1325 1451
neous fission neutron spectrum [85] are given in the table 1.     1325 1451
                                                      Table 1     1325 1451
  1325 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  1325 1451
  1325 1451
 U-235 neutron fission     4.0768         4.133 +- 0.074  [81]  1325 1451
                                          3.914 +- 0.070  [82]  1325 1451
                                          3.902 +- 0.069  [83]  1325 1451
  1325 1451
 CF-252 spont. fission     4.9070         4.880 +- 0.104  [83]  1325 1451
  1325 1451
                                                                  1325 1451
MF=33                                                             1325 1451
MT=103 - (n,p) cross section cov. matrix                          1325 1451
----------------------------------------                          1325 1451
    Uncertainties in the evaluated  excitation function  for  the 1325 1451
reaction Al-27(n,p)Mg-27 are given in the form of relative covari-1325 1451
ance matrix for the  49-neutron energy groups (LB=5).  Covariance 1325 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  1325 1451
recommended cross section data by means of PADE-2 code [79].      1325 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  1325 1451
given in the 33-file are the following:                           1325 1451
                                                                  1325 1451
    9.62022E-08    9.94230E-08    1.04190E-07    1.09967E-07      1325 1451
    1.15699E-07    1.20174E-07    1.24951E-07    1.30570E-07      1325 1451
    1.37068E-07    1.43716E-07    1.50173E-07    1.58001E-07      1325 1451
    1.68205E-07    1.76372E-07    1.86739E-07    1.94590E-07      1325 1451
    2.06816E-07    2.14527E-07    2.29525E-07    2.40453E-07      1325 1451
    2.53581E-07    2.82127E-07    2.99422E-07    3.49389E-07      1325 1451
    3.68505E-07    4.35939E-07    5.42544E-07    7.20462E-07      1325 1451
    9.45090E-07    2.44284E-06    1.77944E-05    8.88268E-05      1325 1451
    1.54306E-04    2.21515E-04    2.92421E-04    3.30954E-04      1325 1451
    3.45302E-04    3.84750E-04    4.56266E-04    6.07458E-04      1325 1451
    1.31442E-03    1.53210E-03    1.76053E-03    2.87076E-03      1325 1451
    4.69639E-03    5.64972E-03    1.04479E-02    1.48842E-02      1325 1451
    5.19587E-02                                                   1325 1451
                                                                  1325 1451
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    excitation function  in the energy range  from threshold  to  1325 1451
    20 MeV, Obninsk, IPPE, eval. March 2003                       1325 1451
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    st's Meeting on Evaluation and Processing of Covariance Data, 1325 1451
    Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105                1325 1451
81. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 1325 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 1325 1451
82. W.Mannhart  Prog. Report INDC(Ger)-045, pp.40-43, June 1999   1325 1451
83. W.Mannhart  Validation of Differential Cross Sections with    1325 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 1325 1451
    September 2002                                                1325 1451
84. L.W.Weston et al.  Evaluated Neutron Data  for  Uranium-235,  1325 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    1325 1451
85. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        1325 1451
***************************************************************** 1325 1451
**********   End of (N,2N) bibliographical component   ********** 1325 1451
================================================================= 1325 1451
***************************************************************** 1325 1451
  The Q values and threshold energies were updated prior to pro-  1325 1451
  cessing through the codes to comply with the values obtained    1325 1451
  using the NNDC calculation program which is based on the 1995   1325 1451
  Update to the Atomic mass Evaluation.Additional points were     1325 1451
  added near threshold for MF/MT=3/107                            1325 1451
 File 2 added to the pointwise file containing only the effective 1325 1451
 scattering radius with no resonance parameters given.            1325 1451
 Taken from ENDF/B-VI                                             1325 1451
***************************************************************** 1325 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1325 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1325 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1325 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1325 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1325 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1325 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1325 1451
 ***************** Program FIXUP (Version 2002-1) ****************1325 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1325 1451
 Corrected Thresholds-----------------------------------------Yes 1325 1451
 Extended Cross Sections to 20 MeV----------------------------No  1325 1451
 Allow Cross Section Deletion---------------------------------No  1325 1451
 Allow Cross Section Reconstruction---------------------------No  1325 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1325 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1325 1451
 Deleted Duplicate Points-------------------------------------Yes 1325 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1325 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1325 1451
 Allow Creation of Missing Sections---------------------------No  1325 1451
 Allow Insertion of Energy Points-----------------------------No  1325 1451
 Create Uniform Energy Grid-----------------------------------No  1325 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1325 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************1325 1451
 Unshielded Group Averages Using  640 Groups                      1325 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     1325 1451
                                1        451        334          11325 1451
                                3        103         64          11325 1451
                                3        107         60          11325 1451
                               33        103        224          11325 1451
                               33        107        224          11325 1451
                                                                  1325 1  0
                                                                  1325 0  0
 1.30270E+4 2.67497E+1          0          0          0          01325 3103
-1.82797E+6-1.82797E+6          0          0          1        1831325 3103
        183          1                                            1325 3103
 1800000.00 7.5592E-10 1900000.00 5.94923E-7 2000000.00 1.70291E-61325 3103
 2100000.00 2.81090E-6 2200000.00 3.91889E-6 2300000.00 5.02688E-61325 3103
 2400000.00 1.36533E-5 2500000.00 5.31887E-5 2600000.00 9.60656E-51325 3103
 2700000.00 .000197747 2800000.00 .000422101 2900000.00 .0007430851325 3103
 3000000.00 .001163323 3100000.00 .001664055 3200000.00 .0021825181325 3103
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 3600000.00 .005923953 3700000.00 .007376778 3800000.00 .0053420301325 3103
 3900000.00 .006590330 4000000.00 .006400630 4100000.00 .0058240901325 3103
 4200000.00 .010016250 4300000.00 .011382425 4400000.00 .0120621751325 3103
 4500000.00 .017318350 4600000.00 .018967950 4700000.00 .0161955251325 3103
 4800000.00 .015435425 4900000.00 .023968600 5000000.00 .0268780501325 3103
 5100000.00 .028133700 5200000.00 .028140700 5300000.00 .0276018001325 3103
 5400000.00 .036424900 5500000.00 .041505550 5600000.00 .0478750101325 3103
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 6000000.00 .048209582 6100000.00 .051044500 6200000.00 .0536079001325 3103
 6300000.00 .055899200 6400000.00 .057914600 6500000.00 .0596261001325 3103
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 7500000.00 .075839390 7600000.00 .077327370 7700000.00 .0788153501325 3103
 7800000.00 .080303330 7900000.00 .081791310 8000000.00 .0832272701325 3103
 8100000.00 .084611210 8200000.00 .085995150 8300000.00 .0873790901325 3103
 8400000.00 .088763030 8500000.00 .090087063 8600000.00 .0913511881325 3103
 8700000.00 .092615313 8800000.00 .093879438 8900000.00 .0950713751325 3103
 9000000.00 .096191125 9100000.00 .097310875 9200000.00 .0984306251325 3103
 9300000.00 .099468917 9400000.00 .100425750 9500000.00 .1013825831325 3103
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 9900000.00 .104511118 10000000.0 .105097471 10100000.0 .1056838241325 3103
 10200000.0 .106158000 10300000.0 .106520000 10400000.0 .1068820001325 3103
 10500000.0 .107122667 10600000.0 .107242000 10700000.0 .1073613331325 3103
 10800000.0 .107354500 10900000.0 .107221500 11000000.0 .1070885001325 3103
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 11400000.0 .105561500 11500000.0 .104940500 11600000.0 .1043195001325 3103
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 12000000.0 .100992733 12100000.0 .099974200 12200000.0 .0989556671325 3103
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 15000000.0 .063087163 15100000.0 .062010288 15200000.0 .0609334131325 3103
 15300000.0 .059856538 15400000.0 .058833300 15500000.0 .0578637001325 3103
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 18900000.0 .035626400 19000000.0 .035213600 19100000.0 .0348008001325 3103
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 19500000.0 .033208820 19600000.0 .032835500 19700000.0 .0324621801325 3103
 19800000.0 .032088860 19900000.0 .031734358 20000000.0 0.0       1325 3103
                                                                  1325 3  0
 1.30270E+4 2.67497E+1          0          0          0          01325 3107
-3.13282E+6-3.13282E+6          0          0          1        1691325 3107
        169          1                                            1325 3107
 3200000.00 1.42951E-8 3300000.00 1.14142E-7 3400000.00 2.28210E-71325 3107
 3500000.00 3.42279E-7 3600000.00 4.56348E-7 3700000.00 5.70416E-71325 3107
 3800000.00 6.84485E-7 3900000.00 7.98554E-7 4000000.00 9.12622E-71325 3107
 4100000.00 1.02669E-6 4200000.00 1.14076E-6 4300000.00 1.25483E-61325 3107
 4400000.00 1.36890E-6 4500000.00 1.48297E-6 4600000.00 2.92250E-61325 3107
 4700000.00 5.68750E-6 4800000.00 8.45250E-6 4900000.00 1.12175E-51325 3107
 5000000.00 1.79500E-5 5100000.00 2.86500E-5 5200000.00 5.45634E-51325 3107
 5300000.00 9.56901E-5 5400000.00 .000136817 5500000.00 .0002041551325 3107
 5600000.00 .000465000 5700000.00 .000765000 5800000.00 .0010585941325 3107
 5900000.00 .001304898 6000000.00 .001541633 6100000.00 .0021375001325 3107
 6200000.00 .003092500 6300000.00 .004247500 6400000.00 .0056025001325 3107
 6500000.00 .007080000 6600000.00 .008680000 6700000.00 .0107375001325 3107
 6800000.00 .013252500 6900000.00 .015816667 7000000.00 .0184300001325 3107
 7100000.00 .020721208 7200000.00 .022225000 7300000.00 .0237352501325 3107
 7400000.00 .026175000 7500000.00 .028995000 7600000.00 .0319331251325 3107
 7700000.00 .035160000 7800000.00 .038347250 7900000.00 .0403740001325 3107
 8000000.00 .041926000 8100000.00 .043858250 8200000.00 .0467200001325 3107
 8300000.00 .049672375 8400000.00 .053689000 8500000.00 .0581410001325 3107
 8600000.00 .061977625 8700000.00 .064310000 8800000.00 .0665961671325 3107
 8900000.00 .069370000 9000000.00 .072343333 9100000.00 .0753166671325 3107
 9200000.00 .078127833 9300000.00 .079966000 9400000.00 .0816420001325 3107
 9500000.00 .083318000 9600000.00 .084994000 9700000.00 .0865785001325 3107
 9800000.00 .087614000 9900000.00 .088558000 10000000.0 .0895020001325 3107
 10100000.0 .090446000 10200000.0 .091575500 10300000.0 .0938180001325 3107
 10400000.0 .096246000 10500000.0 .098674000 10600000.0 .1011020001325 3107
 10700000.0 .103424000 10800000.0 .105110000 10900000.0 .1066900001325 3107
 11000000.0 .108270000 11100000.0 .109850000 11200000.0 .1113287501325 3107
 11300000.0 .112200000 11400000.0 .112970000 11500000.0 .1137400001325 3107
 11600000.0 .114510000 11700000.0 .115324861 11800000.0 .1164088891325 3107
 11900000.0 .117537778 12000000.0 .118666667 12100000.0 .1197955561325 3107
 12200000.0 .120802500 12300000.0 .121687500 12400000.0 .1225725001325 3107
 12500000.0 .123457500 12600000.0 .124150000 12700000.0 .1246500001325 3107
 12800000.0 .125150000 12900000.0 .125095000 13000000.0 .1244850001325 3107
 13100000.0 .124885000 13200000.0 .126295000 13300000.0 .1266200001325 3107
 13400000.0 .125874500 13500000.0 .125448000 13600000.0 .1248750001325 3107
 13700000.0 .123840000 13800000.0 .122775000 13900000.0 .1225800001325 3107
 14000000.0 .122205000 14100000.0 .121510000 14200000.0 .1204250001325 3107
 14300000.0 .118080000 14400000.0 .116097500 14500000.0 .1147725001325 3107
 14600000.0 .113300000 14700000.0 .112195000 14800000.0 .1110200001325 3107
 14900000.0 .109686154 15000000.0 .108778462 15100000.0 .1078707691325 3107
 15200000.0 .106744258 15300000.0 .105082857 15400000.0 .1033971431325 3107
 15500000.0 .101711429 15600000.0 .100025714 15700000.0 .0983604641325 3107
 15800000.0 .096818000 15900000.0 .095296000 16000000.0 .0937740001325 3107
 16100000.0 .092252000 16200000.0 .090685050 16300000.0 .0888484001325 3107
 16400000.0 .086966800 16500000.0 .085085200 16600000.0 .0832036001325 3107
 16700000.0 .081322000 16800000.0 .079440400 16900000.0 .0775588001325 3107
 17000000.0 .075677200 17100000.0 .073795600 17200000.0 .0719140001325 3107
 17300000.0 .070032400 17400000.0 .068150800 17500000.0 .0666965001325 3107
 17600000.0 .065669500 17700000.0 .064642500 17800000.0 .0636155001325 3107
 17900000.0 .062588500 18000000.0 .061561500 18100000.0 .0605345001325 3107
 18200000.0 .059507500 18300000.0 .058480500 18400000.0 .0574535001325 3107
 18500000.0 .056250500 18600000.0 .054871500 18700000.0 .0534925001325 3107
 18800000.0 .052113500 18900000.0 .050734500 19000000.0 .0493555001325 3107
 19100000.0 .047976500 19200000.0 .046597500 19300000.0 .0452185001325 3107
 19400000.0 .043839500 19500000.0 .042685667 19600000.0 .0417570001325 3107
 19700000.0 .040828333 19800000.0 .039899667 19900000.0 .0389710001325 3107
 20000000.0 0.0                                                   1325 3107
                                                                  1325 3  0
                                                                  1325 0  0
 1.30270E+4 2.67497E+1          0          0          0          1132533103
 0.000000+0 0.000000+0          0        103          0          1132533103
 0.000000+0 0.000000+0          1          5       1326         51132533103
 1.000000-5 1.800000+6 3.000000+6 3.500000+6 3.750000+6 4.000000+6132533103
 4.250000+6 4.500000+6 4.750000+6 5.000000+6 5.250000+6 5.500000+6132533103
 5.750000+6 6.000000+6 6.250000+6 6.500000+6 6.750000+6 7.000000+6132533103
 7.250000+6 7.500000+6 7.750000+6 8.000000+6 8.250000+6 8.500000+6132533103
 8.750000+6 9.000000+6 9.250000+6 9.500000+6 9.750000+6 1.000000+7132533103
 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7 1.300000+7132533103
 1.350000+7 1.400000+7 1.420000+7 1.440000+7 1.460000+7 1.480000+7132533103
 1.500000+7 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.800000+7132533103
 1.900000+7 2.000000+7 2.300000+7 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0132533103
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 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 5.023430-2132533103
 2.443400-3 1.782880-3 4.860520-4 8.639200-4 8.052880-4 4.726720-4132533103
 4.862380-4 4.442470-4-2.425800-4 2.602940-4-2.157420-5 2.935090-4132533103
 8.963730-4 1.417880-3 1.792270-3 1.982240-3 1.975950-3 1.784570-3132533103
 1.438050-3 9.802670-4 4.638270-4-5.502870-5-5.214470-4-8.865960-4132533103
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 1.082040-3 6.945080-4 7.880680-5-3.746550-4-4.760820-4-3.597190-4132533103
-2.383950-4-9.610680-5 5.600390-5 2.080320-4 4.428380-4 6.781570-4132533103
 7.443080-4 6.293090-4 1.379440-4-9.414190-4-2.345910-3-5.451960-3132533103
 1.563950-3 8.562080-4 6.700040-4 5.131610-4 5.156830-4 4.816320-4132533103
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                                                                  132533  0
                                                                  1325 0  0
                                                                     0 0  0
 1.40000+ 4 0.0    + 0         -1          0         34          11400 1451
 0.0    + 0 0.0    + 0          0          0          0          61400 1451
 1.0    + 0 0.0    + 0          0          0         10          21400 1451
 0.0    + 0 0.0    + 0          0          0          8          21400 1451
 14-SI-  0 ASTM       EVAL-   91 P.J.GRIFFIN                      1400 1451
 PRIVATE COM.         DIST-JAN03                                  1400 1451
----IRDF-90/NMF-G     MATERIAL 1400                               1400 1451
-----INCIDENT NEUTRON DATA                                        1400 1451
------ENDF-6 FORMAT                                               1400 1451
              DAMAGE CROSS SECTION OF SILICON (ELECTRONICS DAMAGE)1400 1451
              ASTM E722           MT=900                          1400 1451
                                                                  1400 1451
                                1        451         14          01400 1451
                                3        900        217          01400 1451
                                                                  1400 1  0
                                                                  1400 0  0
 1.40000+ 4 0.0    + 0          0          0          0          01400 3900
 0.0    + 0 0.0    + 0          0          0          1        6411400 3900
        641          1                                            1400 3900
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 9.7000E+06 1.6196E+02 9.8000E+06 1.6406E+02 9.9000E+06 1.6449E+021400 3900
 1.0000E+07 1.6621E+02 1.0100E+07 1.6363E+02 1.0200E+07 1.6336E+021400 3900
 1.0300E+07 1.6467E+02 1.0400E+07 1.5476E+02 1.0500E+07 1.5460E+021400 3900
 1.0600E+07 1.5551E+02 1.0700E+07 1.5900E+02 1.0800E+07 1.6291E+021400 3900
 1.0900E+07 1.6005E+02 1.1000E+07 1.5875E+02 1.1100E+07 1.5561E+021400 3900
 1.1200E+07 1.5952E+02 1.1300E+07 1.6603E+02 1.1400E+07 1.6538E+021400 3900
 1.1500E+07 1.6862E+02 1.1600E+07 1.6579E+02 1.1700E+07 1.6562E+021400 3900
 1.1800E+07 1.6546E+02 1.1900E+07 1.6566E+02 1.2000E+07 1.6817E+021400 3900
 1.2100E+07 1.6795E+02 1.2200E+07 1.6771E+02 1.2300E+07 1.6747E+021400 3900
 1.2400E+07 1.6722E+02 1.2500E+07 1.6727E+02 1.2600E+07 1.6760E+021400 3900
 1.2700E+07 1.6794E+02 1.2800E+07 1.6828E+02 1.2900E+07 1.6862E+021400 3900
 1.3000E+07 1.6866E+02 1.3100E+07 1.6839E+02 1.3200E+07 1.6811E+021400 3900
 1.3300E+07 1.6783E+02 1.3400E+07 1.6753E+02 1.3500E+07 1.6716E+021400 3900
 1.3600E+07 1.6673E+02 1.3700E+07 1.6629E+02 1.3800E+07 1.6586E+021400 3900
 1.3900E+07 1.6540E+02 1.4000E+07 1.6554E+02 1.4100E+07 1.6634E+021400 3900
 1.4200E+07 1.6714E+02 1.4300E+07 1.6794E+02 1.4400E+07 1.6873E+021400 3900
 1.4500E+07 1.6921E+02 1.4600E+07 1.6937E+02 1.4700E+07 1.6952E+021400 3900
 1.4800E+07 1.6966E+02 1.4900E+07 1.6979E+02 1.5000E+07 1.6993E+021400 3900
 1.5100E+07 1.7009E+02 1.5200E+07 1.7025E+02 1.5300E+07 1.7040E+021400 3900
 1.5400E+07 1.7056E+02 1.5500E+07 1.7072E+02 1.5600E+07 1.7086E+021400 3900
 1.5700E+07 1.7132E+02 1.5800E+07 1.7363E+02 1.5900E+07 1.7630E+021400 3900
 1.6000E+07 1.7774E+02 1.6100E+07 1.7789E+02 1.6200E+07 1.7803E+021400 3900
 1.6300E+07 1.7818E+02 1.6400E+07 1.7832E+02 1.6500E+07 1.7827E+021400 3900
 1.6600E+07 1.7801E+02 1.6700E+07 1.7776E+02 1.6800E+07 1.7750E+021400 3900
 1.6900E+07 1.7724E+02 1.7000E+07 1.7749E+02 1.7100E+07 1.7828E+021400 3900
 1.7200E+07 1.7908E+02 1.7300E+07 1.7988E+02 1.7400E+07 1.8067E+021400 3900
 1.7500E+07 1.8124E+02 1.7600E+07 1.8158E+02 1.7700E+07 1.8191E+021400 3900
 1.7800E+07 1.8224E+02 1.7900E+07 1.8259E+02 1.8000E+07 1.8294E+021400 3900
 1.8100E+07 1.8328E+02 1.8200E+07 1.8362E+02 1.8300E+07 1.8397E+021400 3900
 1.8400E+07 1.8431E+02 1.8500E+07 1.8445E+02 1.8600E+07 1.8437E+021400 3900
 1.8700E+07 1.8428E+02 1.8800E+07 1.8420E+02 1.8900E+07 1.8411E+021400 3900
 1.9000E+07 1.8400E+02 1.9100E+07 1.8388E+02 1.9200E+07 1.8375E+021400 3900
 1.9300E+07 1.8363E+02 1.9400E+07 1.8351E+02 1.9500E+07 1.8338E+021400 3900
 1.9600E+07 1.8325E+02 1.9700E+07 1.8312E+02 1.9800E+07 1.8300E+021400 3900
 1.9900E+07 1.8287E+02 2.0000E+07 0.0000E+00                      1400 3900
                                                                  1400 3  0
                                                                  1400 0  0
                                                                     0 0  0
 1.50310E+4 3.07076E+1          0          0         34         101525 1451
 0.0        0.0                 0          0          0          61525 1451
 1.00000E+0 2.05000E+7          0          0         10       20021525 1451
 3.00000E+2 0.0                 1          0         40          31525 1451
 15-P - 31 IRK-VIENNA EVAL-JUN80                                  1525 1451
                      DIST-Feb2004                                1525 1451
----IRDF-2002         MATERIAL 1525                               1525 1451
-----INCIDENT NEUTRON DATA                                        1525 1451
------ENDF-6 FORMAT                                               1525 1451
 *****************************************************************1525 1451
                      DIST-JUN90                                  1525 1451
    IRK-EVAL.NLIB 25  1525                                        1525 1451
 *****************************************************************1525 1451
  The Q values and threshold energies were updated prior to pro-  1525 1451
  cessing through the codes to comply with the values obtained    1525 1451
  using the NNDC calculation program which is based on the 1995   1525 1451
  Update to the Atomic mass Evaluation.                           1525 1451
  Additional points were added near threshold for mf/mt= 3/103    1525 1451
 *****************************************************************1525 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1525 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1525 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1525 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1525 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1525 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1525 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1525 1451
 ***************** Program FIXUP (Version 2002-1) ****************1525 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1525 1451
 Corrected Thresholds-----------------------------------------Yes 1525 1451
 Extended Cross Sections to 20 MeV----------------------------No  1525 1451
 Allow Cross Section Deletion---------------------------------No  1525 1451
 Allow Cross Section Reconstruction---------------------------No  1525 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1525 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1525 1451
 Deleted Duplicate Points-------------------------------------Yes 1525 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1525 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1525 1451
 Allow Creation of Missing Sections---------------------------No  1525 1451
 Allow Insertion of Energy Points-----------------------------No  1525 1451
 Create Uniform Energy Grid-----------------------------------No  1525 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1525 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************1525 1451
 Unshielded Group Averages Using  640 Groups                      1525 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     1525 1451
                                1        451         47          11525 1451
                                3        103         69          11525 1451
                               33        103        108          11525 1451
                                                                  1525 1  0
                                                                  1525 0  0
 1.50310E+4 3.07076E+1          0          0          0          01525 3103
-7.09680E+5-7.09680E+5          0          0          1        1981525 3103
        198          1                                            1525 3103
 720000.000 4.99267E-7 760000.000 2.54695E-6 800000.000 4.70496E-61525 3103
 840000.000 6.86298E-6 880000.000 9.02099E-6 920000.000 1.43222E-51525 3103
 960000.000 2.27667E-5 1000000.00 3.75444E-5 1100000.00 8.10667E-51525 3103
 1200000.00 .000175412 1300000.00 .000440231 1400000.00 .0007334621525 3103
 1500000.00 .001126279 1600000.00 .003710000 1700000.00 .0064587501525 3103
 1800000.00 .007420000 1900000.00 .008352500 2000000.00 .0093375001525 3103
 2100000.00 .014567500 2200000.00 .024042500 2300000.00 .0318600001525 3103
 2400000.00 .038020000 2500000.00 .046500000 2600000.00 .0573000001525 3103
 2700000.00 .062522500 2800000.00 .062167500 2900000.00 .0614850001525 3103
 3000000.00 .060475000 3100000.00 .063037500 3200000.00 .0691725001525 3103
 3300000.00 .070750000 3400000.00 .067770000 3500000.00 .0675950001525 3103
 3600000.00 .070225000 3700000.00 .075490000 3800000.00 .0833900001525 3103
 3900000.00 .086990000 4000000.00 .086290000 4100000.00 .0890725001525 3103
 4200000.00 .095337500 4300000.00 .099325000 4400000.00 .1010350001525 3103
 4500000.00 .106707500 4600000.00 .116342500 4700000.00 .1218066671525 3103
 4800000.00 .123100000 4900000.00 .124393333 5000000.00 .1252008331525 3103
 5100000.00 .125522500 5200000.00 .125844167 5300000.00 .1261658331525 3103
 5400000.00 .126487500 5500000.00 .126809167 5600000.00 .1277855561525 3103
 5700000.00 .129416667 5800000.00 .131047778 5900000.00 .1326788891525 3103
 6000000.00 .134310000 6100000.00 .135941111 6200000.00 .1375722221525 3103
 6300000.00 .139203333 6400000.00 .140834444 6500000.00 .1411960001525 3103
 6600000.00 .140288000 6700000.00 .139380000 6800000.00 .1384720001525 3103
 6900000.00 .137564000 7000000.00 .136656000 7100000.00 .1357480001525 3103
 7200000.00 .134840000 7300000.00 .133932000 7400000.00 .1330240001525 3103
 7500000.00 .132823600 7600000.00 .133330800 7700000.00 .1338380001525 3103
 7800000.00 .134345200 7900000.00 .134852400 8000000.00 .1353596001525 3103
 8100000.00 .135866800 8200000.00 .136374000 8300000.00 .1368812001525 3103
 8400000.00 .137388400 8500000.00 .137895600 8600000.00 .1384028001525 3103
 8700000.00 .138892500 8800000.00 .139277200 8900000.00 .1396444001525 3103
 9000000.00 .140011600 9100000.00 .140378800 9200000.00 .1407460001525 3103
 9300000.00 .141113200 9400000.00 .141480400 9500000.00 .1418476001525 3103
 9600000.00 .142214800 9700000.00 .142582000 9800000.00 .1429492001525 3103
 9900000.00 .143316400 10000000.0 .143224500 10100000.0 .1426735001525 3103
 10200000.0 .142122500 10300000.0 .141571500 10400000.0 .1410205001525 3103
 10500000.0 .140469500 10600000.0 .139918500 10700000.0 .1393675001525 3103
 10800000.0 .138816500 10900000.0 .138265500 11000000.0 .1371955001525 3103
 11100000.0 .135606500 11200000.0 .134017500 11300000.0 .1324285001525 3103
 11400000.0 .130839500 11500000.0 .129250500 11600000.0 .1276615001525 3103
 11700000.0 .126072500 11800000.0 .124483500 11900000.0 .1228945001525 3103
 12000000.0 .121343750 12100000.0 .119831250 12200000.0 .1183187501525 3103
 12300000.0 .116806250 12400000.0 .115293750 12500000.0 .1137812501525 3103
 12600000.0 .112268750 12700000.0 .110756250 12800000.0 .1092437501525 3103
 12900000.0 .107731250 13000000.0 .106218750 13100000.0 .1047062501525 3103
 13200000.0 .103193750 13300000.0 .101681250 13400000.0 .1001687501525 3103
 13500000.0 .098656250 13600000.0 .097143750 13700000.0 .0956312501525 3103
 13800000.0 .094118750 13900000.0 .092606250 14000000.0 .0914250001525 3103
 14100000.0 .090575000 14200000.0 .089725000 14300000.0 .0888750001525 3103
 14400000.0 .088025000 14500000.0 .087175000 14600000.0 .0863250001525 3103
 14700000.0 .085475000 14800000.0 .084625000 14900000.0 .0837750001525 3103
 15000000.0 .082900500 15100000.0 .082001500 15200000.0 .0811025001525 3103
 15300000.0 .080203500 15400000.0 .079304500 15500000.0 .0784055001525 3103
 15600000.0 .077506500 15700000.0 .076607500 15800000.0 .0757085001525 3103
 15900000.0 .074809500 16000000.0 .073979000 16100000.0 .0732170001525 3103
 16200000.0 .072455000 16300000.0 .071693000 16400000.0 .0709310001525 3103
 16500000.0 .070169000 16600000.0 .069407000 16700000.0 .0686450001525 3103
 16800000.0 .067883000 16900000.0 .067121000 17000000.0 .0664115001525 3103
 17100000.0 .065754500 17200000.0 .065097500 17300000.0 .0644405001525 3103
 17400000.0 .063783500 17500000.0 .063126500 17600000.0 .0624695001525 3103
 17700000.0 .061812500 17800000.0 .061155500 17900000.0 .0604985001525 3103
 18000000.0 .059888500 18100000.0 .059325500 18200000.0 .0587625001525 3103
 18300000.0 .058199500 18400000.0 .057636500 18500000.0 .0570735001525 3103
 18600000.0 .056510500 18700000.0 .055947500 18800000.0 .0553845001525 3103
 18900000.0 .054821500 19000000.0 .054300000 19100000.0 .0538200001525 3103
 19200000.0 .053340000 19300000.0 .052860000 19400000.0 .0523800001525 3103
 19500000.0 .051900000 19600000.0 .051420000 19700000.0 .0509400001525 3103
 19800000.0 .050460000 19900000.0 .049980000 20000000.0 0.0       1525 3103
                                                                  1525 3  0
                                                                  1525 0  0
 1.50310E+4 3.07076E+1          0          0          0          1152533103
 0.000000+0 0.000000+0          0        103          0          1152533103
 0.000000+0 0.000000+0          1          5        630         35152533103
 1.000000-5 7.200000+5 1.650000+6 1.800000+6 2.000000+6 2.200000+6152533103
 2.400000+6 2.600000+6 2.800000+6 3.000000+6 3.200000+6 3.400000+6152533103
 3.600000+6 3.800000+6 4.000000+6 4.200000+6 4.400000+6 4.600000+6152533103
 4.800000+6 5.200000+6 6.000000+6 7.000000+6 8.000000+6 9.500000+6152533103
 1.050000+7 1.150000+7 1.250000+7 1.350000+7 1.450000+7 1.550000+7152533103
 1.650000+7 1.750000+7 1.850000+7 1.950000+7 2.000000+7 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
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 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 3.607000-2 7.514000-3 8.546000-3152533103
 2.217000-2 5.044000-3 3.058000-3 0.000000+0 2.796000-3 0.000000+0152533103
 0.000000+0 2.393000-3 0.000000+0 0.000000+0 3.590000-3 0.000000+0152533103
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 4.290000-3 4.277000-3 0.000000+0 0.000000+0 0.000000+0 1.946000-3152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 1.779000-2 1.151000-3 1.034000-2 3.241000-3 1.931000-3 1.425000-3152533103
 0.000000+0 2.246000-3 3.361000-3 1.655000-3 0.000000+0 1.815000-3152533103
 0.000000+0 2.383000-3 0.000000+0 0.000000+0 2.841000-3 2.969000-3152533103
 4.145000-3 4.006000-3 3.253000-3 2.777000-3 0.000000+0 0.000000+0152533103
 0.000000+0 1.190000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 9.411000-3 1.093000-2 1.932000-3 2.345000-3152533103
 1.101000-3 2.315000-3 1.256000-3 2.303000-3 3.088000-3 2.223000-3152533103
 0.000000+0 1.553000-3 0.000000+0 9.231000-4 0.000000+0 1.844000-3152533103
 1.606000-3 2.452000-3 2.399000-3 1.856000-3 1.850000-3 0.000000+0152533103
 0.000000+0 0.000000+0 8.419000-4 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 4.199000-2 3.769000-3 5.643000-3152533103
 2.401000-3 4.436000-3 2.989000-3 5.133000-3 6.142000-3 2.746000-3152533103
 1.214000-3 3.859000-3 1.593000-3 2.294000-3 0.000000+0 6.482000-3152533103
 5.977000-3 8.864000-3 8.639000-3 6.786000-3 6.454000-3 0.000000+0152533103
 0.000000+0 0.000000+0 2.888000-3 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 6.128000-3 5.158000-4 6.021000-4152533103
 6.501000-4 7.921000-4 1.123000-3 9.830000-4 1.077000-3 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 2.624000-3 7.177000-4 1.025000-3 1.021000-3152533103
 1.549000-3 1.617000-3 8.940000-4 3.405000-4 7.031000-4 4.469000-4152533103
 4.179000-4 0.000000+0 1.368000-3 1.284000-3 1.887000-3 1.837000-3152533103
 1.450000-3 1.358000-3 0.000000+0 0.000000+0 0.000000+0 6.044000-4152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 1.916000-3 5.914000-4 9.829000-4 1.584000-3 1.194000-3 1.051000-3152533103
 2.676000-4 2.477000-5 3.510000-4 1.203000-5 2.012000-5 4.213000-4152533103
 4.111000-4 5.738000-4 5.546000-4 4.503000-4 3.844000-4 0.000000+0152533103
 0.000000+0 0.000000+0 1.648000-4 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 3.823000-3 7.588000-4 1.130000-3152533103
 1.405000-3 1.053000-3 2.474000-5 6.807000-4 3.218000-5 4.005000-4152533103
 3.068000-5 8.060000-4 6.650000-4 1.015000-3 9.930000-4 7.682000-4152533103
 7.658000-4 0.000000+0 0.000000+0 0.000000+0 3.485000-4 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.146000-2152533103
 1.983000-3 1.462000-3 1.529000-3 7.614000-4 5.570000-5 9.714000-4152533103
 3.322000-4 4.292000-4 9.439000-4 9.748000-4 9.042000-4 8.740000-4152533103
 7.096000-4 6.058000-4 0.000000+0 0.000000+0 0.000000+0 2.596000-4152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 8.227000-3 2.589000-3 2.177000-3 6.663000-4 1.124000-4 8.736000-4152533103
 5.464000-5 9.134000-5 1.054000-3 9.694000-4 1.353000-3 1.308000-3152533103
 1.062000-3 9.065000-4 0.000000+0 0.000000+0 0.000000+0 3.885000-4152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 5.736000-3 1.907000-3 3.481000-4 6.364000-4 4.562000-4 3.689000-4152533103
 6.992000-5 1.207000-3 1.046000-3 1.535000-3 1.494000-3 1.180000-3152533103
 1.102000-3 0.000000+0 0.000000+0 0.000000+0 4.896000-4 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 5.580000-3152533103
 2.494000-4 3.084000-5 3.080000-4 2.276000-4 2.926000-4 2.168000-4152533103
 2.279000-4 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
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 5.735000-4 4.896000-4 0.000000+0 0.000000+0 0.000000+0 2.099000-4152533103
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 2.069000-3 1.934000-3 0.000000+0 0.000000+0 0.000000+0 8.599000-4152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 3.286000-3 2.544000-3 2.474000-3 1.958000-3 1.818000-3 0.000000+0152533103
 0.000000+0 0.000000+0 8.061000-4 0.000000+0 0.000000+0 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 4.903000-3 3.626000-3 2.867000-3152533103
 2.670000-3 0.000000+0 0.000000+0 0.000000+0 1.186000-3 0.000000+0152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 4.647000-3152533103
 2.789000-3 2.599000-3 0.000000+0 0.000000+0 0.000000+0 1.154000-3152533103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103
 3.739000-3 3.378000-3 4.381000-3 4.071000-3 3.459000-3 1.514000-3152533103
 1.984000-4-1.096000-3-2.519000-3-3.665000-3-4.531000-3-5.258000-3152533103
 3.602000-3 5.156000-3 4.879000-3 4.114000-3 1.819000-3 2.486000-4152533103
-1.353000-3-3.211000-3-4.813000-3-6.048000-3-7.052000-3 8.150000-3152533103
 7.875000-3 6.715000-3 3.008000-3 4.616000-4-2.013000-3-4.899000-3152533103
-7.395000-3-9.318000-3-1.087000-2 7.667000-3 6.600000-3 3.022000-3152533103
 5.465000-4-1.729000-3-4.440000-3-6.795000-3-8.611000-3-1.007000-2152533103
 5.775000-3 2.746000-3 6.217000-4-1.125000-3-3.301000-3-5.194000-3152533103
-6.656000-3-7.841000-3 1.439000-3 4.791000-4-7.138000-5-9.163000-4152533103
-1.679000-3-2.276000-3-2.766000-3 3.212000-4 5.254000-4 5.616000-4152533103
 5.582000-4 5.450000-4 5.245000-4 1.899000-3 3.081000-3 4.045000-3152533103
 4.771000-3 5.329000-3 5.515000-3 7.581000-3 9.161000-3 1.039000-2152533103
 1.064000-2 1.300000-2 1.484000-2 1.597000-2 1.828000-2 2.098000-2152533103
                                                                  152533  0
                                                                  1525 0  0
                                                                     0 0  0
 1.60320E+4 3.16974E+1          0          0         34         101625 1451
 0.0        0.0                 0          0          0          61625 1451
 1.00000E+0 2.00000E+7          0          0         10       20021625 1451
 3.00000E+2 0.0                 1          0         45          31625 1451
 16-S - 32                                                        1625 1451
                      DIST-Feb2004                                1625 1451
----IRDF-2002         MATERIAL 1625                               1625 1451
-----INCIDENT NEUTRON DATA                                        1625 1451
------ENDF-6 FORMAT                                               1625 1451
******************************************************************1625 1451
US EVALUATION 1989    DIST-FEB91                                  1625 1451
    ENDF/B-VI         MATERIAL 1625                               1625 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1625 1451
DOSIMETRY                                                         1625 1451
******************************************************************1625 1451
 The Q values and threshold energies were updated prior to pro-   1625 1451
 cessing through the codes to comply with the values obtained     1625 1451
 using the NNDC calculation program which is based on the 1995    1625 1451
 Update to the Atomic mass Evaluation.                            1625 1451
                                                                  1625 1451
 File 2 added to the pointwise file containing only the effective 1625 1451
 scattering radius with no resonance parameters given.            1625 1451
 Taken from ENDF/B-VI                                             1625 1451
******************************************************************1625 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************1625 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     1625 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     1625 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************1625 1451
 Data Doppler Broadened to 300.000000 Kelvin                      1625 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     1625 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     1625 1451
 ***************** Program FIXUP (Version 2002-1) ****************1625 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 1625 1451
 Corrected Thresholds-----------------------------------------Yes 1625 1451
 Extended Cross Sections to 20 MeV----------------------------No  1625 1451
 Allow Cross Section Deletion---------------------------------No  1625 1451
 Allow Cross Section Reconstruction---------------------------No  1625 1451
 Make All Cross Sections Non-Negative-------------------------Yes 1625 1451
 Delete Energies Not in Ascending Order-----------------------Yes 1625 1451
 Deleted Duplicate Points-------------------------------------Yes 1625 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 1625 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 1625 1451
 Allow Creation of Missing Sections---------------------------No  1625 1451
 Allow Insertion of Energy Points-----------------------------No  1625 1451
 Create Uniform Energy Grid-----------------------------------No  1625 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 1625 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************1625 1451
 Unshielded Group Averages Using  640 Groups                      1625 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     1625 1451
                                1        451         52          11625 1451
                                3        103         68          11625 1451
                               33        103        183          11625 1451
                                                                  1625 1  0
                                                                  1625 0  0
 1.60320E+4 3.16974E+1          0          0          0          01625 3103
 0.0       -9.28310E+5          0          0          1        1931625 3103
        193          1                                            1625 3103
 920000.000 2.17952E-8 960000.000 6.76274E-6 1000000.00 1.89000E-51625 3103
 1100000.00 3.75000E-5 1200000.00 8.10000E-5 1300000.00 .0001710001625 3103
 1400000.00 .000350000 1500000.00 .000688000 1600000.00 .0014590001625 3103
 1700000.00 .002604000 1800000.00 .006027000 1900000.00 .0090950001625 3103
 2000000.00 .013095000 2100000.00 .031680400 2200000.00 .0543095001625 3103
 2300000.00 .068355500 2400000.00 .071605250 2500000.00 .0723635001625 3103
 2600000.00 .066761000 2700000.00 .079601500 2800000.00 .0940746001625 3103
 2900000.00 .112995500 3000000.00 .160350000 3100000.00 .1612500001625 3103
 3200000.00 .147160000 3300000.00 .203350000 3400000.00 .2110525001625 3103
 3500000.00 .227087500 3600000.00 .226475000 3700000.00 .1787000001625 3103
 3800000.00 .178425000 3900000.00 .251806667 4000000.00 .3165733331625 3103
 4100000.00 .340450000 4200000.00 .349000000 4300000.00 .3469636361625 3103
 4400000.00 .296862864 4500000.00 .278430000 4600000.00 .2780800001625 3103
 4700000.00 .272954286 4800000.00 .257940851 4900000.00 .2369763641625 3103
 5000000.00 .220432500 5100000.00 .213717500 5200000.00 .2020425001625 3103
 5300000.00 .185407500 5400000.00 .200845000 5500000.00 .2483550001625 3103
 5600000.00 .278015000 5700000.00 .289825000 5800000.00 .2966925001625 3103
 5900000.00 .298617500 6000000.00 .300240000 6100000.00 .3015600001625 3103
 6200000.00 .302880000 6300000.00 .304200000 6400000.00 .3055200001625 3103
 6500000.00 .307460000 6600000.00 .310020000 6700000.00 .3125800001625 3103
 6800000.00 .315140000 6900000.00 .317700000 7000000.00 .3208110001625 3103
 7100000.00 .324473000 7200000.00 .328135000 7300000.00 .3317970001625 3103
 7400000.00 .335459000 7500000.00 .334650000 7600000.00 .3293700001625 3103
 7700000.00 .324090000 7800000.00 .318810000 7900000.00 .3135300001625 3103
 8000000.00 .312126000 8100000.00 .314598000 8200000.00 .3170700001625 3103
 8300000.00 .319542000 8400000.00 .322014000 8500000.00 .3238300001625 3103
 8600000.00 .324990000 8700000.00 .326150000 8800000.00 .3273100001625 3103
 8900000.00 .328470000 9000000.00 .330531000 9100000.00 .3334930001625 3103
 9200000.00 .336455000 9300000.00 .339417000 9400000.00 .3423790001625 3103
 9500000.00 .345761000 9600000.00 .349563000 9700000.00 .3533650001625 3103
 9800000.00 .357167000 9900000.00 .360969000 10000000.0 .3643230001625 3103
 10100000.0 .367229000 10200000.0 .370135000 10300000.0 .3730410001625 3103
 10400000.0 .375947000 10500000.0 .378436000 10600000.0 .3805080001625 3103
 10700000.0 .382580000 10800000.0 .384652000 10900000.0 .3867240001625 3103
 11000000.0 .387142000 11100000.0 .385906000 11200000.0 .3846700001625 3103
 11300000.0 .383434000 11400000.0 .382198000 11500000.0 .3822350001625 3103
 11600000.0 .383545000 11700000.0 .384855000 11800000.0 .3861650001625 3103
 11900000.0 .387475000 12000000.0 .384360000 12100000.0 .3768200001625 3103
 12200000.0 .369280000 12300000.0 .361740000 12400000.0 .3542000001625 3103
 12500000.0 .347640000 12600000.0 .342060000 12700000.0 .3364800001625 3103
 12800000.0 .330900000 12900000.0 .325320000 13000000.0 .3194190001625 3103
 13100000.0 .313197000 13200000.0 .306975000 13300000.0 .3007530001625 3103
 13400000.0 .294531000 13500000.0 .287643000 13600000.0 .2800890001625 3103
 13700000.0 .272535000 13800000.0 .264981000 13900000.0 .2574270001625 3103
 14000000.0 .250686000 14100000.0 .244758000 14200000.0 .2388300001625 3103
 14300000.0 .232902000 14400000.0 .226974000 14500000.0 .2226780001625 3103
 14600000.0 .220014000 14700000.0 .217350000 14800000.0 .2146860001625 3103
 14900000.0 .212022000 15000000.0 .208361500 15100000.0 .2037045001625 3103
 15200000.0 .199047500 15300000.0 .194390500 15400000.0 .1897335001625 3103
 15500000.0 .185076500 15600000.0 .180419500 15700000.0 .1757625001625 3103
 15800000.0 .171105500 15900000.0 .166448500 16000000.0 .1622415001625 3103
 16100000.0 .158484500 16200000.0 .154727500 16300000.0 .1509705001625 3103
 16400000.0 .147213500 16500000.0 .143456500 16600000.0 .1396995001625 3103
 16700000.0 .135942500 16800000.0 .132185500 16900000.0 .1284285001625 3103
 17000000.0 .125255000 17100000.0 .122665000 17200000.0 .1200750001625 3103
 17300000.0 .117485000 17400000.0 .114895000 17500000.0 .1123050001625 3103
 17600000.0 .109715000 17700000.0 .107125000 17800000.0 .1045350001625 3103
 17900000.0 .101945000 18000000.0 .099952300 18100000.0 .0985569001625 3103
 18200000.0 .097161500 18300000.0 .095766100 18400000.0 .0943707001625 3103
 18500000.0 .092975300 18600000.0 .091579900 18700000.0 .0901845001625 3103
 18800000.0 .088789100 18900000.0 .087393700 19000000.0 .0861977001625 3103
 19100000.0 .085201100 19200000.0 .084204500 19300000.0 .0832079001625 3103
 19400000.0 .082211300 19500000.0 .081214700 19600000.0 .0802181001625 3103
 19700000.0 .079221500 19800000.0 .078224900 19900000.0 .0772283001625 3103
 20000000.0 0.0                                                   1625 3103
                                                                  1625 3  0
                                                                  1625 0  0
 1.60320E+4 3.16974E+1          0          0          0          1162533103
 0.0000E+00 0.0000E+00          0        103          0          6162533103
 0.0000E+00 0.0000E+00         77          4        308        154162533103
 1.0000E-05 0.0000E+00 9.5700E+05-6.0000E-01 1.5800E+06-1.0000E+00162533103
 1.6300E+06-1.0000E+00 1.6800E+06-1.0000E+00 1.7800E+06-1.0000E+00162533103
 1.8300E+06-1.0000E+00 1.8800E+06-1.0000E+00 1.9200E+06-1.0000E+00162533103
 1.9600E+06-1.0000E+00 2.0000E+06-1.0000E+00 2.0400E+06-1.0000E+00162533103
 2.0800E+06-1.0000E+00 2.1200E+06-1.0000E+00 2.1600E+06-1.0000E+00162533103
 2.2000E+06-1.0000E+00 2.2500E+06-1.0000E+00 2.3000E+06-1.0000E+00162533103
 2.3500E+06-1.0000E+00 2.4000E+06-1.0000E+00 2.4500E+06-1.0000E+00162533103
 2.5000E+06-1.0000E+00 2.5500E+06-1.0000E+00 2.6000E+06-1.0000E+00162533103
 2.6500E+06-1.0000E+00 2.7000E+06-1.0000E+00 2.7500E+06-1.0000E+00162533103
 2.8000E+06-1.0000E+00 2.8400E+06-1.0000E+00 2.9000E+06-1.0000E+00162533103
 2.9500E+06-1.0000E+00 3.0000E+06-1.0000E+00 3.0500E+06-1.0000E+00162533103
 3.1000E+06-1.0000E+00 3.1500E+06-1.0000E+00 3.2000E+06-1.0000E+00162533103
 3.2400E+06-1.0000E+00 3.2800E+06-1.0000E+00 3.3300E+06-1.0000E+00162533103
 3.3700E+06-1.0000E+00 3.4000E+06-1.0000E+00 3.4400E+06-1.0000E+00162533103
 3.4700E+06-1.0000E+00 3.5100E+06-1.0000E+00 3.5500E+06-1.0000E+00162533103
 3.6000E+06-1.0000E+00 3.6500E+06-1.0000E+00 3.7000E+06-1.0000E+00162533103
 3.7500E+06-1.0000E+00 3.8000E+06-1.0000E+00 3.8500E+06-1.0000E+00162533103
 3.9000E+06-1.0000E+00 3.9400E+06-1.0000E+00 3.9800E+06-1.0000E+00162533103
 4.0400E+06-1.0000E+00 4.1000E+06-1.0000E+00 4.1900E+06-1.0000E+00162533103
 4.3000E+06-9.7000E-01 5.2000E+06-3.2000E-01 5.6000E+06-1.0000E+00162533103
 5.8000E+06-1.0000E+00 6.0000E+06-1.0000E+00 6.5000E+06-1.0000E+00162533103
 7.0000E+06-5.7000E-01 8.5000E+06-1.0000E+00 9.0000E+06-1.0000E+00162533103
 9.5000E+06-1.0000E+00 1.0000E+07-1.0000E+00 1.0500E+07-4.5000E-01162533103
 1.2000E+07-7.3000E-01 1.3500E+07-1.0000E+00 1.4000E+07-8.4000E-01162533103
 1.6000E+07-1.0000E+00 1.7000E+07-1.0000E+00 1.8000E+07-1.0000E+00162533103
 1.9000E+07-1.0000E+00 2.0000E+07 0.0000E+00 1.0000E-05 0.0000E+00162533103
 9.5700E+05 4.3725E-01 1.5800E+06 4.0602E-01 1.6300E+06 2.2757E-01162533103
 1.6800E+06 2.9271E-01 1.7800E+06 4.3861E-01 1.8300E+06 2.2645E-01162533103
 1.8800E+06 2.8648E-01 1.9200E+06 3.5525E-01 1.9600E+06 4.1335E-01162533103
 2.0000E+06 4.9340E-01 2.0400E+06 6.0066E-01 2.0800E+06 5.4379E-01162533103
 2.1200E+06 4.7401E-01 2.1600E+06 1.4893E-01 2.2000E+06 9.0955E-02162533103
 2.2500E+06 1.2562E-01 2.3000E+06 1.1533E-01 2.3500E+06 2.0570E-01162533103
 2.4000E+06 1.0432E-01 2.4500E+06 1.2650E-01 2.5000E+06 1.6181E-01162533103
 2.5500E+06 1.0266E-01 2.6000E+06 1.6317E-01 2.6500E+06 1.2862E-01162533103
 2.7000E+06 2.5907E-01 2.7500E+06 1.6019E-01 2.8000E+06 1.0416E-01162533103
 2.8400E+06 1.2622E-01 2.9000E+06 1.1459E-01 2.9500E+06 1.3695E-01162533103
 3.0000E+06 1.4834E-01 3.0500E+06 1.8369E-01 3.1000E+06 1.6221E-01162533103
 3.1500E+06 1.9655E-01 3.2000E+06 1.5083E-01 3.2400E+06 2.3966E-01162533103
 3.2800E+06 1.7087E-01 3.3300E+06 1.8401E-01 3.3700E+06 1.4962E-01162533103
 3.4000E+06 1.3864E-01 3.4400E+06 1.6099E-01 3.4700E+06 1.7271E-01162533103
 3.5100E+06 1.4944E-01 3.5500E+06 1.0342E-01 3.6000E+06 1.5026E-01162533103
 3.6500E+06 1.2707E-01 3.7000E+06 1.7492E-01 3.7500E+06 2.3126E-01162533103
 3.8000E+06 1.8419E-01 3.8500E+06 2.1541E-01 3.9000E+06 2.8759E-01162533103
 3.9400E+06 3.7783E-01 3.9800E+06 2.6294E-01 4.0400E+06 1.3806E-01162533103
 4.1000E+06 3.4473E-01 4.1900E+06 3.8025E-01 4.3000E+06 2.3099E-01162533103
 5.2000E+06 5.1845E-01 5.6000E+06 1.2626E-01 5.8000E+06 1.2742E-01162533103
 6.0000E+06 8.9710E-02 6.5000E+06 5.4451E-02 7.0000E+06 3.8618E-02162533103
 8.5000E+06 3.7710E-02 9.0000E+06 3.8965E-02 9.5000E+06 3.9717E-02162533103
 1.0000E+07 3.6543E-02 1.0500E+07 3.6762E-02 1.2000E+07 4.0115E-02162533103
 1.3500E+07 4.2948E-02 1.4000E+07 1.1829E-01 1.6000E+07 1.3460E-01162533103
 1.7000E+07 1.3626E-01 1.8000E+07 1.4464E-01 1.9000E+07 1.5243E-01162533103
 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103
 0.0000E+00 0.0000E+00          0          1        198         99162533103
 1.0000E-05 0.0000E+00 9.5700E+05 1.0324E-01 1.0000E+06 1.0324E-01162533103
 1.1000E+06 1.0324E-01 1.2000E+06 1.0324E-01 1.3000E+06 1.0324E-01162533103
 1.4000E+06 1.0324E-01 1.5000E+06 1.0324E-01 1.5800E+06 1.4837E-01162533103
 1.6300E+06 4.6608E-02 1.6800E+06 7.7112E-02 1.7800E+06 1.7314E-01162533103
 1.8300E+06 4.6150E-02 1.8800E+06 7.3861E-02 1.9200E+06 1.1358E-01162533103
 1.9600E+06 1.5377E-01 2.0000E+06 2.1910E-01 2.0400E+06 3.2471E-01162533103
 2.0800E+06 2.6614E-01 2.1200E+06 2.0222E-01 2.1600E+06 1.9961E-02162533103
 2.2000E+06 7.4455E-03 2.2500E+06 1.4202E-02 2.3000E+06 1.1971E-02162533103
 2.3500E+06 3.8080E-02 2.4000E+06 9.7937E-03 2.4500E+06 1.4403E-02162533103
 2.5000E+06 2.3565E-02 2.5500E+06 9.4860E-03 2.6000E+06 2.3963E-02162533103
 2.6500E+06 1.4888E-02 2.7000E+06 6.0406E-02 2.7500E+06 2.3094E-02162533103
 2.8000E+06 9.7642E-03 2.8400E+06 1.4338E-02 2.9000E+06 1.1818E-02162533103
 2.9500E+06 1.6879E-02 3.0000E+06 1.9804E-02 3.0500E+06 3.0368E-02162533103
 3.1000E+06 2.3681E-02 3.1500E+06 3.4770E-02 3.2000E+06 2.0475E-02162533103
 3.2400E+06 5.1692E-02 3.2800E+06 2.6278E-02 3.3300E+06 3.0475E-02162533103
 3.3700E+06 2.0147E-02 3.4000E+06 1.7299E-02 3.4400E+06 2.3325E-02162533103
 3.4700E+06 2.6845E-02 3.5100E+06 2.0098E-02 3.5500E+06 9.6267E-03162533103
 3.6000E+06 2.0321E-02 3.6500E+06 1.4531E-02 3.7000E+06 2.7538E-02162533103
 3.7500E+06 4.8133E-02 3.8000E+06 3.0535E-02 3.8500E+06 4.1763E-02162533103
 3.9000E+06 7.4438E-02 3.9400E+06 1.2848E-01 3.9800E+06 6.2225E-02162533103
 4.0400E+06 1.7156E-02 4.1000E+06 1.0695E-01 4.1900E+06 1.3013E-01162533103
 4.3000E+06 4.6579E-02 4.4100E+06 4.6579E-02 4.4500E+06 4.6579E-02162533103
 4.6000E+06 4.6579E-02 4.7500E+06 4.6579E-02 4.8200E+06 4.6579E-02162533103
 4.8900E+06 4.6579E-02 5.0000E+06 4.6579E-02 5.2000E+06 7.7413E-02162533103
 5.4000E+06 7.7413E-02 5.6000E+06 1.4347E-02 5.8000E+06 1.4611E-02162533103
 6.0000E+06 7.2431E-03 6.5000E+06 2.6684E-03 7.0000E+06 7.6506E-04162533103
 7.5000E+06 7.6506E-04 8.0000E+06 7.6506E-04 8.5000E+06 1.2798E-03162533103
 9.0000E+06 1.3664E-03 9.5000E+06 1.4197E-03 1.0000E+07 1.2019E-03162533103
 1.0500E+07 5.4733E-04 1.1000E+07 5.4733E-04 1.1500E+07 5.4733E-04162533103
 1.2000E+07 1.0573E-03 1.2500E+07 1.0573E-03 1.3000E+07 1.0573E-03162533103
 1.3500E+07 1.6601E-03 1.4000E+07 1.0578E-02 1.4500E+07 1.0578E-02162533103
 1.5000E+07 1.0578E-02 1.6000E+07 1.6305E-02 1.7000E+07 1.6710E-02162533103
 1.8000E+07 1.8829E-02 1.9000E+07 2.0912E-02 2.0000E+07 0.0000E+00162533103
 0.0000E+00 0.0000E+00          0          8        198         99162533103
 1.0000E-05 0.0000E+00 9.5700E+05 4.6987E-13 1.0000E+06 4.0977E-12162533103
 1.1000E+06 1.6132E-11 1.2000E+06 7.5263E-11 1.3000E+06 3.3543E-10162533103
 1.4000E+06 1.4052E-09 1.5000E+06 4.6255E-09 1.5800E+06 1.8175E-08162533103
 1.6300E+06 1.1652E-08 1.6800E+06 4.5325E-08 1.7800E+06 3.0780E-07162533103
 1.8300E+06 2.0352E-07 1.8800E+06 5.1869E-07 1.9200E+06 9.9963E-07162533103
 1.9600E+06 1.6242E-06 2.0000E+06 2.7096E-06 2.0400E+06 6.4783E-06162533103
 2.0800E+06 1.1072E-05 2.1200E+06 1.9868E-05 2.1600E+06 3.3565E-06162533103
 2.2000E+06 1.9116E-06 2.2500E+06 5.7852E-06 2.3000E+06 5.7230E-06162533103
 2.3500E+06 2.1399E-05 2.4000E+06 5.8037E-06 2.4500E+06 7.8821E-06162533103
 2.5000E+06 1.2628E-05 2.5500E+06 5.9731E-06 2.6000E+06 1.4956E-05162533103
 2.6500E+06 5.6757E-06 2.7000E+06 3.0742E-05 2.7500E+06 2.1495E-05162533103
 2.8000E+06 9.6057E-06 2.8400E+06 1.4095E-05 2.9000E+06 1.4186E-05162533103
 2.9500E+06 2.7937E-05 3.0000E+06 4.9213E-05 3.0500E+06 9.8839E-05162533103
 3.1000E+06 7.4618E-05 3.1500E+06 9.1742E-05 3.2000E+06 4.4272E-05162533103
 3.2400E+06 1.2001E-04 3.2800E+06 9.5391E-05 3.3300E+06 1.4439E-04162533103
 3.3700E+06 1.0161E-04 3.4000E+06 8.4846E-05 3.4400E+06 1.1310E-04162533103
 3.4700E+06 1.3775E-04 3.5100E+06 1.0996E-04 3.5500E+06 5.8369E-05162533103
 3.6000E+06 1.2237E-04 3.6500E+06 7.8251E-05 3.7000E+06 1.1350E-04162533103
 3.7500E+06 1.4525E-04 3.8000E+06 8.7453E-05 3.8500E+06 1.7881E-04162533103
 3.9000E+06 4.5482E-04 3.9400E+06 9.2900E-04 3.9800E+06 6.0536E-04162533103
 4.0400E+06 2.0134E-04 4.1000E+06 1.3697E-03 4.1900E+06 1.7611E-03162533103
 4.3000E+06 6.2745E-04 4.4100E+06 5.0611E-04 4.4500E+06 4.0547E-04162533103
 4.6000E+06 4.0395E-04 4.7500E+06 3.7820E-04 4.8200E+06 3.4763E-04162533103
 4.8900E+06 3.0327E-04 5.0000E+06 2.5545E-04 5.2000E+06 3.9126E-04162533103
 5.4000E+06 4.9800E-04 5.6000E+06 1.2711E-04 5.8000E+06 1.4110E-04162533103
 6.0000E+06 7.1879E-05 6.5000E+06 2.8119E-05 7.0000E+06 8.8116E-06162533103
 7.5000E+06 8.4801E-06 8.0000E+06 8.4522E-06 8.5000E+06 1.5461E-05162533103
 9.0000E+06 1.7429E-05 9.5000E+06 2.0092E-05 1.0000E+07 1.8530E-05162533103
 1.0500E+07 9.0846E-06 1.1000E+07 9.2760E-06 1.1500E+07 9.1510E-06162533103
 1.2000E+07 1.6018E-05 1.2500E+07 1.3364E-05 1.3000E+07 1.1289E-05162533103
 1.3500E+07 1.4055E-05 1.4000E+07 6.9831E-05 1.4500E+07 5.7844E-05162533103
 1.5000E+07 4.2055E-05 1.6000E+07 3.8698E-05 1.7000E+07 2.3904E-05162533103
 1.8000E+07 1.8309E-05 1.9000E+07 1.5467E-05 2.0000E+07 0.0000E+00162533103
 0.0000E+00 0.0000E+00          0          1        118         59162533103
 1.0000E-05 0.0000E+00 9.5700E+05 1.9119E-01 1.5800E+06 1.6485E-01162533103
 1.6300E+06 5.1787E-02 1.6800E+06 8.5681E-02 1.7800E+06 1.9237E-01162533103
 1.8300E+06 5.1277E-02 1.8800E+06 8.2068E-02 1.9200E+06 1.2620E-01162533103
 1.9600E+06 1.7086E-01 2.0000E+06 2.4344E-01 2.0400E+06 3.6079E-01162533103
 2.0800E+06 2.9571E-01 2.1200E+06 2.2469E-01 2.1600E+06 2.2179E-02162533103
 2.2000E+06 8.2727E-03 2.2500E+06 1.5780E-02 2.3000E+06 1.3301E-02162533103
 2.3500E+06 4.2311E-02 2.4000E+06 1.0882E-02 2.4500E+06 1.6003E-02162533103
 2.5000E+06 2.6183E-02 2.5500E+06 1.0540E-02 2.6000E+06 2.6626E-02162533103
 2.6500E+06 1.6542E-02 2.7000E+06 6.7118E-02 2.7500E+06 2.5660E-02162533103
 2.8000E+06 1.0849E-02 2.8400E+06 1.5931E-02 2.9000E+06 1.3131E-02162533103
 2.9500E+06 1.8754E-02 3.0000E+06 2.2004E-02 3.0500E+06 3.3742E-02162533103
 3.1000E+06 2.6312E-02 3.1500E+06 3.8633E-02 3.2000E+06 2.2750E-02162533103
 3.2400E+06 5.7436E-02 3.2800E+06 2.9198E-02 3.3300E+06 3.3861E-02162533103
 3.3700E+06 2.2386E-02 3.4000E+06 1.9221E-02 3.4400E+06 2.5917E-02162533103
 3.4700E+06 2.9828E-02 3.5100E+06 2.2332E-02 3.5500E+06 1.0696E-02162533103
 3.6000E+06 2.2579E-02 3.6500E+06 1.6146E-02 3.7000E+06 3.0597E-02162533103
 3.7500E+06 5.3481E-02 3.8000E+06 3.3928E-02 3.8500E+06 4.6404E-02162533103
 3.9000E+06 8.2709E-02 3.9400E+06 1.4276E-01 3.9800E+06 6.9139E-02162533103
 4.0400E+06 1.9062E-02 4.1000E+06 1.1884E-01 4.1900E+06 1.4459E-01162533103
 4.3000E+06 5.3355E-02 5.2000E+06 0.0                             162533103
 0.0        0.0                 1          5        210         20162533103
 1.0000E-05 5.2000E+06 5.6000E+06 5.8000E+06 6.0000E+06 6.5000E+06162533103
 7.0000E+06 8.5000E+06 9.0000E+06 9.5000E+06 1.0000E+07 1.0500E+07162533103
 1.2000E+07 1.3500E+07 1.4000E+07 1.6000E+07 1.7000E+07 1.8000E+07162533103
 1.9000E+07 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103
 0.0000E+00 0.0000E+00 0.0000E+00 2.6879E-01-1.6917E-04 7.6244E-05162533103
 7.6222E-04 6.2428E-04 5.0197E-04 5.2466E-04 5.5936E-04 5.5109E-04162533103
 5.3023E-04 5.2522E-04 4.1067E-04 3.3731E-04 7.0923E-04 7.5401E-04162533103
 1.2360E-03 1.2362E-03 1.2365E-03 1.5941E-02 1.3651E-02 9.8297E-04162533103
 6.7245E-04 5.9382E-04 6.0350E-04 6.3004E-04 6.2252E-04 6.0748E-04162533103
 5.6306E-04 3.8830E-04 2.8840E-04 7.8421E-04 8.4368E-04 1.5155E-03162533103
 1.5156E-03 1.5158E-03 1.6235E-02 9.6110E-04 6.8671E-04 5.8948E-04162533103
 6.0468E-04 6.3316E-04 6.2636E-04 6.1117E-04 5.6685E-04 3.9017E-04162533103
 2.9064E-04 7.9613E-04 8.5614E-04 1.5382E-03 1.5384E-03 1.5389E-03162533103
 8.0479E-03 2.5416E-03 6.7123E-04 7.8051E-04 8.0128E-04 7.5961E-04162533103
 7.2126E-04 6.5442E-04 4.3721E-04 3.8278E-04 5.5604E-04 5.8133E-04162533103
 8.6206E-04 8.6210E-04 8.6223E-04 2.9649E-03 5.5056E-04 7.2580E-04162533103
 7.6567E-04 7.1700E-04 6.7961E-04 6.3216E-04 4.3992E-04 3.8498E-04162533103
 4.6967E-04 4.8689E-04 6.5527E-04 6.5528E-04 6.5537E-04 1.4913E-03162533103
 6.4250E-04 6.5678E-04 6.2474E-04 5.9695E-04 5.2029E-04 3.8122E-04162533103
 3.4843E-04 3.9503E-04 4.0583E-04 5.1030E-04 5.1045E-04 5.1050E-04162533103
 1.4220E-03 7.3842E-04 6.8011E-04 7.0157E-04 6.0537E-04 4.1142E-04162533103
 3.7075E-04 4.1240E-04 4.2421E-04 5.3492E-04 5.3508E-04 5.3511E-04162533103
 1.5183E-03 9.1191E-04 6.9948E-04 6.0628E-04 4.2545E-04 3.8548E-04162533103
 4.3714E-04 4.4949E-04 5.7461E-04 5.7469E-04 5.7485E-04 1.5774E-03162533103
 7.0082E-04 6.9341E-04 4.1684E-04 3.6597E-04 4.2464E-04 4.3794E-04162533103
 5.6898E-04 5.6912E-04 5.6922E-04 1.3354E-03 8.5551E-04 4.5574E-04162533103
 3.6290E-04 4.0894E-04 4.2291E-04 5.4441E-04 5.4452E-04 5.4461E-04162533103
 1.3514E-03 5.3505E-04 3.6596E-04 3.9729E-04 4.1346E-04 5.1956E-04162533103
 5.1969E-04 5.1975E-04 1.6092E-03 3.8766E-04 3.3510E-04 3.4532E-04162533103
 3.5854E-04 3.5862E-04 3.5868E-04 1.8445E-03-1.5721E-03-4.2629E-04162533103
 2.6490E-04 2.6489E-04 2.6498E-04 1.3992E-02 5.9123E-03 7.7589E-04162533103
 7.7619E-04 7.7627E-04 1.8117E-02 8.3496E-04 8.3499E-04 8.3513E-04162533103
 1.8567E-02 1.3578E-02 1.3579E-02 2.0922E-02 1.3581E-02 2.3235E-02162533103
 0.0000E+00 0.0000E+00          0          1          8          4162533103
 1.0000E-05 0.0000E+00 9.2000E+05 1.9119E-01 9.5700E+05 0.0000E+00162533103
 2.0000E+07 0.0000E+00                                            162533103
                                                                  162533  0
                                                                  1625 0  0
                                                                     0 0  0
 2.10450E+4 4.45700E+1          2          0          1          12126 1451
 0.0        0.0                 0          0          0          62126 1451
 1.00000E+0 2.00000E+7          0          0         10       20022126 1451
 3.00000E+2 0.0                 1          0        104          32126 1451
 21-Sc- 45 CNDC       EVAL-FEB91 Z.X.ZHAO                         2126 1451
                      DIST-                                       2126 1451
 ----- DOSIMETRY FILE                                             2126 1451
 ----- INCIDENT NEUTRON DATA                                      2126 1451
 ----- ENDF-6 FORMAT                                              2126 1451
    THIS EVALUATION IS MADE MAINLY BASED ON THE RESONANCE         2126 1451
 PARAMETERS GIVEN BY REF.[1], MEASURED DATA OF REFS.[2-10]        2126 1451
 AND THE SYSTEMATICS OF EXCITATION FUNCTION FOR (N,GAMMA)         2126 1451
 REACTION OF REF.[11]. A SET OF RESOLVED RESONANCE PARAMETERS     2126 1451
 ARE GIVEN FOR EN < 100 KEV. IN THE REGION OF EN > 100 KEV,       2126 1451
 THE SMOOTH CROSS SECTIONS ARE GIVEN.                             2126 1451
    MF=2 MT=251     RESONANCE PARAMETER                           2126 1451
       POSITIVE ENERGY RESONANCE PARAMETERS LISTED BELOW ARE      2126 1451
    BASED ON THOSE OF REF.[1]. TO FIT KENNY'S AVERAGE CROSS       2126 1451
    SECTIONS [2] (CORRECTED IN REF.[3]), THE CAPTURE WIDTHS       2126 1451
    HAVE BEEN ADJUSTED. THE NEGATIVE ENERGY LEVEL PARAMETERS      2126 1451
    ARE TAKEN FROM JENDL-3 EVALUATION [12] TO REPRODUCE           2126 1451
    THERMAL CROSS SECTION 27.2 B WHICH IS IN GOOD AGREEMENT       2126 1451
    WITH 27.2 +- 0.2 B GIVEN BY REF.[1].                          2126 1451
                                                                  2126 1451
    MF=3 MT=102     SMOOTH CROSS SECTION                          2126 1451
       THE SMOOTH CROSS SECTION CALCULATED FROM ZHAO'S            2126 1451
    SYSTEMATICS [11] ARE CONSISTENT WITH THE MEASURED DATA OF     2126 1451
    REFS.[2-9] FROM 1 KEV TO 20 MEV. THEREFORE THE SYSTEMATICS    2126 1451
    IS USED IN THIS EVALUATION TO GIVE THE SMOOTH CROSS           2126 1451
    SECTIONS FOR EN > 100 KEV.                                    2126 1451
                                                                  2126 1451
    MF=8 MT=102    RADIOACTIVE DECAY DATA                         2126 1451
       HALF-TIME OF SC-46 IS TAKEN FROM REF.[13]                  2126 1451
                                                                  2126 1451
    MF=32 AND 33  MT=102     COVARIANCE DATA                      2126 1451
       LONG-RANGE COMPONENT (GIVEN IN FILE 32):                   2126 1451
          10-5 EV TO 1 KEV----- THE ERROR OF THERMAL CROSS        2126 1451
                                SECTION.                          2126 1451
          1 KEV TO 100 KEV----- THE NORMALIZATION ERROR OF        2126 1451
                                KENNY'S MEASUREMENT.              2126 1451
          100 KEV TO 20 MEV---- THE SHAPE ERROR OF THE            2126 1451
                                SYSTEMATICS.                      2126 1451
       SHORT-RANGE COMPONENT:                                     2126 1451
          FOR EN < 100 KEV, THIS COMPONENT IS GIVEN IN FILE 32.   2126 1451
       IN THIS FILE, ONLY THE ERRORS OF SOME LARGE S-WAVE         2126 1451
       RESONANCE PARAMETERS ARE CONSIDERED. THE SHORT-RANGE       2126 1451
       COMPONENT FOR EN > 100 KEV ARE GIVEN IN FILE 33 BASED ON   2126 1451
       THE ERRORS OF PERKIN'S MEASUREMENT [6].                    2126 1451
                                                                  2126 1451
    REFERENCES                                                    2126 1451
     [1]  S. F. MUGHABGHAB ET AL., BNL-325, 4TH, VOL.1 (1981)     2126 1451
     [2]  M. J. KENNY ET AL., AUST. J. PHYS., 30, 605 (1977)      2126 1451
     [3]  B. J. ALLEN ET AL., NUCL. SCI. AND ENG., 82, 230 (1982) 2126 1451
     [4]  S. A. ROMANOV ET AL., YF, 1, 229 (1965)                 2126 1451
     [5]  R. BOOTH ET AL., PHYS. REV., 112, 226 (1958)            2126 1451
     [6]  J. L. PERKIN, J. NUCL. ENERGY, 17, 349 (1963)           2126 1451
     [7]  J. CSIKAI ET AL., NUCL. PHYS., A95, 229 (1967)          2126 1451
     [8]  M. WAGNER ET AL., APA, 52, 30 (1980)                    2126 1451
     [9]  M. BUDNAR ET AL., INDC(YUG)-6 (1979)                    2126 1451
     [10] W. MANNHART, ZP/A, 272, 273 (1975)                      2126 1451
     [11] Z. X. ZHAO ET AL., CHINESE NUCL. PHYS.,VOL.11, 71 (1989)2126 1451
     [12] T. NAKAGAWA ET AL., JAERI-M, 90-099 (1990)              2126 1451
     [13] J. TULI, NUCLEAR WALLET CARDS, NNDC, 1990               2126 1451
                                                                  2126 1451
 ================================================================ 2126 1451
  The Q values and threshold energies were updated prior to pro-  2126 1451
  cessing through the codes to comply with the values obtained    2126 1451
  using the NNDC calculation program which is based on the 1995   2126 1451
  Update to the Atomic mass Evaluation.                           2126 1451
                                                                  2126 1451
  Total and Elastic cross sections were only given upto 100 Kev   2126 1451
  in the original data.For these two reactions the energy region  2126 1451
  between 100 KeV and 20 Mev was taken the ENDF/B-6 library,      2126 1451
  linearised and added to this file.                              2126 1451
                                                                  2126 1451
  The original SC45G cross section set had uncertainties for the  2126 1451
  resonance parameters in file MF=32, which were converted in the 2126 1451
  extended SAND-II group structure and inserted in file MF=33 as  2126 1451
  extra "NI type" sub-subsections.                                2126 1451
                                                                  2126 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2126 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2126 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2126 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2126 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2126 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2126 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2126 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2126 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2126 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2126 1451
 ***************** Program FIXUP (Version 2002-1) ****************2126 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2126 1451
 Corrected Thresholds-----------------------------------------Yes 2126 1451
 Extended Cross Sections to 20 MeV----------------------------No  2126 1451
 Allow Cross Section Deletion---------------------------------No  2126 1451
 Allow Cross Section Reconstruction---------------------------No  2126 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2126 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2126 1451
 Deleted Duplicate Points-------------------------------------Yes 2126 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2126 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2126 1451
 Allow Creation of Missing Sections---------------------------No  2126 1451
 Allow Insertion of Energy Points-----------------------------No  2126 1451
 Create Uniform Energy Grid-----------------------------------No  2126 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2126 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2126 1451
 Unshielded Group Averages Using  640 Groups                      2126 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2126 1451
                                1        451        111          12126 1451
                                3        102        217          12126 1451
                               33        102        150          12126 1451
                                                                  2126 1  0
                                                                  2126 0  0
 2.10450E+4 4.45700E+1          0         99          0          02126 3102
 8.76070E+6 8.76070E+6          0          0          1        6412126 3102
        641          1                                            2126 3102
 .000100000 427.732022 .000105000 417.757985 .000110000 408.2825332126 3102
 .000115000 399.551148 .000120000 389.344069 .000127500 378.0196042126 3102
 .000135000 367.672957 .000142500 358.164586 .000150000 347.8738902126 3102
 .000160000 337.153670 .000170000 327.399593 .000180000 318.4460852126 3102
 .000190000 310.115605 .000200000 302.509818 .000210000 295.3540522126 3102
 .000220000 288.707485 .000230000 282.532179 .000240000 275.2837032126 3102
 .000255000 267.318279 .000270000 261.171754 .000280000 254.3295382126 3102
 .000300000 245.985271 .000320000 238.429839 .000340000 231.5102902126 3102
 .000360000 225.136972 .000380000 219.318956 .000400000 213.2533622126 3102
 .000425000 207.046482 .000450000 201.375468 .000475000 196.1761222126 3102
 .000500000 191.286471 .000525000 186.817655 .000550000 182.5961802126 3102
 .000575000 178.670096 .000600000 174.646076 .000630000 170.5063722126 3102
 .000660000 166.718482 .000690000 163.084846 .000720000 159.2185512126 3102
 .000760000 155.072987 .000800000 151.223913 .000840000 147.6964292126 3102
 .000880000 144.343750 .000920000 141.270264 .000960000 138.3268932126 3102
 .001000000 135.276149 .001050000 132.091272 .001100000 129.1054142126 3102
 .001150000 126.354163 .001200000 123.108711 .001275000 119.5418022126 3102
 .001350000 116.280701 .001425000 113.237250 .001500000 110.0038652126 3102
 .001600000 106.622987 .001700000 103.536643 .001800000 100.6821852126 3102
 .001900000 98.0744223 .002000000 95.6566068 .002100000 93.38785342126 3102
 .002200000 91.3101135 .002300000 89.3232969 .002400000 87.05636512126 3102
 .002550000 84.5374098 .002700000 82.5671182 .002800000 80.42620262126 3102
 .003000000 77.7907406 .003200000 75.3924522 .003400000 73.19925192126 3102
 .003600000 71.1988409 .003800000 69.3518591 .004000000 67.42681312126 3102
 .004250000 65.4722046 .004500000 63.6883355 .004750000 62.02128022126 3102
 .005000000 60.4916571 .005250000 59.0739197 .005500000 57.73388342126 3102
 .005750000 56.5059449 .006000000 55.2164104 .006300000 53.92750782126 3102
 .006600000 52.7077560 .006900000 51.5764124 .007200000 50.34611112126 3102
 .007600000 49.0302904 .008000000 47.8267536 .008400000 46.69722242126 3102
 .008800000 45.6461844 .009200000 44.6696000 .009600000 43.74041822126 3102
 .010000000 42.7785373 .010500000 41.7617263 .011000000 40.83179922126 3102
 .011500000 39.9447085 .012000000 38.9281135 .012750000 37.80274412126 3102
 .013500000 36.7603688 .014250000 35.8041717 .015000000 34.78448212126 3102
 .016000001 33.7142471 .017000001 32.7330001 .017999999 31.83717382126 3102
 .018999999 31.0136319 .020000000 30.2408478 .021000000 29.53324562126 3102
 .022000000 28.8662888 .023000000 28.2445562 .024000000 27.51402492126 3102
 .025500000 26.7268640 .027000001 26.1034659 .028000001 25.42725642126 3102
 .029999999 24.5962584 .032000002 23.8398196 .034000002 23.15038232126 3102
 .035999998 22.5274529 .037999999 21.9273945 .039999999 21.32193392126 3102
 .042500000 20.7051126 .045000002 20.1339940 .047499999 19.61071772126 3102
 .050000001 19.1296299 .052499998 18.6755453 .055000000 18.26028702126 3102
 .057500001 17.8638739 .059999999 17.4620091 .063000001 17.05114812126 3102
 .066000000 16.6648597 .068999998 16.3093321 .071999997 15.91553562126 3102
 .075999998 15.5030358 .079999998 15.1191611 .083999999 14.76037672126 3102
 .088000000 14.4312230 .092000000 14.1162905 .096000001 13.82840922126 3102
 .100000001 13.5183752 .104999997 13.2019650 .109999999 12.90435962126 3102
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 2.1604E-06 9.8596E-07 5.4781E-07 3.3988E-07 8.8484E-08 1.7330E-07212633102
 3.7429E-07 1.2899E-06 1.5365E-07 6.5522E-08 5.5334E-07 5.5353E-07212633102
 6.8596E-08 5.6597E-07 2.4030E-07 2.8137E-07 3.4784E-07 4.5981E-07212633102
 7.5598E-08 1.0838E-06 5.1546E-06 4.4775E-05 9.6111E-07 2.3401E-06212633102
 3.9112E-06 1.2800E-05 2.1702E-05 7.8454E-06 2.4343E-06 1.1124E-06212633102
 6.1856E-07 3.8377E-07 9.9739E-08 1.9382E-07 4.0395E-07 1.3752E-06212633102
 1.6756E-07 7.2161E-08 5.8994E-07 5.8000E-07 7.2761E-08 5.9177E-07212633102
 2.5209E-07 4.3328E-07 5.7922E-07 9.6332E-08 1.4114E-06 6.9676E-06212633102
 6.1270E-05 1.2794E-06 2.8602E-06 4.6076E-06 1.4919E-05 2.5270E-05212633102
 9.1462E-06 2.8447E-06 1.3028E-06 7.2551E-07 4.5029E-07 1.1680E-07212633102
 2.2470E-07 4.4550E-07 1.4893E-06 1.8766E-07 8.2002E-08 6.3885E-07212633102
 6.1425E-07 7.8482E-08 6.2455E-07 2.6740E-07 7.8714E-07 1.3309E-07212633102
 2.0096E-06 1.0413E-05 9.2939E-05 1.8739E-06 3.7044E-06 5.6069E-06212633102
 1.7800E-05 3.0087E-05 1.0910E-05 3.4069E-06 1.5661E-06 8.7465E-07212633102
 5.4357E-07 1.4078E-07 2.6761E-07 4.9614E-07 1.6146E-06 2.1370E-07212633102
 9.5360E-08 6.9271E-07 6.4845E-07 8.5139E-08 6.5563E-07 2.8290E-07212633102
 2.2866E-08 3.5517E-07 1.9201E-06 1.7342E-05 3.3943E-07 5.9449E-07212633102
 8.3507E-07 2.5821E-06 4.3514E-06 1.5812E-06 4.9629E-07 2.2925E-07212633102
 1.2853E-07 8.0069E-08 2.0730E-08 3.9037E-08 6.7924E-08 2.1488E-07212633102
 2.9936E-08 1.3649E-08 9.2255E-08 8.4481E-08 1.1418E-08 8.4861E-08212633102
 3.6933E-08 5.7820E-06 3.3369E-05 3.0616E-04 5.7305E-06 8.0793E-06212633102
 9.4646E-06 2.7073E-05 4.5174E-05 1.6515E-05 5.2597E-06 2.4645E-06212633102
 1.3982E-06 8.7856E-07 2.2830E-07 4.2504E-07 6.6183E-07 1.9753E-06212633102
 3.0563E-07 1.4544E-07 8.5161E-07 7.5803E-07 1.0883E-07 7.4987E-07212633102
 3.3273E-07 2.0963E-04 1.9379E-03 3.4155E-05 3.4355E-05 2.3371E-05212633102
 4.3038E-05 6.6431E-05 2.5386E-05 8.9490E-06 4.5915E-06 2.7974E-06212633102
 1.8542E-06 4.9895E-07 9.2670E-07 1.0662E-06 2.4945E-06 5.8495E-07212633102
 3.1928E-07 1.1255E-06 9.7470E-07 1.7988E-07 8.8768E-07 4.3584E-07212633102
 1.8310E-02 3.2576E-04 2.8938E-04 1.3803E-04 1.1212E-04 1.2320E-04212633102
 5.7938E-05 2.8618E-05 1.8113E-05 1.2562E-05 9.0537E-06 2.5688E-06212633102
 4.8481E-06 3.9722E-06 5.1392E-06 2.7835E-06 1.7502E-06 2.8066E-06212633102
 2.5018E-06 7.5487E-07 1.7106E-06 1.1796E-06 6.0844E-06 6.9310E-06212633102
 6.2869E-06 1.5456E-05 2.5212E-05 9.3414E-06 3.0736E-06 1.4842E-06212633102
 8.6228E-07 5.5061E-07 1.4373E-07 2.5891E-07 2.7967E-07 6.2654E-07212633102
 1.5322E-07 8.3156E-08 2.7751E-07 2.2733E-07 4.3656E-08 2.0351E-07212633102
 1.0156E-07 2.1420E-05 3.9156E-05 1.3226E-04 2.2500E-04 8.1404E-05212633102
 2.5305E-05 1.1547E-05 6.3848E-06 3.9080E-06 9.8334E-07 1.7063E-06212633102
 1.7295E-06 3.7268E-06 9.3236E-07 4.9759E-07 1.6066E-06 1.2634E-06212633102
 2.4505E-07 1.1182E-06 5.5889E-07 8.2376E-05 2.8835E-04 4.9153E-04212633102
 1.7730E-04 5.4890E-05 2.4924E-05 1.3711E-05 8.3456E-06 2.0843E-06212633102
 3.5510E-06 3.0519E-06 5.2665E-06 1.7699E-06 9.9046E-07 2.3172E-06212633102
 1.8166E-06 4.2118E-07 1.4473E-06 8.0185E-07 1.0186E-03 1.7344E-03212633102
 6.2471E-04 1.9346E-04 8.7819E-05 4.8271E-05 2.9334E-05 7.3006E-06212633102
 1.2276E-05 8.9112E-06 1.0675E-05 5.6377E-06 3.3209E-06 5.0173E-06212633102
 4.1306E-06 1.2302E-06 2.6350E-06 1.8117E-06 2.9922E-03 1.0875E-03212633102
 3.3573E-04 1.5214E-04 8.3555E-05 5.0747E-05 1.2622E-05 2.1180E-05212633102
 1.4928E-05 1.6361E-05 9.5985E-06 5.7041E-06 7.8220E-06 6.5105E-06212633102
 2.0601E-06 3.8693E-06 2.8536E-06 3.9777E-04 1.2262E-04 5.5543E-05212633102
 3.0517E-05 1.8559E-05 4.6303E-06 7.8604E-06 6.4671E-06 1.0325E-05212633102
 3.8327E-06 2.1689E-06 4.3923E-06 3.1927E-06 8.5061E-07 2.3383E-06212633102
 1.4256E-06 3.7889E-05 1.7207E-05 9.4876E-06 5.8048E-06 1.4671E-06212633102
 2.6099E-06 3.3811E-06 9.0932E-06 1.6277E-06 7.8228E-07 3.3150E-06212633102
 1.9151E-06 3.7767E-07 1.8012E-06 8.8737E-07 7.8447E-06 4.3534E-06212633102
 2.6974E-06 7.0139E-07 1.3760E-06 3.0715E-06 1.0712E-05 1.2243E-06212633102
 4.6912E-07 3.4699E-06 1.6492E-06 2.7246E-07 1.7667E-06 7.9571E-07212633102
 2.4461E-06 1.5543E-06 4.2703E-07 9.8619E-07 3.5720E-06 1.3965E-05212633102
 1.2612E-06 3.8683E-07 3.9869E-06 1.5685E-06 2.4285E-07 1.8345E-06212633102
 7.9547E-07 1.0380E-06 3.1496E-07 9.1294E-07 4.7837E-06 1.9682E-05212633102
 1.5729E-06 3.9925E-07 4.7579E-06 1.5139E-06 2.3795E-07 1.9312E-06212633102
 8.2324E-07 1.1255E-07 4.2344E-07 2.8478E-06 1.1982E-05 8.9628E-07212633102
 1.9429E-07 2.2608E-06 5.5343E-07 9.5443E-08 7.8940E-07 3.3449E-07212633102
 2.0742E-06 1.6466E-05 6.9671E-05 4.9909E-06 9.2554E-07 7.9133E-06212633102
 1.3090E-06 2.8949E-07 2.2393E-06 9.5624E-07 1.4433E-04 5.9024E-04212633102
 4.0501E-05 6.5832E-06 1.5000E-05 2.4583E-06 7.9553E-07 3.1192E-06212633102
 1.5155E-06 2.5686E-03 1.8228E-04 2.8441E-05 3.1009E-05 8.8389E-06212633102
 2.6695E-06 5.5532E-06 3.2812E-06 1.3371E-05 2.5234E-06 4.0262E-05212633102
 6.1624E-06 5.0010E-07 4.1452E-06 1.7801E-06 1.8179E-06 1.4569E-04212633102
 5.2352E-05 5.4077E-08 4.7205E-06 1.9613E-06 1.7427E-02 9.9573E-03212633102
 8.4587E-05 7.9004E-06 3.3483E-06 9.0737E-03 8.8884E-05 4.9176E-05212633102
 2.2947E-05 4.7205E-06 4.3502E-05 1.7438E-05 4.8803E-04 1.9935E-04212633102
 8.4828E-05                                                       212633102
 0.0        0.0                 1          5         10          4212633102
 1.1000E+04 1.1500E+04 1.2000E+04 1.2750E+04 3.0436E-05 1.0200E-04212633102
 7.2626E-06 3.8928E-04 2.8587E-05 2.1192E-06                      212633102
 0.0        0.0                 1          5         15          5212633102
 2.3000E+04 2.4000E+04 2.5500E+04 2.7000E+04 2.8000E+04 2.0221E-03212633102
 1.1416E-04 2.6389E-07 8.7044E-08 1.0575E-03 8.1002E-07 1.7944E-07212633102
 6.8298E-04 3.0746E-05 9.1324E-05                                 212633102
 0.0        0.0                 0          1         12          6212633102
 3.4000E+04 1.5819E-05 3.6000E+04 4.9680E-12 3.8000E+04 6.5496E-05212633102
 4.0000E+04 4.8828E-05 4.2500E+04 1.6762E-05 4.5000E+04 0.0000E+00212633102
 0.0        0.0                 0          1         12          6212633102
 6.9000E+04 1.6282E-05 7.2000E+04 2.4062E-09 7.6000E+04 2.6452E-08212633102
 8.0000E+04 1.7513E-08 8.4000E+04 2.4018E-04 8.8000E+04 0.0000E+00212633102
                                                                  212633  0
                                                                  2126 0  0
                                                                     0 0  0
 2.20460E+4 4.55579E+1          0          0         34         102225 1451
 0.0        0.0                 0          0          0          62225 1451
 1.00000E+0 2.00000E+7          0          0         10       20022225 1451
 3.00000E+2 0.0                 1          0        233          52225 1451
 22-Ti- 46 FEI        EVAL-Jan02 K.I.Zolotarev                    2225 1451
                      DIST-Feb2004                                2225 1451
----IRDF-2002         MATERIAL 2225                               2225 1451
-----INCIDENT NEUTRON DATA                                        2225 1451
------ENDF-6 FORMAT                                               2225 1451
***************************************************************** 2225 1451
 22-TI- 46 FEI        EVAL-Jan02 K.I.Zolotarev                    2225 1451
                      DIST-Feb02                       20020203   2225 1451
----BROND-2           MATERIAL 2225                               2225 1451
-----INCIDENT NEUTRON DATA                                        2225 1451
------ENDF-6 FORMAT                                               2225 1451
------Russian Reactor Dosimetry File RRDF-2002                    2225 1451
***************************************************************** 2225 1451
Author  of evaluation: K.I.Zolotarev                              2225 1451
***************************************************************** 2225 1451
MF=3                                                              2225 1451
MT= 16 -(n,2n) cross section                                      2225 1451
-------------------------------------                             2225 1451
    Excitation function for the  Ti-46(n,2n)Ti-45 reaction in the 2225 1451
energy region from threshold to 20 MeV was evaluated by  means of 2225 1451
statistical analysis  of experimental cross section data [1-18].  2225 1451
    All experimental data were renormalized to  the new standards 2225 1451
for monitor reactions cross sections and decay data. In the measu-2225 1451
rements [5,14-16] uncertainties were increased until the data had 2225 1451
become mutually consistent.                                       2225 1451
    The final procedure of evaluation the Ti-46(n,2n)Ti-45  reac- 2225 1451
tion excitation function  has been carried out  within the frame- 2225 1451
work of generalized least squares method. Statistical analysis of 2225 1451
input cross section data  was carried out by means of PADE-2 code 2225 1451
[19].  Rational function was used as the model function [20].     2225 1451
    The evaluated  Ti-46(n,2n)Ti-45  excitation function averaged 2225 1451
on U-235 thermal neutron fission spectrum [21] and Cf-252 sponta- 2225 1451
neous fission neutron spectrum [22] gives the following values :  2225 1451
  2225 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2225 1451
  2225 1451
 U-235 neutron fission    0.0044686                             2225 1451
  2225 1451
 CF-252 spont. fission    0.011982        0.093 +- 0.031 [23]   2225 1451
  2225 1451
                                                                  2225 1451
MT=103 -(n,p) cross section                                       2225 1451
-------------------------------------                             2225 1451
    Excitation function for the  Ti-46(n,p)Sc-46m+g  reaction in  2225 1451
the energy region from threshold to 20 MeV was evaluated by means 2225 1451
of statistical analysis of experimental cross section data [6,17, 2225 1451
24-38] and data from GNASH calculation [39].                      2225 1451
    Analyzed microscopic experimental data  were  renormalized to 2225 1451
the new recommended standards for monitor reaction cross sections 2225 1451
and decay data. All experiments performed with Ti-natural samples 2225 1451
were corrected for the contribution of the Ti-47(n,np+pn)Sc-46m+g 2225 1451
and  Ti-47(n,d)Sc-46m+g  reactions. Total cross section for these 2225 1451
reactions were taken from ref.[40].  Experimental data  from ref. 2225 1451
[6,27] were used partially. Data of Bormann et al. [6] were taken 2225 1451
only for 14.1 MeV point. In the work of Lukic and Carroll [27] it 2225 1451
were used data measured relative Fe56(n,p)Mn56 and Al27(n,a)Na24  2225 1451
monitor reactions. These data were renormalized to the cross sec- 2225 1451
sections obtained in the ref.[28]. Data of Smith and Meadows [28] 2225 1451
measured  with using  D(d,n)He3 neutron source  were renormalized 2225 1451
in the energy range  6.996-9.950 MeV to the preliminary evaluated 2225 1451
integral of cross section from new experimental data [34,37,38].  2225 1451
    Experimental cross section  data [8] , [41-46] were  rejected 2225 1451
due to their discrepancy  with the main bulk of experimental data 2225 1451
[6,17,24-38] and data from theoretical model calculation.         2225 1451
    Statistical analysis of input cross section data was carried  2225 1451
out by means of PADE-2 code [19].  Rational function was used as  2225 1451
the model function [20].                                          2225 1451
    U-235 thermal fission [21]  and  Cf-252  spontaneous fission  2225 1451
neutron spectra [22] averaged cross sections calculated from the  2225 1451
evaluated  Ti-46(n,p)Sc-46m+g  reaction excitation function  are  2225 1451
the following:                                                    2225 1451
  2225 1451
   TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)   2225 1451
  2225 1451
U-235 neutron fission      11.447        11.57 +- 0.37  [47]    2225 1451
                                         11.51 +- 0.40  [48]    2225 1451
                                         11.55 +- 0.20  [49]    2225 1451
  2225 1451
CF-252 spont. fission      13.818        14.20 +- 0.24  [50]    2225 1451
                                         14.07 +- 0.25  [51]    2225 1451
  2225 1451
                                                                  2225 1451
MF=33                                                             2225 1451
MT= 16 -(n,2n) cross section cov. matrix                          2225 1451
----------------------------------------                          2225 1451
    Uncertainties in the evaluated  excitation function  for the  2225 1451
reaction Ti-46(n,2n)Ti-45 are given in the form of relative cova- 2225 1451
riance matrix for the 17-neutron energy groups (LB=5). Covariance 2225 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2225 1451
recommended cross section data by means of PADE-2 code.           2225 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2225 1451
given in the 33-file are the following:                           2225 1451
                                                                  2225 1451
    2.26418E-07    2.55017E-07    2.97541E-07    3.62667E-07      2225 1451
    4.74369E-07    6.90723E-07    1.00928E-06    1.38635E-06      2225 1451
    2.05902E-06    3.42154E-06    6.73918E-06    1.77794E-05      2225 1451
    1.16721E-04    5.82882E-03    7.49965E-03    2.28734E-02      2225 1451
    4.20028E-02                                                   2225 1451
                                                                  2225 1451
MT=103 -(n,p) cross section cov. matrix                           2225 1451
----------------------------------------                          2225 1451
    Uncertainties in the evaluated  excitation function  for the  2225 1451
reaction Ti-46(n,p)Sc-45m+g are given in the form of relative co- 2225 1451
variance matrix for the 31-neutron energy groups (LB=5).  Covari- 2225 1451
ance matrix of uncertainties  was calculated simultaneously  with 2225 1451
recommended cross section data by means of PADE-2 code.           2225 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2225 1451
given in the 33-file are the following:                           2225 1451
                                                                  2225 1451
    3.10934E-07    3.16123E-07    3.24666E-07    3.37022E-07      2225 1451
    3.51765E-07    3.72727E-07    3.91781E-07    4.23365E-07      2225 1451
    4.47112E-07    4.91264E-07    5.29779E-07    6.15458E-07      2225 1451
    6.75168E-07    7.93660E-07    9.79888E-07    1.29384E-06      2225 1451
    1.98373E-06    3.65835E-06    2.36258E-04    5.76816E-04      2225 1451
    7.66259E-04    1.00346E-03    1.10942E-03    1.43365E-03      2225 1451
    2.89439E-03    3.32523E-03    4.20534E-03    7.47772E-03      2225 1451
    1.06722E-02    1.67016E-02    7.87709E-02                     2225 1451
                                                                  2225 1451
 References :                                                     2225 1451
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22. W.Mannhart   IAEA-TECDOC-410, p.158, 1987                     2225 1451
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    Model Code for Calculation of Cross Section and Emission      2225 1451
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    Obninsk, May 1999                                             2225 1451
40. K.I.Zolotrarev  RRDF-2002, MAT=2222, eval. January 2002       2225 1451
41. D.L.Allan  Nucl. Phys., v.24, p.274, April 1961               2225 1451
42. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2225 1451
43. S.M.Qaim, N.I.Molla  Nucl. Phys., v.A283, p.269, June 1977    2225 1451
44. I.Ribansky, S.Gmuca  J. Phys. G, v.9, p.1537, December 1983   2225 1451
45. R.Pepelnik et al.  Progress Report, NEANDC(E)-262U,(5), p.32, 2225 1451
    June 1985                                                     2225 1451
46. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990    2225 1451
47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2225 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 2225 1451
48. W.Mannhart  Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry,  2225 1451
    Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985    2225 1451
49. W.Mannhart  Progress Report INDC(Ger)-045, pp.40-43, 1999     2225 1451
50. W.Mannhart  Handbook on Nuclear Activation Data . IAEA Tech-  2225 1451
    nical Report series No.273, p.413, 1987                       2225 1451
51. W.Mannhart  Validation of Differential Cross Sections with    2225 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2225 1451
    September 2002                                                2225 1451
***************************************************************** 2225 1451
 The Q values and threshold energies were updated prior to pro-   2225 1451
 cessing through the codes to comply with the values obtained     2225 1451
 using the NNDC calculation program which is based on the 1995    2225 1451
 Update to the Atomic mass Evaluation.                            2225 1451
                                                                  2225 1451
 File 2 added to the pointwise file containing only the effective 2225 1451
 scattering radius with no resonance parameters given.            2225 1451
 Taken from ENDF/B-VI                                             2225 1451
                                                                  2225 1451
 Psuedo Threshold of 0.0 added at 2.1E+6 ev to MF/MT=3/103        2225 1451
 in original evaluation                                           2225 1451
                                                                  2225 1451
***************************************************************** 2225 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2225 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2225 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2225 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2225 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2225 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2225 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2225 1451
 ***************** Program FIXUP (Version 2002-1) ****************2225 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2225 1451
 Corrected Thresholds-----------------------------------------Yes 2225 1451
 Extended Cross Sections to 20 MeV----------------------------No  2225 1451
 Allow Cross Section Deletion---------------------------------No  2225 1451
 Allow Cross Section Reconstruction---------------------------No  2225 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2225 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2225 1451
 Deleted Duplicate Points-------------------------------------Yes 2225 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2225 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2225 1451
 Allow Creation of Missing Sections---------------------------No  2225 1451
 Allow Insertion of Energy Points-----------------------------No  2225 1451
 Create Uniform Energy Grid-----------------------------------No  2225 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2225 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2225 1451
 Unshielded Group Averages Using  640 Groups                      2225 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2225 1451
                                1        451        242          12225 1451
                                3         16         26          12225 1451
                                3        103         63          12225 1451
                               33         16         35          12225 1451
                               33        103         97          12225 1451
                                                                  2225 1  0
                                                                  2225 0  0
 2.20460E+4 4.55579E+1          0          0          0          02225 3 16
-1.31898E+7-1.31898E+7          0          0          1         672225 3 16
         67          1                                            2225 3 16
 13400000.0 5.44722E-6 13500000.0 .000269054 13600000.0 .0013538942225 3 16
 13700000.0 .003167185 13800000.0 .005539635 13900000.0 .0084628702225 3 16
 14000000.0 .011885800 14100000.0 .015756100 14200000.0 .0200212002225 3 16
 14300000.0 .024629400 14400000.0 .029530600 14500000.0 .0346768502225 3 16
 14600000.0 .040022950 14700000.0 .045526750 14800000.0 .0511660252225 3 16
 14900000.0 .056872075 15000000.0 .062608313 15100000.0 .0683747382225 3 16
 15200000.0 .074141163 15300000.0 .079907588 15400000.0 .0855952332225 3 16
 15500000.0 .091204100 15600000.0 .096812967 15700000.0 .1022853332225 3 16
 15800000.0 .107621200 15900000.0 .112957067 16000000.0 .1181235002225 3 16
 16100000.0 .123120500 16200000.0 .128117500 16300000.0 .1329310002225 3 16
 16400000.0 .137561000 16500000.0 .142191000 16600000.0 .1466356672225 3 16
 16700000.0 .150895000 16800000.0 .155154333 16900000.0 .1592341672225 3 16
 17000000.0 .163134500 17100000.0 .167034833 17200000.0 .1707660002225 3 16
 17300000.0 .174328000 17400000.0 .177890000 17500000.0 .1812951672225 3 16
 17600000.0 .184543500 17700000.0 .187791833 17800000.0 .1908743752225 3 16
 17900000.0 .193791125 18000000.0 .196707875 18100000.0 .1996246252225 3 16
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 19700000.0 .234977375 19800000.0 .236689625 19900000.0 .2384018752225 3 16
 20000000.0 0.0                                                   2225 3 16
                                                                  2225 3  0
 2.20460E+4 4.55579E+1          0          0          0          02225 3103
-1.58430E+6-1.58430E+6          0          0          1        1802225 3103
        180          1                                            2225 3103
 2100000.00 3.42206E-6 2200000.00 1.02662E-5 2300000.00 1.71103E-52225 3103
 2400000.00 2.39544E-5 2500000.00 3.98904E-5 2600000.00 9.93846E-52225 3103
 2700000.00 .000237889 2800000.00 .000478634 2900000.00 .0008521572225 3103
 3000000.00 .001398970 3100000.00 .002172540 3200000.00 .0032406352225 3103
 3300000.00 .004678875 3400000.00 .006545195 3500000.00 .0088332052225 3103
 3600000.00 .011449050 3700000.00 .014282900 3800000.00 .0173307502225 3103
 3900000.00 .020716350 4000000.00 .024597800 4100000.00 .0290779002225 3103
 4200000.00 .034170000 4300000.00 .039792900 4400000.00 .0457663252225 3103
 4500000.00 .051856975 4600000.00 .057775050 4700000.00 .0632341002225 3103
 4800000.00 .068060950 4900000.00 .072219050 5000000.00 .0757965502225 3103
 5100000.00 .079116450 5200000.00 .082497900 5300000.00 .0861876502225 3103
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 5700000.00 .108255000 5800000.00 .115777000 5900000.00 .1239030002225 3103
 6000000.00 .132481167 6100000.00 .141251500 6200000.00 .1500218332225 3103
 6300000.00 .158677500 6400000.00 .167038500 6500000.00 .1749920002225 3103
 6600000.00 .182459000 6700000.00 .189393500 6800000.00 .1957795002225 3103
 6900000.00 .201623000 7000000.00 .206947500 7100000.00 .2116860002225 3103
 7200000.00 .216090000 7300000.00 .220127250 7400000.00 .2237977502225 3103
 7500000.00 .227143500 7600000.00 .230164500 7700000.00 .2331855002225 3103
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 8100000.00 .243425875 8200000.00 .245762125 8300000.00 .2479403752225 3103
 8400000.00 .250118625 8500000.00 .252296875 8600000.00 .2544751252225 3103
 8700000.00 .256653375 8800000.00 .258831625 8900000.00 .2610098752225 3103
 9000000.00 .263188125 9100000.00 .265366375 9200000.00 .2675446252225 3103
 9300000.00 .269722875 9400000.00 .271901125 9500000.00 .2740793752225 3103
 9600000.00 .276257625 9700000.00 .278435875 9800000.00 .2804955832225 3103
 9900000.00 .282436750 10000000.0 .284377917 10100000.0 .2863190832225 3103
 10200000.0 .288260250 10300000.0 .290201417 10400000.0 .2919556252225 3103
 10500000.0 .293522875 10600000.0 .295090125 10700000.0 .2966573752225 3103
 10800000.0 .297992875 10900000.0 .299096625 11000000.0 .3002003752225 3103
 11100000.0 .301304125 11200000.0 .302152667 11300000.0 .3027460002225 3103
 11400000.0 .303339333 11500000.0 .303725000 11600000.0 .3039030002225 3103
 11700000.0 .303857833 11800000.0 .303589500 11900000.0 .3033211672225 3103
 12000000.0 .302775333 12100000.0 .301952000 12200000.0 .3011286672225 3103
 12300000.0 .300027833 12400000.0 .298649500 12500000.0 .2972711672225 3103
 12600000.0 .295628333 12700000.0 .293721000 12800000.0 .2918136672225 3103
 12900000.0 .289664667 13000000.0 .287274000 13100000.0 .2848833332225 3103
 13200000.0 .282253250 13300000.0 .279383750 13400000.0 .2765142502225 3103
 13500000.0 .273644750 13600000.0 .270544600 13700000.0 .2672138002225 3103
 13800000.0 .263883000 13900000.0 .260552200 14000000.0 .2572214002225 3103
 14100000.0 .253720375 14200000.0 .250049125 14300000.0 .2463778752225 3103
 14400000.0 .242706625 14500000.0 .239035375 14600000.0 .2353641252225 3103
 14700000.0 .231692875 14800000.0 .228021625 14900000.0 .2243471882225 3103
 15000000.0 .220669563 15100000.0 .216991938 15200000.0 .2133143132225 3103
 15300000.0 .209636688 15400000.0 .205959063 15500000.0 .2022814382225 3103
 15600000.0 .198603813 15700000.0 .195037600 15800000.0 .1915828002225 3103
 15900000.0 .188128000 16000000.0 .184673200 16100000.0 .1812184002225 3103
 16200000.0 .177889500 16300000.0 .174686500 16400000.0 .1714835002225 3103
 16500000.0 .168280500 16600000.0 .165077500 16700000.0 .1619997502225 3103
 16800000.0 .159047250 16900000.0 .156094750 17000000.0 .1531422502225 3103
 17100000.0 .150303750 17200000.0 .147579250 17300000.0 .1448547502225 3103
 17400000.0 .142130250 17500000.0 .139517250 17600000.0 .1370157502225 3103
 17700000.0 .134514250 17800000.0 .132012750 17900000.0 .1296180002225 3103
 18000000.0 .127330000 18100000.0 .125042000 18200000.0 .1227540002225 3103
 18300000.0 .120566000 18400000.0 .118478000 18500000.0 .1163900002225 3103
 18600000.0 .114302000 18700000.0 .112306625 18800000.0 .1104038752225 3103
 18900000.0 .108501125 19000000.0 .106598375 19100000.0 .1047807882225 3103
 19200000.0 .103048363 19300000.0 .101315938 19400000.0 .0995835132225 3103
 19500000.0 .097928538 19600000.0 .096351013 19700000.0 .0947734882225 3103
 19800000.0 .093195963 19900000.0 .091663500 20000000.0 0.0       2225 3103
                                                                  2225 3  0
                                                                  2225 0  0
 2.20460E+4 4.55579E+1          0          0          0          1222533 16
 0.000000+0 0.000000+0          0         16          0          1222533 16
 0.000000+0 0.000000+0          1          5        190         19222533 16
 1.000000-5 1.340000+7 1.400000+7 1.425000+7 1.450000+7 1.475000+7222533 16
 1.500000+7 1.525000+7 1.550000+7 1.575000+7 1.600000+7 1.650000+7222533 16
 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7222533 16
 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533 16
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 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533 16
 0.000000+0 2.781420-2 1.283920-2 7.150180-3 4.197760-3 2.624760-3222533 16
 1.798060-3 1.390020-3 1.219080-3 1.180400-3 1.246260-3 1.393920-3222533 16
 1.529020-3 1.618830-3 1.660180-3 1.659830-3 1.626720-3 1.569240-3222533 16
 8.069150-3 5.546650-3 3.876040-3 2.695150-3 1.848660-3 1.244700-3222533 16
 8.222770-4 5.378570-4 3.076620-4 2.302510-4 3.187630-4 5.002420-4222533 16
 7.302930-4 9.821299-4 1.239760-3 1.493850-3 4.441590-3 3.461700-3222533 16
 2.659900-3 2.009820-3 1.496660-3 1.103050-3 8.113270-4 5.333680-4222533 16
 3.719090-4 3.756580-4 4.829110-4 6.520230-4 8.556370-4 1.076080-3222533 16
 1.302130-3 2.992840-3 2.525270-3 2.115430-3 1.766930-3 1.480710-3222533 16
 1.252600-3 1.007340-3 8.189470-4 7.457220-4 7.499250-4 8.043140-4222533 16
 8.900250-4 9.942610-4 1.108480-3 2.356520-3 2.171160-3 1.996940-3222533 16
 1.836860-3 1.693860-3 1.511730-3 1.326660-3 1.194050-3 1.101810-3222533 16
 1.039690-3 9.996680-4 9.757380-4 9.634650-4 2.189000-3 2.170040-3222533 16
 2.129830-3 2.072140-3 1.961190-3 1.796150-3 1.623460-3 1.453990-3222533 16
 1.293310-3 1.143860-3 1.006340-3 8.805220-4 2.287180-3 2.348570-3222533 16
 2.367250-3 2.325030-3 2.189140-3 1.994320-3 1.770100-3 1.535250-3222533 16
 1.301110-3 1.074200-3 8.580140-4 2.496620-3 2.577930-3 2.597300-3222533 16
 2.495160-3 2.293670-3 2.036390-3 1.752220-3 1.459490-3 1.169360-3222533 16
 8.883550-4 2.712310-3 2.783360-3 2.716800-3 2.521360-3 2.250200-3222533 16
 1.939420-3 1.612380-3 1.283690-3 9.621380-4 2.917810-3 2.904460-3222533 16
 2.737660-3 2.476860-3 2.163820-3 1.826070-3 1.481200-3 1.140080-3222533 16
 2.958400-3 2.849680-3 2.639550-3 2.369860-3 2.069060-3 1.755710-3222533 16
 1.441570-3 2.814730-3 2.683000-3 2.492950-3 2.269720-3 2.030300-3222533 16
 1.785710-3 2.645740-3 2.555400-3 2.433170-3 2.292510-3 2.142650-3222533 16
 2.575470-3 2.565690-3 2.536570-3 2.494660-3 2.673410-3 2.760570-3222533 16
 2.831970-3 2.964830-3 3.149770-3 3.446400-3                      222533 16
                                                                  222533  0
 2.20460E+4 4.55579E+1          0          0          0          1222533103
 0.000000+0 0.000000+0          0        103          0          1222533103
 0.000000+0 0.000000+0          1          5        561         33222533103
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 1.150000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7 1.400000+7222533103
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 3.485020-5 7.171340-5 7.966920-5 5.982700-5 1.883050-5-3.412800-5222533103
-8.970580-5-1.398910-4-1.786300-4-2.044440-4-1.784490-4-7.872330-5222533103
 1.445120-3 1.083580-3 6.775910-4 3.228050-4 6.477870-5-8.405100-5222533103
-1.334110-4-1.076300-4-3.862740-5 4.120460-5 1.054510-4 1.378040-4222533103
 1.330070-4 9.502850-5 3.371470-5-3.860100-5-1.097120-4-1.690850-4222533103
-2.150220-4-1.835830-4-2.528830-5 9.449480-4 7.149460-4 4.759940-4222533103
 2.760980-4 1.382640-4 6.658500-5 5.147420-5 7.484400-5 1.155630-4222533103
 1.546680-4 1.791070-4 1.831920-4 1.679270-4 1.389590-4 1.042630-4222533103
 7.221500-5 5.034500-5 5.195700-5 1.295790-4 3.101290-4 6.811540-4222533103
 6.084510-4 5.221670-4 4.374450-4 3.632780-4 3.038860-4 2.601880-4222533103
 2.310620-4 2.144760-4 2.083910-4 2.113000-4 2.224300-4 2.416450-4222533103
 2.692310-4 3.056580-4 3.514030-4 4.379840-4 5.881100-4 7.771760-4222533103
 6.991220-4 7.367750-4 7.216420-4 6.605220-4 5.672850-4 4.601860-4222533103
 3.585400-4 2.790880-4 2.332240-4 2.259000-4 2.562190-4 3.191390-4222533103
 4.074230-4 5.132960-4 6.295330-4 8.071950-4 1.037690-3 1.238670-3222533103
 8.790310-4 9.295870-4 8.911490-4 7.816890-4 6.311210-4 4.740970-4222533103
 3.424850-4 2.593850-4 2.363030-4 2.736960-4 3.638140-4 4.942810-4222533103
 6.512370-4 8.215070-4 1.072590-3 1.374030-3 1.596580-3 1.030230-3222533103
 1.017920-3 9.115610-4 7.451500-4 5.608290-4 3.988850-4 2.897880-4222533103
 2.500630-4 2.824190-4 3.789560-4 5.254750-4 7.054170-4 9.026670-4222533103
 1.194860-3 1.543810-3 1.794240-3 1.031340-3 9.444440-4 7.900760-4222533103
 6.093440-4 4.433830-4 3.246260-4 2.720760-4 2.908640-4 3.750690-4222533103
 5.118720-4 6.855040-4 8.801580-4 1.175590-3 1.539350-3 1.814440-3222533103
 8.888590-4 7.683740-4 6.179780-4 4.727510-4 3.619820-4 3.042870-4222533103
 3.066400-4 3.663150-4 4.741640-4 6.179180-4 7.847390-4 1.048310-3222533103
 1.391040-3 1.675830-3 6.951350-4 5.922860-4 4.861340-4 3.987540-4222533103
 3.458160-4 3.349860-4 3.668070-4 4.367340-4 5.373920-4 6.604250-4222533103
 8.666360-4 1.155910-3 1.425790-3 5.437350-4 4.853040-4 4.317030-4222533103
 3.935570-4 3.777810-4 3.872170-4 4.214350-4 4.778520-4 5.527710-4222533103
 6.898540-4 9.026440-4 1.128670-3 4.746400-4 4.584020-4 4.431900-4222533103
 4.334040-4 4.325120-4 4.425760-4 4.644360-4 4.980230-4 5.687730-4222533103
 6.952440-4 8.510190-4 4.782020-4 4.903100-4 4.974660-4 5.014420-4222533103
 5.045490-4 5.088950-4 5.162030-4 5.356930-4 5.801400-4 6.480210-4222533103
 5.318150-4 5.641130-4 5.877260-4 6.024600-4 6.093140-4 6.096570-4222533103
 6.009780-4 5.800660-4 5.550780-4 6.279110-4 6.821970-4 7.248890-4222533103
 7.532320-4 7.662530-4 7.557490-4 6.950360-4 5.859180-4 7.756650-4222533103
 8.574140-4 9.223740-4 9.650720-4 9.782880-4 9.082420-4 7.360870-4222533103
 9.863660-4 1.097040-3 1.182050-3 1.240980-3 1.193460-3 9.889720-4222533103
 1.260120-3 1.394240-3 1.515950-3 1.521600-3 1.321900-3 1.583350-3222533103
 1.778520-3 1.865200-3 1.710910-3 2.097680-3 2.348220-3 2.342350-3222533103
 2.885560-3 3.210870-3 4.018350-3                                 222533103
                                                                  222533  0
                                                                  2225 0  0
                                                                     0 0  0
 2.20470E+4 4.65484E+1          0          0         34         102228 1451
 0.0        0.0                 0          0          0          62228 1451
 1.00000E+0 2.00000E+7          0          0         10       20022228 1451
 3.00000E+2 0.0                 1          0        163          62228 1451
 22-Ti- 47 FEI/ANL    EVAL-Oct95 K.I.Zolotarev, C.Philis et al.   2228 1451
                      DIST-Feb2004                                2228 1451
----IRDF-2002         MATERIAL 2228                               2228 1451
-----INCIDENT NEUTRON DATA                                        2228 1451
------ENDF-6 FORMAT                                               2228 1451
***************************************************************** 2228 1451
 22-TI-47  FEI        EVAL-Oct95 K.I.Zolotarev                    2228 1451
                      DIST-Feb02 Mod1                  20020203   2228 1451
----BROND-2           MATERIAL 2228         REVISION 1            2228 1451
-----INCIDENT NEUTRON DATA                                        2228 1451
------ENDF-6 FORMAT                                               2228 1451
------Russian Reactor Dosimetry File RRDF-2002                    2228 1451
***************************************************************** 2228 1451
 22-TI- 47 ANL        EVAL-JAN77  C.PHILIS,O.BERSILLON,D.SMITH,ETC2228 1451
                      DIST-FEB91                       910201     2228 1451
----IRDF-90           MATERIAL 2228                               2228 1451
    ENDF/B-VI         MATERIAL 2228                               2228 1451
 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC             2228 1451
***************************************************************** 2228 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2228 1451
DOSIMETRY                                                         2228 1451
***************************************************************** 2228 1451
********     Start of (N,NP) bibliographical component    ******* 2228 1451
***************************************************************** 2228 1451
Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko           2228 1451
                                                                  2228 1451
 For this special purpose library it was decided the reaction     2228 1451
 MF=3 MT=28+104 would use MF=10, MT=5 to present this data.       2228 1451
 This was done after processing through the codes. The            2228 1451
 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5    2228 1451
******************************************************************2228 1451
MF=3                                                              2228 1451
MT= 28 -(n,np+pn+d) cross section data                            2228 1451
--------------------------------------                            2228 1451
    In this section is given the sum of cross section of the reac-2228 1451
tions Ti47(n,np)Sc46m+g , Ti47(n,pn)Sc46m+g and Ti47(n,d)Sc46m+g. 2228 1451
    Excitation function for the Ti47(n,x)Sc46m+g reaction  in the 2228 1451
energy region from threshold to 20 MeV was evaluated  by means of 2228 1451
statistical analysis of experimental cross section data [1-6] and 2228 1451
data from STAPRE [7] calculation.                                 2228 1451
    All experimental data were renormalized to  the new standards 2228 1451
for monitor reactions cross sections and decay data.              2228 1451
    The final procedure of evaluation Ti47(n,x)Sc46m+g excitation 2228 1451
function from threshold to 20 MeV has been carried out within the 2228 1451
framework of generalized least squares method.  Rational function 2228 1451
was used  as  model function [8].  Calculations  was performed by 2228 1451
means of  Pade-2 code [9].                                        2228 1451
    U-235 thermal fission [10]  and  Cf-252 spontaneous  fission  2228 1451
neutron spectra [11] averaged cross-sections calculated from the  2228 1451
evaluated Ti47(n,x)Sc46m+g excitation function are the following: 2228 1451
                                                                  2228 1451
         --------------------------------------------             2228 1451
              TYPE OF SPECTRUM     I <CS>, mb (calc.)             2228 1451
         --------------------------I-----------------             2228 1451
         U-235  neutron  fission   I    8.1158E-3                 2228 1451
         CF-252 spontan. fission   I    1.9201E-2                 2228 1451
                                                                  2228 1451
MF=33                                                             2228 1451
MT= 28 -(n,np+pn+d) cross section cov. matrix                     2228 1451
---------------------------------------------                     2228 1451
    Uncertainties in the evaluated  excitation function  for the  2228 1451
reaction  Ti-47(n,x)Sc-46m+g  are given in the form  of relative  2228 1451
covariance matrix for the 17-neutron energy groups (LB=5).  Cova- 2228 1451
riance matrix of uncertainties was calculated simultaneously with 2228 1451
recommended cross section data by means of PADE-2 code.           2228 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2228 1451
given in the 33-file are the following:                           2228 1451
                                                                  2228 1451
    3.91044E-07    4.46920E-07    5.37497E-07    7.02044E-07      2228 1451
    9.85160E-07    1.54860E-06    2.86775E-06    6.97782E-06      2228 1451
    2.67223E-05    1.90995E-04    9.14779E-04    5.12707E-03      2228 1451
    8.27566E-03    1.15029E-02    1.68918E-02    5.65241E-02      2228 1451
    8.27240E-01                                                   2228 1451
                                                                  2228 1451
 References :                                                     2228 1451
 1. W.G.Cross, H.L.Pai Progress Rep. EANDC(CAN)-16, p.1, Jan.1963 2228 1451
 2. H.L.Pai  Canadian J. of Physics, v.44, p.2337, 1966           2228 1451
 3. I.Ribansky, S.Gmuca  J. Phys.G, v.9, p.1537, December 1983    2228 1451
 4. N.I.Molla et al.  Report INDC(BAN)-003, September 1986        2228 1451
 5. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2228 1451
 6. Y.Uno et al.  Report JAERI-M-93-046, p.247-256, 1993          2228 1451
 7. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2228 1451
    Activation Cross Section and Related Quantities, Report       2228 1451
    IRK 76-01, Vienna, 1976                                       2228 1451
 8. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2228 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2228 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2228 1451
 9. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2228 1451
10. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2228 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2228 1451
11. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2228 1451
******************************************************************2228 1451
********      End of (N,NP) bibliographical component     ******* 2228 1451
***************************************************************** 2228 1451
***************************************************************** 2228 1451
********     Start of (N,P) bibliographical component     ******* 2228 1451
***************************************************************** 2228 1451
 **                                                               2228 1451
  DOCUMENTATION ANL/NDM-27 (N,P)                                  2228 1451
 **                                                               2228 1451
  COMMENTS- (N,P) EVAL IN ENERGY RANGES 1.10E+06 TO 1.0E+07 AND   2228 1451
  1.35E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA.INTERPOLATION AND2228 1451
  EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCULATION2228 1451
  PERFORMED AT BRUYERES LE CHATEL USING CODE DEVELOPED BY JARRE.  2228 1451
  PSEUDO-THRESHOLD=0.5 MEV PER P.C. MACFARLANE TO MAGURNO 6/20/83 2228 1451
 **                                                               2228 1451
  REFERENCES                                                      2228 1451
   AL61 ALLAN D. NP24,274,61                                      2228 1451
   AR78 ARTHUR E. PC 1978                                         2228 1451
   AR75 ARMITAGE P.C. NNDC                                        2228 1451
   CR62 CROSS W.AND PAI H. EANDC(CAN)-16,2,62                     2228 1451
   FU79 C.Y.FU PRIVATE COMM 1979                                  2228 1451
   GH71 GHORAI S.AND CARROLL E. NSE 43,233,71                     2228 1451
   GO62 GONZALES ET.AL. PR126,271,62                              2228 1451
   HI   HILLMAN M. NP37,78,62                                     2228 1451
   LE69 LEVKOVSKII V. Y.F.10,44,69                                2228 1451
   PA66 PAI H.CJP 44 2337,66                                      2228 1451
   PO59 POULARIKIS A. AND FINK R. PR115,989,59                    2228 1451
   SM75 SMITH D.L.AND MEADOWS J. ANL/NDM10,75                     2228 1451
   TI68 TIKKU ET.AL. SYMP ON NUCL CROSS SECT AND TECH WASH.1968   2228 1451
******************************************************************2228 1451
********      End of (N,P) bibliographical component      ******* 2228 1451
***************************************************************** 2228 1451
 The Q values and threshold energies were updated prior to pro-   2228 1451
 cessing through the codes to comply with the values obtained     2228 1451
 using the NNDC calculation program which is based on the 1995    2228 1451
 Update to the Atomic mass Evaluation.                            2228 1451
                                                                  2228 1451
 File 2 added to the pointwise file containing only the effective 2228 1451
 scattering radius with no resonance parameters given.            2228 1451
 Taken from ENDF/B-VI                                             2228 1451
                                                                  2228 1451
 Threshold of 0.0 added at 1.0E-5 ev to MF/MT=3/103 after         2228 1451
 processing thro. the pre-procssing codes.                        2228 1451
                                                                  2228 1451
                                                                  2228 1451
***************************************************************** 2228 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2228 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2228 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2228 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2228 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2228 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2228 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2228 1451
 ***************** Program FIXUP (Version 2002-1) ****************2228 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2228 1451
 Corrected Thresholds-----------------------------------------Yes 2228 1451
 Extended Cross Sections to 20 MeV----------------------------No  2228 1451
 Allow Cross Section Deletion---------------------------------No  2228 1451
 Allow Cross Section Reconstruction---------------------------No  2228 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2228 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2228 1451
 Deleted Duplicate Points-------------------------------------Yes 2228 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2228 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2228 1451
 Allow Creation of Missing Sections---------------------------No  2228 1451
 Allow Insertion of Energy Points-----------------------------No  2228 1451
 Create Uniform Energy Grid-----------------------------------No  2228 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2228 1451
***************************************************************** 2228 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2228 1451
 Unshielded Group Averages Using  640 Groups                      2228 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2228 1451
                                1        451        173          12228 1451
                                3        103         70          02228 1451
                                8          5          2          12228 1451
                               10          5         42          12228 1451
                               33        103        115          12228 1451
                               40          5         36          02228 1451
                                                                  2228 1  0
                                                                  2228 0  0
 2.20470E+4 4.65484E+1          0          0          0          02228 3103
 1.82300E+5 1.82300E+5          0          0          1        1992228 3103
        199          1                                            2228 3103
 690000.000 5.16780E-5 720000.000 .000310068 760000.000 .0006201362228 3103
 800000.000 .000805043 840000.000 .000864789 880000.000 .0009245352228 3103
 920000.000 .000984281 960000.000 .001044027 1000000.00 .0013463902228 3103
 1100000.00 .001891370 1200000.00 .002436350 1300000.00 .0029813302228 3103
 1400000.00 .003526310 1500000.00 .004697020 1600000.00 .0064934602228 3103
 1700000.00 .008289900 1800000.00 .010086340 1900000.00 .0118827802228 3103
 2000000.00 .013905700 2100000.00 .016155100 2200000.00 .0184045002228 3103
 2300000.00 .020653900 2400000.00 .022903300 2500000.00 .0250525002228 3103
 2600000.00 .027101500 2700000.00 .029097500 2800000.00 .0310405002228 3103
 2900000.00 .032983500 3000000.00 .034727500 3100000.00 .0362725002228 3103
 3200000.00 .037817500 3300000.00 .039362500 3400000.00 .0409075002228 3103
 3500000.00 .042653900 3600000.00 .044601700 3700000.00 .0465495002228 3103
 3800000.00 .048497300 3900000.00 .050445100 4000000.00 .0520001002228 3103
 4100000.00 .053162300 4200000.00 .054324500 4300000.00 .0554867002228 3103
 4400000.00 .056648900 4500000.00 .057830600 4600000.00 .0590318002228 3103
 4700000.00 .060233000 4800000.00 .061434200 4900000.00 .0626354002228 3103
 5000000.00 .064098300 5100000.00 .065822900 5200000.00 .0675475002228 3103
 5300000.00 .069272100 5400000.00 .070996700 5500000.00 .0723986002228 3103
 5600000.00 .073477800 5700000.00 .074557000 5800000.00 .0756362002228 3103
 5900000.00 .076715400 6000000.00 .078082700 6100000.00 .0797381002228 3103
 6200000.00 .081393500 6300000.00 .083048900 6400000.00 .0847043002228 3103
 6500000.00 .086108400 6600000.00 .087261200 6700000.00 .0884140002228 3103
 6800000.00 .089566800 6900000.00 .090719600 7000000.00 .0919561002228 3103
 7100000.00 .093276300 7200000.00 .094596500 7300000.00 .0959167002228 3103
 7400000.00 .097236900 7500000.00 .098933300 7600000.00 .1010059002228 3103
 7700000.00 .103078500 7800000.00 .105151100 7900000.00 .1072237002228 3103
 8000000.00 .109065000 8100000.00 .110675000 8200000.00 .1122850002228 3103
 8300000.00 .113895000 8400000.00 .115505000 8500000.00 .1168730002228 3103
 8600000.00 .117999000 8700000.00 .119125000 8800000.00 .1202510002228 3103
 8900000.00 .121377000 9000000.00 .122315000 9100000.00 .1230650002228 3103
 9200000.00 .123815000 9300000.00 .124565000 9400000.00 .1253150002228 3103
 9500000.00 .126237000 9600000.00 .127331000 9700000.00 .1284250002228 3103
 9800000.00 .129519000 9900000.00 .130613000 10000000.0 .1312550002228 3103
 10100000.0 .131445000 10200000.0 .131635000 10300000.0 .1318250002228 3103
 10400000.0 .132015000 10500000.0 .132205000 10600000.0 .1323950002228 3103
 10700000.0 .132585000 10800000.0 .132775000 10900000.0 .1329650002228 3103
 11000000.0 .132870000 11100000.0 .132490000 11200000.0 .1321100002228 3103
 11300000.0 .131730000 11400000.0 .131350000 11500000.0 .1309700002228 3103
 11600000.0 .130590000 11700000.0 .130210000 11800000.0 .1298300002228 3103
 11900000.0 .129450000 12000000.0 .128879750 12100000.0 .1281192502228 3103
 12200000.0 .127358750 12300000.0 .126598250 12400000.0 .1258377502228 3103
 12500000.0 .125077250 12600000.0 .124316750 12700000.0 .1235562502228 3103
 12800000.0 .122795750 12900000.0 .122035250 13000000.0 .1212747502228 3103
 13100000.0 .120514250 13200000.0 .119753750 13300000.0 .1189932502228 3103
 13400000.0 .118232750 13500000.0 .117472250 13600000.0 .1167117502228 3103
 13700000.0 .115951250 13800000.0 .115190750 13900000.0 .1144302502228 3103
 14000000.0 .113575000 14100000.0 .112625000 14200000.0 .1116750002228 3103
 14300000.0 .110725000 14400000.0 .109775000 14500000.0 .1087300002228 3103
 14600000.0 .107590000 14700000.0 .106450000 14800000.0 .1053100002228 3103
 14900000.0 .104170000 15000000.0 .103124400 15100000.0 .1021732002228 3103
 15200000.0 .101222000 15300000.0 .100270800 15400000.0 .0993196002228 3103
 15500000.0 .098273800 15600000.0 .097133400 15700000.0 .0959930002228 3103
 15800000.0 .094852600 15900000.0 .093712200 16000000.0 .0926786752228 3103
 16100000.0 .091752025 16200000.0 .090825375 16300000.0 .0898987252228 3103
 16400000.0 .088972075 16500000.0 .088045425 16600000.0 .0871187752228 3103
 16700000.0 .086192125 16800000.0 .085265475 16900000.0 .0843388252228 3103
 17000000.0 .083412175 17100000.0 .082485525 17200000.0 .0815588752228 3103
 17300000.0 .080632225 17400000.0 .079705575 17500000.0 .0787789252228 3103
 17600000.0 .077852275 17700000.0 .076925625 17800000.0 .0759989752228 3103
 17900000.0 .075072325 18000000.0 .074409400 18100000.0 .0740102002228 3103
 18200000.0 .073611000 18300000.0 .073211800 18400000.0 .0728126002228 3103
 18500000.0 .072413400 18600000.0 .072014200 18700000.0 .0716150002228 3103
 18800000.0 .071215800 18900000.0 .070816600 19000000.0 .0704174002228 3103
 19100000.0 .070018200 19200000.0 .069619000 19300000.0 .0692198002228 3103
 19400000.0 .068820600 19500000.0 .068411900 19600000.0 .0679937002228 3103
 19700000.0 .067575500 19800000.0 .067157300 19900000.0 .0667391002228 3103
 20000000.0 0.0                                                   2228 3103
                                                                  2228 3  0
                                                                  2228 0  0
 2.20470E+4 4.65484E+1          0          0          1          02228 3231
-8.23750E+6-8.23750E+6      21046          0          1        1172228 3231
        117          1                                            2228 3231
 8400000.00 1.03337E-8 8500000.00 3.81554E-8 8600000.00 6.62625E-82228 3231
 8700000.00 9.43696E-8 8800000.00 1.22477E-7 8900000.00 1.50584E-72228 3231
 9000000.00 1.78691E-7 9100000.00 2.06798E-7 9200000.00 2.34905E-72228 3231
 9300000.00 2.63012E-7 9400000.00 2.91119E-7 9500000.00 3.73532E-62228 3231
 9600000.00 8.18206E-6 9700000.00 9.01171E-6 9800000.00 8.63786E-62228 3231
 9900000.00 9.48327E-6 10000000.0 1.39527E-5 10100000.0 2.43832E-52228 3231
 10200000.0 4.30199E-5 10300000.0 7.19698E-5 10400000.0 .0001131522228 3231
 10500000.0 .000168247 10600000.0 .000238639 10700000.0 .0003253602228 3231
 10800000.0 .000429044 10900000.0 .000549884 11000000.0 .0006876222228 3231
 11100000.0 .000841563 11200000.0 .001010656 11300000.0 .0011936252228 3231
 11400000.0 .001389225 11500000.0 .001596590 11600000.0 .0018156902228 3231
 11700000.0 .002047940 11800000.0 .002296875 11900000.0 .0025688152228 3231
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 17700000.0 .174035700 17800000.0 .176967100 17900000.0 .1798985002228 3231
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 19800000.0 .226149500 19900000.0 .228098500 20000000.0 0.0       2228 3231
                                                                  2228 3  0
                                                                  2228 0  0
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                                                                  222833  0
                                                                  2228 0  0
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 2.378610-3 3.491600-3 3.386500-3 3.181570-3 2.911070-3 2.602730-3222833231
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 2.860660-3 3.016960-3 2.911930-3 2.923980-3                      222833231
                                                                  222833  0
                                                                  2228 0  0
                                                                     0 0  0
 2.20480E+4 4.75360E+1          0          0         34         102231 1451
 0.0        0.0                 0          0          0          62231 1451
 1.00000E+0 2.00000E+7          0          0         10       20022231 1451
 3.00000E+2 0.0                 1          0        242          62231 1451
 22-Ti- 48 FEI        EVAL-Feb02 K.I.Zolotarev                    2231 1451
                      DIST-Feb2004                                2231 1451
----IRDF-2002         MATERIAL 2231                               2231 1451
-----INCIDENT NEUTRON DATA                                        2231 1451
------ENDF-6 FORMAT                                               2231 1451
***************************************************************** 2231 1451
**********************  SPECIAL PURPOSE FILE   *****************  2231 1451
DOSIMETRY                                                         2231 1451
                                                                  2231 1451
 For this special purpose library it was decided the reaction     2231 1451
 MF=3 MT=28+104 would use MF=10, MT=5 to present this data.       2231 1451
 This was done after processing through the codes. The            2231 1451
 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5    2231 1451
***************************************************************** 2231 1451
 22-TI-48  FEI        EVAL-Feb02 K.I.Zolotarev                    2231 1451
                      DIST-Feb02                       20020214   2231 1451
----BROND-2           MATERIAL 2231                               2231 1451
-----INCIDENT NEUTRON DATA                                        2231 1451
------ENDF-6 FORMAT                                               2231 1451
------Russian Reactor Dosimetry File RRDF-2002                    2231 1451
***************************************************************** 2231 1451
Author  of evaluation: K.I.Zolotarev                              2231 1451
***************************************************************** 2231 1451
MF= 3                                                             2231 1451
MT= 28 -(n,np+pn+d) cross section data                            2231 1451
--------------------------------------                            2231 1451
    In this section is given the sum of cross sections of the re- 2231 1451
actions Ti-48(n,np)Sc-47 , Ti-48(n,pn)Sc-47 and Ti-48(n,d)Sc-47.  2231 1451
    Excitation function for the  Ti-48(n,x)Sc-47  reaction in the 2231 1451
energy region from threshold to 20 MeV was evaluated  by means of 2231 1451
statistical analysis  of  experimental cross section data [1-12]  2231 1451
and data from STAPRE [13] calculation.                            2231 1451
    Analised  experimental  data were corrected  to the new stan- 2231 1451
dards for monitor reactions cross sections and decay data.        2231 1451
    The final procedure of evaluation Ti-48(n,x)Sc-47  excitation 2231 1451
function from threshold to 20 MeV has been carried out within the 2231 1451
framework of generalized least squares method.  Rational function 2231 1451
was used  as model function [14].  Calculations  was performed by 2231 1451
means of Pade-2 code [15].                                        2231 1451
     U-235 thermal fission [16] and  Cf-252 spontanious  fission  2231 1451
neutron spectra [17] averaged cross-sections calculated from the  2231 1451
evaluated Ti-48(n,x)Sc-47 excitation function are the following:  2231 1451
                                                                  2231 1451
         --------------------------------------------             2231 1451
              TYPE OF SPECTRUM     I <CS>, mb (calc.)             2231 1451
         --------------------------I-----------------             2231 1451
         U-235  neutron  fission   I    1.6558E-3                 2231 1451
         CF-252 spontan. fission   I    4.2891E-3                 2231 1451
                                                                  2231 1451
MT=103 -(n,p) cross section                                       2231 1451
-------------------------------------                             2231 1451
    Excitation function for the  Ti-48(n,p)Sc-48  reaction in the 2231 1451
energy region from threshold to 20 MeV was evaluated by means of  2231 1451
of statistical analysis of experimental cross section data [1-2], 2231 1451
[4], [6-11], [18-41] and data from STAPRE [13] calculation.       2231 1451
    The energy dependence of cross-section from 4.7 MeV to thres- 2231 1451
hold was extrapolated  with  L=0  penetrability function  for the 2231 1451
the outgoing  p + Sc48  channal [42].                             2231 1451
    Analised microscopic experimental data  were  renormalized to 2231 1451
the new recommended standards for monitor reaction cross sections 2231 1451
and decay data. All experiments performed with Ti-natural samples 2231 1451
were  corrected for  the contribution of the  Ti-49(n,np+pn)Sc-48 2231 1451
and Ti-49(n,d)Sc-48 reactions. Total cross section for these reac-2231 1451
tions were taken from ref. [43]. Experimental data from ref. [24] 2231 1451
were used partially. It was taken into account only data measured 2231 1451
relative  Fe-56(n,p)Mn-56  monitor reaction.  Data of Firkin  [6] 2231 1451
obtained in the experiment with neutrons from  D(d,n)He3 reaction 2231 1451
were renormalized  to the results  of his measurements  with 14.1 2231 1451
MeV neutrons from T(d,n)He4 reaction. Cross section data of Smith 2231 1451
and Meadows [26]  measured in the energy range  5.964 - 9.952 MeV 2231 1451
with  using  D(d,n)He3  neutron source  were renormalized  to the 2231 1451
value  30.7 mb at 10 MeV [38].  The correction  factors  for  the 2231 1451
experimental data  [6] and [26]  were Fc=0.91853 and Fc=1.11111 , 2231 1451
respectively                                                      2231 1451
    Experimental cross section data [44-51] were rejected  due to 2231 1451
their discrepancy with the main bulk  of experimental data [1-2], 2231 1451
[4], [6-11], [18-41] and data from theoretical model calculatiod. 2231 1451
    Statistical analysis of input cross section data was carried  2231 1451
out by means of PADE-2 code [15].  Rational function was used as  2231 1451
the model function [14].                                          2231 1451
     U-235 thermal fission [16] and  Cf-252 spontanious  fission  2231 1451
neutron spectra [17] averaged cross-sections calculated from the  2231 1451
evaluated Ti-48(n,p)Sc-48 excitation function are the following:  2231 1451
                                                                  2231 1451
  2231 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2231 1451
  2231 1451
 U-235 neutron fission     0.30430        0.305 +- 0.020  [52]  2231 1451
                                          0.302 +- 0.010  [53]  2231 1451
                                          0.3007+- 0.0054 [54]  2231 1451
  2231 1451
 Cf-252 spont. fission     0.42629        0.4275+- 0.0078 [55]  2231 1451
                                          0.4247+- 0.0080 [56]  2231 1451
  2231 1451
                                                                  2231 1451
MF=33                                                             2231 1451
MT= 28 -(n,np+pn+d) cross section cov. matrix                     2231 1451
---------------------------------------------                     2231 1451
    Uncertainties in the evaluated  excitation function  for  the 2231 1451
reaction  Ti-48(n,x)Sc-47 are given in the form of relative cova- 2231 1451
riance matrix for the 15-neutron energy groups (LB=5). Covariance 2231 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2231 1451
recommended cross section data by means of PADE-2 code.           2231 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2231 1451
given in the 33-file are the following:                           2231 1451
                                                                  2231 1451
    3.25031E-07    3.96552E-07    7.16579E-07    1.55391E-06      2231 1451
    4.04176E-06    1.39260E-05    8.49003E-05    2.17726E-03      2231 1451
    8.18710E-03    1.48647E-02    2.37004E-02    3.37652E-02      2231 1451
    4.90298E-02    9.66009E-02    5.46438E-01                     2231 1451
                                                                  2231 1451
MT=103 -(n,p) cross section cov. matrix                           2231 1451
---------------------------------------------                     2231 1451
    Uncertainties in the evaluated  excitation function  for  the 2231 1451
reaction  Ti-48(n,p)Sc-48 are given in the form of relative cova- 2231 1451
riance matrix for the 26-neutron energy groups (LB=5). Covariance 2231 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2231 1451
recommended cross section data by means of PADE-2 code.           2231 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2231 1451
given in the 33-file are the following:                           2231 1451
                                                                  2231 1451
    5.39341E-09    5.73858E-09    5.82529E-09    6.42044E-09      2231 1451
    6.80480E-09    7.28300E-09    7.51108E-09    8.50118E-09      2231 1451
    9.53827E-09    1.09813E-08    1.24867E-08    1.39806E-08      2231 1451
    1.87772E-08    2.72244E-08    4.05820E-08    6.49966E-08      2231 1451
    1.23779E-07    3.87183E-07    6.43873E-06    1.17950E-03      2231 1451
    6.38433E-03    8.03793E-03    1.04937E-02    1.43313E-02      2231 1451
    2.41167E-02    3.79706E-02                                    2231 1451
                                                                  2231 1451
 References :                                                     2231 1451
 1. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2231 1451
 2. H.L.Pai  Canadian J. of Physics, v.44, p.2337, 1966           2231 1451
 3. S.M.Grimes et al.  Nucl. Sci. Eng., v.62, p.187, Feb. 1977    2231 1451
 4. S.M.Qaim, N.I.Molla  Nucl. Phys., v.A283, p.269, June 1977    2231 1451
 5. M.Viennot et al.  Progress Report MOH-5, p.10, 1982           2231 1451
 6. S.Firkin  Report AERE-M-3350, Harwell, September 1983         2231 1451
 7. I.Ribansky, S.Gmuca  J. Phys.G, v.9, p.1537, December 1983    2231 1451
 8. N.I.Molla et al.  Report INDC(BAN)-003, September 1986        2231 1451
 9. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2231 1451
10. W.V.Hecker et al.  Nucl. Inst. Meth., Sec.B, v.40/41, p.478,  2231 1451
    April 1989                                                    2231 1451
11. M.Viennot et al.  Nucl. Sci. Eng., v.108, p.289, July 1991    2231 1451
12. Y.Uno et al.  Report JAERI-M-93-046, p.247-256, 1993          2231 1451
13. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2231 1451
    Activation Cross Section and Related Quantities, Report       2231 1451
    IRK 76-01, Vienna, 1976                                       2231 1451
14. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2231 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2231 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2231 1451
15. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2231 1451
16. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2231 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2231 1451
17. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2231 1451
18. A.Poularikas, R.W.Fink  Physical Review, v.115, p.989, 1959   2231 1451
19. F.Gabbard, B.D.Kern  Physical Review, v.128, p.1276, 1962     2231 1451
20. M.Hillman  Nucl. Phys., v.37, p.78, 1962                      2231 1451
21. M.Bormann et al.  Nucl. Phys., v.63, p.438, March 1965        2231 1451
22. V.N.Levkovskij et al.  Yadernaja Fizika (Sov.), v.10, n.1,    2231 1451
    p.44, July 1969                                               2231 1451
23. S.K.Ghorai et al.  J. Nucl. Energy, v.25, p.319, August 1971  2231 1451
24. Y.Lukic, E.E.Carroll  Nucl. Sci. Eng., v.43, p.233, 1971      2231 1451
25. J.Prasad, D.C.Sarcar  Nuovo Cimento, v.A3, p.467, June 1971   2231 1451
26. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2231 1451
27. W.Mannhart, H.Vonach   Zeitsch. fur Physik, Sect.A, v.272,    2231 1451
    p.279, March 1975                                             2231 1451
28. K.Kayashima e.a.  Prog. Report NEANDC(J)-61U, p.94, Sep. 1979 2231 1451
29. M.Viennot e.a. Proc. of Int. Conf.on Nuclear Data for Science 2231 1451
    and Technology, Antwerpen, Belgium, 6-10 Sep. 1982, p.406     2231 1451
30. L.R.Greenwood  Report DOE-ER-0046-21, p.15, May 1985          2231 1451
31. R.Pepelnik et al. Progress Report NEANDC(E)-262U, (5), p.32,  2231 1451
    June 1985                                                     2231 1451
32. J.P.Gupta et al.  Pramana, v.24, p.637, 1985                  2231 1451
33. Hoang Dac Luc et al.  Progress Report INDC(VN)-5, Sep. 1986   2231 1451
34. K.Kobayashi, I.Kimura  Proc. of an Intern. Conf. on Nuclear   2231 1451
    Data for Science and Technology, 30 May - 3 June 1988, Mito,  2231 1451
    Japan, Saikon Publishing Co., LTD, pp.261-265, 1989           2231 1451
35. Lu Han-Lin et al. Report INDC(CPR)-16, August 1989            2231 1451
36. Y.Ikeda et al.  Report JAERI-M-91-032, p.281, March 1991      2231 1451
37. M.Viennot et al.  Nucl. Sci. Eng., v.108, p.289, July 1991    2231 1451
38. S.M.Qaim et al. Proc. of Int. Conference on Nuclear Data for  2231 1451
    Science and Technology, Julich, FRG, 13-17 May 1991. Springer 2231 1451
    Verlag, Berlin - Heidelberg, 1992, p.297-300                  2231 1451
39. I.Garlea et al.  J. Rev. Roum. Phys., v.37, n.1, p.19, 1992   2231 1451
40. Yuan Junqian et al.  High Energy Physics and Nuclear Physics  2231 1451
    (China), v.16, n.1, p.57, January 1992                        2231 1451
41. A.Grallert et al.  Progress Report,INDC(NDS)-286, p.131, 1993 2231 1451
42. S.A.Badikov, A.B.Pashchenko Voprosy Atomnoy Nauki i Tekhniki  2231 1451
    Ser. Jadernye Konstanty, 2(53), p.70, 1987                    2231 1451
43. K.Zolotarev RRDF-2002, MAT=2243, eval. February 2002          2231 1451
44. E.B.Paul, R.L.Clarke  Canadian J. Phys., v.31, p.267, 1953    2231 1451
45. D.L.Allan  Nucl. Phys., v.24, p.274, April 1961               2231 1451
46. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965          2231 1451
47. H.K.Vonach et al. Proc. of 2-nd Conference on Nuclear Cross-  2231 1451
    Sections and Technology, Washington D.C., 4-7 March 1968,     2231 1451
    v.2, p.885                                                    2231 1451
48. D.Crumpton  J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969    2231 1451
49. V.K.Tikku et al.  Proc. of Nucl. Phys. and Solid State Phys.  2231 1451
    Symp., Chandigarh, v.2, p.115, December 1972                  2231 1451
50. R.Spangler et al.  J. Trans. Amer. Nucl. Soc., v.22, p.818,   2231 1451
    November 1975                                                 2231 1451
51. K.T.Osman, F.I.Habbani  Report, INDC(SUD)-001, October 1996   2231 1451
52. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2231 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 2231 1451
53. W.Mannhart  Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry,  2231 1451
    Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985    2231 1451
54. W.Mannhart  Progress Report INDC(Ger)-045, pp.40-43, 1999     2231 1451
55. W.Mannhart  Handbook on Nuclear Activation Cross Sections ,   2231 1451
    IAEA Tech. Report Ser. No.273, p.413, 1987                    2231 1451
56. W.Mannhart  Validation of Differential Cross Sections with    2231 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2231 1451
    September 2002                                                2231 1451
***************************************************************** 2231 1451
 File 2 added to the pointwise file containing only the effective 2231 1451
 scattering radius with no resonance parameters given.            2231 1451
 Taken from ENDF/B-VI                                             2231 1451
***************************************************************** 2231 1451
                                                                  2231 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2231 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2231 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2231 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2231 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2231 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2231 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2231 1451
 ***************** Program FIXUP (Version 2002-1) ****************2231 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2231 1451
 Corrected Thresholds-----------------------------------------Yes 2231 1451
 Extended Cross Sections to 20 MeV----------------------------No  2231 1451
 Allow Cross Section Deletion---------------------------------No  2231 1451
 Allow Cross Section Reconstruction---------------------------No  2231 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2231 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2231 1451
 Deleted Duplicate Points-------------------------------------Yes 2231 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2231 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2231 1451
 Allow Creation of Missing Sections---------------------------No  2231 1451
 Allow Insertion of Energy Points-----------------------------No  2231 1451
 Create Uniform Energy Grid-----------------------------------No  2231 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2231 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2231 1451
 Unshielded Group Averages Using  640 Groups                      2231 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2231 1451
                                1        451        252          12231 1451
                                3        103         60          12231 1451
                                8          5          2          12231 1451
                               10          5         39          12231 1451
                               33        103         71          12231 1451
                               40          5         30          12231 1451
                                                                  2231 1  0
                                                                  2231 0  0
 2.20480E+4 4.75360E+1          0          0          0          02231 3103
-3.21180E+6-3.21180E+6          0          0          1        1692231 3103
        169          1                                            2231 3103
 3200000.00 4.44705E-9 3300000.00 1.46148E-7 3400000.00 3.52717E-72231 3103
 3500000.00 5.59287E-7 3600000.00 7.65856E-7 3700000.00 9.72426E-72231 3103
 3800000.00 1.17900E-6 3900000.00 1.75487E-6 4000000.00 2.60258E-62231 3103
 4100000.00 3.40719E-6 4200000.00 4.52734E-6 4300000.00 6.44311E-62231 3103
 4400000.00 9.78106E-6 4500000.00 1.53460E-5 4600000.00 2.41598E-52231 3103
 4700000.00 3.75024E-5 4800000.00 5.69529E-5 4900000.00 8.44353E-52231 3103
 5000000.00 .000122266 5100000.00 .000173193 5200000.00 .0002404202231 3103
 5300000.00 .000327615 5400000.00 .000438894 5500000.00 .0005787552231 3103
 5600000.00 .000751970 5700000.00 .000963413 5800000.00 .0012178252231 3103
 5900000.00 .001519525 6000000.00 .001872085 6100000.00 .0022779952231 3103
 6200000.00 .002738340 6300000.00 .003252590 6400000.00 .0038184952231 3103
 6500000.00 .004432120 6600000.00 .005088075 6700000.00 .0057798702231 3103
 6800000.00 .006500350 6900000.00 .007244910 7000000.00 .0080010502231 3103
 7100000.00 .008760743 7200000.00 .009523988 7300000.00 .0102872332231 3103
 7400000.00 .011050478 7500000.00 .011802500 7600000.00 .0125433002231 3103
 7700000.00 .013284100 7800000.00 .014013106 7900000.00 .0147303172231 3103
 8000000.00 .015447528 8100000.00 .016164739 8200000.00 .0168819502231 3103
 8300000.00 .017599161 8400000.00 .018316372 8500000.00 .0190335832231 3103
 8600000.00 .019750794 8700000.00 .020481567 8800000.00 .0212259002231 3103
 8900000.00 .021970233 9000000.00 .022732733 9100000.00 .0235134002231 3103
 9200000.00 .024294067 9300000.00 .025097267 9400000.00 .0259230002231 3103
 9500000.00 .026748733 9600000.00 .027599500 9700000.00 .0284753002231 3103
 9800000.00 .029351100 9900000.00 .030252433 10000000.0 .0311793002231 3103
 10100000.0 .032106167 10200000.0 .033060613 10300000.0 .0340426382231 3103
 10400000.0 .035024663 10500000.0 .036006688 10600000.0 .0370165502231 3103
 10700000.0 .038054250 10800000.0 .039091950 10900000.0 .0401296502231 3103
 11000000.0 .041167350 11100000.0 .042205050 11200000.0 .0432453922231 3103
 11300000.0 .044288375 11400000.0 .045331358 11500000.0 .0463743422231 3103
 11600000.0 .047417325 11700000.0 .048460308 11800000.0 .0494691382231 3103
 11900000.0 .050443813 12000000.0 .051418488 12100000.0 .0523931632231 3103
 12200000.0 .053319817 12300000.0 .054198450 12400000.0 .0550770832231 3103
 12500000.0 .055898950 12600000.0 .056664050 12700000.0 .0574291502231 3103
 12800000.0 .058125350 12900000.0 .058752650 13000000.0 .0593799502231 3103
 13100000.0 .059929217 13200000.0 .060400450 13300000.0 .0608716832231 3103
 13400000.0 .061259683 13500000.0 .061564450 13600000.0 .0618692172231 3103
 13700000.0 .062089767 13800000.0 .062226100 13900000.0 .0623624332231 3103
 14000000.0 .062443900 14100000.0 .062418567 14200000.0 .0623413002231 3103
 14300000.0 .062264033 14400000.0 .062115133 14500000.0 .0618946002231 3103
 14600000.0 .061674067 14700000.0 .061391650 14800000.0 .0610473502231 3103
 14900000.0 .060703050 15000000.0 .060300825 15100000.0 .0598406752231 3103
 15200000.0 .059380525 15300000.0 .058920375 15400000.0 .0584077502231 3103
 15500000.0 .057842650 15600000.0 .057277550 15700000.0 .0567124502231 3103
 15800000.0 .056147350 15900000.0 .055545431 16000000.0 .0549066942231 3103
 16100000.0 .054267956 16200000.0 .053629219 16300000.0 .0529904812231 3103
 16400000.0 .052351744 16500000.0 .051713006 16600000.0 .0510742692231 3103
 16700000.0 .050433667 16800000.0 .049791200 16900000.0 .0491487332231 3103
 17000000.0 .048506267 17100000.0 .047863800 17200000.0 .0472213332231 3103
 17300000.0 .046578867 17400000.0 .045936400 17500000.0 .0452939332231 3103
 17600000.0 .044674017 17700000.0 .044076650 17800000.0 .0434792832231 3103
 17900000.0 .042881917 18000000.0 .042284550 18100000.0 .0416871832231 3103
 18200000.0 .041113110 18300000.0 .040562330 18400000.0 .0400115502231 3103
 18500000.0 .039460770 18600000.0 .038909990 18700000.0 .0383818002231 3103
 18800000.0 .037876200 18900000.0 .037370600 19000000.0 .0368650002231 3103
 19100000.0 .036359400 19200000.0 .035876120 19300000.0 .0354151602231 3103
 19400000.0 .034954200 19500000.0 .034493240 19600000.0 .0340322802231 3103
 19700000.0 .033588467 19800000.0 .033161800 19900000.0 .0327351332231 3103
 20000000.0 0.0                                                   2231 3103
                                                                  2231 3  0
                                                                  2231 0  0
 2.20480E+4 4.75360E+1          0          0          1          02231 3231
-9.22000E+6-9.22000E+6      21047          0          1        1072231 3231
        107          1                                            2231 3231
 9400000.00 2.09726E-7 9500000.00 7.70813E-7 9600000.00 1.33742E-62231 3231
 9700000.00 1.90403E-6 9800000.00 2.47063E-6 9900000.00 3.03724E-62231 3231
 10000000.0 3.60385E-6 10100000.0 4.17046E-6 10200000.0 5.32343E-62231 3231
 10300000.0 6.96840E-6 10400000.0 8.47552E-6 10500000.0 9.97182E-62231 3231
 10600000.0 1.16052E-5 10700000.0 1.35478E-5 10800000.0 1.59992E-52231 3231
 10900000.0 1.91905E-5 11000000.0 2.33890E-5 11100000.0 2.89037E-52231 3231
 11200000.0 3.60910E-5 11300000.0 4.53630E-5 11400000.0 5.71957E-52231 3231
 11500000.0 7.21354E-5 11600000.0 9.08205E-5 11700000.0 .0001139832231 3231
 11800000.0 .000142472 11900000.0 .000177270 12000000.0 .0002195112231 3231
 12100000.0 .000270512 12200000.0 .000331792 12300000.0 .0004051062231 3231
 12400000.0 .000492478 12500000.0 .000596228 12600000.0 .0007190072231 3231
 12700000.0 .000863820 12800000.0 .001034037 12900000.0 .0012333902231 3231
 13000000.0 .001465935 13100000.0 .001735970 13200000.0 .0020478952231 3231
 13300000.0 .002406025 13400000.0 .002814295 13500000.0 .0032759202231 3231
 13600000.0 .003793045 13700000.0 .004366360 13800000.0 .0049948652231 3231
 13900000.0 .005675790 14000000.0 .006404750 14100000.0 .0071762152231 3231
 14200000.0 .007984180 14300000.0 .008822960 14400000.0 .0096879952231 3231
 14500000.0 .010576400 14600000.0 .011487400 14700000.0 .0124223002231 3231
 14800000.0 .013384250 14900000.0 .014377950 15000000.0 .0154092002231 3231
 15100000.0 .016484400 15200000.0 .017610150 15300000.0 .0187930502231 3231
 15400000.0 .020039300 15500000.0 .021354500 15600000.0 .0227436502231 3231
 15700000.0 .024211050 15800000.0 .025760250 15900000.0 .0273939502231 3231
 16000000.0 .029113900 16100000.0 .030921100 16200000.0 .0328156502231 3231
 16300000.0 .034796700 16400000.0 .036862600 16500000.0 .0390300752231 3231
 16600000.0 .041257225 16700000.0 .043556600 16800000.0 .0459282002231 3231
 16900000.0 .048360800 17000000.0 .050854400 17100000.0 .0534042502231 3231
 17200000.0 .056010350 17300000.0 .058616450 17400000.0 .0612563002231 3231
 17500000.0 .063929900 17600000.0 .066603500 17700000.0 .0692771002231 3231
 17800000.0 .071950700 17900000.0 .074624300 18000000.0 .0772979002231 3231
 18100000.0 .079971500 18200000.0 .082581663 18300000.0 .0851283882231 3231
 18400000.0 .087675113 18500000.0 .090221838 18600000.0 .0926924672231 3231
 18700000.0 .095087000 18800000.0 .097481533 18900000.0 .0997890752231 3231
 19000000.0 .102009625 19100000.0 .104230175 19200000.0 .1064507252231 3231
 19300000.0 .108572375 19400000.0 .110595125 19500000.0 .1126178752231 3231
 19600000.0 .114640625 19700000.0 .116584333 19800000.0 .1184490002231 3231
 19900000.0 .120313667 20000000.0 0.0                             2231 3231
                                                                  2231 3  0
                                                                  2231 0  0
 2.20480E+4 4.75360E+1          0          0          0          1223133103
 0.000000+0 0.000000+0          0        103          0          1223133103
 0.000000+0 0.000000+0          1          5        406         28223133103
 1.000000-5 3.200000+6 6.000000+6 6.500000+6 7.000000+6 7.500000+6223133103
 8.000000+6 8.500000+6 9.000000+6 9.500000+6 1.000000+7 1.050000+7223133103
 1.100000+7 1.150000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7223133103
 1.400000+7 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7223133103
 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0223133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133103
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 0.000000+0 1.244150-2 7.346840-3 3.817820-3 2.023600-3 1.670350-3223133103
 1.914370-3 2.193410-3 2.302740-3 2.218330-3 1.979000-3 1.639120-3223133103
 1.253720-3 8.737600-4 5.426420-4 2.916640-4 1.355170-4 7.050570-5223133103
 7.734160-5 1.278750-4 1.929260-4 2.482490-4 2.773030-4 2.712070-4223133103
 1.881820-4-3.611490-5-3.653190-4 7.757040-3 6.098810-3 4.154940-3223133103
 2.757920-3 2.029510-3 1.740490-3 1.661580-3 1.649290-3 1.626740-3223133103
 1.556430-3 1.424190-3 1.231820-3 9.938690-4 7.346730-4 4.837150-4223133103
 2.692840-4 1.123370-4 2.294990-5 4.162250-7 3.615880-5 1.176120-4223133103
 2.315270-4 4.363480-4 7.328280-4 1.020310-3 6.503610-3 5.489650-3223133103
 3.976130-3 2.699250-3 1.867760-3 1.414650-3 1.213220-3 1.152120-3223133103
 1.147310-3 1.138840-3 1.087580-3 9.745160-4 8.006670-4 5.850100-4223133103
 3.585880-4 1.556880-4 5.163970-6-7.495940-5-7.848490-5-8.736890-6223133103
 1.250630-4 4.204940-4 9.271180-4 1.503660-3 5.552780-3 4.728610-3223133103
 3.677500-3 2.763700-3 2.078490-3 1.597730-3 1.269150-3 1.042580-3223133103
 8.773390-4 7.431170-4 6.198910-4 4.976790-4 3.756340-4 2.597860-4223133103
 1.595420-4 8.401720-5 3.941690-5 2.808660-5 4.893890-5 9.857880-5223133103
 2.175880-4 4.321980-4 6.863320-4 4.790110-3 4.371670-3 3.743620-3223133103
 3.070070-3 2.427020-3 1.846360-3 1.342280-3 9.227950-4 5.930010-4223133103
 3.544680-4 2.030560-4 1.273520-4 1.091970-4 1.266340-4 1.581090-4223133103
 1.860910-4 1.988820-4 1.905280-4 1.595630-4 7.340970-5-9.489670-5223133103
-3.035830-4 4.559690-3 4.328510-3 3.834850-3 3.204670-3 2.525770-3223133103
 1.860850-3 1.258050-3 7.551470-4 3.784270-4 1.388780-4 2.896740-5223133103
 2.339780-5 8.492700-5 1.733170-4 2.536830-4 3.013780-4 3.028680-4223133103
 2.538460-4 8.842740-5-2.600790-4-7.124240-4 4.443610-3 4.193460-3223133103
 3.704470-3 3.080450-3 2.402690-3 1.737410-3 1.139940-3 6.539380-4223133103
 3.068680-4 1.051810-4 3.310610-5 5.730040-5 1.360700-4 2.293020-4223133103
 3.054380-4 3.441220-4 3.353790-4 2.263220-4-6.184520-5-4.772290-4223133103
 4.184480-3 3.901610-3 3.433910-3 2.855910-3 2.231760-3 1.619200-3223133103
 1.069660-3 6.241580-4 3.072850-4 1.229730-4 5.521990-5 7.418210-5223133103
 1.449290-4 2.352440-4 3.202190-4 3.834970-4 4.158710-4 3.503750-4223133103
 1.706540-4 3.841480-3 3.579830-3 3.168340-3 2.655440-3 2.090130-3223133103
 1.523280-3 1.004800-3 5.770040-4 2.671880-4 8.324800-5 1.472830-5223133103
 3.861770-5 1.268670-4 2.526630-4 3.940650-4 5.981290-4 8.250830-4223133103
 9.753670-4 3.533590-3 3.316190-3 2.953300-3 2.478450-3 1.935820-3223133103
 1.378990-3 8.644510-4 4.414440-4 1.425100-4-2.100990-5-5.701670-5223133103
 1.375670-5 1.651580-4 3.714370-4 7.336580-4 1.246250-3 1.732250-3223133103
 3.289420-3 3.089100-3 2.730660-3 2.247570-3 1.691990-3 1.128910-3223133103
 6.243380-4 2.317070-4-1.757530-5-1.160950-4-7.567700-5 8.046170-5223133103
 3.251350-4 8.001730-4 1.534130-3 2.293640-3 3.042310-3 2.810680-3223133103
 2.415160-3 1.901810-3 1.337270-3 7.964780-4 3.467360-4 3.457190-5223133103
-1.195210-4-1.182760-4 1.987030-5 2.689400-4 7.909510-4 1.642760-3223133103
 2.565630-3 2.702120-3 2.413220-3 1.980270-3 1.464370-3 9.402680-4223133103
 4.799640-4 1.379780-4-5.622000-5-9.732580-5 2.238020-6 2.202720-4223133103
 7.149690-4 1.561400-3 2.510920-3 2.239130-3 1.916880-3 1.495720-3223133103
 1.041080-3 6.199920-4 2.868080-4 7.473390-5-4.977680-6 4.173930-5223133103
 1.985470-4 5.968160-4 1.317270-3 2.154950-3 1.722740-3 1.430710-3223133103
 1.089500-3 7.526010-4 4.664730-4 2.625740-4 1.553380-4 1.449180-4223133103
 2.220540-4 4.739210-4 9.741100-4 1.586440-3 1.284100-3 1.083510-3223133103
 8.630930-4 6.555010-4 4.857210-4 3.683870-4 3.084850-4 3.039140-4223133103
 3.897200-4 6.216970-4 9.436640-4 1.029260-3 9.394040-4 8.298590-4223133103
 7.159590-4 6.098560-4 5.194640-4 4.487370-4 3.839040-4 3.563470-4223133103
 3.847270-4 9.747120-4 9.683430-4 9.248700-4 8.519660-4 7.582400-4223133103
 6.516400-4 4.833150-4 2.585650-4 5.063660-5 1.056860-3 1.089950-3223133103
 1.069780-3 1.003410-3 9.001680-4 6.972650-4 3.769060-4 3.602680-5223133103
 1.198100-3 1.245680-3 1.236370-3 1.178080-3 1.018650-3 7.236530-4223133103
 3.779790-4 1.370840-3 1.444020-3 1.469220-3 1.429200-3 1.281260-3223133103
 1.063020-3 1.619290-3 1.760150-3 1.905790-3 2.014100-3 2.044310-3223133103
 2.041250-3 2.425520-3 2.879870-3 3.259940-3 3.234920-3 4.327790-3223133103
 5.371820-3 6.404900-3 8.492620-3 1.170350-2                      223133103
                                                                  223133  0
                                                                  2231 0  0
 2.20480E+4 4.75360E+1          0          0          0          1223133231
 0.0000E+00 0.0000E+00          0        231          0          1223133231
 0.0000E+00 0.0000E+00          1          5        153         17223133231
 1.000000-5 9.400000+6 1.300000+7 1.350000+7 1.400000+7 1.450000+7223133231
 1.500000+7 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7223133231
 1.800000+7 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0223133231
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133231
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133231
 0.000000+0 0.000000+0 0.000000+0 4.707110-1 1.759780-1 1.697230-2223133231
-8.146730-3 1.038100-2-5.224470-3-2.034860-2-1.446150-2-9.242090-4223133231
 6.358550-3 3.918520-3-4.555240-3-1.358030-2-1.920200-2-1.987550-2223133231
 1.052700-1 2.855580-2-1.423330-3 4.824040-3 5.676170-3-7.073830-4223133231
-3.600630-3-2.063640-3 5.308030-4 1.841570-3 1.461860-3 1.058170-4223133231
-1.304640-3-2.135350-3 2.285560-2 7.762240-3 1.086850-3 2.996610-3223133231
 4.013350-3 2.410030-3 5.183160-4-1.889870-4 3.334230-4 1.449850-3223133231
 2.490700-3 3.052080-3 3.021780-3 1.008370-2 4.078560-3-7.285120-4223133231
-1.046370-3 3.213810-4 1.205820-3 1.175750-3 5.878370-4-9.490980-5223133231
-5.712350-4-7.367790-4-6.212280-4 8.396540-3 7.438880-3 3.738740-3223133231
 1.061670-3 2.405110-4 6.122910-4 1.371180-3 2.005320-3 2.316480-3223133231
 2.311780-3 2.087370-3 1.621120-2 1.515040-2 8.632830-3 3.121490-3223133231
 1.138420-3 2.178660-3 4.619640-3 6.965430-3 8.348450-3 8.521370-3223133231
 1.895660-2 1.478520-2 8.727250-3 4.770970-3 3.793980-3 4.917860-3223133231
 6.814000-3 8.444710-3 9.280860-3 1.554920-2 1.270370-2 8.999380-3223133231
 6.138550-3 4.681970-3 4.463460-3 5.026080-3 5.915220-3 1.336970-2223133231
 1.145390-2 8.468040-3 5.709430-3 3.904690-3 3.230420-3 3.514540-3223133231
 1.161810-2 1.015980-2 8.083180-3 6.198510-3 4.939490-3 4.404970-3223133231
 1.095260-2 1.089590-2 1.022320-2 9.212560-3 8.099470-3 1.315310-2223133231
 1.425810-2 1.409830-2 1.289180-2 1.700900-2 1.794020-2 1.714270-2223133231
 1.989720-2 1.988000-2 2.083550-2                                 223133231
                                                                  223133  0
                                                                  2231 0  0
                                                                     0 0  0
 2.20490E+4 4.85274E+1          0          0         34         202234 1451
 0.0        0.0                 0          0          0          62234 1451
 1.00000E+0 2.00000E+7          0          0         10       20022234 1451
 3.00000E+2 0.0                 1          0        116          42234 1451
 22-Ti- 49 FEI        EVAL-Feb02 K.I.Zolotarev                    2234 1451
                      DIST-Feb2004                                2234 1451
----IRDF-2002         MATERIAL 2234                               2234 1451
-----INCIDENT NEUTRON DATA                                        2234 1451
------ENDF-6 FORMAT                                               2234 1451
***************************************************************** 2234 1451
**********************  SPECIAL PURPOSE FILE   *****************  2234 1451
DOSIMETRY                                                         2234 1451
                                                                  2234 1451
 For this special purpose library it was decided the reaction     2234 1451
 MF=3 MT=28+104 would use MF=10, MT=5 to present this data.       2234 1451
 This was done after processing through the codes. The            2234 1451
 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5    2234 1451
***************************************************************** 2234 1451
 22-TI-49  FEI        EVAL-Feb02 K.I.Zolotarev                    2234 1451
                      DIST-Feb02                                  2234 1451
----BROND-2           MATERIAL 2234                               2234 1451
-----INCIDENT NEUTRON DATA                                        2234 1451
------ENDF-6 FORMAT                                               2234 1451
------Russian Reactor Dosimetry File RRDF-2002                    2234 1451
***************************************************************** 2234 1451
Author  of evaluation: K.I.Zolotarev                              2234 1451
***************************************************************** 2234 1451
MF= 3                                                             2234 1451
MT= 28 -(n,np+pn+d) cross section data                            2234 1451
--------------------------------------                            2234 1451
    In this section is given the sum of cross section of the reac-2234 1451
tions Ti-49(n,np)Sc-48 , Ti-49(n,pn)Sc-48 and Ti-49(n,d)Sc-48.    2234 1451
    Excitation function for the  Ti-49(n,x)Sc-48  reaction in the 2234 1451
energy region from threshold to 20 MeV was evaluated  by means of 2234 1451
statistical analysis of experimental cross section data [1-5] and 2234 1451
data from STAPRE [6] calculation.                                 2234 1451
    Analysed experimental data were renormalized to the new stan- 2234 1451
dards for monitor reactions cross sections  and decay data.  Data 2234 1451
of Pai [2]  measured in the energy  region 16-19.5 MeV with using 2234 1451
Van de Graaff accelerator  were  renormalized  to the results  of 2234 1451
the theoretical model calculation.                                2234 1451
    The final procedure of evaluation Ti-49(n,x)Sc-48  excitation 2234 1451
function from threshold to 20 MeV has been carried out within the 2234 1451
framework of generalized least squares method.  Rational function 2234 1451
was used  as  model function [7].  Calculations  was performed by 2234 1451
means of  Pade-2 code [8].                                        2234 1451
    U-235 thermal fission [9]  and   Cf-252 spontaneous  fission  2234 1451
neutron spectra [10] averaged cross-sections calculated from the  2234 1451
evaluated Ti-49(n,x)Sc-48  excitation function are the following: 2234 1451
                                                                  2234 1451
         --------------------------------------------             2234 1451
              TYPE OF SPECTRUM     I <CS>, mb (calc.)             2234 1451
         --------------------------I-----------------             2234 1451
         U-235  neutron  fission   I    1.0041E-3                 2234 1451
         CF-252 spontan. fission   I    2.6070E-3                 2234 1451
                                                                  2234 1451
MF=33                                                             2234 1451
MT= 28 -(n,np+pn+d) cross section cov. matrix                     2234 1451
---------------------------------------------                     2234 1451
    Uncertainties in the evaluated  excitation function  for  the 2234 1451
reaction  Ti-49(n,x)Sc-48 are given in the form of relative cova- 2234 1451
riance matrix for the 14-neutron energy groups (LB=5). Covariance 2234 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2234 1451
recommended cross section data by means of PADE-2 code.           2234 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2234 1451
given in the 33-file are the following:                           2234 1451
                                                                  2234 1451
    2.34704E-08    3.44912E-08    5.30971E-08    9.47934E-08      2234 1451
    1.99158E-07    5.01449E-07    1.49592E-06    5.55617E-06      2234 1451
    1.56887E-04    2.23251E-03    8.77261E-03    1.50447E-02      2234 1451
    3.12585E-02    8.36148E-02                                    2234 1451
                                                                  2234 1451
 References :                                                     2234 1451
 1. W.G.Cross, H.L.Pai  Progress Report, EANDC(CAN)-16, p.1,      2234 1451
    January 1963                                                  2234 1451
 2. H.L.Pai  Canadian J. of Physics, v.44, p.2337, 1966           2234 1451
 3. S.M.Qaim   Nucl. Phys., v.A382, n.2, p.255, July 1982         2234 1451
 4. I.Ribansky, S.Gmuca  J. Phys.G, v.9, p.1537, December 1983    2234 1451
 5. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2234 1451
 6. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2234 1451
    Activation Cross Section and Related Quantities, Report       2234 1451
    IRK 76-01, Vienna, 1976                                       2234 1451
 7. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2234 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2234 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2234 1451
 8. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2234 1451
 9. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2234 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2234 1451
10. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2234 1451
***************************************************************** 2234 1451
 File 2 added to the pointwise file containing only the effective 2234 1451
 scattering radius with no resonance parameters given.            2234 1451
 Taken from JENDL-3.2                                             2234 1451
***************************************************************** 2234 1451
                                                                  2234 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2234 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2234 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2234 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2234 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2234 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2234 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2234 1451
 ***************** Program FIXUP (Version 2002-1) ****************2234 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2234 1451
 Corrected Thresholds-----------------------------------------Yes 2234 1451
 Extended Cross Sections to 20 MeV----------------------------No  2234 1451
 Allow Cross Section Deletion---------------------------------No  2234 1451
 Allow Cross Section Reconstruction---------------------------No  2234 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2234 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2234 1451
 Deleted Duplicate Points-------------------------------------Yes 2234 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2234 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2234 1451
 Allow Creation of Missing Sections---------------------------No  2234 1451
 Allow Insertion of Energy Points-----------------------------No  2234 1451
 Create Uniform Energy Grid-----------------------------------No  2234 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2234 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2234 1451
 Unshielded Group Averages Using  640 Groups                      2234 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2234 1451
                                1        451        124          12234 1451
                                8          5          2          12234 1451
                               10          5         39          12234 1451
                               40          5         27          12234 1451
                                                                  2234 1  0
                                                                  2234 0  0
 2.20490E+4 4.85274E+1          0          0          1          02234 3231
-9.12956E+6-9.12956E+6      21048          0          1        1082234 3231
        108          1                                            2234 3231
 9300000.00 1.10601E-9 9400000.00 4.31560E-9 9500000.00 7.57585E-92234 3231
 9600000.00 1.08361E-8 9700000.00 1.40964E-8 9800000.00 1.73566E-82234 3231
 9900000.00 2.06169E-8 10000000.0 4.97413E-8 10100000.0 1.04730E-72234 3231
 10200000.0 1.59719E-7 10300000.0 2.14707E-7 10400000.0 2.69696E-72234 3231
 10500000.0 7.67187E-7 10600000.0 1.70718E-6 10700000.0 2.64718E-62234 3231
 10800000.0 3.58717E-6 10900000.0 4.52716E-6 11000000.0 6.74413E-62234 3231
 11100000.0 1.09550E-5 11200000.0 1.67516E-5 11300000.0 2.44527E-52234 3231
 11400000.0 3.44093E-5 11500000.0 4.70069E-5 11600000.0 6.26681E-52234 3231
 11700000.0 8.18556E-5 11800000.0 .000105075 11900000.0 .0001328762234 3231
 12000000.0 .000165858 12100000.0 .000204674 12200000.0 .0002500282234 3231
 12300000.0 .000302684 12400000.0 .000363468 12500000.0 .0004332672234 3231
 12600000.0 .000513036 12700000.0 .000603801 12800000.0 .0007066562234 3231
 12900000.0 .000822770 13000000.0 .000953387 13100000.0 .0010998252234 3231
 13200000.0 .001263485 13300000.0 .001445825 13400000.0 .0016483902234 3231
 13500000.0 .001872795 13600000.0 .002120700 13700000.0 .0023938452234 3231
 13800000.0 .002694020 13900000.0 .003023020 14000000.0 .0033827102234 3231
 14100000.0 .003774980 14200000.0 .004201675 14300000.0 .0046646352234 3231
 14400000.0 .005165665 14500000.0 .005706495 14600000.0 .0062887702234 3231
 14700000.0 .006914025 14800000.0 .007583640 14900000.0 .0082988252234 3231
 15000000.0 .009060585 15100000.0 .009869710 15200000.0 .0107266502234 3231
 15300000.0 .011631650 15400000.0 .012584700 15500000.0 .0135853002234 3231
 15600000.0 .014632650 15700000.0 .015725600 15800000.0 .0168627502234 3231
 15900000.0 .018042250 16000000.0 .019270825 16100000.0 .0205278752234 3231
 16200000.0 .021818075 16300000.0 .023141425 16400000.0 .0244962002234 3231
 16500000.0 .025882400 16600000.0 .027268600 16700000.0 .0286769782234 3231
 16800000.0 .030107533 16900000.0 .031538089 17000000.0 .0329686442234 3231
 17100000.0 .034399200 17200000.0 .035829756 17300000.0 .0372603112234 3231
 17400000.0 .038690867 17500000.0 .040121422 17600000.0 .0415165752234 3231
 17700000.0 .042876325 17800000.0 .044236075 17900000.0 .0455958252234 3231
 18000000.0 .046914233 18100000.0 .048191300 18200000.0 .0494683672234 3231
 18300000.0 .050704483 18400000.0 .051899650 18500000.0 .0530948172234 3231
 18600000.0 .054247567 18700000.0 .055357900 18800000.0 .0564682332234 3231
 18900000.0 .057530588 19000000.0 .058544963 19100000.0 .0595593382234 3231
 19200000.0 .060573713 19300000.0 .061538125 19400000.0 .0624525752234 3231
 19500000.0 .063367025 19600000.0 .064281475 19700000.0 .0651578332234 3231
 19800000.0 .065996100 19900000.0 .066834367 20000000.0 0.0       2234 3231
                                                                  2234 3  0
                                                                  2234 0  0
 2.20490E+4 4.85274E+1          0          0          0          1223433231
 0.0000E+00 0.0000E+00          0        231          0          1223433231
 0.0000E+00 0.0000E+00          1          5        136         16223433231
 1.000000-5 9.300000+6 1.350000+7 1.400000+7 1.450000+7 1.500000+7223433231
 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7223433231
 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0223433231
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223433231
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223433231
 0.000000+0 2.340470-2 1.573760-2 1.165090-2 8.156660-3 5.541340-3223433231
 3.969430-3 3.457650-3 3.824990-3 4.757050-3 5.897230-3 6.966960-3223433231
 7.813930-3 8.420550-3 8.858660-3 1.315090-2 1.087230-2 8.474760-3223433231
 6.164580-3 4.207140-3 2.837340-3 2.197190-3 2.273200-3 2.909620-3223433231
 3.862550-3 4.881200-3 5.757010-3 6.348110-3 9.682370-3 8.120140-3223433231
 6.337310-3 4.558730-3 3.043450-3 2.025410-3 1.624510-3 1.820480-3223433231
 2.467110-3 3.358720-3 4.283100-3 5.063930-3 7.358580-3 6.232520-3223433231
 4.886070-3 3.524410-3 2.383810-3 1.652690-3 1.421190-3 1.664730-3223433231
 2.277620-3 3.115020-3 4.034000-3 5.813390-3 5.091770-3 4.162120-3223433231
 3.178860-3 2.326160-3 1.761320-3 1.580970-3 1.811250-3 2.427040-3223433231
 3.374180-3 5.065940-3 4.757790-3 4.186930-3 3.465820-3 2.759470-3223433231
 2.250120-3 2.087010-3 2.377510-3 3.180650-3 5.131690-3 5.149110-3223433231
 4.804660-3 4.208600-3 3.567080-3 3.106570-3 3.039490-3 3.527780-3223433231
 5.806850-3 6.020090-3 5.794240-3 5.284890-3 4.721960-3 4.359860-3223433231
 4.419540-3 6.857770-3 7.193930-3 7.072420-3 6.651570-3 6.153380-3223433231
 5.798070-3 8.165580-3 8.611450-3 8.566600-3 8.159510-3 7.554320-3223433231
 9.678720-3 1.017690-2 1.011670-2 9.575680-3 1.129210-2 1.182470-2223433231
 1.177030-2 1.316420-2 1.408870-2 1.651520-2                      223433231
                                                                  223433  0
                                                                  2234 0  0
                                                                     0 0  0
 2.30510E+4 5.05063E+1          0          0         34         102328 1451
 0.0        0.0                 0          0          0          62328 1451
 1.00000E+0 2.00000E+7          0          0         10       20022328 1451
 3.00000E+2 0.0                 1          0        185          32328 1451
 23-V - 51 FEI        EVAL-Dec01 K.I.Zolotarev                    2328 1451
                      DIST-Feb2004                                2328 1451
----IRDF-2002         MATERIAL 2328                               2328 1451
-----INCIDENT NEUTRON DATA                                        2328 1451
------ENDF-6 FORMAT                                               2328 1451
***************************************************************** 2328 1451
 23-V-51   FEI        EVAL-Dec01 K.I.Zolotarev                    2328 1451
                      DIST-Jan02                       20020118   2328 1451
----BROND-2           MATERIAL 2328                               2328 1451
-----INCIDENT NEUTRON DATA                                        2328 1451
------ENDF-6 FORMAT                                               2328 1451
------Russian Reactor Dosimetry File RRDF-2002                    2328 1451
***************************************************************** 2328 1451
Author  of evaluation: K.I.Zolotarev                              2328 1451
***************************************************************** 2328 1451
MF=3                                                              2328 1451
MT=107 -(n,a) cross section                                       2328 1451
-------------------------------------                             2328 1451
    Excitation function for the  V-51(n,a)Sc-48  reaction in the  2328 1451
energy region from threshold to 20 MeV was evaluated by means of  2328 1451
statistical analysis of  experimental cross section  data [1-21]  2328 1451
and data from GNASH calculation [45].                             2328 1451
    All analysed microscopic experimental data [1-38] and integ-  2328 1451
ral experimental data [39-44] were renormalized to the new recom- 2328 1451
mended standards for monitor reactions cross sections and decay   2328 1451
data.                                                             2328 1451
    Experimental cross section data from ref. [22-38] were rejec- 2328 1451
ted due to their big discrepancy with the main bulk of experimen- 2328 1451
al data [1-21], data from theoretical model calculation and data  2328 1451
from (n,a) cross section systematics.                             2328 1451
    Statistical analysis of input cross section data was carried  2328 1451
out by means of PADE-2 code [46].  Rational function was used as  2328 1451
the model function [47].                                          2328 1451
    The evaluated   V-51(n,a)Sc-48   excitation function averaged 2328 1451
on  U-235  neutron fission spectrum [48]  and  Cf-252 spontaneous 2328 1451
fission neutron spectrum [49] gives the next values :             2328 1451
  2328 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2328 1451
  2328 1451
 U-235 neutron fission    0.024414       0.0235 +-0.0015  [40]  2328 1451
                                         0.0215 +-0.0013  [41]  2328 1451
                                         0.02438+-0.00056 [42]  2328 1451
  2328 1451
 CF-252 spont. fission    0.038514       0.03904+-0.00086 [43]  2328 1451
  2328 1451
                                                                  2328 1451
MF=33                                                             2328 1451
MT=107 -(n,a) cross section cov. matrix                           2328 1451
---------------------------------------                           2328 1451
    Uncertainties in the evaluated  excitation function  for the  2328 1451
reaction V-51(n,a)Sc-48 are given in the form of relative covari- 2328 1451
ance matrix for the  26-neutron energy groups (LB=5).  Covariance 2328 1451
matrix of uncertainties was calculated simultaneously with recom- 2328 1451
mended cross section data by means of PADE-2 code.                2328 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2328 1451
given in the 33-file are the following:                           2328 1451
                                                                  2328 1451
    4.46067E-09    4.97872E-09    5.09809E-09    5.50951E-09      2328 1451
    6.22582E-09    6.26808E-09    6.80068E-09    7.58613E-09      2328 1451
    9.63723E-09    1.29149E-08    1.82432E-08    2.72742E-08      2328 1451
    4.31460E-08    7.37989E-08    1.54144E-07    3.75108E-07      2328 1451
    2.13791E-06    1.13740E-04    6.78127E-04    9.37900E-04      2328 1451
    1.40076E-03    2.54577E-03    4.01102E-03    5.22789E-03      2328 1451
    1.00172E-02    1.34609E-02                                    2328 1451
                                                                  2328 1451
References :                                                      2328 1451
 1. H.Vonach, H.Muenzer  Oesterr. Akad. Wiss., Math + Naturw.     2328 1451
    Anzeiger, v.95, p.199, November 1958                          2328 1451
 2. H.K.Vonach et al.  Proc. 2nd Conf. on Nuclear Cross-Sections  2328 1451
    and Technology, Washington D.C., 4-7 March 1968, v.2, p.885   2328 1451
 3. V.N.Levkovskiy et al. Yadernaja Fizika (Sov.), v.10, n.1,     2328 1451
    p.44, July 1968                                               2328 1451
 4. J.C.Robertson,B.Audric,P.Kolkowski  J. of Nucl. Energy, v.27, 2328 1451
    p.531, August 1973                                            2328 1451
 5. A.Paulsen et al.  Atomkernenergie, v.22, p.291, Sep. 1974     2328 1451
 6. W.Mannhart, H.Vonach  Zeitschrift fuer Physik, Section A,     2328 1451
    v.272, p.279, March 1975                                      2328 1451
 7. S.M.Qaim, N.I.Molla  Proc. of 9th Symposium on Fusion Techno- 2328 1451
    logy, Garmisch, 14-18 June 1976, FRG, Pergamon Press, p.589   2328 1451
 8. O.Schwerer et al.  Oesterr. Akad. Wiss., Math + Naturw.       2328 1451
    Anzeiger, v.113, p.153, June 1976                             2328 1451
 9. E.Zupranska et al.  Progress Report INR-1821/I/PL/A, 1980     2328 1451
    E.Zupranska et al.  Acta. Phys. Pol., v.B11, p.853, Nov. 1980 2328 1451
10. I.Kanno, J.W.Meadows, D.L.Smith  Annals of Nuclear Energy,    2328 1451
    v.11, p.623, December 1984                                    2328 1451
11. J.W.Meadows, D.L.Smith et al. Annals of Nuclear Energy, v.14, 2328 1451
    p.489, September 1987                                         2328 1451
12. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2328 1451
13. K.Kobayashi, I.Kimura  Proc. of an Intern. Conf. on Nuclear   2328 1451
    Data for Science and Technology, 30 May - 3 June 1988, Mito,  2328 1451
    Japan, Saikon Publishing Co., LTD, pp.261-265, 1989  ;        2328 1451
14. Lu Hanlin et al.  Report INDC(CPR)-16, August 1989            2328 1451
15. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990    2328 1451
16. N.I.Molla et al. Proc. of International Conference on Nuclear 2328 1451
    Data for Science and Technology, Gatlinburg, Tennessee, USA,  2328 1451
    9-13 May 1994, v.2, pp.938-940                                2328 1451
17. A.A.Filatenkov et al.  VANT, Ser.: Yadernye Konstanty, v.2,   2328 1451
    p.8, Moscow, 1996                                             2328 1451
18. D.Subasi et al.  Nucl. Sci. Eng., to be published in 1997     2328 1451
19. A.D.Majdeddin  "Measurement and analysis of excitation func-  2328 1451
    tions for fast  neutron induced reactions from threshold to   2328 1451
    20 MeV".  PHD dissertation, Inst. of Experimental Physics ,   2328 1451
    Kossuth University, Hungary, June 1998                        2328 1451
20. A.A.Filatenkov et al.  Report RI-252, St.Petersburg, May 1999 2328 1451
21. P.Reimer et al. Progress Report On Nuclear Data Research in   2328 1451
    the FRG, INDC(Ger)-047, July 2001                             2328 1451
22. I.Kumabe  J. Phys. Soc. Jap., v.13, p.325, 1958               2328 1451
23. E.B.Paul, R.L.Clarke  Canadian J. of Phys., v.31, p.267, 1953 2328 1451
24. A.Poularikas, R.W.Fink   Progress Report A-ARK-60, p.3, 1960  2328 1451
25. F.Gabbard et al.  Bull. Amer. Phys. Soc., v.5, p.42(C8), 1960 2328 1451
26. M.Bormann et al.  Phys. Rev., v.22, p.602, October 1961       2328 1451
27. M.Hillman  Nucl. Phys., v.37, p.78, 1962                      2328 1451
28. E.T.Bramlitt, R.W.Fink  Phys. Rev., v.131, p.2649, Sep. 1963  2328 1451
29. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965          2328 1451
30. D.Crumpton  J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969    2328 1451
31. R.Spangler et al.  Transactions of the American Nuclear       2328 1451
    Society, v.22, p.818, November 1975                           2328 1451
32. U.Garuska et al. Progress Report INDC(POL)-9, p.16, Sep. 1978 2328 1451
33. W.H.Warren, W.L.Alford Annals of Nucl. Energy, v.9,p.369,1982 2328 1451
34. G.Helfer et al.  Czechoslovak Journal of Physics, Part B,     2328 1451
    v.34, p.30, 1984                                              2328 1451
35. R.Pepelnik et al.  Int. Conf. on Nuclear Data for Basic and   2328 1451
    Applied Science, Santa Fe, N.M., 13-17 May 1985, v.1, p.211   2328 1451
36. N.I.Molla et al.  Progress Report INDC(BAN)-003 , Sep. 1986   2328 1451
37. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24,Aug. 1992 2328 1451
38. J.Cezar Suita et al.  Nucl. Sci. Eng., v.126, p.101, 1997     2328 1451
39. F.Nasyrov,B.D.Sciborskij  Atomnaya Energija (Sov.), v.25, n.5,2328 1451
    p.437, November 1968                                          2328 1451
40. I.Kimura, K.Kobayashi, T.Shibata  Nucl. Sci. Technology., v.8,2328 1451
    p.59, February 1971                                           2328 1451
41. K.Kobayashi, I.Kimura et al.  Nucl. Sci. Technology., v.13,   2328 1451
    p.531, October 1976                                           2328 1451
42. W.Mannhart   Progress Report INDC(Ger)-045, pp.40-43, 1999    2328 1451
43. W.Mannhart   Proc. of 5-th ASTM-EURATOM Symposium on Reactor  2328 1451
    Dosimetry. Geesthacht, Sep.24-28, 1984, Vol.2, p.813, 1985    2328 1451
44. J.Csikai, Z.Dezso   Proc. of 4th All Union Conf. on Neutron   2328 1451
    Physics, Kiev, 18-22 April 1977, v.3, p.32                    2328 1451
45. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear    2328 1451
    Model Code for Calculation of Cross Section and Emission      2328 1451
    Spectra. Report LA-6947, Los Alamos, 1977  ;                  2328 1451
    E.L.Trykiv, G.Ya.Tertychnyi  Private communication, IPPE,     2328 1451
    Obninsk, May 1999                                             2328 1451
46. S.A.Badikov et.al.  Preprint FEI-1686, Obninsk, 1985          2328 1451
47. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2328 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2328 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2328 1451
48. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2328 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2328 1451
49. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2328 1451
***************************************************************** 2328 1451
 The Q values and threshold energies were updated prior to pro-   2328 1451
 cessing through the codes to comply with the values obtained     2328 1451
 using the NNDC calculation program which is based on the 1995    2328 1451
 Update to the Atomic mass Evaluation.                            2328 1451
                                                                  2328 1451
 File 2 added to the pointwise file containing only the effective 2328 1451
 scattering radius with no resonance parameters given.            2328 1451
 Taken from ENDF/B-VI                                             2328 1451
***************************************************************** 2328 1451
                                                                  2328 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2328 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2328 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2328 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2328 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2328 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2328 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2328 1451
 ***************** Program FIXUP (Version 2002-1) ****************2328 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2328 1451
 Corrected Thresholds-----------------------------------------Yes 2328 1451
 Extended Cross Sections to 20 MeV----------------------------No  2328 1451
 Allow Cross Section Deletion---------------------------------No  2328 1451
 Allow Cross Section Reconstruction---------------------------No  2328 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2328 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2328 1451
 Deleted Duplicate Points-------------------------------------Yes 2328 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2328 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2328 1451
 Allow Creation of Missing Sections---------------------------No  2328 1451
 Allow Insertion of Energy Points-----------------------------No  2328 1451
 Create Uniform Energy Grid-----------------------------------No  2328 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2328 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2328 1451
 Unshielded Group Averages Using  640 Groups                      2328 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2328 1451
                                1        451        192          12328 1451
                                3        107         64          12328 1451
                               33        107         71          12328 1451
                                                                  2328 1  0
                                                                  2328 0  0
 2.30510E+4 5.05063E+1          0          0          0          02328 3107
-2.05800E+6-2.05800E+6          0          0          1        1812328 3107
        181          1                                            2328 3107
 2000000.00 1.8941E-13 2100000.00 1.23745E-9 2200000.00 3.65187E-92328 3107
 2300000.00 6.06629E-9 2400000.00 8.48070E-9 2500000.00 1.08951E-82328 3107
 2600000.00 1.33095E-8 2700000.00 1.57240E-8 2800000.00 1.81384E-82328 3107
 2900000.00 2.05528E-8 3000000.00 2.17904E-8 3100000.00 2.18512E-82328 3107
 3200000.00 2.27955E-8 3300000.00 2.90406E-8 3400000.00 3.61691E-82328 3107
 3500000.00 4.62735E-8 3600000.00 5.93538E-8 3700000.00 7.29986E-82328 3107
 3800000.00 9.00308E-8 3900000.00 1.07628E-7 4000000.00 1.27030E-72328 3107
 4100000.00 1.48239E-7 4200000.00 1.70077E-7 4300000.00 1.95692E-72328 3107
 4400000.00 2.21935E-7 4500000.00 2.54196E-7 4600000.00 2.92475E-72328 3107
 4700000.00 3.34447E-7 4800000.00 3.98587E-7 4900000.00 4.66421E-72328 3107
 5000000.00 5.51411E-7 5100000.00 6.72556E-7 5200000.00 8.29793E-72328 3107
 5300000.00 1.03209E-6 5400000.00 1.29020E-6 5500000.00 1.61995E-62328 3107
 5600000.00 2.05269E-6 5700000.00 2.65886E-6 5800000.00 3.58291E-62328 3107
 5900000.00 5.05849E-6 6000000.00 7.35895E-6 6100000.00 1.07116E-52328 3107
 6200000.00 1.52802E-5 6300000.00 2.12329E-5 6400000.00 2.88160E-52328 3107
 6500000.00 3.83819E-5 6600000.00 5.03943E-5 6700000.00 6.54289E-52328 3107
 6800000.00 8.41770E-5 6900000.00 .000107451 7000000.00 .0001361842328 3107
 7100000.00 .000171432 7200000.00 .000214354 7300000.00 .0002661842328 3107
 7400000.00 .000328186 7500000.00 .000401585 7600000.00 .0004874772328 3107
 7700000.00 .000586733 7800000.00 .000699892 7900000.00 .0008270702328 3107
 8000000.00 .000967893 8100000.00 .001121482 8200000.00 .0012870882328 3107
 8300000.00 .001461652 8400000.00 .001644565 8500000.00 .0018322112328 3107
 8600000.00 .002024416 8700000.00 .002217381 8800000.00 .0024103462328 3107
 8900000.00 .002603311 9000000.00 .002796276 9100000.00 .0029885792328 3107
 9200000.00 .003176905 9300000.00 .003364567 9400000.00 .0035522292328 3107
 9500000.00 .003739891 9600000.00 .003927553 9700000.00 .0041152152328 3107
 9800000.00 .004302877 9900000.00 .004490539 10000000.0 .0046821802328 3107
 10100000.0 .004877800 10200000.0 .005073420 10300000.0 .0052744232328 3107
 10400000.0 .005480810 10500000.0 .005687197 10600000.0 .0059001172328 3107
 10700000.0 .006119570 10800000.0 .006339023 10900000.0 .0065655302328 3107
 11000000.0 .006799090 11100000.0 .007032650 11200000.0 .0072732432328 3107
 11300000.0 .007520870 11400000.0 .007768497 11500000.0 .0080236842328 3107
 11600000.0 .008286431 11700000.0 .008549179 11800000.0 .0088119262328 3107
 11900000.0 .009082530 12000000.0 .009360990 12100000.0 .0096394502328 3107
 12200000.0 .009917910 12300000.0 .010196370 12400000.0 .0104803252328 3107
 12500000.0 .010769775 12600000.0 .011059225 12700000.0 .0113486752328 3107
 12800000.0 .011638125 12900000.0 .011927575 13000000.0 .0122170252328 3107
 13100000.0 .012506475 13200000.0 .012795925 13300000.0 .0130853752328 3107
 13400000.0 .013374825 13500000.0 .013664275 13600000.0 .0139537252328 3107
 13700000.0 .014243175 13800000.0 .014524710 13900000.0 .0147983302328 3107
 14000000.0 .015071950 14100000.0 .015345570 14200000.0 .0156191902328 3107
 14300000.0 .015882675 14400000.0 .016136025 14500000.0 .0163893752328 3107
 14600000.0 .016642725 14700000.0 .016883825 14800000.0 .0171126752328 3107
 14900000.0 .017341525 15000000.0 .017570375 15100000.0 .0177839752328 3107
 15200000.0 .017982325 15300000.0 .018180675 15400000.0 .0183790252328 3107
 15500000.0 .018561667 15600000.0 .018728600 15700000.0 .0188955332328 3107
 15800000.0 .019047283 15900000.0 .019183850 16000000.0 .0193204172328 3107
 16100000.0 .019440483 16200000.0 .019544050 16300000.0 .0196476172328 3107
 16400000.0 .019733583 16500000.0 .019801950 16600000.0 .0198703172328 3107
 16700000.0 .019920333 16800000.0 .019952000 16900000.0 .0199836672328 3107
 17000000.0 .020002900 17100000.0 .019997250 17200000.0 .0199791502328 3107
 17300000.0 .019961050 17400000.0 .019924583 17500000.0 .0198697502328 3107
 17600000.0 .019814917 17700000.0 .019742550 17800000.0 .0196526502328 3107
 17900000.0 .019562750 18000000.0 .019456550 18100000.0 .0193340502328 3107
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                                                                  232833  0
                                                                  2328 0  0
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 24-CR-  0 PETTEN     EVAL-MAR85 W.L.ZIJP                         2400 1451
PRIVATE COMM.         DIST-MAR85                       850320     2400 1451
----IRDF-90/NMF-G     MATERIAL 2400                               2400 1451
-----INCIDENT-NEUTRON DATA                                        2400 1451
------ENDF-6 FORMAT                                               2400 1451
                                                                  2400 1451
DAMAGE CROSS SECTIONS OF CHROMIUM IN STEEL                        2400 1451
ASTM STANDARD                                                     2400 1451
                                                                  2400 1451
 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2400 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.100 PER-CENT      2400 1451
 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2400 1451
 UNSHIELDED GROUP AVERAGES USING  640 GROUPS (SAND-II EXTEND)     2400 1451
                                                                  2400 1451
                                                                  2400 1451
                                1        451         21          02400 1451
                                3        900        116          02400 1451
                                                                  2400 1  0
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 1.63000+ 7 2.92900+ 3 1.64000+ 7 2.92900+ 3 1.65000+ 7 2.92900+ 32400 3900
 1.66000+ 7 2.92900+ 3 1.67000+ 7 2.92900+ 3 1.68000+ 7 2.92900+ 32400 3900
 1.69000+ 7 2.92900+ 3 1.70000+ 7 2.95400+ 3 1.71000+ 7 2.95400+ 32400 3900
 1.72000+ 7 2.95400+ 3 1.73000+ 7 2.95400+ 3 1.74000+ 7 2.95400+ 32400 3900
 1.75000+ 7 2.95400+ 3 1.76000+ 7 2.95400+ 3 1.77000+ 7 2.95400+ 32400 3900
 1.78000+ 7 2.95400+ 3 1.79000+ 7 2.95400+ 3 1.80000+ 7 2.96500+ 32400 3900
 1.81000+ 7 2.96500+ 3 1.82000+ 7 2.96500+ 3 1.83000+ 7 2.96500+ 32400 3900
 1.84000+ 7 2.96500+ 3 1.85000+ 7 2.96500+ 3 1.86000+ 7 2.96500+ 32400 3900
 1.87000+ 7 2.96500+ 3 1.88000+ 7 2.96500+ 3 1.89000+ 7 2.96500+ 32400 3900
 1.90000+ 7 2.98900+ 3 1.91000+ 7 2.98900+ 3 1.92000+ 7 2.98900+ 32400 3900
 1.93000+ 7 2.98900+ 3 1.94000+ 7 2.98900+ 3 1.95000+ 7 2.98900+ 32400 3900
 1.96000+ 7 2.98900+ 3 1.97000+ 7 2.98900+ 3 1.98000+ 7 2.98900+ 32400 3900
 1.99000+ 7 2.98900+ 3 2.00000+ 7 0.0    + 0                      2400 3900
                                                                  2400 3  0
                                                                  2400 0  0
                                                                     0 0  0
 2.40520E+4 5.14943E+1          0          0         34         102431 1451
 0.0        0.0                 0          0          0          62431 1451
 1.00000E+0 2.00000E+7          0          0         10       20022431 1451
 3.00000E+2 0.0                 1          0         38          32431 1451
 24-Cr- 52 IRK-VIENNA EVAL-APR90                                  2431 1451
                      DIST-Feb2004                                2431 1451
----IRDF-2002         MATERIAL 2431                               2431 1451
-----INCIDENT NEUTRON DATA                                        2431 1451
------ENDF-6 FORMAT                                               2431 1451
 *****************************************************************2431 1451
                      DIST-JUN90                       910201     2431 1451
    IRK-EVAL.NLIB 25  2431                                        2431 1451
 *****************************************************************2431 1451
  The Q values and threshold energies were updated prior to pro-  2431 1451
  cessing through the codes to comply with the values obtained    2431 1451
  using the NNDC calculation program which is based on the 1995   2431 1451
  Update to the Atomic mass Evaluation.                           2431 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2431 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2431 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2431 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2431 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2431 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2431 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2431 1451
 ***************** Program FIXUP (Version 2002-1) ****************2431 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2431 1451
 Corrected Thresholds-----------------------------------------Yes 2431 1451
 Extended Cross Sections to 20 MeV----------------------------No  2431 1451
 Allow Cross Section Deletion---------------------------------No  2431 1451
 Allow Cross Section Reconstruction---------------------------No  2431 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2431 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2431 1451
 Deleted Duplicate Points-------------------------------------Yes 2431 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2431 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2431 1451
 Allow Creation of Missing Sections---------------------------No  2431 1451
 Allow Insertion of Energy Points-----------------------------No  2431 1451
 Create Uniform Energy Grid-----------------------------------No  2431 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2431 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2431 1451
 Unshielded Group Averages Using  640 Groups                      2431 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2431 1451
                                1        451         45          12431 1451
                                3         16         30          12431 1451
                               33         16         26          12431 1451
                                                                  2431 1  0
                                                                  2431 0  0
 2.40520E+4 5.14943E+1          0          0          0          02431 3 16
-1.20394E+7-1.20394E+7          0          0          1         792431 3 16
         79          1                                            2431 3 16
 12200000.0 .000136222 12300000.0 .002913303 12400000.0 .0090350002431 3 16
 12500000.0 .018505000 12600000.0 .029945000 12700000.0 .0421942862431 3 16
 12800000.0 .054457131 12900000.0 .070114000 13000000.0 .0871593332431 3 16
 13100000.0 .104204667 13200000.0 .121073583 13300000.0 .1368840002431 3 16
 13400000.0 .152518000 13500000.0 .168152000 13600000.0 .1837860002431 3 16
 13700000.0 .200229500 13800000.0 .221530000 13900000.0 .2436400002431 3 16
 14000000.0 .265750000 14100000.0 .287860000 14200000.0 .3092252502431 3 16
 14300000.0 .326122000 14400000.0 .342274000 14500000.0 .3584260002431 3 16
 14600000.0 .374578000 14700000.0 .390333833 14800000.0 .4037126672431 3 16
 14900000.0 .416695333 15000000.0 .429678000 15100000.0 .4426606672431 3 16
 15200000.0 .455643333 15300000.0 .468626000 15400000.0 .4816086672431 3 16
 15500000.0 .491881500 15600000.0 .499444500 15700000.0 .5070075002431 3 16
 15800000.0 .514570500 15900000.0 .522133500 16000000.0 .5296965002431 3 16
 16100000.0 .537259500 16200000.0 .544822500 16300000.0 .5523855002431 3 16
 16400000.0 .559948500 16500000.0 .566084500 16600000.0 .5707935002431 3 16
 16700000.0 .575502500 16800000.0 .580211500 16900000.0 .5849205002431 3 16
 17000000.0 .589629500 17100000.0 .594338500 17200000.0 .5990475002431 3 16
 17300000.0 .603756500 17400000.0 .608465500 17500000.0 .6119980002431 3 16
 17600000.0 .614354000 17700000.0 .616710000 17800000.0 .6190660002431 3 16
 17900000.0 .621422000 18000000.0 .623778000 18100000.0 .6261340002431 3 16
 18200000.0 .628490000 18300000.0 .630846000 18400000.0 .6332020002431 3 16
 18500000.0 .635069000 18600000.0 .636447000 18700000.0 .6378250002431 3 16
 18800000.0 .639203000 18900000.0 .640581000 19000000.0 .6419590002431 3 16
 19100000.0 .643337000 19200000.0 .644715000 19300000.0 .6460930002431 3 16
 19400000.0 .647471000 19500000.0 .648849000 19600000.0 .6502270002431 3 16
 19700000.0 .651605000 19800000.0 .652983000 19900000.0 .6543610002431 3 16
 20000000.0 0.0                                                   2431 3 16
                                                                  2431 3  0
                                                                  2431 0  0
 2.40520E+4 5.14943E+1          0          0          0          1243133 16
 0.000000+0 0.000000+0          0         16          0          1243133 16
 0.000000+0 0.000000+0          1          5        136         16243133 16
 1.000000-5 1.220000+7 1.245000+7 1.255000+7 1.265000+7 1.275000+7243133 16
 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.600000+7243133 16
 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 4.031000-2 3.015000-2 2.011000-2 4.441000-3 1.005000-2243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 2.256000-2 1.504000-2 3.322000-3243133 16
 7.519000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.003000-2 2.216000-3243133 16
 5.015000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.088000-3 1.107000-3243133 16
 7.104000-4 1.385000-4 3.054000-4 1.269000-4 1.050000-3 6.131000-4243133 16
 5.750000-4 5.345000-4 5.072000-4 2.506000-3 0.000000+0 0.000000+0243133 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16
 0.000000+0 1.823000-3 2.416000-4 5.329000-4 2.213000-4 1.832000-3243133 16
 1.070000-3 1.003000-3 9.325000-4 8.849000-4 4.904000-4 3.163000-4243133 16
 3.318000-4 3.571000-4 2.085000-4 1.956000-4 1.818000-4 1.725000-4243133 16
 7.342000-4 2.818000-4 7.875000-4 4.599000-4 4.313000-4 4.009000-4243133 16
 3.805000-4 4.986000-4 3.270000-4 1.910000-4 1.791000-4 1.665000-4243133 16
 1.580000-4 3.980000-3 1.581000-3 1.482000-3 1.378000-3 1.308000-3243133 16
 1.358000-3 8.658000-4 8.048000-4 7.637000-4 1.194000-3 7.548000-4243133 16
 7.163000-4 1.032000-3 6.658000-4 9.292000-4                      243133 16
                                                                  243133  0
                                                                  2431 0  0
                                                                     0 0  0
 2.50550E+4 5.44661E+1          0          0         34         102525 1451
 0.0        0.0                 0          0          0          62525 1451
 1.00000E+0 2.00000E+7          0          5         10       20022525 1451
 3.00000E+2 0.0                 1          0        157          32525 1451
 25-Mn- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA                        2525 1451
                      DIST-Feb2004                                2525 1451
----IRDF-2002         MATERIAL 2525                               2525 1451
-----INCIDENT NEUTRON DATA                                        2525 1451
------ENDF-6 FORMAT                                               2525 1451
******************************************************************2525 1451
 25-MN- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA                        2525 1451
                      DIST-SEP 1 REV2-SEP98            20010926   2525 1451
----ENDF/B-VI         MATERIAL 2525         REVISION 2            2525 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2525 1451
DOSIMETRY                                                         2525 1451
******************************************************************2525 1451
                                                                  2525 1451
 **************************************************************** 2525 1451
                                                                  2525 1451
 ENDF/B-VI MOD 3 Revision, June 2000, S.C. Frankle, R.C. Reedy,   2525 1451
                 P.G. Young (LANL)                                2525 1451
                                                                  2525 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   2525 1451
 MT 102) has been updated for new experimental data at incident   2525 1451
 neutron energies up to 100 keV.                                  2525 1451
 The previous (MOD 2) pure continuum at thermal neutron energy    2525 1451
 is replaced by 321 discrete photons.                             2525 1451
 The Q-value for radiative capture was also updated in File 3.    2525 1451
 Details of these changes are described in Frankel et al. [Fr01]. 2525 1451
                                                                  2525 1451
REFERENCES                                                        2525 1451
                                                                  2525 1451
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      2525 1451
   National Laboratory Report, LA-13812 (2001).                   2525 1451
                                                                  2525 1451
 **************************************************************** 2525 1451
                                                                  2525 1451
 ENDF/B-VI MOD 2 Revision, October 1997, V. McLane (NNDC)         2525 1451
                                                                  2525 1451
 1. Corrected residual nucleus and AWR in File 6, MT=22,107.      2525 1451
 2. Updated file 1 comments and and corrected references.         2525 1451
                                                                  2525 1451
 **************************************************************** 2525 1451
                                                                  2525 1451
 ENDF/B-VI MOD 1 Evaluation, March 1988, K. Shibata (ORNL,JAERI)  2525 1451
                                                                  2525 1451
 File 1  General Information ------------------------------------ 2525 1451
   MT=451   Descriptive data and dictionary                       2525 1451
                                                                  2525 1451
 File 2  Resonance Parameters ----------------------------------- 2525 1451
   MT=151   Resolved resonance parameters for MLBW formula.       2525 1451
     The parameters of the lowest four resonances were taken      2525 1451
     from the work of Macklin [1].  Others were taken from the    2525 1451
     compilation of mughabghab et al.[2] except that the          2525 1451
     parameters of two negative resonances were adjusted so as    2525 1451
     to fit to experimental thermal cross sections.               2525 1451
     Resonance region : 1.0E-5 eV to 100 keV.                     2525 1451
                                                                  2525 1451
     Scattering radius: 5.15 fm                                   2525 1451
                                                                  2525 1451
     Calculated 2200-m/s cross sections and res. integrals        2525 1451
                  2200-M/S         RES. INTEG.                    2525 1451
      Elastic      2.167 b            -                           2525 1451
      Capture     13.413 b          11.73  b                      2525 1451
      Total       15.579 b            -                           2525 1451
                                                                  2525 1451
 File 3  Neutron Cross Sections --------------------------------- 2525 1451
   MT=1 Total                                                     2525 1451
      Below 100 keV   : no background                             2525 1451
      Above 100 keV   : based on the experimental data [3,4,5].   2525 1451
   MT=2 Elastic scattering                                        2525 1451
      (total) - (nonelastic cross section).                       2525 1451
                                                                  2525 1451
   MT=102  Radiative capture cross section                        2525 1451
      Below 100 keV: Resonance parameters given (no background).  2525 1451
      Above 100 keV: Based on the experimental data [11-15]       2525 1451
        (taken from JENDL-2 [20]).                                2525 1451
                                                                  2525 1451
                                                                  2525 1451
 File 33.  Covariance data                                        2525 1451
   MT=102                                                         2525 1451
      Estimated from experimental data and calculation.           2525 1451
                                                                  2525 1451
 ---------------------------------------------------------------- 2525 1451
 REFERENCES                                                       2525 1451
                                                                  2525 1451
 [1] R.L. Macklin, Nucl.Sci.Eng. 89, 362 (1985).                  2525 1451
 [2] S. F. Mughabghab, M. Divadeenam, and N.E. Holden, Neutron    2525 1451
     Cross Sections, Vol. 1, Part A (Academic Press, 1981).       2525 1451
 [3] S. Cierjacks, P. Forti, D. Kopsch et al., "High Resolution   2525 1451
     Total Cross Sections for Na, Cl, K, V, Mn and Co between 0.5 2525 1451
     and 30 MeV", report KFK-1000 (1968).                         2525 1451
 [4] W.F.E. Pineo, M. Divadeenam, E.G. Bilpuch et al., Ann.Phys.  2525 1451
     84, 165 (1974).                                              2525 1451
 [5] J.B. Garg, J. Rainwater, and W.W. Havens Jr., Nucl.Sci.Eng.  2525 1451
     65, 76 (1978).                                               2525 1451
 [6] C.Y. Fu, "A Consistent Nuclear Model for Compound and Pre-   2525 1451
     compound Reactions with Conservation of Angular Momentum,"   2525 1451
     Oak Ridge National Laboratory report ORNL/TM-7042 (1980).    2525 1451
 [7] C.Y. Fu, private communication (1985).                       2525 1451
 [8] Zhou Enchen, Huo Junde, Zhou Chunmei et al., Nucl.Data       2525 1451
     Sheets 44, 463 (1985).                                       2525 1451
 [9] F.G. Perey, Phys.Rev. 131, 745 (1963).                       2525 1451
[10] J.R. Huizenga,  and G.J. Igo, Nucl.Phys. 29, 462 (1962).     2525 1451
[11] J.B. Garg, R.L. Macklin, and J. Halperin, Phys.Rev. C18,     2525 1451
     2079 (1978).                                                 2525 1451
[12] A.G. Dovbenko, V.E. Kolesov, V.P. Koroleva et al., At.En. 26,2525 1451
     67 (1969).                                                   2525 1451
[13] H.O. Menlove, K.L. Coop, H.A. Grench et al. Phys.Rev. 163,   2525 1451
     1299 (1967).                                                 2525 1451
[14] O. Schwerer, M. Winkler-Rohatsch, H. Warhanek et al., Nucl.  2525 1451
     Phys. A264, 105 (1976).                                      2525 1451
[15] M. Budnar, F. Cvelbar, E. Hodgson et al., "Prompt Gamma-ray  2525 1451
     Spectra and Integarted Cross Sections for the Radiative      2525 1451
     Capture of 14 MeV Neutrons for 28 Natural Targets ...,"      2525 1451
     report INDC(YUG)-6 (1979).                                   2525 1451
[16] L. Colli, I. Iori, S. Micheletti et al., Nuovo.cim. 21, 966  2525 1451
     (1962).                                                      2525 1451
[17] S. Sudar and Csikai, Nucl.Phys. A319, 157 (1979).            2525 1451
[18] M. Diksic, P. Strohal, and I. Slaus, J.Inorg.Nucl.Chem. 36,  2525 1451
     477 (1974).                                                  2525 1451
[19] C.H. Wu, R. Woelfle, and S.M. Qaim, Nucl.Phys. A329, 63      2525 1451
     (1979).                                                      2525 1451
[20] T. Nakagawa, "Summary of JENDL-2 General Purpose File,"      2525 1451
     report JAERI-M 84-103 (1984).                                2525 1451
                                                                  2525 1451
******************************************************************2525 1451
                                                                  2525 1451
                                                                  2525 1451
                                                                  2525 1451
                                                                  2525 1451
******************************************************************2525 1451
 ************************ C O N T E N T S *********************** 2525 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2525 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2525 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2525 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2525 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2525 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2525 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2525 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2525 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2525 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2525 1451
 ***************** Program FIXUP (Version 2002-1) ****************2525 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2525 1451
 Corrected Thresholds-----------------------------------------Yes 2525 1451
 Extended Cross Sections to 20 MeV----------------------------No  2525 1451
 Allow Cross Section Deletion---------------------------------No  2525 1451
 Allow Cross Section Reconstruction---------------------------No  2525 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2525 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2525 1451
 Deleted Duplicate Points-------------------------------------Yes 2525 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2525 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2525 1451
 Allow Creation of Missing Sections---------------------------No  2525 1451
 Allow Insertion of Energy Points-----------------------------No  2525 1451
 Create Uniform Energy Grid-----------------------------------No  2525 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2525 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2525 1451
 Unshielded Group Averages Using  640 Groups                      2525 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2525 1451
                                1        451        164          32525 1451
                                3        102        217          32525 1451
                               33        102         17          12525 1451
                                                                  2525 1  0
                                                                  2525 0  0
 2.50550E+4 5.44661E+1          0          0          0          02525 3102
 7.27050E+6 7.27050E+6          0          0          1        6412525 3102
        641          1                                            2525 3102
 .000100000 210.836355 .000105000 205.919990 .000110000 201.2493862525 3102
 .000115000 196.945550 .000120000 191.914305 .000127500 186.3322702525 3102
 .000135000 181.232244 .000142500 176.545422 .000150000 171.4729812525 3102
 .000160000 166.188822 .000170000 161.380891 .000180000 156.9675742525 3102
 .000190000 152.861356 .000200000 149.112342 .000210000 145.5851502525 3102
 .000220000 142.308952 .000230000 139.265033 .000240000 135.6921232525 3102
 .000255000 131.765845 .000270000 128.736134 .000280000 125.3635062525 3102
 .000300000 121.250491 .000320000 117.526305 .000340000 114.1155532525 3102
 .000360000 110.974047 .000380000 108.106258 .000400000 105.1164352525 3102
 .000425000 102.056964 .000450000 99.2616168 .000475000 96.69878032525 3102
 .000500000 94.2886053 .000525000 92.0858615 .000550000 90.00503242525 3102
 .000575000 88.0698021 .000600000 86.0862994 .000630000 84.04577582525 3102
 .000660000 82.1786685 .000690000 80.3875951 .000720000 78.48183612525 3102
 .000760000 76.4384216 .000800000 74.5411534 .000840000 72.80240492525 3102
 .000880000 71.1498209 .000920000 69.6348527 .000960000 68.18401602525 3102
 .001000000 66.6802433 .001050000 65.1103649 .001100000 63.63859712525 3102
 .001150000 62.2824700 .001200000 60.6827442 .001275000 58.92456722525 3102
 .001350000 57.3171279 .001425000 55.8169724 .001500000 54.22319982525 3102
 .001600000 52.5567297 .001700000 51.0354372 .001800000 49.62844332525 3102
 .001900000 48.3430548 .002000000 47.1512904 .002100000 46.03299822525 3102
 .002200000 45.0088670 .002300000 44.0295564 .002400000 42.91218462525 3102
 .002550000 41.6705962 .002700000 40.6994470 .002800000 39.64420442525 3102
 .003000000 38.3451997 .003200000 37.1630951 .003400000 36.08207962525 3102
 .003600000 35.0960903 .003800000 34.1857100 .004000000 33.23684052525 3102
 .004250000 32.2734093 .004500000 31.3941221 .004750000 30.57240332525 3102
 .005000000 29.8184172 .005250000 29.1195793 .005500000 28.45904112525 3102
 .005750000 27.8537655 .006000000 27.2181353 .006300000 26.58283802525 3102
 .006600000 25.9816432 .006900000 25.4240492 .007200000 24.81770612525 3102
 .007600000 24.1692362 .008000000 23.5761125 .008400000 23.01947612525 3102
 .008800000 22.5015420 .009200000 22.0202748 .009600000 21.56235132525 3102
 .010000000 21.0882598 .010500000 20.5870570 .011000000 20.12861052525 3102
 .011500000 19.6912520 .012000000 19.1899854 .012750000 18.63505402525 3102
 .013500000 18.1210420 .014250000 17.6495482 .015000000 17.14682352525 3102
 .016000001 16.6192227 .017000001 16.1355174 .017999999 15.69395772525 3102
 .018999999 15.2880961 .020000000 14.9073261 .021000000 14.55883792525 3102
 .022000000 14.2305260 .023000000 13.9246739 .024000000 13.56547322525 3102
 .025500000 13.1784088 .027000001 12.8717830 .028000001 12.53936132525 3102
 .029999999 12.1291901 .032000002 11.7547360 .034000002 11.41274262525 3102
 .035999998 11.1005590 .037999999 10.8126558 .039999999 10.51231402525 3102
 .042500000 10.2075815 .045000002 9.92927915 .047499999 9.670116892525 3102
 .050000001 9.43185152 .052499998 9.21135883 .055000000 9.002585112525 3102
 .057500001 8.81134205 .059999999 8.61018310 .063000001 8.408838552525 3102
 .066000000 8.21903201 .068999998 8.04215889 .071999997 7.850354602525 3102
 .075999998 7.64490573 .079999998 7.45726121 .083999999 7.280971782525 3102
 .088000000 7.11715255 .092000000 6.96557614 .096000001 6.820888452525 3102
 .100000001 6.67151668 .104999997 6.51337408 .109999999 6.368653332525 3102
 .115000002 6.23031740 .119999997 6.07166588 .127499998 5.896271252525 3102
 .135000005 5.73360932 .142499998 5.58432775 .150000006 5.425221582525 3102
 .159999996 5.25797835 .170000002 5.10499237 .180000007 4.965220922525 3102
 .189999998 4.83697678 .200000003 4.71666452 .209999993 4.606885102525 3102
 .219999999 4.50328170 .230000004 4.40673826 .239999995 4.295008192525 3102
 .254999995 4.16988289 .270000011 4.07410511 .280000001 3.967682092525 3102
 .300000012 3.83731116 .319999993 3.71906586 .340000004 3.611487862525 3102
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 1000.00000 .405906404 1050.00000 6.82454289 1100.00000 5.669811202525 3102
 1150.00000 .362922980 1200.00000 .135326257 1275.00000 .0814805222525 3102
 1350.00000 .067185336 1425.00000 .063830493 1500.00000 .0678397682525 3102
 1600.00000 2.00709610 1700.00000 .091643683 1800.00000 .1131801142525 3102
 1900.00000 .155612894 2000.00000 .231560731 2100.00000 .3637798062525 3102
 2200.00000 .532485764 2300.00000 .564672971 2400.00000 .3616478762525 3102
 2550.00000 .173581391 2700.00000 .101383890 2800.00000 .0626053172525 3102
 3000.00000 .037614764 3200.00000 .025780154 3400.00000 .0193484292525 3102
 3600.00000 .015524039 3800.00000 .013933781 4000.00000 .0121013302525 3102
 4250.00000 .017858518 4500.00000 .010019391 4750.00000 .4341479382525 3102
 5000.00000 .011191569 5250.00000 .011343136 5500.00000 .0131607792525 3102
 5750.00000 .016506710 6000.00000 .024240580 6300.00000 .1880253652525 3102
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 7600.00000 .047031546 8000.00000 .051335651 8400.00000 .2067577862525 3102
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 10000.0000 .008594289 10500.0000 .071149886 11000.0000 .0037810102525 3102
 11500.0000 .022854740 12000.0000 .069200279 12750.0000 .0020981072525 3102
 13500.0000 .001627971 14250.0000 .046109203 15000.0000 .0016402262525 3102
 16000.0000 .029458063 17000.0000 .130854113 18000.0000 .0502274762525 3102
 19000.0000 .009456944 20000.0000 .070705073 21000.0000 .0243172022525 3102
 22000.0000 .038367457 23000.0000 .036942031 24000.0000 .0226151772525 3102
 25500.0000 .069917782 27000.0000 .038371766 28000.0000 .0150495922525 3102
 30000.0000 .021166131 32000.0000 .026460201 34000.0000 .0565505602525 3102
 36000.0000 .023746549 38000.0000 .004205812 40000.0000 .0324913042525 3102
 42500.0000 .020375006 45000.0000 .011843840 47500.0000 .0081271672525 3102
 50000.0000 .011519836 52500.0000 .022264406 55000.0000 .0169532242525 3102
 57500.0000 .020321384 60000.0000 .016414420 63000.0000 .0229886212525 3102
 66000.0000 .018082623 69000.0000 .032203939 72000.0000 .0126870832525 3102
 76000.0000 .008990626 80000.0000 .015362294 84000.0000 .0230386352525 3102
 88000.0000 .013447072 92000.0000 .007114579 96000.0000 .0126574952525 3102
 100000.000 .013806739 105000.000 .013550256 110000.000 .0136299922525 3102
 115000.000 .013850078 120000.000 .014341383 127500.000 .0145210592525 3102
 135000.000 .011994399 142500.000 .007991463 150000.000 .0078895692525 3102
 160000.000 .008206275 170000.000 .008248789 180000.000 .0077906912525 3102
 190000.000 .007096819 200000.000 .006509434 210000.000 .0061033842525 3102
 220000.000 .006045272 230000.000 .005877501 240000.000 .0054732062525 3102
 255000.000 .005270608 270000.000 .005672634 280000.000 .0062395712525 3102
 300000.000 .005588775 320000.000 .005429058 340000.000 .0054481402525 3102
 360000.000 .004134608 380000.000 .004576443 400000.000 .0054508182525 3102
 425000.000 .004852925 450000.000 .005400591 475000.000 .0045650072525 3102
 500000.000 .003995238 525000.000 .004095028 550000.000 .0044449782525 3102
 575000.000 .004432601 600000.000 .003668962 630000.000 .0037604722525 3102
 660000.000 .004102263 690000.000 .003633306 720000.000 .0033332162525 3102
 760000.000 .003279851 800000.000 .003243539 840000.000 .0032055942525 3102
 880000.000 .003123043 920000.000 .002977128 960000.000 .0027922012525 3102
 1000000.00 .002620025 1100000.00 .002490003 1200000.00 .0023800022525 3102
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 3100000.00 .001490001 3200000.00 .001451001 3300000.00 .0014180012525 3102
 3400000.00 .001386001 3500000.00 .001356001 3600000.00 .0013280012525 3102
 3700000.00 .001300509 3800000.00 .001276061 3900000.00 .0012521212525 3102
 4000000.00 .001228181 4100000.00 .001204868 4200000.00 .0011828282525 3102
 4300000.00 .001160882 4400000.00 .001139808 4500000.00 .0011190302525 3102
 4600000.00 .001098252 4700000.00 .001077733 4800000.00 .0010584212525 3102
 4900000.00 .001039262 5000000.00 .001020102 5100000.00 .0010009422525 3102
 5200000.00 .000982035 5300000.00 .000964328 5400000.00 .0009467772525 3102
 5500000.00 .000930649 5600000.00 .000915945 5700000.00 .0009012422525 3102
 5800000.00 .000886609 5900000.00 .000873163 6000000.00 .0008601392525 3102
 6100000.00 .000847563 6200000.00 .000835951 6300000.00 .0008243712525 3102
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 8800000.00 .000693433 8900000.00 .000692700 9000000.00 .0006919672525 3102
 9100000.00 .000691233 9200000.00 .000690500 9300000.00 .0006897672525 3102
 9400000.00 .000689033 9500000.00 .000688300 9600000.00 .0006875672525 3102
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 10000000.0 .000684825 10100000.0 .000684475 10200000.0 .0006841252525 3102
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 12700000.0 .000673278 12800000.0 .000672649 12900000.0 .0006720192525 3102
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 14800000.0 .000655383 14900000.0 .000654330 15000000.0 .0006532762525 3102
 15100000.0 .000652135 15200000.0 .000650786 15300000.0 .0006494292525 3102
 15400000.0 .000648072 15500000.0 .000646714 15600000.0 .0006453572525 3102
 15700000.0 .000643970 15800000.0 .000642400 15900000.0 .0006408002525 3102
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 18400000.0 .000479172 18500000.0 .000467322 18600000.0 .0004539702525 3102
 18700000.0 .000440765 18800000.0 .000428381 18900000.0 .0004161262525 3102
 19000000.0 .000401843 19100000.0 .000385857 19200000.0 .0003706312525 3102
 19300000.0 .000355930 19400000.0 .000341972 19500000.0 .0003253282525 3102
 19600000.0 .000306371 19700000.0 .000288758 19800000.0 .0002721502525 3102
 19900000.0 .000256641 20000000.0 0.0                             2525 3102
                                                                  2525 3  0
                                                                  2525 0  0
 2.50550E+4 5.44661E+1          0          0          0          1252533102
 0.000000+0 0.000000+0          0        102          0          3252533102
 0.000000+0 0.000000+0          0          1          8          4252533102
 1.000000-5 1.600000-3 1.000000+3 2.500000-3 2.000000+6 1.000000-2252533102
 2.000000+7 0.000000+0                                            252533102
 0.000000+0 0.000000+0          0          1         26         13252533102
 1.000000-5 3.960000-4 2.530000-2 2.227500-4 5.000000+1 3.960000-4252533102
 2.000000+3 3.564000-3 6.000000+4 6.336000-3 1.500000+5 9.900000-3252533102
 1.000000+6 2.227500-2 2.000000+6 2.227500-2 5.000000+6 3.960000-2252533102
 7.000000+6 3.960000-2 1.000000+7 8.910000-2 1.450000+7 8.910000-2252533102
 2.000000+7 0.000000+0                                            252533102
 0.000000+0 0.000000+0          0          8         26         13252533102
 1.000000-5 1.960000-2 2.530000-2 6.692000-3 5.000000+1 1.090900-6252533102
 2.000000+3 2.851600-3 6.000000+4 1.747200-7 1.500000+5 5.975300-9252533102
 1.000000+6 1.802000-9 2.000000+6 7.37120-10 5.000000+6 4.24360-10252533102
 7.000000+6 2.29830-10 1.000000+7 4.22300-10 1.450000+7 3.92040-10252533102
 2.000000+7 0.000000+0                                            252533102
                                                                  252533  0
                                                                  2525 0  0
                                                                     0 0  0
 2.60000+ 4 0.0    + 0         -1          0         34          12600 1451
 0.0    + 0 0.0    + 0          0          0          0          62600 1451
 1.0    + 0 0.0    + 0          0          0         10          22600 1451
 0.0    + 0 0.0    + 0          0          0         17          32600 1451
 26-FE-  0 PETTEN     EVAL-   79 W.L.ZIJP  P.GRIFFIN              2600 1451
 PRIVATE COM.         DIST-JAN82                       811211     2600 1451
----IRDF-90/NMF-G     MATERIAL 2600                               2600 1451
-----INCIDENT NEUTRON DATA                                        2600 1451
------ENDF-6 FORMAT                                               2600 1451
              DAMAGE CROSS SECTION OF IRON IN STEEL               2600 1451
     E693     ASTM DPA FOR IRON   MT=900   (FROM P.GRIFFIN 2002)  2600 1451
              DAMAGE ENERGY TO DISPLACEMENT CONVERSION CONSISTENT 2600 1451
              WITH ASTM E521 AND E281                             2600 1451
              EURATOM STANDARD    MT=901                          2600 1451
 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2600 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.100 PER-CENT      2600 1451
 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2600 1451
 UNSHIELDED GROUP AVERAGES USING  640 GROUPS (SAND-II EXTEND)     2600 1451
                                                                  2600 1451
                                                                  2600 1451
                                                                  2600 1451
                                1        451         24          02600 1451
                                3        900        217          02600 1451
                                3        901        217          02600 1451
                                                                  2600 1  0
                                                                  2600 0  0
 2.60000+ 4 0.0    + 0          0          0          0          02600 3900
 0.0    + 0 0.0    + 0          0          0          1        6412600 3900
        641          1                                            2600 3900
 1.0000E-04 1.5835E+02 1.0500E-04 1.5462E+02 1.1000E-04 1.5114E+022600 3900
 1.1500E-04 1.4789E+02 1.2000E-04 1.4411E+02 1.2750E-04 1.3992E+022600 3900
 1.3500E-04 1.3609E+02 1.4250E-04 1.3254E+02 1.5000E-04 1.2875E+022600 3900
 1.6000E-04 1.2479E+02 1.7000E-04 1.2117E+02 1.8000E-04 1.1785E+022600 3900
 1.9000E-04 1.1481E+02 2.0000E-04 1.1196E+02 2.1000E-04 1.0932E+022600 3900
 2.2000E-04 1.0686E+02 2.3000E-04 1.0457E+02 2.4000E-04 1.0189E+022600 3900
 2.5500E-04 9.8933E+01 2.7000E-04 9.6660E+01 2.8000E-04 9.4127E+012600 3900
 3.0000E-04 9.1052E+01 3.2000E-04 8.8249E+01 3.4000E-04 8.5688E+012600 3900
 3.6000E-04 8.3339E+01 3.8000E-04 8.1173E+01 4.0000E-04 7.8925E+012600 3900
 4.2500E-04 7.6636E+01 4.5000E-04 7.4537E+01 4.7500E-04 7.2599E+012600 3900
 5.0000E-04 7.0818E+01 5.2500E-04 6.9148E+01 5.5000E-04 6.7592E+012600 3900
 5.7500E-04 6.6138E+01 6.0000E-04 6.4642E+01 6.3000E-04 6.3120E+012600 3900
 6.6000E-04 6.1702E+01 6.9000E-04 6.0373E+01 7.2000E-04 5.8927E+012600 3900
 7.6000E-04 5.7397E+01 8.0000E-04 5.5979E+01 8.4000E-04 5.4660E+012600 3900
 8.8000E-04 5.3432E+01 9.2000E-04 5.2287E+01 9.6000E-04 5.1215E+012600 3900
 1.0000E-03 5.0077E+01 1.0500E-03 4.8896E+01 1.1000E-03 4.7796E+012600 3900
 1.1500E-03 4.6769E+01 1.2000E-03 4.5571E+01 1.2750E-03 4.4250E+012600 3900
 1.3500E-03 4.3037E+01 1.4250E-03 4.1918E+01 1.5000E-03 4.0717E+012600 3900
 1.6000E-03 3.9463E+01 1.7000E-03 3.8320E+01 1.8000E-03 3.7270E+012600 3900
 1.9000E-03 3.6310E+01 2.0000E-03 3.5407E+01 2.1000E-03 3.4574E+012600 3900
 2.2000E-03 3.3797E+01 2.3000E-03 3.3070E+01 2.4000E-03 3.2224E+012600 3900
 2.5500E-03 3.1289E+01 2.7000E-03 3.0570E+01 2.8000E-03 2.9770E+012600 3900
 3.0000E-03 2.8798E+01 3.2000E-03 2.7910E+01 3.4000E-03 2.7101E+012600 3900
 3.6000E-03 2.6359E+01 3.8000E-03 2.5674E+01 4.0000E-03 2.4963E+012600 3900
 4.2500E-03 2.4240E+01 4.5000E-03 2.3575E+01 4.7500E-03 2.2963E+012600 3900
 5.0000E-03 2.2399E+01 5.2500E-03 2.1871E+01 5.5000E-03 2.1380E+012600 3900
 5.7500E-03 2.0920E+01 6.0000E-03 2.0447E+01 6.3000E-03 1.9965E+012600 3900
 6.6000E-03 1.9517E+01 6.9000E-03 1.9097E+01 7.2000E-03 1.8640E+012600 3900
 7.6000E-03 1.8156E+01 8.0000E-03 1.7707E+01 8.4000E-03 1.7290E+012600 3900
 8.8000E-03 1.6902E+01 9.2000E-03 1.6541E+01 9.6000E-03 1.6202E+012600 3900
 1.0000E-02 1.5842E+01 1.0500E-02 1.5469E+01 1.1000E-02 1.5121E+012600 3900
 1.1500E-02 1.4796E+01 1.2000E-02 1.4419E+01 1.2750E-02 1.4001E+012600 3900
 1.3500E-02 1.3617E+01 1.4250E-02 1.3263E+01 1.5000E-02 1.2884E+012600 3900
 1.6000E-02 1.2488E+01 1.7000E-02 1.2126E+01 1.8000E-02 1.1794E+012600 3900
 1.9000E-02 1.1490E+01 2.0000E-02 1.1208E+01 2.1000E-02 1.0943E+012600 3900
 2.2000E-02 1.0697E+01 2.3000E-02 1.0468E+01 2.4000E-02 1.0201E+012600 3900
 2.5500E-02 9.9067E+00 2.7000E-02 9.6794E+00 2.8000E-02 9.4270E+002600 3900
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 1.9900E+07 3.0502E+03 2.0000E+07 0.0000E+00                      2600 3900
                                                                  2600 3  0
 2.60000+ 4 0.0    + 0          0          0          0          02600 3901
 0.0    + 0 0.0    + 0          0          0          1        6412600 3901
        641          1                                            2600 3901
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 1.57000+ 7 2.17517+ 3 1.58000+ 7 2.17686+ 3 1.59000+ 7 2.17856+ 32600 3901
 1.60000+ 7 2.18025+ 3 1.61000+ 7 2.18025+ 3 1.62000+ 7 2.17856+ 32600 3901
 1.63000+ 7 2.17686+ 3 1.64000+ 7 2.17517+ 3 1.65000+ 7 2.17347+ 32600 3901
 1.66000+ 7 2.17178+ 3 1.67000+ 7 2.17008+ 3 1.68000+ 7 2.16839+ 32600 3901
 1.69000+ 7 2.16669+ 3 1.70000+ 7 2.16500+ 3 1.71000+ 7 2.16330+ 32600 3901
 1.72000+ 7 2.16161+ 3 1.73000+ 7 2.15991+ 3 1.74000+ 7 2.15822+ 32600 3901
 1.75000+ 7 2.15652+ 3 1.76000+ 7 2.15483+ 3 1.77000+ 7 2.15313+ 32600 3901
 1.78000+ 7 2.15144+ 3 1.79000+ 7 2.14974+ 3 1.80000+ 7 2.14890+ 32600 3901
 1.81000+ 7 2.14890+ 3 1.82000+ 7 2.14890+ 3 1.83000+ 7 2.14890+ 32600 3901
 1.84000+ 7 2.14890+ 3 1.85000+ 7 2.14890+ 3 1.86000+ 7 2.14890+ 32600 3901
 1.87000+ 7 2.14890+ 3 1.88000+ 7 2.14890+ 3 1.89000+ 7 2.14890+ 32600 3901
 1.90000+ 7 2.14890+ 3 1.91000+ 7 2.14890+ 3 1.92000+ 7 2.14890+ 32600 3901
 1.93000+ 7 2.14890+ 3 1.94000+ 7 2.14889+ 3 1.95000+ 7 2.14889+ 32600 3901
 1.96000+ 7 2.14889+ 3 1.97000+ 7 2.14889+ 3 1.98000+ 7 2.14889+ 32600 3901
 1.99000+ 7 2.14889+ 3 2.00000+ 7 0.0    + 0                      2600 3901
                                                                  2600 3  0
                                                                  2600 0  0
                                                                     0 0  0
 2.60540E+4 5.34762E+1          0          0         34         102625 1451
 0.0        0.0                 0          0          0          62625 1451
 1.00000E+0 2.00000E+7          0          0         10       20022625 1451
 3.00000E+2 0.0                 1          0        527          72625 1451
 26-Fe- 54 FEI/LANL   EVAL-Jun01 K.I.Zolotarev,M.Chadwick et al.  2625 1451
                      DIST-Feb2004                                2625 1451
----IRDF-2002         MATERIAL 2625                               2625 1451
-----INCIDENT NEUTRON DATA                                        2625 1451
------ENDF-6 FORMAT                                               2625 1451
***************************************************************** 2625 1451
 26-Fe- 54 FEI        EVAL-Jun01 K.I.Zolotarev                    2625 1451
                      DIST-May03 Mod1                  20030512   2625 1451
----BROND-2           MATERIAL 2625         REVISION 1            2625 1451
-----INCIDENT NEUTRON DATA                                        2625 1451
------ENDF-6 FORMAT                                               2625 1451
------Russian Reactor Dosimetry File RRDF-2002                    2625 1451
***************************************************************** 2625 1451
Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko           2625 1451
MF=3                                                              2625 1451
MT= 16 - (n,2n) cross section                                     2625 1451
MT=107 - (n,a) cross section                                      2625 1451
***************************************************************** 2625 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2625 1451
DOSIMETRY                                                         2625 1451
******************************************************************2625 1451
  26-Fe- 54 LANL,ORNL EVAL-SEP96 M.B.CHADWICK,P.G.YOUNG,D.HETRICK 2625 1451
 Ch99,He91,Fu91       DIST-SEP 1 REV4-                 20010926   2625 1451
----ENDF/B-VI         MATERIAL 2625         REVISION 4            2625 1451
MF=3                                                              2625 1451
MT= 103 - (n,p) cross section                                     2625 1451
******************************************************************2625 1451
******** Start of (N,2N), (N,A) bibliographical component ******* 2625 1451
***************************************************************** 2625 1451
------Russian Reactor Dosimetry File RRDF-2002                    2625 1451
***************************************************************** 2625 1451
Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko           2625 1451
MF=3                                                              2625 1451
MT= 16 - (n,2n) cross section                                     2625 1451
-------------------------------------                             2625 1451
    Excitation function for the  Fe-54(n,2n)Fe-53m+g  reaction in 2625 1451
the energy region from threshold to 21 MeV was evaluated by means 2625 1451
of statistical analysis of experimental cross section data [1-16] 2625 1451
and data from GNASH [31-32] calculation.                          2625 1451
    All experimental data if it was possible were renormalized to 2625 1451
the new standards for  monitor reactions cross sections and decay 2625 1451
data. Decay data for the iron-53 were taken from ref. [17]. Recom-2625 1451
mended cross sections for the reaction Al-27(n,p)Mg-27  used as a 2625 1451
monitor  in the measurements  [7-8], [11-14] and [16]  were taken 2625 1451
from new evaluation [18].                                         2625 1451
    Special correction was done with experimental data [1], [6].  2625 1451
    Data of Terrel and Holm [1] measured  at  16.89 and 17.89 MeV 2625 1451
have been corrected introducing the  Fc= 0.5 coefficient, because 2625 1451
of authors used double decreased value for positron yield.        2625 1451
    Data of Andreev et al.[6] were renormalized using preliminary 2625 1451
evaluated integral of cross section value in  the  neutron energy 2625 1451
range from  15.0 to 16.5 MeV.  Renormalised experimental data  of 2625 1451
Andreev et al. are well consistent with Ryves measurements [9-10] 2625 1451
and with new experimental data for  Fe-54(n,2n)Fe-53m+g  reaction 2625 1451
obtained by Fessler [16].                                         2625 1451
    Experimental data from ref.[19-27] were rejected due to their 2625 1451
discrepancy  with  the main bulk of experimental data [1-16]  and 2625 1451
data from theoretical model calculation.                          2625 1451
    The final procedure of evaluation Fe54(n,2n)Fe53m+g excitati- 2625 1451
on function from threshold to 21 MeV  has been carried out within 2625 1451
the framework of generalized least squares method. Rational func- 2625 1451
tion was used as model function [28].                             2625 1451
    U-235 thermal fission [29]  and  Cf-252  spontaneous fission  2625 1451
neutron spectra [30] averaged cross sections calculated  from the 2625 1451
evaluated Fe54(n,2n)Fe53m+g excitation function are the following:2625 1451
                                                                  2625 1451
        ----------------------------------------------            2625 1451
              TYPE OF SPECTRUM     I <CS>, mb (calc.)             2625 1451
        ---------------------------I------------------            2625 1451
         U-235  neutron  fission   I    1.2839E-3                 2625 1451
         CF-252 spontan. fission   I    3.6219E-3                 2625 1451
        ----------------------------------------------            2625 1451
                                                                  2625 1451
                                                                  2625 1451
MT=107 - (n,a) cross section                                      2625 1451
-------------------------------------                             2625 1451
    Excitation function for the  Fe-54(n,a)Cr-51  reaction in the 2625 1451
energy region from  3 MeV to 20 MeV  was  evaluated  by  means of 2625 1451
statistical analysis of experimental cross section data [3], [4], 2625 1451
[11], [15], [21], [34-49] and data from GNASH [31-32] calculation.2625 1451
    All experimental data were renormalized to  the new standards 2625 1451
for monitor reactions cross sections and decay data. Experimental 2625 1451
data of  S.R.Salisbury and R.A.Chalmers [5]  were rejected due to 2625 1451
their inconsistency  with  theoretical model calculations data in 2625 1451
the energy region 3-5 MeV.  Data  of  Y.M.Gledenov et al. [50] at 2625 1451
7 MeV were rejected due to their very big discrepancy with measu- 2625 1451
rements [39,46,49] and data from theoretical model calculation.   2625 1451
    The final procedure of evaluation  Fe-54(n,a)Cr-51 excitation 2625 1451
function from  3 MeV to 20 MeV  have been carried out  within the 2625 1451
framework of generalized least squares method.  Rational function 2625 1451
was used as a model function [28].                                2625 1451
    Integral experimental data for U-235 neutron fission spectrum 2625 1451
[51-53] was used for testing evaluated Fe-54(n,a)Cr-51 excitation 2625 1451
function. The results of testing are given in Table 1.            2625 1451
    Data for  U-235 thermal fission neutron spectrum  and  Cf-252 2625 1451
spontaneous fission neutron spectrum were taken from ref.[29] and 2625 1451
[30],  respectively.                                              2625 1451
                                                     Table 1.     2625 1451
 ---------------------------------------------------------------  2625 1451
    TYPE OF SPECTRUM   I <CS>,MB (calc.) I   <CS>, MB (measured)  2625 1451
 ----------------------I-----------------I----------------------  2625 1451
                       I                 I                        2625 1451
 U-235 neutron fission I    0.8459       I  0.790 +- 0.119 [51]   2625 1451
                       I                 I  0.850 +- 0.100 [52]   2625 1451
                       I                 I  0.891 +- 0.070 [53]   2625 1451
                       I                 I  0.850 +- 0.050 [* ]   2625 1451
                       I                 I                        2625 1451
 ----------------------I-----------------I----------------------  2625 1451
                       I                 I                        2625 1451
 CF-252 spont. fission I    1.1114       I                        2625 1451
                       I                 I                        2625 1451
 ---------------------------------------------------------------  2625 1451
 [*] -   Averaged cross section value obtained from ref. [51-53]  2625 1451
                                                                  2625 1451
MF=33                                                             2625 1451
MT= 16 - (n,2n) cross section cov. matrix                         2625 1451
------------------------------------------                        2625 1451
    Uncertainties in the evaluated  excitation function  for the  2625 1451
reaction Fe-54(n,2n)Fe-53m+g are given in the form of relative co-2625 1451
variance matrix for the 12-neutron energy groups (LB=5).  Covari- 2625 1451
ance matrix of uncertainties  was calculated  simultaneously with 2625 1451
recommended cross section data by means of PADE-2 code [33].      2625 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2625 1451
given in the 33-file are the following:                           2625 1451
                                                                  2625 1451
    3.65349E-06    4.84130E-06    6.69851E-06    1.07852E-05      2625 1451
    2.27150E-05    7.28051E-05    4.72794E-04    5.38548E-03      2625 1451
    1.13987E-02    1.32443E-02    3.40784E-02    1.48451E-01      2625 1451
                                                                  2625 1451
MT=107 - (n,a) cross section cov. matrix                          2625 1451
------------------------------------------                        2625 1451
    Uncertainties in the evaluated  excitation function  for  the 2625 1451
reaction Fe-54(n,a)Cr-51 are given in the form of relative covari-2625 1451
ance matrix for the  17-neutron energy groups (LB=5).  Covariance 2625 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2625 1451
recommended cross section data by means of PADE-2 code [33].      2625 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2625 1451
given in the 33-file are the following:                           2625 1451
                                                                  2625 1451
    1.85377E-09    2.23510E-09    2.38367E-09    2.64130E-09      2625 1451
    3.38901E-09    3.88584E-09    5.57195E-09    9.37333E-09      2625 1451
    2.56042E-08    1.44977E-06    8.05497E-04    1.66197E-03      2625 1451
    2.03640E-03    2.27545E-03    4.21154E-03    1.42809E-02      2625 1451
    2.58358E-02                                                   2625 1451
                                                                  2625 1451
References :                                                      2625 1451
 1. J.Terrell, D.M.Holm  Physical Review, v.109, p.2031, 1958     2625 1451
 2. M.J.Depraz et al.  Journal de Physique, v.21, p.377, May 1960 2625 1451
 3. D.M.Chittenden et al.  Physical Review, v.122, p.860, 1961    2625 1451
 4. W.G.Cross et al. Prog. Report EANDC(CAN)-16, p., January 1963 2625 1451
 3. W.G.Cross et al.  Prog. Report EANDC(CAN)-16, p.1, Jan. 1963  2625 1451
 5. S.R.Salisbury, R.A.Chalmers  Phys. Rev., v.B140, p.305, 1965  2625 1451
 6. M.F.Andreev, V.I.Serov  Jadernaja Fizika (Sov.), v.7, n.4,    2625 1451
    p.745, April 1968                                             2625 1451
 7. S.M.Qaim  Nuclear Physics, v.A185, p.614, May 1972            2625 1451
 8. R.A.Sigg, P.K.Kuroda  Inorg. Nucl. Chem., v.37, p.631, 1975   2625 1451
 9. T.B.Ryves et al.  J. Metrologia, v.14, n.3, p.127, June 1978  2625 1451
10. T.B.Ryves et al.  J. of Physics, pt.G, v.4, p.1783, 1978      2625 1451
11. B.M.Bahal, R.Pepelnik  Report GKSS-84-E-, 1984 ;              2625 1451
    B.M.Bahal, R.Pepelnik  Progress Report NEANDC(E)-252/U,(5),   2625 1451
    p.28, June 1984                                               2625 1451
12. L.R.Greenwood, R.K.Smither  Proc. of Intern. Conf. on Nuclear 2625 1451
    Data for Basic and Applied Science, Santa Fe, New Mexico, USA,2625 1451
    13-17 May 1985, v.1, p.163                                    2625 1451
13. T.Katoh et al.  Report JAERI-M-89-083, 1989                   2625 1451
14. M.Viennot et al.  Nucl. Sci. Eng., v.108, p.289, July 1991    2625 1451
15. A.Ercan et al.  Proc. of International Conference on Nuclear  2625 1451
    Data for Science and Technology, 13-17 May 1991, Julich, FRG, 2625 1451
    Springer-Verlag, 1992, p.376-377                              2625 1451
16. A.Fessler  Report JUL-3502, FZ Julich GmbH, Germany, 1998     2625 1451
17. E.Browne, R.B.Firestone   Table of  Radioactive  Isotopes,    2625 1451
    John Wiley & Sons, New York, 1986 ;                           2625 1451
    R.B.Firestone   Table of Isotopes, Eighth edition, Vol. 1,    2625 1451
    John Wiley & Sons, Inc., New York, 1995                       2625 1451
18. K.I.Zolotarev  Report INDC(CCP)-431, Distr.: J+R/EL, IAEA,    2625 1451
    Vienna, August 2002                                           2625 1451
19. D.L.Allan   J. Proc. Phys. Soc., v.70, p.195, 1956            2625 1451
20. L.A.Rayburn   Phys. Rev., v.122, p.168, 1961                  2625 1451
21. H.Pollehn, H.Neuert  Zeitschrift f. Naturforschung, sect.A,   2625 1451
    v.16, p.227, 1961                                             2625 1451
22. C.Carles  Comptes Rendus, v.257, p.659, July 1963             2625 1451
23. J.Csikai  Report EANDC-50S, v.2, p.102, July 1965             2625 1451
24. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965          2625 1451
25. J.Araminowicz, J.Dresler Rep.INR-1464, Swierk-Warsew,May 1973 2625 1451
26. M.Bormann et al.  Zeitsch. f. Physik, sect.A, v.277, p.203,   2625 1451
    June 1976                                                     2625 1451
27. Zhou Muyao et al. Chinese J. of Nucl. Phys., v.9, p.34, 1987  2625 1451
28. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2625 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2625 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2625 1451
29. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2625 1451
    Library, MAT-9228, MF=5, MT=18, eval. April 1989              2625 1451
30. W.Mannhart  Report IAEA-TECDOC-410, p.158, 1987               2625 1451
31. P.G.Young, E.D.Arthur  A Preequilibrium Statistical Nuclear   2625 1451
    Model Code for Calculation of Cross Section and Emission      2625 1451
    Spectra. Report LA-6947, Los Alamos, 1977                     2625 1451
32. E.L.Trykov, G.Ya.Tertychnyi  Private communication, IPPE,     2625 1451
    Obninsk, May 1999                                             2625 1451
33. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2625 1451
34. P.Venugopala Rao, R.W.Fink  Phys. Rev., v.154, p.1023, 1967   2625 1451
35. S.M.Qaim et al.  Conf. on Chemical Nuclear Data, Measurements 2625 1451
    and Applicat., Univ. of Kent, Canterbury,20-22 September 1971,2625 1451
    p.121                                                         2625 1451
36. J.J.Singh  Trans. Amer. Nucl. Soc., v.15, p.147, June 1972    2625 1451
37. G.N.Maslov et al. Sov. J. Yadernye Konstanty, v.9, p.50, 1972 2625 1451
38. K.Fukuda et al.  Prog. Report NEANDC(J)-56/U, p.44, Sep. 1978 2625 1451
39. A.Paulsen et al.  Nucl. Sci. Eng., v.72, v.1, p.113, 1979     2625 1451
40. O.I.Artem'ev et al.  Atomnaja Energija (Sov.), v.49, no.3,    2625 1451
    p.195, September 1980                                         2625 1451
41. L.R.Greenwood  Progress Report DOE-ER-0046-21, p.15, May 1985 2625 1451
42. J.W.Meadows et al.  Annals of Nuclear Energy, v.14, p.489,    2625 1451
    September 1987                                                2625 1451
43. Y.Ikeda, C.Konno, K.Oishi et al.  Report JAERI-1312, 1988     2625 1451
44. Lu Hanlin et al.  Report INDC(CPR)-16, August 1989            2625 1451
45. S.K.Saraf et al.  Nucl. Sci. Eng., v.107, p.365, April 1991   2625 1451
46. J.W.Meadows, W.Mannhart et.al.  Proc. of Intern. Conference   2625 1451
    on Nuclear Data for Science and Technology, 13-17 May 1991,   2625 1451
    Julich, FRG, Springer-Verlag, 1992, p.288-290                 2625 1451
47. A.Grallert et al.  Report INDC(NDS)-286, p.193, IAEA, 1993    2625 1451
48. J.W.Meadows et al.  Annals of Nuclear Energy, v.23, p.877,    2625 1451
    July 1996                                                     2625 1451
49. Lu Hanlin et al.  Report INDC(CPR)-045, IAEA, Vienna, 1998    2625 1451
50. Y.M.Gledenov et al.  Report INDC(CPR)-042, IAEA, Vienna, 1997 2625 1451
51. H.Braun, L.Nagy  Radiochimica Acta, v.10, p.15, 1968          2625 1451
52. A.V.Bushuev et al.  Atomnaja Energija (Sov.), v.63, no.3,     2625 1451
    p.207, September 1987                                         2625 1451
53. O.Horibe et al.   Proc. of Conference: 50 Years with  Nuclear 2625 1451
    Fission, Washington D.C., 25-28 April 1989, v.2, p.923        2625 1451
******************************************************************2625 1451
********  End of (N,2N), (N,A) bibliographical component ******** 2625 1451
***************************************************************** 2625 1451
******************************************************************2625 1451
*********    Start of (N,P) bibliographical component   ********* 2625 1451
***************************************************************** 2625 1451
                                                                  2625 1451
 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy,    2625 1451
                 P.G. Young (LANL)                                2625 1451
                                                                  2625 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   2625 1451
 MT 102) has been updated for new experimental data at incident   2625 1451
 neutron energies up to 1 keV.                                    2625 1451
 The MF=12, MT=102 yields above 1 keV were adjusted slightly to   2625 1451
 force energy conservation.                                       2625 1451
 The Q-value for radiative capture was also updated in File 3.    2625 1451
 Details of these changes are described in Frankel et al. [Fr01]. 2625 1451
                                                                  2625 1451
******************************************************************2625 1451
                                                                  2625 1451
 ENDF/B-VI MOD 4 Evaluation, September 1997, M.B. Chadwick,       2625 1451
                 P.G. Young (LANL) and A.J. Koning (ECN)          2625 1451
                                                                  2625 1451
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     2625 1451
   in cooperation with ECN Petten.                                2625 1451
                                                                  2625 1451
  This evaluation provides a complete representation of the       2625 1451
nuclear data needed for transport, damage, heating,               2625 1451
radioactivity, and shielding applications over the incident       2625 1451
neutron energy range from 1.0E-11 to 150 MeV.  The discussion     2625 1451
here is divided into the region below and above 20 MeV.           2625 1451
                                                                  2625 1451
INCIDENT NEUTRON ENERGIES < 20 MeV                                2625 1451
                                                                  2625 1451
  Below 20 MeV the evaluation is based completely on the ENDF/B-  2625 1451
VI.1 (Release 1) evaluation by [He91] (see also [Fu86]).  The     2625 1451
following modifications were made to the ENDF/B-VI.1 evaluation:  2625 1451
                                                                  2625 1451
1.  The covariance files (MF=33) were removed from the file.      2625 1451
                                                                  2625 1451
2.  The derived MF=3 files for MT=203,205,207 were removed.       2625 1451
                                                                  2625 1451
                                                                  2625 1451
INCIDENT NEUTRON ENERGIES > 20 MeV                                2625 1451
                                                                  2625 1451
  The evaluation above 20 MeV utilizes MF=6, MT=5 to represent    2625 1451
all reaction data.  Production cross sections and emission        2625 1451
spectra are given for neutrons, protons, deuterons, tritons,      2625 1451
alpha particles, gamma rays, and all residual nuclides produced   2625 1451
(A>5) in the reaction chains. To summarize, the ENDF sections     2625 1451
with non-zero data above En = 20 MeV are:                         2625 1451
                                                                  2625 1451
  MF=3 MT=  1  Total Cross Section                                2625 1451
       MT=  2  Elastic Scattering Cross Section                   2625 1451
       MT=  3  Nonelastic Cross Section                           2625 1451
       MT=  5  Sum of Binary (n,n') and (n,x) Reactions           2625 1451
                                                                  2625 1451
  MF=4 MT=  2  Elastic Angular Distributions                      2625 1451
                                                                  2625 1451
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         2625 1451
               Distributions for Emission Neutrons, Protons,      2625 1451
               Deuterons, and Alphas; and Angle-Integrated        2625 1451
               Spectra for Gamma Rays and Residual Nuclei That    2625 1451
               Are Stable Against Particle Emission               2625 1451
                                                                  2625 1451
  The evaluation is based on nuclear model calculations that      2625 1451
have been benchmarked to experimental data, especially for n +    2625 1451
Fe56 and p + Fe56 reactions [Ch96a]. We use the GNASH code        2625 1451
system [Yo92], which utilizes Hauser-Feshbach statistical,        2625 1451
preequilibrium and direct-reaction theories. Spherical optical    2625 1451
model calculations are used to obtain particle transmission       2625 1451
coefficients for the Hauser-Feshbach calculations, as well as     2625 1451
for the elastic neutron angular distributions.                    2625 1451
  Cross sections and spectra for producing individual residual    2625 1451
nuclei are included for reactions that exceed a cross section of  2625 1451
approximately 1 nb at any energy. The energy-angle-correlations   2625 1451
for all outgoing particles are based on Kalbach systematics       2625 1451
[Ka88].                                                           2625 1451
  A model was developed to calculate the energy distributions of  2625 1451
all recoil nuclei in the GNASH calculations [Ch96b]. The recoil   2625 1451
energy distributions are represented in the laboratory system in  2625 1451
MT=5, MF=6, and are given as isotropic in the lab system. Note    2625 1451
that all other data in MT=5,MF=6 are given in the center-of-mass  2625 1451
system. This method of representation requires a modification of  2625 1451
the original ENDF-6 format.                                       2625 1451
  Preequilibrium corrections were performed in the course of the  2625 1451
GNASH calculations using the exciton model of Kalbach [Ka77,      2625 1451
Ka85], validated by comparison with calculations using Feshbach,  2625 1451
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     2625 1451
nuclear data sheets were matched to continuum level densities     2625 1451
using the formulation of Ignatyuk [Ig75] and pairing and shell    2625 1451
parameters from the Cook [Co67] analysis. Neutron and charged-    2625 1451
particle transmission coefficients were obtained from the         2625 1451
optical potentials, as discussed below.  Gamma-ray transmission   2625 1451
coefficients were calculated using the Kopecky-Uhl model [Ko90].  2625 1451
  The neutron optical model potential of Arthur et al. [Ar80]     2625 1451
was used to calculate transmission coefficients and cross         2625 1451
sections with the SCAT2 code [Be92] up to a neutron energy of 26  2625 1451
MeV.  Between 26 and 52 MeV, the imaginary volume component of    2625 1451
Arthur's potential was modified to better account for nonelastic  2625 1451
cross section measurements, and above 52 MeV the Semmering        2625 1451
potential of Madland [Ma88] was used.  For protons, the           2625 1451
Beccetti-Greenlees potential [Be69] was utilized below 28 MeV,    2625 1451
and the Madland potential [Ma88] was used at higher energies.     2625 1451
The global spherical potential of Perey [Pe63] was utilized for   2625 1451
deuterons, and the potential of Beccetti-Greenlees [Be71]         2625 1451
was adopted for tritons.  Finally, the alpha potential of Lemos   2625 1451
[Le72], as adapted by Arthur et al. [Ar80], was used for alpha    2625 1451
particles.                                                        2625 1451
  Direct reaction cross sections to discrete states were          2625 1451
calculated with the ECIS95 code [Re95] using deformation          2625 1451
parameters compiled in Nuclear Data Sheets.                       2625 1451
  We used the same values for the total, elastic, and nonelastic  2625 1451
cross sections (MF=3, MT=1,2,3) above 20 MeV as were used in the  2625 1451
n + 56Fe evaluation.                                              2625 1451
                                                                  2625 1451
***************************************************************** 2625 1451
                                                                  2625 1451
REFERENCES                                                        2625 1451
                                                                  2625 1451
[Ar80]  E.D. Arthur and P.G. Young, 'Evaluation of Neutron        2625 1451
   Cross Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron   2625 1451
   Cross Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven   2625 1451
   National Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-   2625 1451
   NCS-51245, 1980] p. 731.                                       2625 1451
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182    2625 1451
   1190 (1969)                                                    2625 1451
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees, "Polarization    2625 1451
   Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and       2625 1451
   W. Haeberli, The University of Wisconsin Press, 1971) p.682    2625 1451
[Be92]  O. Bersillon, "SCAT2 - A Spherical Optical Model Code,"   2625 1451
   in Proc. ICTP Workshop on Computation and Analysis of Nuclear  2625 1451
   Data Relevant to Nuclear Energy and Safety, 10 February-13     2625 1451
  March, 1992, Trieste, Italy, to be published in World Scientific2625 1451
   Press, and Progress Report of the Nuclear Physics Division,    2625 1451
   Bruyeres-le-Chatel 1977, CEA-N-2037, p.111 (1978).             2625 1451
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2625 1451
[Ch96a] M.B. Chadwick and P.G. Young, "GNASH Calculations of      2625 1451
   n,p + 54,56,57,58Fe and Benchmarking of Results" in APT        2625 1451
   progress report: 1 August - 1 September 1996, internal Los     2625 1451
   Alamos National Laboratory memo T-2-96/MS-52, 6 Aug. 1996 from 2625 1451
   R.E. MacFarlane to L. Waters.                                  2625 1451
[Ch96b] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      2625 1451
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   2625 1451
   Driven Technologies:  Calculational Method for Heavy Recoils," 2625 1451
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    2625 1451
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996     2625 1451
[Ch99]  M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2625 1451
   National Laboratory report, LA-UR-99-1222 (1999)               2625 1451
[Co67]  J.L. Cook, H. Ferguson, and A.R. deL Musgrove, Aust.J.    2625 1451
   Phys. 20, 477 (1967)                                           2625 1451
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      2625 1451
   National Laboratory Report, LA-13812 (2001).                   2625 1451
[Fu86]  C.Y. Fu and D.M. Hetrick, report ORNL/TM-9964 [ENDF-341]  2625 1451
   (1986)                                                         2625 1451
[He91]  D.M. Hetrick, C.Y. Fu, and N.M. Larson, ENDF/B-VI.1       2625 1451
   Evaluation of n + 54Fe, personal comm. (1991)                  2625 1451
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    2625 1451
   Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485     2625 1451
   (1975)                                                         2625 1451
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     2625 1451
[Ka85]  C. Kalbach, Los Alamos report LA-10248-MS (1985)          2625 1451
[Ka88]  C. Kalbach, Phys.Rev. C 37, 2350 (1988); see also         2625 1451
   C. Kalbach and F.M. Mann, Phys.Rev. C 23, 112 (1981)           2625 1451
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev. C 41, 1941 (1990)        2625 1451
[Le72]  O.F. Lemos, "Diffusion Elastique de Particules Alpha      2625 1451
   de 21 a 29.6 MeV sur des Noyaux de la Region Ti-Zn," Orsay     2625 1451
   report, Series A, No. 136 (1976).                              2625 1451
[Ma88]  D.G. Madland, Proc. OECD/NEANDC Specialist's Meeting on   2625 1451
   Preequilibrium Nuclear Reactions, Semmering, Austria, Feb.     2625 1451
   1988, NEANDC-245 (1988) p.103                                  2625 1451
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      2625 1451
[Re95]  J. Raynal, "Notes on ECIS94," CEA informal                2625 1451
   report,Saclay (1995).                                          2625 1451
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos    2625 1451
   National Laboratory report LA-12343-MS (1992)                  2625 1451
                                                                  2625 1451
 **************************************************************** 2625 1451
                                                                  2625 1451
 ENDF/B-VI MOD 3 Revision, October 1997, V. McLane (NNDC).        2625 1451
                                                                  2625 1451
 1. Corrected residual nucleus in File 6, MT=16,103.              2625 1451
 2. Updated File 1 comments and corrected references.             2625 1451
                                                                  2625 1451
 **************************************************************** 2625 1451
                                                                  2625 1451
 ENDF/B-VI MOD 2 Revision, July 1991, (ORNL)                      2625 1451
                                                                  2625 1451
 The elastic transformation matrix was removed.                   2625 1451
                                                                  2625 1451
 **************************************************************** 2625 1451
                                                                  2625 1451
 ENDF/B-VI MOD 1 Evaluation, November 1989, D.M. Hetrick, C.Y. Fu,2625 1451
                 and N.M. Larson (ORNL)                           2625 1451
                                                                  2625 1451
    This work employed nuclear model codes including the          2625 1451
 Distorted Wave Born Approximation (DWBA) program DWUCK [1]       2625 1451
 and the Hauser-Feshbach code TNG [2,3,4].  The TNG code provides 2625 1451
 energy and angular distributions of particles emitted in the     2625 1451
 compound and pre-compound reactions, ensures consistency among   2625 1451
 all reactions, and maintains energy balance. details pertinent   2625 1451
 to the contents of this evaluation will be published at a later  2625 1451
 date.                                                            2625 1451
    Resonance  parameters and thermal cross sections are from     2625 1451
 Mughabghab [5] and adjusted through SAMMY by N. Larson using J.A.2625 1451
 Harvey data [11] between 110 and 700 keV, capture cross section  2625 1451
 background in File 3 was based on Allen [6].   Above 700 keV,    2625 1451
 54Fe data from J.A. Harvey [11] used.                            2625 1451
                                                                  2625 1451
 DESCRIPTION OF FILES                                             2625 1451
                                                                  2625 1451
 File 1  GENERAL INFORMATION ------------------------------------ 2625 1451
   MT=451 General information, references, and definitions.       2625 1451
                                                                  2625 1451
                                                                  2625 1451
 File 3  NEUTRON CROSS SECTIONS --------------------------------- 2625 1451
   MT=103 (n,p) cross sections                                    2625 1451
      Taken from the GLUCS [8] calculation in which this reaction 2625 1451
        was studied simultaneously with 12 other dosimetry        2625 1451
        reaction cross sections [9].                              2625 1451
                                                                  2625 1451
 File 33  UNCERTAINTY FILES ------------------------------------- 2625 1451
        An LB=8 section is included for all non-derived files as  2625 1451
        required by ENDF/B-VI.                                    2625 1451
   MT=103  (n,p) covariances were taken from the GLUCS [8]        2625 1451
          calculation in which this reaction was studied          2625 1451
          simultaneously with 12 other dosimetry reaction cross   2625 1451
          sections [9].                                           2625 1451
 ---------------------------------------------------------------- 2625 1451
 REFERENCES:                                                      2625 1451
                                                                  2625 1451
 [1] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of Colorado, 2625 1451
     unpublished (1972).                                          2625 1451
 [2] C.Y. Fu, "A Consistent Nuclear Model For Compound and Pre-   2625 1451
     compound Reactions with Conservation of Angular Momentum,"   2625 1451
     report ORNL/TM-7042 (1980), and Nucl.Sci.Eng. 100, 61 (1988).2625 1451
 [3] C.Y Fu, "Development and Application of Multi-step           2625 1451
     Hauser-Feshbach/Pre-equilibrium Model Theory," Symp.         2625 1451
     Neutron Cross Sections from 10 to 50 MeV, Upton, N.Y.,       2625 1451
     May 12-14,1980, report BNL-NCS-51425, p.675.                 2625 1451
 [4] K. Shibata and C.Y. Fu, "Recent Improvements of the TNG      2625 1451
     Statistical Model Code", report ORNL/TM-10093 (1986).        2625 1451
 [5] S.F. Mughabghab, Neutron Cross Sections, Vol. 1: Resonance   2625 1451
     Parameters and Thermal Cross Sections, Part A (Academic      2625 1451
     Press, 1981).                                                2625 1451
 [6] B.J. Allen, Geel Conference, 1977.                           2625 1451
 [7] P.T. Guenther, D.L. Smith, A.B. Smith et al., Ann.Nucl. En.  2625 1451
     13 (11), 601-610 (1986).                                     2625 1451
 [8] D.M. Hetrick and C.Y. Fu, "GLUCS: A Generalized Least Squares2625 1451
     Program for Updating Cross Section Evaluations with          2625 1451
     Correlated Data Sets," report ORNL/TM-7341 [ENDF-303] (1980).2625 1451
 [9] C.Y. Fu and D.M. Hetrick, "Experience in Using the           2625 1451
     Covariances of Some ENDF/B-V Dosimetry Cross Sections:       2625 1451
     Proposed Improvements and Additions of Cross-reaction        2625 1451
     Covariances," Proc. Fourth ASTM-Euratom Symp. on Reactor     2625 1451
     Dosimetry, Gaithersburg, Maryland, March 22-26,1982,         2625 1451
     (US National Bureau Of Standards, 1982) p.877.               2625 1451
([0] S.M. Grimes, R.C. Haight, K.R. Alvar et al., Phys.Rev., C19, 2625 1451
     2127 (1979).                                                 2625 1451
[11] J.A. Harvey, private communication, 1989.                    2625 1451
                                                                  2625 1451
******************************************************************2625 1451
*********     End of (N,P) bibliographical component    ********* 2625 1451
***************************************************************** 2625 1451
 The Q values and threshold energies were updated prior to pro-   2625 1451
 cessing through the codes to comply with the values obtained     2625 1451
 using the NNDC calculation program which is based on the 1995    2625 1451
 Update to the Atomic mass Evaluation.                            2625 1451
                                                                  2625 1451
 File 2 added to the pointwise file containing only the effective 2625 1451
 scattering radius with no resonance parameters given.            2625 1451
 Taken from ENDF/B-VI                                             2625 1451
 ************************ C O N T E N T S *********************** 2625 1451
                                                                  2625 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2625 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2625 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2625 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2625 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2625 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2625 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2625 1451
 ***************** Program FIXUP (Version 2002-1) ****************2625 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2625 1451
 Corrected Thresholds-----------------------------------------Yes 2625 1451
 Extended Cross Sections to 20 MeV----------------------------No  2625 1451
 Allow Cross Section Deletion---------------------------------No  2625 1451
 Allow Cross Section Reconstruction---------------------------No  2625 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2625 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2625 1451
 Deleted Duplicate Points-------------------------------------Yes 2625 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2625 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2625 1451
 Allow Creation of Missing Sections---------------------------No  2625 1451
 Allow Insertion of Energy Points-----------------------------No  2625 1451
 Create Uniform Energy Grid-----------------------------------No  2625 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2625 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2625 1451
 Unshielded Group Averages Using  640 Groups                      2625 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2625 1451
                                1        451        538          12625 1451
                                3         16         25          12625 1451
                                3        103         70          12625 1451
                                3        107         62          12625 1451
                               33         16         21          12625 1451
                               33        103         57          12625 1451
                               33        107         35          12625 1451
                                                                  2625 1  0
                                                                  2625 0  0
 2.60540E+4 5.34762E+1          0          0          0          02625 3 16
-1.33785E+7-1.33785E+7          0          0          1         652625 3 16
         65          1                                            2625 3 16
 13600000.0 .000103649 13700000.0 .000301866 13800000.0 .0005359992625 3 16
 13900000.0 .000884865 14000000.0 .001354120 14100000.0 .0019488152625 3 16
 14200000.0 .002673245 14300000.0 .003530825 14400000.0 .0045239552625 3 16
 14500000.0 .005653915 14600000.0 .006920745 14700000.0 .0083231552625 3 16
 14800000.0 .009858505 14900000.0 .011522700 15000000.0 .0133103002625 3 16
 15100000.0 .015214450 15200000.0 .017227050 15300000.0 .0193389002625 3 16
 15400000.0 .021539650 15500000.0 .023818050 15600000.0 .0261741002625 3 16
 15700000.0 .028571900 15800000.0 .031006167 15900000.0 .0334769002625 3 16
 16000000.0 .035947633 16100000.0 .038418367 16200000.0 .0408891002625 3 16
 16300000.0 .043359833 16400000.0 .045778600 16500000.0 .0481454002625 3 16
 16600000.0 .050512200 16700000.0 .052809775 16800000.0 .0550381252625 3 16
 16900000.0 .057197825 17000000.0 .059288875 17100000.0 .0613041002625 3 16
 17200000.0 .063243500 17300000.0 .065103375 17400000.0 .0668837252625 3 16
 17500000.0 .068583625 17600000.0 .070203075 17700000.0 .0717433752625 3 16
 17800000.0 .073204525 17900000.0 .074571333 18000000.0 .0758438002625 3 16
 18100000.0 .077116267 18200000.0 .078284383 18300000.0 .0793481502625 3 16
 18400000.0 .080411917 18500000.0 .081382933 18600000.0 .0822612002625 3 16
 18700000.0 .083139467 18800000.0 .083937067 18900000.0 .0846540002625 3 16
 19000000.0 .085370933 19100000.0 .086008663 19200000.0 .0865671882625 3 16
 19300000.0 .087125713 19400000.0 .087684238 19500000.0 .0881690752625 3 16
 19600000.0 .088580225 19700000.0 .088991375 19800000.0 .0894025252625 3 16
 19900000.0 .089749640 20000000.0 0.0                             2625 3 16
                                                                  2625 3  0
 2.60540E+4 5.34762E+1          0          0          0          02625 3103
 8.53000E+4 8.53000E+4          0          0          1        1992625 3103
        199          1                                            2625 3103
 690000.000 3.86480E-7 720000.000 2.31888E-6 760000.000 4.63776E-62625 3103
 800000.000 2.08835E-5 840000.000 5.10560E-5 880000.000 8.12286E-52625 3103
 920000.000 .000111401 960000.000 .000141574 1000000.00 .0003158642625 3103
 1100000.00 .000634272 1200000.00 .001003305 1300000.00 .0016760882625 3103
 1400000.00 .002399496 1500000.00 .003358460 1600000.00 .0045529802625 3103
 1700000.00 .006521960 1800000.00 .013137700 1900000.00 .0205279002625 3103
 2000000.00 .027669000 2100000.00 .034561000 2200000.00 .0417467502625 3103
 2300000.00 .050695000 2400000.00 .059937000 2500000.00 .0719424002625 3103
 2600000.00 .086711200 2700000.00 .103066400 2800000.00 .1289400002625 3103
 2900000.00 .156400000 3000000.00 .175595909 3100000.00 .1865277272625 3103
 3200000.00 .188037864 3300000.00 .194305500 3400000.00 .1967550002625 3103
 3500000.00 .207098000 3600000.00 .226334000 3700000.00 .2456455002625 3103
 3800000.00 .265410000 3900000.00 .285250000 4000000.00 .2962360002625 3103
 4100000.00 .298368000 4200000.00 .301528500 4300000.00 .3108600002625 3103
 4400000.00 .321220000 4500000.00 .334012000 4600000.00 .3492360002625 3103
 4700000.00 .364460000 4800000.00 .379684000 4900000.00 .3949080002625 3103
 5000000.00 .405579000 5100000.00 .411697000 5200000.00 .4178150002625 3103
 5300000.00 .423933000 5400000.00 .430051000 5500000.00 .4371370002625 3103
 5600000.00 .445191000 5700000.00 .453245000 5800000.00 .4612990002625 3103
 5900000.00 .469353000 6000000.00 .473782000 6100000.00 .4745860002625 3103
 6200000.00 .475390000 6300000.00 .476194000 6400000.00 .4769980002625 3103
 6500000.00 .477802000 6600000.00 .478606000 6700000.00 .4794100002625 3103
 6800000.00 .480214000 6900000.00 .481018000 7000000.00 .4814590002625 3103
 7100000.00 .481537000 7200000.00 .481615000 7300000.00 .4816930002625 3103
 7400000.00 .481771000 7500000.00 .481849000 7600000.00 .4819270002625 3103
 7700000.00 .482005000 7800000.00 .482083000 7900000.00 .4821610002625 3103
 8000000.00 .482303000 8100000.00 .482509000 8200000.00 .4827150002625 3103
 8300000.00 .482921000 8400000.00 .483127000 8500000.00 .4833330002625 3103
 8600000.00 .483539000 8700000.00 .483745000 8800000.00 .4839510002625 3103
 8900000.00 .484157000 9000000.00 .483918000 9100000.00 .4832340002625 3103
 9200000.00 .482550000 9300000.00 .481866000 9400000.00 .4811820002625 3103
 9500000.00 .480498000 9600000.00 .479814000 9700000.00 .4791300002625 3103
 9800000.00 .478446000 9900000.00 .477762000 10000000.0 .4770780002625 3103
 10100000.0 .476394000 10200000.0 .475710000 10300000.0 .4750260002625 3103
 10400000.0 .474342000 10500000.0 .473600000 10600000.0 .4728000002625 3103
 10700000.0 .472000000 10800000.0 .471200000 10900000.0 .4704000002625 3103
 11000000.0 .469500000 11100000.0 .468500000 11200000.0 .4675000002625 3103
 11300000.0 .466500000 11400000.0 .465500000 11500000.0 .4636200002625 3103
 11600000.0 .460860000 11700000.0 .458100000 11800000.0 .4553400002625 3103
 11900000.0 .452580000 12000000.0 .449127500 12100000.0 .4449825002625 3103
 12200000.0 .440837500 12300000.0 .436692500 12400000.0 .4325475002625 3103
 12500000.0 .428402500 12600000.0 .424257500 12700000.0 .4201125002625 3103
 12800000.0 .415967500 12900000.0 .411822500 13000000.0 .4070590002625 3103
 13100000.0 .401677000 13200000.0 .396295000 13300000.0 .3909130002625 3103
 13400000.0 .385531000 13500000.0 .378722000 13600000.0 .3704860002625 3103
 13700000.0 .362250000 13800000.0 .354014000 13900000.0 .3457780002625 3103
 14000000.0 .338064000 14100000.0 .330872000 14200000.0 .3236800002625 3103
 14300000.0 .316488000 14400000.0 .309296000 14500000.0 .3021040002625 3103
 14600000.0 .294912000 14700000.0 .287720000 14800000.0 .2805280002625 3103
 14900000.0 .273336000 15000000.0 .267023000 15100000.0 .2615890002625 3103
 15200000.0 .256155000 15300000.0 .250721000 15400000.0 .2452870002625 3103
 15500000.0 .239853000 15600000.0 .234419000 15700000.0 .2289850002625 3103
 15800000.0 .223551000 15900000.0 .218117000 16000000.0 .2137710002625 3103
 16100000.0 .210513000 16200000.0 .207255000 16300000.0 .2039970002625 3103
 16400000.0 .200739000 16500000.0 .197481000 16600000.0 .1942230002625 3103
 16700000.0 .190965000 16800000.0 .187707000 16900000.0 .1844490002625 3103
 17000000.0 .181191000 17100000.0 .177933000 17200000.0 .1746750002625 3103
 17300000.0 .171417000 17400000.0 .168159000 17500000.0 .1649010002625 3103
 17600000.0 .161643000 17700000.0 .158385000 17800000.0 .1551270002625 3103
 17900000.0 .151869000 18000000.0 .149438250 18100000.0 .1478347502625 3103
 18200000.0 .146231250 18300000.0 .144627750 18400000.0 .1430242502625 3103
 18500000.0 .141420750 18600000.0 .139817250 18700000.0 .1382137502625 3103
 18800000.0 .136610250 18900000.0 .135006750 19000000.0 .1334032502625 3103
 19100000.0 .131799750 19200000.0 .130196250 19300000.0 .1285927502625 3103
 19400000.0 .126989250 19500000.0 .125385750 19600000.0 .1237822502625 3103
 19700000.0 .122178750 19800000.0 .120575250 19900000.0 .1189717502625 3103
 20000000.0 0.0                                                   2625 3103
                                                                  2625 3  0
 2.60540E+4 5.34762E+1          0          0          0          02625 3107
 8.42740E+5 8.42740E+5          0          0          1        1762625 3107
        176          1                                            2625 3107
 2500000.00 2.10284E-7 2600000.00 6.30852E-7 2700000.00 1.05142E-62625 3107
 2800000.00 1.47199E-6 2900000.00 1.89256E-6 3000000.00 2.52321E-62625 3107
 3100000.00 3.98553E-6 3200000.00 7.02915E-6 3300000.00 1.23952E-52625 3107
 3400000.00 2.09633E-5 3500000.00 3.37715E-5 3600000.00 5.20401E-52625 3107
 3700000.00 7.71963E-5 3800000.00 .000110897 3900000.00 .0001550542625 3107
 4000000.00 .000211864 4100000.00 .000283836 4200000.00 .0003738112625 3107
 4300000.00 .000484983 4400000.00 .000620912 4500000.00 .0007855312625 3107
 4600000.00 .000983134 4700000.00 .001218350 4800000.00 .0014960952625 3107
 4900000.00 .001821495 5000000.00 .002199795 5100000.00 .0026362052625 3107
 5200000.00 .003135735 5300000.00 .003703005 5400000.00 .0043420002625 3107
 5500000.00 .005055840 5600000.00 .005846530 5700000.00 .0067147202625 3107
 5800000.00 .007659540 5900000.00 .008678465 6000000.00 .0097672152625 3107
 6100000.00 .010919900 6200000.00 .012129050 6300000.00 .0133860002625 3107
 6400000.00 .014686800 6500000.00 .016009800 6600000.00 .0173421132625 3107
 6700000.00 .018683738 6800000.00 .020025363 6900000.00 .0213669882625 3107
 7000000.00 .022684150 7100000.00 .023976850 7200000.00 .0252385002625 3107
 7300000.00 .026469100 7400000.00 .027664000 7500000.00 .0288232002625 3107
 7600000.00 .029946275 7700000.00 .031033225 7800000.00 .0320792002625 3107
 7900000.00 .033084200 8000000.00 .034089200 8100000.00 .0350546172625 3107
 8200000.00 .035980450 8300000.00 .036906283 8400000.00 .0377991702625 3107
 8500000.00 .038659110 8600000.00 .039519050 8700000.00 .0403789902625 3107
 8800000.00 .041238930 8900000.00 .042086331 9000000.00 .0429211922625 3107
 9100000.00 .043756054 9200000.00 .044590916 9300000.00 .0454257772625 3107
 9400000.00 .046260639 9500000.00 .047095501 9600000.00 .0479303622625 3107
 9700000.00 .048765224 9800000.00 .049600086 9900000.00 .0504588442625 3107
 10000000.0 .051341509 10100000.0 .052224174 10200000.0 .0531068382625 3107
 10300000.0 .053989503 10400000.0 .054872168 10500000.0 .0557861212625 3107
 10600000.0 .056731364 10700000.0 .057676607 10800000.0 .0586218502625 3107
 10900000.0 .059567093 11000000.0 .060512336 11100000.0 .0614575792625 3107
 11200000.0 .062429903 11300000.0 .063429309 11400000.0 .0644287162625 3107
 11500000.0 .065428122 11600000.0 .066427528 11700000.0 .0674269342625 3107
 11800000.0 .068426341 11900000.0 .069425747 12000000.0 .0704251532625 3107
 12100000.0 .071424559 12200000.0 .072423966 12300000.0 .0734233722625 3107
 12400000.0 .074422778 12500000.0 .075422184 12600000.0 .0764215912625 3107
 12700000.0 .077420997 12800000.0 .078378170 12900000.0 .0792931102625 3107
 13000000.0 .080208050 13100000.0 .081122990 13200000.0 .0820379302625 3107
 13300000.0 .082895300 13400000.0 .083695100 13500000.0 .0844949002625 3107
 13600000.0 .085294700 13700000.0 .086021900 13800000.0 .0866765002625 3107
 13900000.0 .087331100 14000000.0 .087985700 14100000.0 .0885615832625 3107
 14200000.0 .089058750 14300000.0 .089555917 14400000.0 .0899754172625 3107
 14500000.0 .090317250 14600000.0 .090659083 14700000.0 .0909162332625 3107
 14800000.0 .091088700 14900000.0 .091261167 15000000.0 .0913746002625 3107
 15100000.0 .091368817 15200000.0 .091302850 15300000.0 .0912368832625 3107
 15400000.0 .091080733 15500000.0 .090834400 15600000.0 .0905880672625 3107
 15700000.0 .090254433 15800000.0 .089833500 15900000.0 .0894125672625 3107
 16000000.0 .088910100 16100000.0 .088326100 16200000.0 .0877421002625 3107
 16300000.0 .087084817 16400000.0 .086354250 16500000.0 .0856236832625 3107
 16600000.0 .084820888 16700000.0 .083945863 16800000.0 .0830708382625 3107
 16900000.0 .082195813 17000000.0 .081250670 17100000.0 .0802354102625 3107
 17200000.0 .079220150 17300000.0 .078204890 17400000.0 .0771896302625 3107
 17500000.0 .076121279 17600000.0 .074999836 17700000.0 .0738783932625 3107
 17800000.0 .072756950 17900000.0 .071635507 18000000.0 .0705140642625 3107
 18100000.0 .069392621 18200000.0 .068261360 18300000.0 .0671202802625 3107
 18400000.0 .065979200 18500000.0 .064838120 18600000.0 .0636970402625 3107
 18700000.0 .062555960 18800000.0 .061414880 18900000.0 .0602738002625 3107
 19000000.0 .059132720 19100000.0 .057991640 19200000.0 .0568819102625 3107
 19300000.0 .055803530 19400000.0 .054725150 19500000.0 .0536467702625 3107
 19600000.0 .052568390 19700000.0 .051517117 19800000.0 .0504929502625 3107
 19900000.0 .049468783 20000000.0 0.0                             2625 3107
                                                                  2625 3  0
                                                                  2625 0  0
 2.60540E+4 5.34762E+1          0          0          0          1262533 16
 0.000000+0 0.000000+0          0         16          0          1262533 16
 0.000000+0 0.000000+0          1          5        105         14262533 16
 1.000000-5 1.360000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7262533 16
 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.900000+7262533 16
 2.000000+7 2.100000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533 16
 0.000000+0 0.000000+0 0.000000+0 1.385970-1 2.336600-2 1.635960-2262533 16
 1.098730-2 6.859980-3 3.890130-3 2.017890-3 1.086200-3 8.687900-4262533 16
 1.423320-3 2.801040-3 4.345230-3 2.881570-2 1.450720-2 8.110020-3262533 16
 4.285820-3 2.078820-3 9.630210-4 5.664710-4 6.084130-4 1.088950-3262533 16
 1.903710-3 2.652220-3 1.135380-2 7.712790-3 4.909480-3 2.752250-3262533 16
 1.331250-3 5.993020-4 4.183060-4 8.557850-4 1.979310-3 3.264690-3262533 16
 6.479190-3 4.726220-3 3.163780-3 1.948420-3 1.163340-3 7.775520-4262533 16
 7.826550-4 1.330260-3 2.129450-3 4.106440-3 3.245180-3 2.440780-3262533 16
 1.770330-3 1.272470-3 8.389780-4 6.427510-4 6.994730-4 3.070150-3262533 16
 2.674900-3 2.198150-3 1.696700-3 9.962081-4 2.498150-4-2.924550-4262533 16
 2.643570-3 2.369400-3 1.959520-3 1.217450-3 2.525980-4-5.803090-4262533 16
 2.305690-3 2.046740-3 1.465260-3 6.096450-4-1.909900-4 1.992790-3262533 16
 1.710520-3 1.215950-3 7.055910-4 2.041300-3 2.371640-3 2.606230-3262533 16
 3.933320-3 5.314650-3 7.812890-3                                 262533 16
                                                                  262533  0
 2.60540E+4 5.34760E+1          0          0          0          1262533103
 0.0000E+00 0.0000E+00          0        103          0          4262533103
 0.0000E+00 0.0000E+00          1          5         78         12262533103
 1.0000E-05 6.9000E+05 1.0000E+06 2.0000E+06 4.0000E+06 6.0000E+06262533103
 7.0000E+06 8.0000E+06 1.2000E+07 1.4000E+07 1.6000E+07 2.0000E+07262533103
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00262533103
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 3.0636E-02262533103
 7.4355E-03 1.0657E-04 2.9882E-04 1.8815E-04 1.8663E-04 1.6565E-04262533103
 2.3528E-04 1.7206E-04 6.2483E-04 6.4762E-03 3.5286E-04 2.8481E-04262533103
 2.1734E-04 2.1479E-04 1.6072E-04 1.3802E-04 1.0328E-04 1.8866E-04262533103
 1.0632E-03 3.3244E-04 2.2580E-04 2.2490E-04 1.8391E-04 6.6682E-05262533103
 5.3549E-05 7.1187E-05 7.5034E-04 3.3845E-04 3.3483E-04 2.3650E-04262533103
 9.7942E-05 7.7582E-05 5.3421E-05 6.7111E-04 5.1950E-04 2.9475E-04262533103
 3.7441E-05 3.7615E-05 9.3780E-05 6.4010E-04 2.9008E-04 4.1529E-05262533103
 4.1092E-05 9.1147E-05 6.1427E-04 1.4060E-04 1.1416E-04-1.7842E-05262533103
 2.0209E-03 1.0198E-03 1.0894E-03 1.1388E-03 7.7129E-04 9.9624E-03262533103
 0.0000E+00 0.0000E+00         12          4         48         24262533103
 1.0000E-05 0.0000E+00 6.9000E+05-9.0000E-02 1.0000E+06-1.0000E-01262533103
 2.0000E+06-2.7000E-01 4.0000E+06-2.5000E-01 6.0000E+06-1.0000E+00262533103
 7.0000E+06-1.0000E+00 8.0000E+06-3.1000E-01 1.2000E+07-2.9000E-01262533103
 1.4000E+07-3.3000E-01 1.6000E+07-1.1000E-01 2.0000E+07 0.0000E+00262533103
 1.0000E-05 0.0000E+00 5.0000E+05 1.7503E-01 1.0000E+06 8.0475E-02262533103
 2.0000E+06 3.2607E-02 4.0000E+06 2.7392E-02 6.0000E+06 2.5906E-02262533103
 7.0000E+06 2.5300E-02 8.0000E+06 2.4784E-02 1.2000E+07 4.4955E-02262533103
 1.4000E+07 3.3747E-02 1.6000E+07 9.9812E-02 2.0000E+07 0.0000E+00262533103
 0.000000+0 0.000000+0          0          1         88         44262533103
 1.000000-5 0.000000+0 6.900000+5 2.481500-3 8.000000+5 2.481500-3262533103
 1.000000+6 5.828500-4 1.250000+6 5.828500-4 1.500000+6 5.828500-4262533103
 1.750000+6 5.828500-4 2.000000+6 2.583600-4 2.250000+6 2.583600-4262533103
 2.500000+6 2.583600-4 2.750000+6 2.583600-4 3.000000+6 2.583600-4262533103
 3.220000+6 2.583600-4 3.250000+6 2.583600-4 3.350000+6 2.583600-4262533103
 3.360000+6 2.583600-4 3.400000+6 2.583600-4 3.450000+6 2.583600-4262533103
 3.500000+6 2.583600-4 3.750000+6 2.583600-4 4.000000+6 1.688300-4262533103
 4.250000+6 1.688300-4 4.500000+6 1.688300-4 5.000000+6 1.688300-4262533103
 5.500000+6 1.688300-4 6.000000+6 6.040000-4 7.000000+6 5.760900-4262533103
 8.000000+6 1.713800-4 8.500000+6 1.713800-4 9.000000+6 1.713800-4262533103
 9.500000+6 1.713800-4 1.000000+7 1.713800-4 1.050000+7 1.713800-4262533103
 1.100000+7 1.713800-4 1.150000+7 1.713800-4 1.200000+7 5.274600-4262533103
 1.250000+7 5.274600-4 1.300000+7 5.274600-4 1.350000+7 5.274600-4262533103
 1.400000+7 3.382400-4 1.500000+7 3.382400-4 1.600000+7 9.862800-4262533103
 1.800000+7 9.862800-4 2.000000+7 0.000000+0                      262533103
 0.000000+0 0.000000+0          0          8         88         44262533103
 1.000000-5 0.000000+0 6.900000+5 2.35040-15 8.000000+5 1.71840-12262533103
 1.000000+6 1.95530-11 1.250000+6 2.22450-10 1.500000+6 1.147100-9262533103
 1.750000+6 1.375300-8 2.000000+6 2.922700-8 2.250000+6 7.618500-8262533103
 2.500000+6 1.917900-7 2.750000+6 5.167200-7 3.000000+6 9.244100-7262533103
 3.220000+6 1.009500-6 3.250000+6 1.023200-6 3.350000+6 1.076900-6262533103
 3.360000+6 1.086200-6 3.400000+6 1.080800-6 3.450000+6 1.070700-6262533103
 3.500000+6 1.331000-6 3.750000+6 2.066100-6 4.000000+6 1.708300-6262533103
 4.250000+6 1.882200-6 4.500000+6 2.511700-6 5.000000+6 3.306500-6262533103
 5.500000+6 3.915200-6 6.000000+6 1.542800-5 7.000000+6 1.456600-5262533103
 8.000000+6 4.226500-6 8.500000+6 4.273700-6 9.000000+6 4.251200-6262533103
 9.500000+6 4.188000-6 1.000000+7 4.185100-6 1.050000+7 4.171800-6262533103
 1.100000+7 4.161300-6 1.150000+7 3.840400-6 1.200000+7 1.040900-5262533103
 1.250000+7 9.461600-6 1.300000+7 8.428100-6 1.350000+7 7.345100-6262533103
 1.400000+7 3.401000-6 1.500000+7 2.165000-6 1.600000+7 3.813700-6262533103
 1.800000+7 2.035300-6 2.000000+7 0.000000+0                      262533103
                                                                  262533  0
 2.60540E+4 5.34762E+1          0          0          0          1262533107
 0.000000+0 0.000000+0          0        107          0          1262533107
 0.000000+0 0.000000+0          1          5        190         19262533107
 1.000000-5 2.500000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6262533107
 9.000000+6 1.000000+7 1.100000+7 1.200000+7 1.300000+7 1.400000+7262533107
 1.450000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7262533107
 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107
 0.000000+0 1.967980-2 9.633730-3 1.642170-3-7.873680-4 1.763730-4262533107
 1.190170-3 1.391200-3 9.990920-4 4.107560-4-7.196720-6-7.635850-5262533107
 2.317610-5 3.165150-4 6.879890-4 6.296170-4-2.098400-4-1.980280-3262533107
 6.920690-3 2.102260-3-7.406730-5 2.004840-4 8.283510-4 1.023790-3262533107
 8.128000-4 4.201790-4 9.153950-5-9.428760-6 2.000020-5 1.747130-4262533107
 4.071900-4 4.020070-4-9.273340-5-1.202550-3 1.693190-3 1.036540-3262533107
 6.190390-4 4.273150-4 3.244890-4 2.414200-4 1.586830-4 8.148000-5262533107
 3.617670-5 1.695870-5 9.143770-6 3.129460-5 7.794820-5 1.265680-4262533107
 1.582870-4 1.363610-3 1.012880-3 5.698220-4 2.544810-4 7.642630-5262533107
 1.509480-6-8.578530-6 3.519450-6 1.342820-5 2.123660-5 1.402950-5262533107
-8.208480-6-2.787740-5-2.663740-5 1.050540-3 8.480750-4 5.713450-4262533107
 3.081410-4 1.087930-4 6.133330-6-4.996150-6 1.701860-5 7.081030-5262533107
 1.229820-4 8.175890-5-1.015020-4-4.373240-4 9.279640-4 8.104860-4262533107
 5.821060-4 3.289850-4 1.294250-4 4.657870-5 3.281980-5 6.404150-5262533107
 1.406320-4 1.699830-4 6.664270-5-2.138110-4 8.554680-4 7.401580-4262533107
 5.261530-4 2.858140-4 1.331290-4 6.311590-5 2.314560-5 7.289320-5262533107
 2.015140-4 3.382760-4 4.238150-4 7.539170-4 6.328030-4 4.126190-4262533107
 2.242130-4 1.093590-4-4.162330-6-8.076930-6 1.879820-4 5.566300-4262533107
 1.043390-3 6.135990-4 4.640330-4 2.936590-4 1.706290-4 2.345590-5262533107
-2.956160-5 1.618240-4 6.136620-4 1.290690-3 4.150570-4 3.211380-4262533107
 2.403130-4 1.283090-4 6.183300-5 1.676890-4 4.809320-4 9.913180-4262533107
 3.109670-4 2.900050-4 2.492490-4 2.077720-4 2.190780-4 3.096660-4262533107
 4.856670-4 3.172600-4 3.412840-4 3.374690-4 2.897020-4 2.071610-4262533107
 1.028700-4 4.633580-4 5.410820-4 4.564190-4 1.841470-4-2.618490-4262533107
 7.451530-4 7.686590-4 5.507190-4 8.915960-5 1.151220-3 1.553840-3262533107
 1.941310-3 3.226410-3 5.491560-3 1.062090-2                      262533107
                                                                  262533  0
                                                                  2625 0  0
                                                                     0 0  0
 2.60560E+4 5.54544E+1          0          0         34         102631 1451
 0.0        0.0                 0          0          0          62631 1451
 1.00000E+0 2.00000E+7          0          0         10       20022631 1451
 3.00000E+2 0.0                 1          0        261          32631 1451
 26-Fe- 56 FEI        EVAL-Oct02 K.I.Zolotarev                    2631 1451
                      DIST-Feb2004                                2631 1451
----IRDF-2002         MATERIAL 2631                               2631 1451
-----INCIDENT NEUTRON DATA                                        2631 1451
------ENDF-6 FORMAT                                               2631 1451
***************************************************************** 2631 1451
 26-FE- 56 FEI        EVAL-Oct02 K.I.Zolotarev                    2631 1451
                      DIST-Nov03                       20031105   2631 1451
----BROND-3           MATERIAL 2631         Revision 2            2631 1451
-----INCIDENT NEUTRON DATA                                        2631 1451
------ENDF-6 FORMAT                                               2631 1451
------Russian Reactor Dosimetry File RRDF-2002                    2631 1451
***************************************************************** 2631 1451
Author  of evaluation: K.I.Zolotarev                              2631 1451
***************************************************************** 2631 1451
MF=3                                                              2631 1451
MT=103 -(n,p) cross section                                       2631 1451
-------------------------------------                             2631 1451
    Excitation function for the  Fe-56(n,p)Mn-56  reaction in the 2631 1451
energy region from threshold to 20 MeV  was evaluated by means of 2631 1451
statistical analysis of experimental cross section data [1-40].   2631 1451
    The energy dependence of cross-section  from  4.0 MeV  to the 2631 1451
threshold was extrapolated with  L=0  penetrability function  for 2631 1451
the outgoing  p + Mn-56 channel [41].                             2631 1451
    Analised  microscopic experimental data [1-40], [42-69]  were 2631 1451
renormalized if it were possible to the new recommended standards 2631 1451
for monitor reaction cross sections and decay data.               2631 1451
    Special correction was done with experimental data [2],[9-10] 2631 1451
and [17]. Experimental data of Terrell and Holm [2] corresponding 2631 1451
to neutron energies 6.54 MeV,7.41 MeV, 8.21 MeV were renormalized 2631 1451
to the preliminary evaluated  cross section value at  En=8.2 MeV. 2631 1451
Data of  Liskien and Paulsen [9]  measured  in  the energy range  2631 1451
12.60 - 19.58 MeV  were corrected  to  the preliminary  evaluated 2631 1451
integral of cross section in the energy interval 14-15 MeV.  Data 2631 1451
obtained by Liskien and Paulsen in the energy range 6.06-8.20 MeV 2631 1451
[10] were renormalized to the preliminary evaluated cross section 2631 1451
value at 8.0 MeV.  The correction factors  for  the  experimental 2631 1451
data [2], [9] and [10] were Fc=1.23487, Fc=1.09674 and Fc=1.05835,2631 1451
respectively. Data of Smith and Meadows [17] obtained in the expe-2631 1451
riment with neutrons from  D(d,n)He3 reaction in the energy range 2631 1451
6.486 - 9.945 MeV were renormalized  to the preliminary evaluated 2631 1451
cross section value at 9.945 MeV.  The correction factor for this 2631 1451
data was Fc=1.14066 .                                             2631 1451
    Experimental information about Fe-56(n,p)Mn-56 reaction exci- 2631 1451
tation function are given in the ref. [13],[21],[23] in the form  2631 1451
of cross section ratios to monitor reactions. The results of pre- 2631 1451
cise relative measurements  of Vonach et al. [13]  in  the energy 2631 1451
range 13.6-14.7 MeV were normalized  to the preliminary evaluated 2631 1451
absolute cross section value 107.13 mb  for Fe-56(n,p)Mn-56 reac- 2631 1451
tion at 14.7 MeV point. Raics et al. [21]  and  Antov et al. [23] 2631 1451
measured ratios of  Fe-56(n,p)Mn-56 cross section  to  U-238(n,f) 2631 1451
and Al-27(n,a)Na-24 reactions cross section. Recommended absolute 2631 1451
cross section data for U-238(n,f)  and  Al-27(n,a)Na-24 reactions 2631 1451
were taken from ref. [70] and [71], respectively.                 2631 1451
    Experimental data from refs.[5] and [11] were used partially. 2631 1451
It were used only data obtained at 14.5 MeV [5] and in the energy 2631 1451
range 3.95 - 10.0 MeV [11]. Cross et al. data [5] for the neutron 2631 1451
energies  13.78 MeV, 14.07 MeV and 14.73 MeV  were  rejected  due 2631 1451
to their  systematically  underestimation  Fe-56(n,p)Mn-56  cross 2631 1451
section.  The result of Grundl measurements at 14.1 MeV [11]  was 2631 1451
not taken into account due to a very  overestimated cross section 2631 1451
value obtained for this energy point.                             2631 1451
    Experimental cross section data [42-69] were rejected  due to 2631 1451
their discrepancy with the main bulk of experimental data [1-40]. 2631 1451
In  the  rejected  experiments  [42-51], [53], [56-58], [60-62],  2631 1451
[64-67] and [69] the cross section values were measured only in a 2631 1451
one energy point in the interval 14 - 15 MeV.                     2631 1451
    Statistical analysis of input cross section data was carried  2631 1451
out by means of PADE-2 code [72].  Rational function was used as  2631 1451
the model function [73].                                          2631 1451
    U-235 thermal fission [74]  and  Cf-252 spontaneous  fission  2631 1451
neutron spectra [75] averaged cross-sections calculated from the  2631 1451
evaluated Fe-56(n,p)Mn-56 excitation function are the following:  2631 1451
                                                                  2631 1451
  2631 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2631 1451
  2631 1451
 U-235 neutron fission     1.0997         1.09  +- 0.04   [76]  2631 1451
                                          1.13  +- 0.07   [77]  2631 1451
                                          1.083 +- 0.017  [78]  2631 1451
  2631 1451
 Cf-252 spont. fission     1.4626         1.471 +- 0.025  [79]  2631 1451
                                          1.465 +- 0.026  [80]  2631 1451
  2631 1451
                                                                  2631 1451
MT=33                                                             2631 1451
MT=103 -(n,p) cross section cov. matrix                           2631 1451
---------------------------------------                           2631 1451
    Uncertainties in the evaluated  excitation function  for  the 2631 1451
reaction Fe-56(n,p)Mn-56 are given in the form of relative covari-2631 1451
ance matrix for the  25-neutron energy groups (LB=5).  Covariance 2631 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  2631 1451
recommended cross section data by means of PADE-2 code.           2631 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2631 1451
given in the 33-file are the following:                           2631 1451
                                                                  2631 1451
    4.35195E-09    4.49367E-09    4.76919E-09    5.19445E-09      2631 1451
    5.64595E-09    6.15408E-09    7.11594E-09    7.89561E-09      2631 1451
    9.40785E-09    1.19394E-08    1.48739E-08    1.86574E-08      2631 1451
    2.18905E-08    1.32640E-06    2.77587E-04    3.86559E-04      2631 1451
    6.52483E-04    1.02062E-03    1.26078E-03    1.33119E-03      2631 1451
    1.81407E-03    2.05264E-03    3.00443E-03    4.57086E-03      2631 1451
    8.12954E-03                                                   2631 1451
                                                                  2631 1451
 References :                                                     2631 1451
 1. G.Brown  Philosophical Magazine, v.2, p.785, 1957             2631 1451
 2. J.Terrell, D.M.Holm  Phys. Rev., v.109, p.2031, 1958          2631 1451
 3. H.Pollehn, H.Neuert  Zeitschrift f. Naturforschung, sect. A,  2631 1451
    v.16, p.227, 1961                                             2631 1451
 4. F.Gabbard, B.D.Kern  Phys. Rev., v.128, p.1276, 1962          2631 1451
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13. H.K.Vonach et al. Proc. of 2nd Conference on Nuclear Cross-   2631 1451
    Sections and Technology, Washington D.C., 4-7 March 1968,     2631 1451
    v.2, p.885                                                    2631 1451
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20. Ju.A.Nemilov, Ju.N.Trofimov  Progress Report YFI-26, p.25,    2631 1451
    November 1978                                                 2631 1451
21. P.Raics et al.  Proc. of the 5th All Union Conference on      2631 1451
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    Reference Data, Geel, Belgium, 1984                           2631 1451
25. B.M.Bahal, R.Pepelnik  Report GKSS-84-E-, 1984                2631 1451
26. I.Garlea et al. Proc. of the 14-th International Symposium on 2631 1451
    Nuclear Physics, Gaussig, GDR, 19-23 November 1984, ZFK-562,  2631 1451
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28. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2631 1451
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30. S.Cabral, G.Boerker, H.Klein, W.Mannhart  Nucl. Sci. Eng.,    2631 1451
    v.106, p.308, 1990                                            2631 1451
31. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990    2631 1451
32. S.K.Saraf et al.  Nucl. Sci. Eng., v.107, p.365, April 1991   2631 1451
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34. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24, Aug.1992 2631 1451
35. W.Mannhart, G.Boerker  Progress Report, INDC(Ger)-036/L, p.59 2631 1451
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36. Bao Zongyu et al. China J. Nucl. Phys., v.15,no.4,p.341, 1993 2631 1451
37. Y.Ikeda et al.  Journal of Nuclear Science and Technology,    2631 1451
    v.30, n.9, pp.870-880, September 1993                         2631 1451
38. Lu Hanlin et al.  Report INDC(CPR)-045, IAEA, October 1998    2631 1451
39. A.A.Filatenkov et al.  Report, RI-252, May 1999 ;             2631 1451
    A.A.Filatenkov et al.  VANT, Ser.: Yadernye Konstanty, v.2,   2631 1451
    p.8, Moscow, 1996                                             2631 1451
40. A.Fessler Report JUL-3502, FZ Julich GmbH, Germany, 1998 ;    2631 1451
    A.Fessler et al.  Nucl. Sci. Eng., v.134, no.2, p.171, 2000   2631 1451
41. S.A.Badikov, A.B.Pashchenko Voprosy Atomnoy Nauki i Tekhniki  2631 1451
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43. E.B.Paul, R.L.Clarke  Canadian J. Phys., v.31, p.267, 1953    2631 1451
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45. D.L.Allan  Proc. Physical Society, section A, v.70, p.195,    2631 1451
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46. S.Yasumi  J. Phys. Soc. Japan, v.12, p.443, May 1957          2631 1451
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48. B.D.Kern et al.  Nucl. Phys., v.10, p.226, May 1959           2631 1451
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50. R.S.Storey et al.  Proc. Phys. Soc., v.75, p.526, 1960        2631 1451
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53. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965          2631 1451
54. J.D.Hemingway et al.  Proc. Royal Society, section A, v.292,  2631 1451
    p.180, May 1966                                               2631 1451
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56. S.M.Qaim et al. Proc. of Conference on Chemical Nuclear Data, 2631 1451
    Measurements and Applicat., Univ. of Kent, Canterbury, 20-22  2631 1451
    September 1971, p.121                                         2631 1451
57. J.J.Singh  Trans. Amer. Nucl. Soc., v.15, p.147, June 1972    2631 1451
58. R.Spangler et al. Ann. Nucl. Sci., v.22, p.818, November 1975 2631 1451
59. A.B.M.G.Mostafa  Nuclear Science and Applications, Series B:  2631 1451
    Physical sciences, v.9, p.10, October  1976                   2631 1451
60. Z.A.Ramendik et al.  Atomnaja Energija (Sov.), v.42, no.2,    2631 1451
    p.136, February 1977                                          2631 1451
61. S.Sothras  Dissertation Abstracts, v.B38, p.280, July 1978    2631 1451
62. D.Sharma et al. Proc. of the 21-st Nuclear Physics and Solid  2631 1451
    State Physics Symposium, Bombay, India, 28-31 December 1978 , 2631 1451
    v.2, p.349                                                    2631 1451
63. P.N.Ngoc et al. Progress Report INDC(HUN)-20,p.3, March 1983; 2631 1451
    P.N.Ngoc et al.  Nucleonika (Pol), v.29, p.87, 1984           2631 1451
64. P.N.Ngoc et al.  Report INDC(VN)-2, November 1983             2631 1451
65. I.Garlea et al. Rev. Roum. Phys., v.29, p.421, 1984           2631 1451
66. J.P.Gupta et al.  Pramana, v.24, p.637, 1985                  2631 1451
67. M.Viennot et al.  Nucl. Sci. Eng., v.108, p.289, July 1991    2631 1451
68. A.Ercan et al.  Proc. of an Int. Conference on Nuclear Data   2631 1451
    for Science and Technology, Juelich, FRG, 13-17 May 1991,     2631 1451
    Springer-Verlag, 1992                                         2631 1451
69. L.I.Klochkova et al.  Voprosy Atomnoy Nauki i Techniki, Ser.: 2631 1451
    Yadernye Konstanty, v.1, p.27, 1992                           2631 1451
70. ENDF/B-VI library,MAT=9237, eval. L.W.Weston et al. Nov.1989; 2631 1451
    H.Conde editor   Nuclear Data Standards for Nuclear Measu-    2631 1451
    rements, Report NEANDC-311 U, pp.70-74, OECD, Paris, 1992.    2631 1451
71. M.Wagner et al. Evaluation of cross sections for 14 important 2631 1451
    neutron-dosimetry reactions. Physics Data No.13-5, 1990       2631 1451
72. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2631 1451
73. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2631 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2631 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2631 1451
74. L.W.Weston  et al.  Evaluated Neutron Data for Uranium-235,   2631 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    2631 1451
75. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2631 1451
76. W.Mannhart  Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry,  2631 1451
    Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985    2631 1451
77. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2631 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 2631 1451
78. W.Mannhart  Progress Report INDC(Ger)-045, pp.40-43, 1999     2631 1451
79. W.Mannhart  Handbook on Nuclear Activation Cross Sections ,   2631 1451
    IAEA Tech. Report Ser. No.273, p.413, 1987                    2631 1451
80. W.Mannhart  Validation of Differential Cross Sections with    2631 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2631 1451
    September 2002                                                2631 1451
***************************************************************** 2631 1451
 The Q values and threshold energies were updated prior to pro-   2631 1451
 cessing through the codes to comply with the values obtained     2631 1451
 using the NNDC calculation program which is based on the 1995    2631 1451
 Update to the Atomic mass Evaluation.                            2631 1451
                                                                  2631 1451
 File 2 added to the pointwise file containing only the effective 2631 1451
 scattering radius with no resonance parameters given.            2631 1451
 Taken from ENDF/B-VI                                             2631 1451
***************************************************************** 2631 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2631 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2631 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2631 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2631 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2631 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2631 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2631 1451
 ***************** Program FIXUP (Version 2002-1) ****************2631 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2631 1451
 Corrected Thresholds-----------------------------------------Yes 2631 1451
 Extended Cross Sections to 20 MeV----------------------------No  2631 1451
 Allow Cross Section Deletion---------------------------------No  2631 1451
 Allow Cross Section Reconstruction---------------------------No  2631 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2631 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2631 1451
 Deleted Duplicate Points-------------------------------------Yes 2631 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2631 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2631 1451
 Allow Creation of Missing Sections---------------------------No  2631 1451
 Allow Insertion of Energy Points-----------------------------No  2631 1451
 Create Uniform Energy Grid-----------------------------------No  2631 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2631 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2631 1451
 Unshielded Group Averages Using  640 Groups                      2631 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2631 1451
                                1        451        268          12631 1451
                                3        103         61          12631 1451
                               33        103         66          12631 1451
                                                                  2631 1  0
                                                                  2631 0  0
 2.60560E+4 5.54544E+1          0         99          0          02631 3103
-2.91310E+6-2.91310E+6          0          0          1        1722631 3103
        172          1                                            2631 3103
 2900000.00 5.9516E-12 3000000.00 8.5021E-11 3100000.00 1.8579E-102631 3103
 3200000.00 2.8657E-10 3300000.00 3.8734E-10 3400000.00 4.8811E-102631 3103
 3500000.00 3.19099E-9 3600000.00 2.37217E-8 3700000.00 2.30222E-72631 3103
 3800000.00 8.58242E-7 3900000.00 3.26945E-6 4000000.00 9.50803E-62631 3103
 4100000.00 2.13588E-5 4200000.00 4.12059E-5 4300000.00 7.22242E-52631 3103
 4400000.00 .000118637 4500000.00 .000186046 4600000.00 .0002818582631 3103
 4700000.00 .000415817 4800000.00 .000600658 4900000.00 .0008528442631 3103
 5000000.00 .001193315 5100000.00 .001648055 5200000.00 .0022481552631 3103
 5300000.00 .003028750 5400000.00 .004026080 5500000.00 .0052718752631 3103
 5600000.00 .006784800 5700000.00 .008560220 5800000.00 .0105616952631 3103
 5900000.00 .012519200 6000000.00 .014238450 6100000.00 .0161200002631 3103
 6200000.00 .018181150 6300000.00 .020072250 6400000.00 .0217811502631 3103
 6500000.00 .023826650 6600000.00 .025746950 6700000.00 .0270865002631 3103
 6800000.00 .028396700 6900000.00 .029706900 7000000.00 .0310583252631 3103
 7100000.00 .032450975 7200000.00 .033907625 7300000.00 .0354282752631 3103
 7400000.00 .037005975 7500000.00 .038640725 7600000.00 .0403027402631 3103
 7700000.00 .041992020 7800000.00 .043681300 7900000.00 .0453705802631 3103
 8000000.00 .047059860 8100000.00 .048700150 8200000.00 .0502914502631 3103
 8300000.00 .051829775 8400000.00 .053315125 8500000.00 .0547361002631 3103
 8600000.00 .056092700 8700000.00 .057449300 8800000.00 .0587423902631 3103
 8900000.00 .059971970 9000000.00 .061201550 9100000.00 .0624311302631 3103
 9200000.00 .063660710 9300000.00 .064888420 9400000.00 .0661142602631 3103
 9500000.00 .067340100 9600000.00 .068565940 9700000.00 .0697917802631 3103
 9800000.00 .071069213 9900000.00 .072398238 10000000.0 .0737272632631 3103
 10100000.0 .075056288 10200000.0 .076448125 10300000.0 .0779027752631 3103
 10400000.0 .079357425 10500000.0 .080812075 10600000.0 .0823283602631 3103
 10700000.0 .083906280 10800000.0 .085484200 10900000.0 .0870621202631 3103
 11000000.0 .088640040 11100000.0 .090238714 11200000.0 .0918581432631 3103
 11300000.0 .093477571 11400000.0 .095097000 11500000.0 .0967164292631 3103
 11600000.0 .098335857 11700000.0 .099955286 11800000.0 .1014997502631 3103
 11900000.0 .102969250 12000000.0 .104438750 12100000.0 .1059082502631 3103
 12200000.0 .107264833 12300000.0 .108508500 12400000.0 .1097521672631 3103
 12500000.0 .110859000 12600000.0 .111829000 12700000.0 .1127990002631 3103
 12800000.0 .113632500 12900000.0 .114329500 13000000.0 .1149057502631 3103
 13100000.0 .115361250 13200000.0 .115689750 13300000.0 .1158912502631 3103
 13400000.0 .115996000 13500000.0 .115908000 13600000.0 .1157240002631 3103
 13700000.0 .115415750 13800000.0 .114983250 13900000.0 .1144076672631 3103
 14000000.0 .113689000 14100000.0 .112970333 14200000.0 .1121030002631 3103
 14300000.0 .111087000 14400000.0 .110071000 14500000.0 .1089375002631 3103
 14600000.0 .107686500 14700000.0 .106435500 14800000.0 .1050908752631 3103
 14900000.0 .103652625 15000000.0 .102214375 15100000.0 .1007761252631 3103
 15200000.0 .099278768 15300000.0 .097722305 15400000.0 .0961658412631 3103
 15500000.0 .094609377 15600000.0 .093052914 15700000.0 .0914964502631 3103
 15800000.0 .089939986 15900000.0 .088383523 16000000.0 .0868270592631 3103
 16100000.0 .085270595 16200000.0 .083714132 16300000.0 .0822083502631 3103
 16400000.0 .080753250 16500000.0 .079298150 16600000.0 .0778430502631 3103
 16700000.0 .076387950 16800000.0 .074994575 16900000.0 .0736629252631 3103
 17000000.0 .072331275 17100000.0 .070999625 17200000.0 .0697291002631 3103
 17300000.0 .068519700 17400000.0 .067310300 17500000.0 .0661009002631 3103
 17600000.0 .064953475 17700000.0 .063868025 17800000.0 .0627825752631 3103
 17900000.0 .061697125 18000000.0 .060671725 18100000.0 .0597063752631 3103
 18200000.0 .058741025 18300000.0 .057775675 18400000.0 .0568670002631 3103
 18500000.0 .056015000 18600000.0 .055163000 18700000.0 .0543110002631 3103
 18800000.0 .053511638 18900000.0 .052764913 19000000.0 .0520181882631 3103
 19100000.0 .051271463 19200000.0 .050573338 19300000.0 .0499238132631 3103
 19400000.0 .049274288 19500000.0 .048624763 19600000.0 .0480200382631 3103
 19700000.0 .047460113 19800000.0 .046900188 19900000.0 .0463402632631 3103
 20000000.0 0.0                                                   2631 3103
                                                                  2631 3  0
                                                                  2631 0  0
 2.60560E+4 5.54544E+1          0          0          0          1263133103
 0.000000+0 0.000000+0          0        103          0          1263133103
 0.000000+0 0.000000+0          1          5        378         27263133103
 1.000000-5 2.900000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6263133103
 9.000000+6 1.000000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7263133103
 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7263133103
 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7263133103
 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0263133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 6.980350-3263133103
 2.338270-3 1.702360-4 4.353220-4-6.002120-5 3.473780-4 1.342800-4263133103
-1.013590-4-1.453170-4-1.032110-4-1.910370-5 5.580590-5 8.566180-5263133103
 6.229070-5 2.961970-6-6.242280-5-1.056550-4-1.094340-4-7.037860-5263133103
 1.788090-6 8.621690-5 1.524240-4 1.605400-4 6.014590-5-2.124840-4263133103
 2.467980-3 6.043630-4 3.578130-4 2.055220-4 1.680110-4 2.128390-4263133103
 1.769540-4 1.366700-4 9.916600-5 7.004770-5 5.015720-5 3.742600-5263133103
 2.898990-5 2.276790-5 1.801980-5 1.502850-5 1.442320-5 1.658450-5263133103
 2.127870-5 2.748400-5 3.331180-5 3.591800-5 3.136190-5 1.435050-5263133103
 2.089910-3 5.811850-4 5.051160-4 4.463010-4 2.985330-4 2.415270-4263133103
 2.109770-4 1.764960-4 1.378030-4 9.933400-5 6.703660-5 4.523120-5263133103
 3.491560-5 3.399020-5 3.868840-5 4.511620-5 5.024920-5 5.232580-5263133103
 5.086090-5 4.651010-5 4.095130-5 3.684360-5 3.791020-5 2.324770-3263133103
 8.909680-4 2.243430-4 3.463970-4 3.586080-4 2.979140-4 2.137770-4263133103
 1.316880-4 7.020240-5 3.717340-5 2.982110-5 3.851070-5 5.200710-5263133103
 6.154930-5 6.257380-5 5.457370-5 4.014340-5 2.400670-5 1.237870-5263133103
 1.271590-5 3.381980-5 8.624440-5 1.319090-3 6.075680-4 2.355910-4263133103
 1.831400-4 1.797820-4 1.702720-4 1.477000-4 1.151050-4 8.057420-5263133103
 5.231750-5 3.522030-5 2.987090-5 3.362030-5 4.237480-5 5.207540-5263133103
 5.950970-5 6.257730-5 6.025220-5 5.245210-5 3.991740-5 2.416230-5263133103
 9.567740-4 6.313680-4 3.548430-4 2.308720-4 1.569670-4 1.197830-4263133103
 1.026490-4 9.147350-5 7.836800-5 6.207430-5 4.573040-5 3.376500-5263133103
 2.952730-5 3.417020-5 4.646570-5 6.302500-5 7.854620-5 8.588480-5263133103
 7.587850-5 3.687780-5 8.246530-4 7.196220-4 5.696170-4 4.048130-4263133103
 2.567500-4 1.473680-4 8.506790-5 6.460670-5 7.167200-5 8.965720-5263133103
 1.051550-4 1.104520-4 1.034970-4 8.671160-5 6.568700-5 4.830190-5263133103
 4.434860-5 6.563950-5 1.265420-4 7.702400-4 6.761850-4 5.244410-4263133103
 3.568230-4 2.113590-4 1.126140-4 6.666460-5 6.340340-5 8.424730-5263133103
 1.104340-4 1.283210-4 1.311310-4 1.184390-4 9.489290-5 6.907830-5263133103
 5.291210-5 6.162700-5 1.143590-4 6.324280-4 5.237220-4 3.833900-4263133103
 2.465290-4 1.404880-4 7.815570-5 5.731600-5 6.563580-5 8.760000-5263133103
 1.098400-4 1.236410-4 1.252570-4 1.152190-4 9.746320-5 7.872510-5263133103
 6.832810-5 7.841040-5 4.660800-4 3.703690-4 2.615780-4 1.637400-4263133103
 9.329270-5 5.585710-5 4.758000-5 5.925580-5 8.050700-5 1.024550-4263133103
 1.187560-4 1.256000-4 1.212720-4 1.056640-4 7.989500-5 4.610700-5263133103
 3.221980-4 2.512470-4 1.743220-4 1.074110-4 6.116180-5 3.911000-5263133103
 3.878810-5 5.417430-5 7.796430-5 1.029630-4 1.226310-4 1.310680-4263133103
 1.227080-4 9.188330-5 3.231030-5 2.173730-4 1.684250-4 1.165000-4263133103
 7.298230-5 4.527820-5 3.576720-5 4.260230-5 6.120560-5 8.556410-5263133103
 1.089960-4 1.244090-4 1.241890-4 9.983000-5 4.136060-5 1.477430-4263133103
 1.185060-4 8.879480-5 6.555160-5 5.284920-5 5.160670-5 6.020940-5263133103
 7.533050-5 9.255920-5 1.067160-4 1.118820-4 1.012050-4 6.649160-5263133103
 1.138400-4 1.048040-4 9.446080-5 8.548980-5 7.961900-5 7.752210-5263133103
 7.899010-5 8.316230-5 8.870200-5 9.388370-5 9.659180-5 9.423730-5263133103
 1.173920-4 1.242110-4 1.246130-4 1.193480-4 1.101060-4 9.908370-5263133103
 8.873970-5 8.171760-5 8.089820-5 8.954360-5 1.115220-4 1.477160-4263133103
 1.605150-4 1.612980-4 1.513370-4 1.337330-4 1.127760-4 9.360740-5263133103
 8.216990-5 8.538600-5 1.115530-4 1.858920-4 1.970250-4 1.936170-4263133103
 1.779120-4 1.539130-4 1.268590-4 1.030300-4 8.983880-5 9.619200-5263133103
 2.208930-4 2.302800-4 2.251120-4 2.070470-4 1.789350-4 1.444760-4263133103
 1.081310-4 7.522490-5 2.563740-4 2.685600-4 2.647750-4 2.436270-4263133103
 2.040130-4 1.447640-4 6.437880-5 3.017630-4 3.184520-4 3.123860-4263133103
 2.767690-4 2.037260-4 8.360900-5 3.586650-4 3.751200-4 3.563580-4263133103
 2.888990-4 1.563190-4 4.204370-4 4.352960-4 4.045150-4 3.095700-4263133103
 5.050460-4 5.556140-4 5.751470-4 7.485460-4 9.914130-4 1.606050-3263133103
                                                                  263133  0
                                                                  2631 0  0
                                                                     0 0  0
 2.60580E+4 5.74356E+1          0          0         34         102637 1451
 0.0        0.0                 0          0          0          62637 1451
 1.00000E+0 2.00000E+7          0          0         10       20022637 1451
 3.00000E+2 0.0                 1          0        162          32637 1451
 26-Fe- 58 JAERI      EVAL-OCT95 T.NAKAGAWA                       2637 1451
                      DIST-Feb2004                                2637 1451
----IRDF-2002         MATERIAL 2637                               2637 1451
-----INCIDENT NEUTRON DATA                                        2637 1451
------ENDF-6 FORMAT                                               2637 1451
******************************************************************2637 1451
 26-FE- 58 JAERI      EVAL-OCT95 T.NAKAGAWA                       2637 1451
                      DIST-JUL98                                  2637 1451
----JENDL/D-99        MATERIAL 2637                               2637 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2637 1451
DOSIMETRY                                                         2637 1451
******************************************************************2637 1451
HISTORY                                                           2637 1451
93-10  JENDL-3.2.                                                 2637 1451
       RE-EVALUATION WAS MADE BY                                  2637 1451
         T.NAKAGAWA (NDC/JAERI): RESONANCE PARAMS, CROSS SECTIONS 2637 1451
         S.IGARASI (NEDAC): GAMMA-RAY PRODUCTION DATA BELOW 10 KEV2637 1451
       COMPILED BY T.NAKAGAWA                                     2637 1451
95-10  COVARIANCES WERE ESTIMATED BY T.NAKAGAWA.                  2637 1451
98-04  COMPILED TO JENDL DOSIMETRY FILE 99.                       2637 1451
02-10  COVARIANCES WERE MODIFIED BY K.SHIBATA.                    2637 1451
                                                                  2637 1451
            ===== ORIGINAL DATA FILE =====                        2637 1451
                                                                  2637 1451
 <DOSIMETRY CROSS SECTION>                                        2637 1451
      FE-58 CAPTURE FE-59 (HALF-LIFE = 44.503 H)                  2637 1451
                                                                  2637 1451
MF=1  GENERAL INFORMATION                                         2637 1451
 MT=451  DESCRIPTIVE DATA AND DICTIONARY                          2637 1451
                                                                  2637 1451
                                                                  2637 1451
MF=2  RESONANCE PARAMETERS                                        2637 1451
  MT = 151 RESOLVED RESONANCES                                    2637 1451
     TAKEM FROM JENDL-3.2/1/.                                     2637 1451
      RESONANCE REGION = 1.0E-5 EV TO 350.0 KEV                   2637 1451
      THE MULTILEVEL BREIT-WIGNER FORMULA WAS USED.  PARAMETERS   2637 1451
      WERE DETERMINED ON THE BASIS OF DATA BY GARG ET AL./2/,     2637 1451
      KAEPPELER ET AL./3/, ALLEN AND MACKLIN/4/.                  2637 1451
                                                                  2637 1451
      CALCULATED 2200-M/S CROSS SECTIONS AND RES. INTEGRALS.      2637 1451
                      2200-M/S     RES. INTEG.                    2637 1451
         ELASTIC       6.470 B         -                          2637 1451
         CAPTURE       1.300 B        1.36 B                      2637 1451
         TOTAL         7.770 B         -                          2637 1451
                                                                  2637 1451
MF = 3 CROSS SECTIONS                                             2637 1451
 MT = 1   TOTAL CROSS SECTION                                     2637 1451
    TAKEN FROM JENDL-3.2/1/.                                      2637 1451
                                                                  2637 1451
 MT=2   ELASTIC                                                   2637 1451
      TOTAL CROSS SECTION - SUM OF PARTIAL CROSS SECTIONS         2637 1451
    TAKEN FROM JENDL-3.2/1/.                                      2637 1451
                                                                  2637 1451
 MT = 102 CAPTURE CROSS SECTION                                   2637 1451
    THE CAPTURE CROSS SECTION WAS ADOPTED FROM JENDL-3.2/1/.      2637 1451
                                                                  2637 1451
    IN THE RESONANCE REGION BELOW 350 KEV, THE CROSS SECTION WAS  2637 1451
  CALCULATED FROM THE MULTILEVEL BREIT-WIGNER FORMULA WITH        2637 1451
  RESONANCE PARAMETERS DETERMINED ON THE BASIS OF DATA BY GARG ET 2637 1451
  AL./2/, KAEPPELER ET AL./3/, AND ALLEN AND MACKLIN/4/. ABOVE    2637 1451
  350 KEV, THE CROSS SECTION WAS CALCULATED WITH THE OPTICAL AND  2637 1451
  STATISTICAL MODEL CODE CASTHY/5/ BY NORMALIZING TO 3 MB AT 500  2637 1451
  KEV/6/.  CONTRIBUTIONS TO THE INELASTIC SCATTERING PROCESSES    2637 1451
  WERE CONSIDERED BY ADOPTING THE EXCITED LEVELS IN THE FOLLOWING 2637 1451
  TABLE.  AND THE CONTRIBUTIONS TO THE (N,2N), (N,N'A), (N,N'P),  2637 1451
  (N,P) AND (N,A) REACTION CROSS SECTIONS WERE ALSO CONSIDERED IN 2637 1451
  THE CASTHY CALCULATION. THESE THRESHOLD REACTION CROSS SECTIONS 2637 1451
  WERE CALCULATED WITH GNASH/7/. THE DIRECT CAPTURE CROSS SECTION 2637 1451
  WAS CALCULATED WITH A SIMPLE FORMULA DERIVED BY BENZI AND       2637 1451
  REFFO/8/ AND ADDED TO THE RESULT OF CASTHY CALCULATION.         2637 1451
                                                                  2637 1451
    OPTICAL POTENTIAL PARAMETERS                                  2637 1451
    -----------------------------                                 2637 1451
         V  = 46.0-0.25*EN (MEV),                                 2637 1451
         WS = 14.0-0.2*EN  (MEV),  (IN THE GAUSSIAN FORM)         2637 1451
         WI = 0.125*E-0.0004*E**2 (MEV),                          2637 1451
         VSO= 6.0          (MEV),                                 2637 1451
         R  = 1.286 (FM), A0 = 0.62  (FM)                         2637 1451
         RS = 1.387 (FM), AS = 0.7   (FM)                         2637 1451
         RSO= 1.07  (FM), ASO= 0.62  (FM)                         2637 1451
                                                                  2637 1451
    LEVEL SCHEME                                                  2637 1451
    ------------                                                  2637 1451
          NO.     ENERGY(MEV)  SPIN-PARITY                        2637 1451
         G.S.      0.0          0 +                               2637 1451
          1.       0.8108       2 +                               2637 1451
          2.       1.6747       2 +                               2637 1451
          3.       2.0765       4 +                               2637 1451
          4.       2.1339       3 +                               2637 1451
          5.       2.2581       0 +                               2637 1451
          6.       2.6004       4 +                               2637 1451
          7.       2.7819       1 +                               2637 1451
          8.       2.8764       2 +                               2637 1451
          9.       3.0840       2 +                               2637 1451
         10.       3.1330       4 +                               2637 1451
         11.       3.2330       2 +                               2637 1451
         12.       3.2440       0 +                               2637 1451
       LEVELS ABOVE 3.389 MEV WERE ASSUMED TO BE OVERLAPPING.     2637 1451
                                                                  2637 1451
MF = 33 COVARIANCES OF NEUTRON CROSS SECTIONS                     2637 1451
 MT=102 CAPTURE CROSS SECTION                                     2637 1451
                                                                  2637 1451
    IN THE RESONANCE REGION, THE VARIANCE OF CROSS SECTION WAS    2637 1451
  ESTIMATED FROM STANDARD DEVIATIONS OF THE RESONANCE PARAMETERS. 2637 1451
  THE STANDARD DEVIATIONS WERE TAKEN FROM THE EXPERIMENTAL DATA   2637 1451
  WHICH WERE CONSIDERED IN THE PARAMETER EVALUATION. FOR THE      2637 1451
  RESONANCE PARAMETERS DETERMINED FROM THE CAPTURE AREA, THEIR    2637 1451
  ERRORS WERE ESTIMATED FROM THE ERROR OF CAPTURE AREA. AN ERROR  2637 1451
  OF 10 % WAS ASSUMED FOR OTHER CASES.                            2637 1451
                                                                  2637 1451
    ABOVE 350 KEV, COVARIANCE MATRIX WAS OBTAINED BY USING        2637 1451
  KALMAN/10/. ERRORS WERE CONSIDERED TO THE OPTICAL POTENTIAL AND 2637 1451
  LEVEL DENSITY PARAMETERS, AND THE NORMALIZATION CROSS SECTION.  2637 1451
                                                                  2637 1451
                                                                  2637 1451
REFERENCES                                                        2637 1451
 1) NAKAGAWA T. ET AL.: J. NUCL. SCI. TECHNOL., 32, 1259 (1995).  2637 1451
 2) GARG J.B. ET AL.: PHYS. REV., C18, 1141 (1978).               2637 1451
 3) KAEPPELER F. ET AL.: NUCL. SCI. ENG., 84, 234 (1983).         2637 1451
 4) ALLEN B.J. AND MACKLIN R.L.: J. PHYS. G., 6, 381 (1980).      2637 1451
 5) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991).                2637 1451
 6) TROFIMOV JU.N.: ATOMNAJA ENERGIJA, 58, 278 (1985).            2637 1451
 7) YOUNG P.G. AND ARTHUR E.D.: LA-6974 (1977).                   2637 1451
 8) BENZI V. AND REFFO G.: CCDN/NW/10 (1969).                     2637 1451
 9) YAMAKOSHI H.: JAERI 1261, P.30 (1979).                        2637 1451
10) KAWANO T. AND SHIBATA K.: JAERI-DATA/CODE 97-037 (1997)       2637 1451
    [IN JAPANESE]                                                 2637 1451
                                                                  2637 1451
******************************************************************2637 1451
 The Q values and threshold energies were updated prior to pro-   2637 1451
 cessing through the codes to comply with the values obtained     2637 1451
 using the NNDC calculation program which is based on the 1995    2637 1451
 Update to the Atomic mass Evaluation.                            2637 1451
******************************************************************2637 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2637 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2637 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2637 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2637 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2637 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2637 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2637 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2637 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2637 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2637 1451
 ***************** Program FIXUP (Version 2002-1) ****************2637 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2637 1451
 Corrected Thresholds-----------------------------------------Yes 2637 1451
 Extended Cross Sections to 20 MeV----------------------------No  2637 1451
 Allow Cross Section Deletion---------------------------------No  2637 1451
 Allow Cross Section Reconstruction---------------------------No  2637 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2637 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2637 1451
 Deleted Duplicate Points-------------------------------------Yes 2637 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2637 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2637 1451
 Allow Creation of Missing Sections---------------------------No  2637 1451
 Allow Insertion of Energy Points-----------------------------No  2637 1451
 Create Uniform Energy Grid-----------------------------------No  2637 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2637 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2637 1451
 Unshielded Group Averages Using  640 Groups                      2637 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2637 1451
                                1        451        169          12637 1451
                                3        102        217          12637 1451
                               33        102        200          12637 1451
                                                                  2637 1  0
                                                                  2637 0  0
 2.60580E+4 5.74356E+1          0          0          0          02637 3102
 6.58043E+6 6.58043E+6          0          0          1        6412637 3102
        641          1                                            2637 3102
 .000100000 20.4392055 .000105000 19.9625958 .000110000 19.50981102637 3102
 .000115000 19.0925823 .000120000 18.6048355 .000127500 18.06369282637 3102
 .000135000 17.5692785 .000142500 17.1149218 .000150000 16.62318192637 3102
 .000160000 16.1109172 .000170000 15.6448196 .000180000 15.21697742637 3102
 .000190000 14.8189064 .000200000 14.4554637 .000210000 14.11352532637 3102
 .000220000 13.7959192 .000230000 13.5008306 .000240000 13.15445962637 3102
 .000255000 12.7738323 .000270000 12.4801215 .000280000 12.15316722637 3102
 .000300000 11.7544371 .000320000 11.3934012 .000340000 11.06275082637 3102
 .000360000 10.7582019 .000380000 10.4801883 .000400000 10.19034412637 3102
 .000425000 9.89374786 .000450000 9.62275629 .000475000 9.374305662637 3102
 .000500000 9.14065477 .000525000 8.92711285 .000550000 8.725389682637 3102
 .000575000 8.53778124 .000600000 8.34549314 .000630000 8.147677252637 3102
 .000660000 7.96667299 .000690000 7.79303972 .000720000 7.608288232637 3102
 .000760000 7.41019189 .000800000 7.22626354 .000840000 7.057702732637 3102
 .000880000 6.89749502 .000920000 6.75062822 .000960000 6.609978432637 3102
 .001000000 6.46419650 .001050000 6.31200628 .001100000 6.169327572637 3102
 .001150000 6.03785946 .001200000 5.88277591 .001275000 5.712331392637 3102
 .001350000 5.55649992 .001425000 5.41106895 .001500000 5.256562372637 3102
 .001600000 5.09500822 .001700000 4.94752804 .001800000 4.811128322637 3102
 .001900000 4.68651763 .002000000 4.57098309 .002100000 4.462571142637 3102
 .002200000 4.36328773 .002300000 4.26834950 .002400000 4.160027262637 3102
 .002550000 4.03966313 .002700000 3.94551646 .002800000 3.843217652637 3102
 .003000000 3.71728764 .003200000 3.60269027 .003400000 3.497892912637 3102
 .003600000 3.40230772 .003800000 3.31405188 .004000000 3.222064332637 3102
 .004250000 3.12866520 .004500000 3.04342270 .004750000 2.963760672637 3102
 .005000000 2.89066447 .005250000 2.82291434 .005500000 2.758877062637 3102
 .005750000 2.70019726 .006000000 2.63857476 .006300000 2.576985012637 3102
 .006600000 2.51870187 .006900000 2.46464639 .007200000 2.405865762637 3102
 .007600000 2.34300200 .008000000 2.28550413 .008400000 2.231544032637 3102
 .008800000 2.18133652 .009200000 2.13468288 .009600000 2.090291782637 3102
 .010000000 2.04433192 .010500000 1.99574288 .011000000 1.951297212637 3102
 .011500000 1.90889566 .012000000 1.86029813 .012750000 1.806500712637 3102
 .013500000 1.75667498 .014250000 1.71097879 .015000000 1.662270692637 3102
 .016000001 1.61116877 .017000001 1.56433412 .017999999 1.521593422637 3102
 .018999999 1.48230911 .020000000 1.44544810 .021000000 1.411690562637 3102
 .022000000 1.37985994 .023000000 1.35017140 .024000000 1.315254582637 3102
 .025500000 1.27755863 .027000001 1.24769142 .028000001 1.215377272637 3102
 .029999999 1.17559431 .032000002 1.13934099 .034000002 1.106250512637 3102
 .035999998 1.07604438 .037999999 1.04817460 .039999999 1.019082342637 3102
 .042500000 .989533432 .045000002 .962518985 .047499999 .9373464302637 3102
 .050000001 .914192463 .052499998 .892773955 .055000000 .8725021872637 3102
 .057500001 .853960255 .059999999 .834472263 .063000001 .8149967642637 3102
 .066000000 .796644674 .068999998 .779548568 .071999997 .7609938862637 3102
 .075999998 .741102563 .079999998 .722908192 .083999999 .7057989722637 3102
 .088000000 .689882490 .092000000 .675140115 .096000001 .6610590892637 3102
 .100000001 .646522061 .104999997 .631136232 .109999999 .6170821452637 3102
 .115000002 .603662108 .119999997 .588304210 .127499998 .5713551142637 3102
 .135000005 .555649753 .142499998 .541234858 .150000006 .5258358342637 3102
 .159999996 .509622987 .170000002 .494776149 .180000007 .4812086142637 3102
 .189999998 .468752593 .200000003 .457057534 .209999993 .4463636572637 3102
 .219999999 .436272716 .230000004 .426874995 .239999995 .4160312402637 3102
 .254999995 .403920329 .270000011 .394680652 .280000001 .3844092962637 3102
 .300000012 .371802308 .319999993 .360327503 .340000004 .3498814882637 3102
 .360000014 .340309396 .379999995 .331389224 .400000006 .3222441372637 3102
 .425000012 .312921512 .449999988 .304303986 .474999994 .2963775622637 3102
 .500000000 .289135566 .524999976 .282315883 .550000012 .2761605822637 3102
 .574999988 .270083638 .600000024 .263997936 .629999995 .2577124582637 3102
 .660000026 .251965198 .689999998 .246489298 .720000029 .2406106652637 3102
 .759999990 .234371121 .800000012 .228527149 .839999974 .2231886452637 3102
 .879999995 .218128077 .920000017 .213447889 .959999979 .2090518092637 3102
 1.00000000 .204421102 1.04999995 .199667372 1.10000002 .1951632682637 3102
 1.14999998 .191015952 1.20000005 .186096938 1.27499998 .1806833862637 3102
 1.35000002 .175751221 1.42499995 .171152860 1.50000000 .1662486742637 3102
 1.60000002 .161127894 1.70000005 .156461795 1.79999995 .1521477362637 3102
 1.89999998 .148190407 2.00000000 .144551899 2.09999990 .1411033972637 3102
 2.20000005 .137937425 2.29999995 .135073444 2.40000010 .1315887782637 3102
 2.54999995 .127822257 2.70000005 .124828118 2.79999995 .1215583732637 3102
 3.00000000 .117559676 3.20000005 .113931804 3.40000010 .1106274152637 3102
 3.59999990 .107601805 3.79999995 .104783493 4.00000000 .1018926942637 3102
 4.25000000 .098950178 4.50000000 .096248213 4.75000000 .0937775462637 3102
 5.00000000 .091435698 5.25000000 .089329983 5.50000000 .0872701052637 3102
 5.75000000 .085409240 6.00000000 .083459368 6.30000019 .0814992132637 3102
 6.59999990 .079670875 6.90000010 .077941727 7.19999981 .0760872572637 3102
 7.59999990 .074090681 8.00000000 .072270196 8.39999962 .0705722052637 3102
 8.80000019 .068974089 9.19999981 .067531286 9.60000038 .0661368912637 3102
 10.0000000 .064655340 10.5000000 .063169263 11.0000000 .0617147532637 3102
 11.5000000 .060400469 12.0000000 .058846385 12.7500000 .0571354822637 3102
 13.5000000 .055572707 14.2500000 .054119249 15.0000000 .0525657072637 3102
 16.0000000 .050953030 17.0000000 .049480908 18.0000000 .0481252172637 3102
 19.0000000 .046897325 20.0000000 .045712369 21.0000000 .0446480872637 3102
 22.0000000 .043667605 23.0000000 .042697774 24.0000000 .0416122252637 3102
 25.5000000 .040404259 27.0000000 .039474319 28.0000000 .0384506972637 3102
 30.0000000 .037187331 32.0000000 .036037593 34.0000000 .0349943802637 3102
 36.0000000 .034053368 38.0000000 .033166688 40.0000000 .0322533352637 3102
 42.5000000 .031321702 45.0000000 .030459793 47.5000000 .0296759772637 3102
 50.0000000 .028941880 52.5000000 .028274037 55.0000000 .0276249962637 3102
 57.5000000 .027037779 60.0000000 .026428804 63.0000000 .0258252842637 3102
 66.0000000 .025234071 69.0000000 .024699507 72.0000000 .0241123992637 3102
 76.0000000 .023507267 80.0000000 .022922779 84.0000000 .0223932642637 3102
 88.0000000 .021896355 92.0000000 .021431239 96.0000000 .0210030712637 3102
 100.000000 .020547156 105.000000 .020083533 110.000000 .0196438812637 3102
 115.000000 .019247720 120.000000 .018782183 127.500000 .0182736412637 3102
 135.000000 .017823060 142.500000 .017420337 150.000000 .0170093692637 3102
 160.000000 .016623682 170.000000 .016350054 180.000000 .0162363562637 3102
 190.000000 .016462641 200.000000 .017611779 210.000000 .0233497882637 3102
 220.000000 1.71520400 230.000000 1.03090169 240.000000 .0210559222637 3102
 255.000000 .015776852 270.000000 .015015936 280.000000 .0152385872637 3102
 300.000000 .017525747 320.000000 .029622690 340.000000 9.487500002637 3102
 360.000000 .624757298 380.000000 .026309083 400.000000 .0153835392637 3102
 425.000000 .012264850 450.000000 .010964901 475.000000 .0102195282637 3102
 500.000000 .009708855 525.000000 .009318123 550.000000 .0090000212637 3102
 575.000000 .008734899 600.000000 .008474226 630.000000 .0082223452637 3102
 660.000000 .007997192 690.000000 .007793375 720.000000 .0075793502637 3102
 760.000000 .007353365 800.000000 .007151738 840.000000 .0069621222637 3102
 880.000000 .006789316 920.000000 .006630916 960.000000 .0064804972637 3102
 1000.00000 .006325455 1050.00000 .006165594 1100.00000 .0060154132637 3102
 1150.00000 .005877807 1200.00000 .005715130 1275.00000 .0055364672637 3102
 1350.00000 .005375247 1425.00000 .005226880 1500.00000 .0050699722637 3102
 1600.00000 .004903878 1700.00000 .004752835 1800.00000 .0046152212637 3102
 1900.00000 .004488776 2000.00000 .004374660 2100.00000 .0042674612637 3102
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 2.634240-4 1.570240-4 4.852240-5 1.467400-4 5.002030-5 3.702840-5263733102
 0.000000+0 8.394770-3 4.979120-3 6.482810-3 2.245710-3 2.337420-3263733102
 1.395740-3 8.010230-4 9.243420-4 4.487630-4 3.300940-4 2.595640-4263733102
 2.197820-4 1.291360-4 3.907760-5 1.182940-4 4.042350-5 2.980260-5263733102
 0.000000+0 5.229450-3 4.204170-3 1.314540-3 1.195570-3 6.062080-4263733102
 3.100550-4 3.685110-4 1.735910-4 1.358180-4 1.053360-4 8.643910-5263733102
 4.986320-5 1.422210-5 4.352510-5 1.480560-5 1.079310-5 0.000000+0263733102
 1.278420-1 2.452270-3 2.539050-3 2.072490-3 1.469510-3 1.366090-3263733102
 5.206170-4 3.119520-4 2.278490-4 1.801430-4 1.022540-4 3.074710-5263733102
 8.714300-5 2.963580-5 2.157570-5 0.000000+0 7.521570-3 3.842640-3263733102
 2.593450-3 2.008520-3 1.647500-3 5.265960-4 2.727500-4 1.854600-4263733102
 1.405310-4 7.614870-5 2.328790-5 6.164280-5 2.115420-5 1.506390-5263733102
 0.000000+0 1.142120-2 1.411770-2 1.208490-2 8.727930-3 2.120050-3263733102
 1.108620-3 7.019000-4 4.881680-4 2.493760-4 6.820590-5 1.806640-4263733102
 6.117650-5 4.249310-5 0.000000+0 3.086340-2 2.499900-2 1.445590-2263733102
 2.723800-3 1.668000-3 1.041820-3 6.895220-4 3.392970-4 8.127800-5263733102
 2.323500-4 7.682050-5 5.373630-5 0.000000+0 3.605820-2 1.334690-2263733102
 2.371710-3 1.411660-3 8.701670-4 5.634090-4 2.775480-4 6.328910-5263733102
 1.843110-4 6.030200-5 4.240900-5 0.000000+0 2.767230-2 2.138060-2263733102
 3.647850-3 1.627810-3 9.254780-4 4.127230-4 9.512600-5 2.498320-4263733102
 8.201170-5 5.665830-5 0.000000+0 2.410570-2 3.291670-3 1.295050-3263733102
 7.262060-4 2.948150-4 7.619040-5 1.652050-4 5.503500-5 3.619310-5263733102
 0.000000+0 4.431960-3 1.083480-3 5.503280-4 1.490210-4 6.371800-5263733102
 8.299840-5 3.021650-5 1.758260-5 0.000000+0 5.737610-3 1.937990-3263733102
 2.286100-4 8.452630-5 8.918570-5 3.293200-5 1.773060-5 0.000000+0263733102
 5.158250-3 5.355720-4 1.926830-4 1.610610-4 5.505400-5 2.715720-5263733102
 0.000000+0 1.985240-3 7.815200-4 3.535600-4 1.006480-4 4.388340-5263733102
 0.000000+0 2.054360-3 9.126950-4 1.862830-4 6.453820-5 0.000000+0263733102
 3.101820-3 1.536580-3 3.321300-4 0.000000+0 4.214860-3 7.661330-4263733102
 0.000000+0 2.318810-3 0.000000+0 0.000000+0                      263733102
 0.000000+0 0.000000+0          1          5        351         26263733102
 1.000000-5 3.500000+5 4.250000+5 5.999990+5 7.499990+5 8.499990+5263733102
 9.499990+5 1.125000+6 1.375000+6 1.625000+6 1.875000+6 2.500000+6263733102
 3.500000+6 4.500000+6 5.500000+6 6.500000+6 7.500000+6 8.500000+6263733102
 9.500000+6 1.100000+7 1.300000+7 1.450000+7 1.550000+7 1.700000+7263733102
 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0263733102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263733102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263733102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263733102
 0.000000+0 0.000000+0 0.000000+0 6.400900-4 6.270690-4 4.454030-4263733102
 3.813650-4 5.728910-4 5.453350-4 4.591580-4 1.188950-3 5.366080-4263733102
 5.889310-4 8.843060-4 8.644480-4 6.824290-4 4.434780-4 2.425270-4263733102
 1.251280-4 8.677850-5 1.043610-4 3.720190-4 1.069640-3 1.617310-3263733102
 2.203690-3 3.112890-3 3.252440-3 6.604900-4 4.616010-4 3.984970-4263733102
 6.302680-4 6.094160-4 5.330270-4 1.353860-3 6.338810-4 6.997410-4263733102
 1.049430-3 1.048230-3 8.740180-4 6.452080-4 4.449160-4 3.301910-4263733102
 2.914300-4 3.070620-4 5.609440-4 1.215170-3 1.734790-3 2.319580-3263733102
 3.155040-3 3.065710-3 6.300100-4 6.532220-4 7.524960-4 7.617110-4263733102
 7.936020-4 6.772240-4 7.573830-4 7.292780-4 5.352740-4 5.011720-4263733102
 5.669470-4 7.002240-4 8.556530-4 9.622720-4 1.032640-3 1.042960-3263733102
 1.020410-3 9.934750-4 9.421310-4 8.841950-4 8.629140-4 9.120660-4263733102
 7.507600-4 7.684990-4 7.962920-4 8.833210-4 4.307810-4 7.942380-4263733102
 7.291470-4 3.649350-4 3.326150-4 4.859060-4 7.565710-4 1.033560-3263733102
 1.226240-3 1.328120-3 1.329670-3 1.178640-3 8.324920-4 5.325110-4263733102
 1.729720-4-2.437260-4-1.493790-4 1.608010-3 1.626470-3 1.640970-3263733102
 1.452110-3 1.526050-3 1.497130-3 1.653870-3 1.591680-3 1.626360-3263733102
 1.745200-3 1.865800-3 1.942930-3 1.967390-3 1.888840-3 1.747240-3263733102
 1.496650-3 1.273130-3 1.079570-3 1.216020-3 2.115460-3 1.755610-3263733102
 1.771520-3 1.369290-3 1.601530-3 1.556760-3 1.699100-3 1.631830-3263733102
 1.746660-3 1.934560-3 2.127160-3 2.235890-3 2.278390-3 2.201960-3263733102
 1.900010-3 1.288550-3 7.795860-4 2.641850-4 0.000000+0 9.788250-4263733102
 1.962490-3 1.117600-3 1.693930-3 1.608710-3 1.609390-3 1.572910-3263733102
 1.792730-3 2.158490-3 2.513250-3 2.720840-3 2.784210-3 2.693820-3263733102
 2.221200-3 1.039010-3 9.561710-5-8.607330-4-1.822950-3-9.269570-4263733102
 7.089630-3 1.465430-3 1.637290-3 2.893500-3 2.716890-3 2.003500-3263733102
 1.195020-3 5.785340-4 3.649000-4 4.572150-4 7.079900-4 1.750060-3263733102
 4.057770-3 5.866950-3 7.838270-3 1.067520-2 1.038370-2 1.780840-3263733102
 1.792260-3 2.161630-3 2.263660-3 2.346080-3 2.495750-3 2.601530-3263733102
 2.643620-3 2.603900-3 2.520760-3 2.290590-3 1.852580-3 1.554630-3263733102
 1.273120-3 9.360720-4 6.909550-4 1.962490-3 2.565350-3 2.718850-3263733102
 2.738130-3 2.737270-3 2.706870-3 2.617010-3 2.543910-3 2.452930-3263733102
 2.311910-3 2.188750-3 2.172640-3 2.191620-3 2.157160-3 1.411970-3263733102
 4.664890-3 4.921800-3 4.558810-3 3.928810-3 3.214320-3 2.671990-3263733102
 2.315360-3 2.166760-3 2.369560-3 3.283990-3 4.245070-3 5.342850-3263733102
 6.652700-3 4.731700-3 5.685160-3 5.591200-3 5.274010-3 4.850020-3263733102
 4.581470-3 4.495990-3 4.685950-3 5.286950-3 4.999010-3 5.436770-3263733102
 6.162770-3 6.981340-3 4.208060-3 7.242010-3 8.144960-3 8.988140-3263733102
 9.918580-3 1.092380-2 1.199030-2 1.333340-2 8.692290-3 6.805680-3263733102
 5.439890-3 4.040780-3 8.571640-4 1.102500-2 1.306150-2 1.509890-2263733102
 1.724460-2 1.918590-2 2.120880-2 1.205000-2 7.614630-3 3.937350-3263733102
 1.748330-4-3.032670-3 1.716100-2 2.007790-2 2.296380-2 2.600090-2263733102
 2.868950-2 1.526500-2 8.626110-3 2.794700-3-2.965560-3-6.111480-3263733102
 2.496400-2 2.837000-2 3.205500-2 3.559640-2 1.889820-2 1.035410-2263733102
 2.874230-3-4.179100-3-6.893270-3 3.422500-2 3.780380-2 4.258630-2263733102
 2.280370-2 1.270380-2 4.104470-3-3.570120-3-5.771380-3 4.452100-2263733102
 4.894050-2 2.666180-2 1.559110-2 6.239220-3-1.641400-3-3.268030-3263733102
 5.904900-2 3.828800-2 2.968610-2 2.309590-2 2.096800-2 2.670880-2263733102
 6.150400-2 7.269370-2 8.635430-2 1.136110-1 1.503490-1 1.004890-1263733102
 1.229590-1 1.670490-1 2.212880-1 1.648360-1 2.211800-1 2.929660-1263733102
 3.226240-1 4.279660-1 6.352090-1                                 263733102
                                                                  263733  0
                                                                  2637 0  0
                                                                     0 0  0
 2.70590E+4 5.84269E+1          0          0         34         102725 1451
 0.0        0.0                 0          0          0          62725 1451
 1.00000E+0 2.00000E+7          0          0         10       20022725 1451
 3.00000E+2 0.0                 1          0        233          72725 1451
 27-Co- 59 IRK/FEI    EVAL-APR90 VONACH ET AL., K.I.Zolotarev     2725 1451
                      DIST-Feb2004                                2725 1451
----IRDF-2002         MATERIAL 2725                               2725 1451
-----INCIDENT NEUTRON DATA                                        2725 1451
------ENDF-6 FORMAT                                               2725 1451
******************************************************************2725 1451
 27-CO- 59 ANL,ORNL   EVAL-JUL89 A.SMITH+,G.DESAUSSURE+           2725 1451
 ANL/NDM-107          DIST-JUN93 REV1-JUN92            19930930   2725 1451
----ENDF/B-VI         MATERIAL 2725         REVISION 1            2725 1451
******************************************************************2725 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2725 1451
DOSIMETRY                                                         2725 1451
******************************************************************2725 1451
***************************************************************** 2725 1451
**********   Start of (N,G) bibliographical component *********** 2725 1451
   MF=2,MT=151, MF=3,MT=1,2,102 taken from ENDF/B-VI Release 2    2725 1451
   MF=33,MT=102 taken from IRDF-90                                2725 1451
***************************************************************** 2725 1451
 REVISION 1 -- NEW RESONANCE REGION EVALUATION BY G.DESAUSSURE,   2725 1451
   N.M.LARSON, J.A.HARVEY, N.W.HILL, ANN. NUCL. ENERGY 19, 393    2725 1451
   (1992)                                                         2725 1451
 ORIGINAL EVALUATION                                              2725 1451
 A.SMITH,D.SMITH,P.GUENTHER,J.MEADOWS, R.LAWSON(ANL),             2725 1451
 R.HOWERTON(LLNL),M.SUGIMOTO(JAERI)                               2725 1451
******************************************************************2725 1451
 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC             2725 1451
 *         *          *          *          *          *         *2725 1451
 R.LAWSON(ANL),R.HOWERTON(LLNL),M.SUGIMOTO(JAERI)                 2725 1451
***************************************************************** 2725 1451
**********   End of (N,G) bibliographical component   *********** 2725 1451
***************************************************************** 2725 1451
***************************************************************** 2725 1451
**********   Start of (N,2N) bibliographical component ********** 2725 1451
***************************************************************** 2725 1451
++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451
         MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING    2725 1451
         EVALUATION                                               2725 1451
                27-CO- 59 IRK-VIENNA EVAL-APR90                   2725 1451
                                     DIST-JUN90                   2725 1451
                IRK-EVAL.NLIB 25  2725                            2725 1451
++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451
***************************************************************** 2725 1451
**********   End of (N,2N) bibliographical component  *********** 2725 1451
***************************************************************** 2725 1451
***************************************************************** 2725 1451
**********   Start of (N,A) bibliographical component  ********** 2725 1451
***************************************************************** 2725 1451
***************************************************************** 2725 1451
------Russian Reactor Dosimetry File RRDF-2002                    2725 1451
***************************************************************** 2725 1451
Author  of evaluation: K.I.Zolotarev                              2725 1451
***************************************************************** 2725 1451
MF=3                                                              2725 1451
MT=107 -(n,a) cross section                                       2725 1451
-------------------------------------                             2725 1451
    Excitation function for the  Co-59(n,a)Mn-56 reaction in the  2725 1451
energy region from 2.5 MeV   to 21 MeV was evaluated by means of  2725 1451
statistical analysis of  experimental cross section  data [1-28], 2725 1451
and data from GNASH [29] calculation.                             2725 1451
    All experimental data were renormalized to the new standards  2725 1451
for monitor reactions cross sections and decay data. Uncertainty  2725 1451
in the monitor reaction  cross section  was added  to  the total  2725 1451
uncertainty for Santry and Butler data [5].   Cross section data  2725 1451
measured by Huang Jianzhou et al. [18]  were renormalized to the  2725 1451
preliminary evaluated  integral  of excitation function  in  the  2725 1451
energy range 13.0 - 15.0 MeV.  Correction factor for  this  data  2725 1451
was equal Fc= 1.0507 .                                            2725 1451
    Experimental data from ref.[30-41] were rejected due to their 2725 1451
discrepancy  with  the main bulk of experimental data [1-28]  and 2725 1451
data from theoretical model calculation.                          2725 1451
    The final procedure of evaluation Co-59(n,a)Mn-56  excitation 2725 1451
function from threshold to 21 MeV has been carried out within the 2725 1451
framework of generalized least squares method.  Rational function 2725 1451
was used  as  model function [42].  Calculations was performed by 2725 1451
means of  Pade-2 code [43].                                       2725 1451
    The evaluated  Co-59(n,a)Mn-56  excitation  function averaged 2725 1451
on  U-235  neutron fission spectrum [44]  and  Cf-252 spontaneous 2725 1451
fission neutron spectrum [45] gives the next values :             2725 1451
  2725 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2725 1451
  2725 1451
                                         0.1500 +- 0.0080 [46]  2725 1451
 U-235 neutron fission      0.15823      0.1700 +- 0.0120 [47]  2725 1451
                                         0.1568 +- 0.0035 [48]  2725 1451
  2725 1451
                                         0.2208 +- 0.0014 [49]  2725 1451
 Cf-252 spont. fission      0.22095      0.2221 +- 0.0023 [50]  2725 1451
                                         0.2218 +- 0.0041 [51]  2725 1451
  2725 1451
                                                                  2725 1451
MT=33                                                             2725 1451
MT=107 -(n,a) cross section cov. matrix                           2725 1451
-------------------------------------------                       2725 1451
    Uncertainties in the evaluated  excitation function  for the  2725 1451
reaction Co-59(n,a)Mn-56  are given in the form of relative cova- 2725 1451
riance matrix for the 32-neutron energy groups (LB=5). Covariance 2725 1451
matrix of uncertainties was calculated simultaneously with recom- 2725 1451
mended cross section data by means of PADE-2 code.                2725 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2725 1451
given in the 33-file are the following:                           2725 1451
                                                                  2725 1451
    1.11275E-08    1.13832E-08    1.19424E-08    1.27401E-08      2725 1451
    1.37409E-08    1.52188E-08    1.67720E-08    1.86812E-08      2725 1451
    2.09005E-08    2.38172E-08    2.69210E-08    3.07595E-08      2725 1451
    3.46853E-08    4.07806E-08    4.51537E-08    5.55085E-08      2725 1451
    6.19014E-08    7.83507E-08    9.39864E-08    1.01329E-07      2725 1451
    2.19793E-07    8.71735E-07    1.71048E-05    1.93185E-04      2725 1451
    1.05065E-03    1.19481E-03    2.01786E-03    2.53733E-03      2725 1451
    3.28491E-03    1.06037E-02    2.32281E-02    9.30559E-01      2725 1451
                                                                  2725 1451
References :                                                      2725 1451
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21. A.Ercan et al.  Proc. of an Intern. Conf. on Nuclear Data for 2725 1451
    Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2725 1451
    Verlag, 1992, p.376-377                                       2725 1451
22. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24, Aug.1992 2725 1451
23. A.Grallert et al. Report INDC(NDS)-286, p.193, IAEA, 1993     2725 1451
24. W.Mannhart et al. Proc.of International Conference on Nuclear 2725 1451
    Data for Science and Technology, Gatlinburg, Tennessee, May   2725 1451
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    Vienna, October 1996                                          2725 1451
27. A.D.Majdeddin Measurement and analysis of excitation func-    2725 1451
    tions for fast neutron induced reactions from threshold to    2725 1451
    20 MeV. PHD dissertation, Inst. of Experimental Physics ,     2725 1451
    Kossuth University, Hungary, June 1998                        2725 1451
28. A.A.Filatenkov et al. Report RI-252, St. Petersburg, May 1999 2725 1451
29. P.G.Young, E.D.Arthur  A Preequilibrium Statistical Nuclear   2725 1451
    Model Code for Calculation of Cross Section and Emission      2725 1451
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    E.L.Trykiv, G.Ya.Tertychnyi  Private communication, IPPE,     2725 1451
    Obninsk, May 1999                                             2725 1451
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34. J.E.Strain, W.J.Ross  Report ORNL-3672, January 1965          2725 1451
35. M.Bormann  et al.  Nucl. Phys., v.63, p.438, March 1965       2725 1451
36. U.Garuska, J.Dresler, H.Malecki  Prog. Report INR-1773/I/PL/A,2725 1451
    p.16, September 1978                                          2725 1451
37. H.M.Agrawal, K.R.Zasadny, G.F.Knoll  Trans. Amer. Nucl. Soc., 2725 1451
    v.47, p.431, November 1984   save                             2725 1451
38. M.Berrada  Progress Report to the IAEA NDS on Research        2725 1451
    Contract -3311.R1/RB, 15 April 1984                           2725 1451
39. I.Garlea et al. Revue Roumaine de Physique, v.30, p.673, 1985 2725 1451
40. N.I.Molla et al.  Proc.of International Conference on Nuclear 2725 1451
    Data for Science and Technology, Gatlinburg, Tennessee, May   2725 1451
    9-13, 1994, v.2, pp.938-940                                   2725 1451
41. R.Doczi et al.  Nucl. Sci. Eng., v.129, p.164, June 1998      2725 1451
42. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2725 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2725 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  2725 1451
43. S.A.Badikov et.al.  Preprint FEI-1686, Obninsk, 1985          2725 1451
44. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2725 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2725 1451
45. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2725 1451
46. K.Kobayashi, I.Kimura  Report NEANDC(J)-61U, p.81, Sep. 1979  2725 1451
47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2725 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 2725 1451
48. W.Mannhart   Progress Report INDC(Ger)-045, pp.40-43, 1999    2725 1451
49. K.Kobayashi,I.Kimura,W.Mannhart  Nucl. Sci. Technology.,v.19, 2725 1451
    n.5, p.341, May 1982                                          2725 1451
50. W.Mannhart  Handbook on Nuclear Activation Data , IAEA        2725 1451
    Technical Report series No.273, p.413, 1987                   2725 1451
51. W.Mannhart  Validation of Differential Cross Sections with    2725 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2725 1451
    September 2002                                                2725 1451
***************************************************************** 2725 1451
**********   End of (N,A) bibliographical component    ********** 2725 1451
***************************************************************** 2725 1451
  The Q values and threshold energies were updated prior to pro-  2725 1451
  cessing through the codes to comply with the values obtained    2725 1451
  using the NNDC calculation program which is based on the 1995   2725 1451
  Update to the Atomic mass Evaluation.                           2725 1451
***************************************************************** 2725 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2725 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2725 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2725 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2725 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2725 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2725 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2725 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2725 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2725 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2725 1451
 ***************** Program FIXUP (Version 2002-1) ****************2725 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2725 1451
 Corrected Thresholds-----------------------------------------Yes 2725 1451
 Extended Cross Sections to 20 MeV----------------------------No  2725 1451
 Allow Cross Section Deletion---------------------------------No  2725 1451
 Allow Cross Section Reconstruction---------------------------No  2725 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2725 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2725 1451
 Deleted Duplicate Points-------------------------------------Yes 2725 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2725 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2725 1451
 Allow Creation of Missing Sections---------------------------No  2725 1451
 Allow Insertion of Energy Points-----------------------------No  2725 1451
 Create Uniform Energy Grid-----------------------------------No  2725 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2725 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2725 1451
 Unshielded Group Averages Using  640 Groups                      2725 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2725 1451
                                1        451        244          12725 1451
                                3         16         35          12725 1451
                                3        102        217          12725 1451
                                3        107         75          12725 1451
                               33         16         42          12725 1451
                               33        102         11          12725 1451
                               33        107        103          12725 1451
                                                                  2725 1  0
                                                                  2725 0  0
 2.70590E+4 5.84269E+1          0          0          0          02725 3 16
-1.04530E+7-1.04530E+7          0          0          1         952725 3 16
         95          1                                            2725 3 16
 10600000.0 .001438929 10700000.0 .007329788 10800000.0 .0135365962725 3 16
 10900000.0 .026794286 11000000.0 .047102857 11100000.0 .0674114292725 3 16
 11200000.0 .088045429 11300000.0 .110632000 11400000.0 .1335440002725 3 16
 11500000.0 .156456000 11600000.0 .179368000 11700000.0 .2030660002725 3 16
 11800000.0 .231480000 11900000.0 .260680000 12000000.0 .2898800002725 3 16
 12100000.0 .319080000 12200000.0 .348198750 12300000.0 .3768300002725 3 16
 12400000.0 .405380000 12500000.0 .433930000 12600000.0 .4624800002725 3 16
 12700000.0 .489566125 12800000.0 .507869000 12900000.0 .5247080002725 3 16
 13000000.0 .541547000 13100000.0 .558386000 13200000.0 .5752250002725 3 16
 13300000.0 .592064000 13400000.0 .608903000 13500000.0 .6257420002725 3 16
 13600000.0 .642581000 13700000.0 .658802125 13800000.0 .6713160002725 3 16
 13900000.0 .683212000 14000000.0 .695108000 14100000.0 .7087826252725 3 16
 14200000.0 .726805000 14300000.0 .744663875 14400000.0 .7545780002725 3 16
 14500000.0 .761242000 14600000.0 .767676500 14700000.0 .7735500002725 3 16
 14800000.0 .779190750 14900000.0 .780272000 15000000.0 .7794880002725 3 16
 15100000.0 .778704000 15200000.0 .777920000 15300000.0 .7771360002725 3 16
 15400000.0 .776352000 15500000.0 .777791000 15600000.0 .7814530002725 3 16
 15700000.0 .785115000 15800000.0 .788777000 15900000.0 .7924390002725 3 16
 16000000.0 .796101000 16100000.0 .799763000 16200000.0 .8034250002725 3 16
 16300000.0 .807087000 16400000.0 .810749000 16500000.0 .8140205002725 3 16
 16600000.0 .816901500 16700000.0 .819782500 16800000.0 .8226635002725 3 16
 16900000.0 .825544500 17000000.0 .828425500 17100000.0 .8313065002725 3 16
 17200000.0 .834187500 17300000.0 .837068500 17400000.0 .8399495002725 3 16
 17500000.0 .842123500 17600000.0 .843590500 17700000.0 .8450575002725 3 16
 17800000.0 .846524500 17900000.0 .847991500 18000000.0 .8494585002725 3 16
 18100000.0 .850925500 18200000.0 .852392500 18300000.0 .8538595002725 3 16
 18400000.0 .855326500 18500000.0 .857348000 18600000.0 .8599240002725 3 16
 18700000.0 .862500000 18800000.0 .865076000 18900000.0 .8676520002725 3 16
 19000000.0 .870228000 19100000.0 .872804000 19200000.0 .8753800002725 3 16
 19300000.0 .877956000 19400000.0 .880532000 19500000.0 .8801300002725 3 16
 19600000.0 .876750000 19700000.0 .873370000 19800000.0 .8699900002725 3 16
 19900000.0 .866610000 20000000.0 0.0                             2725 3 16
                                                                  2725 3  0
 2.70590E+4 5.84269E+1          0          0          0          02725 3102
 7.49193E+6 7.49193E+6          0          0          1        6412725 3102
        641          1                                            2725 3102
 .000100000 584.185484 .000105000 570.563242 .000110000 557.6219562725 3102
 .000115000 545.696918 .000120000 531.756360 .000127500 516.2896652725 3102
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 .000160000 460.476357 .000170000 447.154582 .000180000 434.9262022725 3102
 .000190000 423.548731 .000200000 413.160989 .000210000 403.3878732725 3102
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 .000255000 365.097432 .000270000 356.702740 .000280000 347.3579002725 3102
 .000300000 335.961605 .000320000 325.642666 .000340000 316.1921912725 3102
 .000360000 307.487740 .000380000 299.541707 .000400000 291.2575392725 3102
 .000425000 282.780390 .000450000 275.035058 .000475000 267.9339852725 3102
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 .001000000 184.759123 .001050000 180.409336 .001100000 176.3314052725 3102
 .001150000 172.573888 .001200000 168.141416 .001275000 163.2699132725 3102
 .001350000 158.816071 .001425000 154.659491 .001500000 150.2435252725 3102
 .001600000 145.626137 .001700000 141.411001 .001800000 137.5125602725 3102
 .001900000 133.951067 .002000000 130.648982 .002100000 127.5504692725 3102
 .002200000 124.712865 .002300000 121.999453 .002400000 118.9035222725 3102
 .002550000 115.463431 .002700000 112.772659 .002800000 109.8489012725 3102
 .003000000 106.249757 .003200000 102.974513 .003400000 99.97936642725 3102
 .003600000 97.2475146 .003800000 94.7251396 .004000000 92.09611472725 3102
 .004250000 89.4267548 .004500000 86.9905145 .004750000 84.71376542725 3102
 .005000000 82.6246700 .005250000 80.6883682 .005500000 78.85818172725 3102
 .005750000 77.1811183 .006000000 75.4199577 .006300000 73.65974462725 3102
 .006600000 71.9940432 .006900000 70.4491798 .007200000 68.76929032725 3102
 .007600000 66.9727243 .008000000 65.3295245 .008400000 63.78744202725 3102
 .008800000 62.3526170 .009200000 61.0193559 .009600000 59.75075432725 3102
 .010000000 58.4373145 .010500000 57.0487330 .011000000 55.77855022725 3102
 .011500000 54.5667817 .012000000 53.1779379 .012750000 51.64049772725 3102
 .013500000 50.2165753 .014250000 48.9106874 .015000000 47.51873502725 3102
 .016000001 46.0583058 .017000001 44.7196792 .017999999 43.49778192725 3102
 .018999999 42.3743034 .020000000 41.3198262 .021000000 40.35374692725 3102
 .022000000 39.4428172 .023000000 38.5932562 .024000000 37.59479382725 3102
 .025500000 36.5191359 .027000001 35.6680169 .028000001 34.74991352725 3102
 .029999999 33.6156775 .032000002 32.5790486 .034000002 31.63233972725 3102
 .035999998 30.7679543 .037999999 29.9706518 .039999999 29.13878112725 3102
 .042500000 28.2946696 .045000002 27.5237876 .047499999 26.80597762725 3102
 .050000001 26.1461476 .052499998 25.5356154 .055000000 24.95757972725 3102
 .057500001 24.4280849 .059999999 23.8711254 .063000001 23.31360472725 3102
 .066000000 22.7880144 .068999998 22.2982197 .071999997 21.76707862725 3102
 .075999998 21.1981495 .079999998 20.6785003 .083999999 20.19026492725 3102
 .088000000 19.7365257 .092000000 19.3167529 .096000001 18.91609842725 3102
 .100000001 18.5027050 .104999997 18.0651507 .109999999 17.66481252725 3102
 .115000002 17.2821127 .119999997 16.8430941 .127499998 16.35769482725 3102
 .135000005 15.9075233 .142499998 15.4944076 .150000006 15.05414422725 3102
 .159999996 14.5914061 .170000002 14.1681534 .180000007 13.78148202725 3102
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 1.35000002 5.10402280 1.42499995 4.97655969 1.50000000 4.838975522725 3102
 1.60000002 4.69647418 1.70000005 4.56795002 1.79999995 4.446391762725 3102
 1.89999998 4.33773325 2.00000000 4.23317671 2.09999990 4.138702352725 3102
 2.20000005 4.05025180 2.29999995 3.96751505 2.40000010 3.872481342725 3102
 2.54999995 3.76644166 2.70000005 3.68510595 2.79999995 3.595604182725 3102
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 3.59999990 3.21430758 3.79999995 3.13901263 4.00000000 3.061641802725 3102
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 6.59999990 2.47319201 6.90000010 2.42851466 7.19999981 2.381156052725 3102
 7.59999990 2.33129672 8.00000000 2.28482332 8.39999962 2.242819442725 3102
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 10.0000000 2.09858408 10.5000000 2.06374093 11.0000000 2.029882452725 3102
 11.5000000 2.00059471 12.0000000 1.96527507 12.7500000 1.928926612725 3102
 13.5000000 1.89508432 14.2500000 1.86583401 15.0000000 1.833839692725 3102
 16.0000000 1.80300729 17.0000000 1.77545290 18.0000000 1.753254082725 3102
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 22.0000000 1.68654792 23.0000000 1.67685866 24.0000000 1.665829862725 3102
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 30.0000000 1.64945128 32.0000000 1.65522530 34.0000000 1.665150142725 3102
 36.0000000 1.67976876 38.0000000 1.69920005 40.0000000 1.725084232725 3102
 42.5000000 1.76133671 45.0000000 1.80377452 47.5000000 1.853677912725 3102
 50.0000000 1.91006593 52.5000000 1.97562416 55.0000000 2.050340302725 3102
 57.5000000 2.13426946 60.0000000 2.24029046 63.0000000 2.373197762725 3102
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 76.0000000 3.29803446 80.0000000 3.73022715 84.0000000 4.279924682725 3102
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 135.000000 161.070780 142.500000 34.2835176 150.000000 12.91275602725 3102
 160.000000 6.02280714 170.000000 3.45465685 180.000000 2.230757592725 3102
 190.000000 1.55619348 200.000000 1.14738292 210.000000 .8819609392725 3102
 220.000000 .700077257 230.000000 .570389340 240.000000 .4558813282725 3102
 255.000000 .359732141 270.000000 .302080739 280.000000 .2520402552725 3102
 300.000000 .203642336 320.000000 .169589690 340.000000 .1447280782725 3102
 360.000000 .125984450 380.000000 .111526437 400.000000 .0988766112725 3102
 425.000000 .087844493 450.000000 .079197660 475.000000 .0722661712725 3102
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 660.000000 .045807498 690.000000 .043569285 720.000000 .0413140722725 3102
 760.000000 .039090543 800.000000 .037191255 840.000000 .0355291862725 3102
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 1000.00000 .030435533 1050.00000 .029296449 1100.00000 .0282827922725 3102
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 4250.00000 .766931416 4500.00000 .265557892 4750.00000 .4090593042725 3102
 5000.00000 .403888113 5250.00000 .216112505 5500.00000 .1084731532725 3102
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 11500.0000 .091434505 12000.0000 .020155530 12750.0000 .1367945802725 3102
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 25500.0000 .040126448 27000.0000 .057127672 28000.0000 .0251416672725 3102
 30000.0000 .045085071 32000.0000 .037100272 34000.0000 .0300135722725 3102
 36000.0000 .017753426 38000.0000 .007222453 40000.0000 .0194958662725 3102
 42500.0000 .009261574 45000.0000 .029319495 47500.0000 .0088850252725 3102
 50000.0000 .026627246 52500.0000 .027305394 55000.0000 .0054548832725 3102
 57500.0000 .020371139 60000.0000 .027713474 63000.0000 .0140446352725 3102
 66000.0000 .018176031 69000.0000 .011847318 72000.0000 .0135827212725 3102
 76000.0000 .028023729 80000.0000 .013093373 84000.0000 .0138250172725 3102
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 100000.000 .019674606 105000.000 .019167782 110000.000 .0186598082725 3102
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 135000.000 .015992944 142500.000 .015231690 150000.000 .0145825342725 3102
 160000.000 .014037000 170000.000 .013495000 180000.000 .0129530002725 3102
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 220000.000 .011315000 230000.000 .010985000 240000.000 .0105863762725 3102
 255000.000 .010219410 270000.000 .009945000 280000.000 .0096180282725 3102
 300000.000 .009251546 320000.000 .008949102 340000.000 .0086536292725 3102
 360000.000 .008418254 380000.000 .008196718 400000.000 .0079889382725 3102
 425000.000 .007806250 450000.000 .007623750 475000.000 .0074416502725 3102
 500000.000 .007301663 525000.000 .007189934 550000.000 .0070782042725 3102
 575000.000 .006966475 600000.000 .006871813 630000.000 .0067954662725 3102
 660000.000 .006719122 690000.000 .006647339 720000.000 .0065954332725 3102
 760000.000 .006541062 800000.000 .006496855 840000.000 .0064701212725 3102
 880000.000 .006445406 920000.000 .006440000 960000.000 .0064400032725 3102
 1000000.00 .006464973 1100000.00 .006269954 1200000.00 .0057074862725 3102
 1300000.00 .005022523 1400000.00 .004485038 1500000.00 .0040950332725 3102
 1600000.00 .003806278 1700000.00 .003618774 1800000.00 .0034312692725 3102
 1900000.00 .003243765 2000000.00 .003095013 2100000.00 .0029850122725 3102
 2200000.00 .002875011 2300000.00 .002765010 2400000.00 .0026550092725 3102
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 3100000.00 .002160003 3200000.00 .002100003 3300000.00 .0020400032725 3102
 3400000.00 .001980004 3500000.00 .001920004 3600000.00 .0018600052725 3102
 3700000.00 .001800005 3800000.00 .001740005 3900000.00 .0016800062725 3102
 4000000.00 .001632506 4100000.00 .001597506 4200000.00 .0015625052725 3102
 4300000.00 .001527505 4400000.00 .001492505 4500000.00 .0014575042725 3102
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 4900000.00 .001317503 5000000.00 .001287503 5100000.00 .0012625032725 3102
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 5800000.00 .001087502 5900000.00 .001062502 6000000.00 .0010412522725 3102
 6100000.00 .001023752 6200000.00 .001006252 6300000.00 .0009887522725 3102
 6400000.00 .000971252 6500000.00 .000953753 6600000.00 .0009362532725 3102
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 5.070840-4 4.403860-4 3.788130-4 3.232970-4 2.750060-4 2.352810-4272533107
 2.052130-4 1.851250-4 1.743350-4 1.713250-4 1.741730-4 1.809930-4272533107
 1.902140-4 2.006690-4 2.115680-4 2.224170-4 2.329320-4 2.429660-4272533107
 2.524580-4 8.183640-4 8.118980-4 7.620090-4 6.844360-4 5.926450-4272533107
 4.972630-4 4.067960-4 3.281370-4 2.664180-4 2.243300-4 2.015050-4272533107
 1.946250-4 1.984970-4 2.076070-4 2.174110-4 2.249110-4 2.286000-4272533107
 2.280690-4 2.235730-4 2.156850-4 2.050780-4 1.923990-4 1.782160-4272533107
 8.478330-4 8.276790-4 7.647320-4 6.738120-4 5.686930-4 4.617520-4272533107
 3.639960-4 2.845050-4 2.291860-4 1.995590-4 1.925450-4 2.017460-4272533107
 2.196390-4 2.395840-4 2.569040-4 2.690010-4 2.749150-4 2.747480-4272533107
 2.691630-4 2.590450-4 2.453050-4 2.287740-4 8.395720-4 8.040470-4272533107
 7.322570-4 6.362260-4 5.281700-4 4.203680-4 3.245530-4 2.503860-4272533107
 2.035030-4 1.842990-4 1.884050-4 2.086620-4 2.374990-4 2.686370-4272533107
 2.977680-4 3.224500-4 3.416560-4 3.552660-4 3.636700-4 3.674940-4272533107
 3.674300-4 8.006990-4 7.588770-4 6.856440-4 5.899000-4 4.825730-4272533107
 3.762830-4 2.838360-4 2.156520-4 1.772910-4 1.685220-4 1.844370-4272533107
 2.177930-4 2.612660-4 3.088450-4 3.562780-4 4.009380-4 4.414360-4272533107
 4.772250-4 5.082750-4 5.348470-4 7.504010-4 7.072690-4 6.334420-4272533107
 5.366440-4 4.286200-4 3.239150-4 2.369800-4 1.785680-4 1.532970-4272533107
 1.595740-4 1.915490-4 2.416910-4 3.028130-4 3.691110-4 4.364210-4272533107
 5.020320-4 5.643450-4 6.225410-4 6.763130-4 6.954900-4 6.488650-4272533107
 5.711140-4 4.716320-4 3.648490-4 2.673910-4 1.938160-4 1.529760-4272533107
 1.467730-4 1.714680-4 2.202600-4 2.856770-4 3.610820-4 4.412890-4272533107
 5.225880-4 6.025210-4 6.795740-4 7.529000-4 6.304990-4 5.779750-4272533107
 4.974380-4 4.012420-4 3.056610-4 2.266320-4 1.755580-4 1.571880-4272533107
 1.701860-4 2.093370-4 2.679510-4 3.395610-4 4.187810-4 5.015220-4272533107
 5.848930-4 6.669610-4 7.465100-4 5.538450-4 5.011350-4 4.285190-4272533107
 3.492380-4 2.776720-4 2.253930-4 1.986950-4 1.984520-4 2.216480-4272533107
 2.633770-4 3.184200-4 3.821410-4 4.508370-4 5.217500-4 5.929330-4272533107
 6.630700-4 4.818940-4 4.435680-4 3.942170-4 3.438620-4 3.016870-4272533107
 2.738300-4 2.627190-4 2.677410-4 2.864790-4 3.158430-4 3.528180-4272533107
 3.948190-4 4.398050-4 4.862480-4 5.330440-4 4.454760-4 4.358260-4272533107
 4.183730-4 3.977630-4 3.782510-4 3.628080-4 3.529130-4 3.488350-4272533107
 3.500810-4 3.558030-4 3.650640-4 3.769920-4 3.908400-4 4.060010-4272533107
 4.694740-4 4.919110-4 5.021020-4 5.010390-4 4.910640-4 4.749580-4272533107
 4.552730-4 4.340100-4 4.125720-4 3.918450-4 3.723330-4 3.542780-4272533107
 3.377530-4 5.551840-4 6.013500-4 6.274720-4 6.341600-4 6.243770-4272533107
 6.020350-4 5.709710-4 5.344320-4 4.949130-4 4.542130-4 4.135620-4272533107
 3.737590-4 6.840580-4 7.432410-4 7.770160-4 7.873120-4 7.781630-4272533107
 7.542020-4 7.197360-4 6.783370-4 6.327650-4 5.850490-4 5.366280-4272533107
 8.380390-4 9.069060-4 9.497230-4 9.693530-4 9.700350-4 9.561740-4272533107
 9.317100-4 8.998770-4 8.631920-4 8.235440-4 1.015840-3 1.100660-3272533107
 1.162040-3 1.202680-3 1.226130-3 1.235970-3 1.235370-3 1.226980-3272533107
 1.212910-3 1.233620-3 1.346000-3 1.438180-3 1.512010-3 1.570060-3272533107
 1.614950-3 1.649110-3 1.674650-3 1.515040-3 1.666300-3 1.799460-3272533107
 1.915510-3 2.016160-3 2.103310-3 2.178850-3 1.880800-3 2.078470-3272533107
 2.258530-3 2.421410-3 2.568290-3 2.700660-3 2.343020-3 2.589960-3272533107
 2.818450-3 3.028720-3 3.221760-3 2.904570-3 3.199650-3 3.474650-3272533107
 3.729920-3 3.560920-3 3.900280-3 4.217700-3 4.302840-3 4.681140-3272533107
 5.118740-3                                                       272533107
                                                                  272533  0
                                                                  2725 0  0
                                                                     0 0  0
 2.80000+ 4 0.0    + 0         -1          0         34          12800 1451
 0.0    + 0 0.0    + 0          0          0          0          62800 1451
 1.0    + 0 0.0    + 0          0          0         10          22800 1451
 0.0    + 0 0.0    + 0          0          0         19          22800 1451
 28-NI-  0 PETTEN     EVAL-85    W.L.ZIJP                         2800 1451
PRIVATE COMM.         DIST-MAR85                       850320     2800 1451
----IRDF-90/NMF-G     MATERIAL 2800                               2800 1451
-----INCIDENT NEUTRON DATA                                        2800 1451
------ENDF-6 FORMAT                                               2800 1451
                                                                  2800 1451
DAMAGE CROSS SECTIONS OF NICKEL IN STEEL                          2800 1451
ASTM STANDARD                                                     2800 1451
                                                                  2800 1451
 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2800 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.100 PER-CENT      2800 1451
 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2800 1451
 UNSHIELDED GROUP AVERAGES USING  640 GROUPS (SAND-II EXTEND)     2800 1451
                                                                  2800 1451
##################################################################2800 1451
MODIFICATIONS RELATED TO THE VERSION IRDF-90/NMF-G ARE DESCRIBED  2800 1451
IN THE REPORT INDC(HUN)-31/G+R BY E. J. SZONDI (APR. 1994)        2800 1451
##################################################################2800 1451
                                                                  2800 1451
                                1        451         25          02800 1451
                                3        900        116          02800 1451
                                                                  2800 1  0
                                                                  2800 0  0
 2.80000+ 4 0.0    + 0          0          0          0          02800 3900
 0.0    + 0 0.0    + 0          0          0          1        3382800 3900
        338          1                                            2800 3900
 5.75000+ 2 1.44000+ 0 6.00000+ 2 1.44000+ 0 6.30000+ 2 1.44000+ 02800 3900
 6.60000+ 2 1.44000+ 0 6.90000+ 2 1.44000+ 0 7.20000+ 2 1.44000+ 02800 3900
 7.60000+ 2 4.75900+ 0 8.00000+ 2 4.75900+ 0 8.40000+ 2 4.75900+ 02800 3900
 8.80000+ 2 4.75900+ 0 9.20000+ 2 4.75900+ 0 9.60000+ 2 7.46500+ 02800 3900
 1.00000+ 3 7.46500+ 0 1.05000+ 3 7.46500+ 0 1.10000+ 3 7.46500+ 02800 3900
 1.15000+ 3 7.46500+ 0 1.20000+ 3 7.46500+ 0 1.27500+ 3 8.68999+ 02800 3900
 1.35000+ 3 8.68999+ 0 1.42500+ 3 8.68999+ 0 1.50000+ 3 8.68999+ 02800 3900
 1.60000+ 3 9.68200+ 0 1.70000+ 3 9.68200+ 0 1.80000+ 3 9.68200+ 02800 3900
 1.90000+ 3 9.68200+ 0 2.00000+ 3 1.17400+ 1 2.10000+ 3 1.17400+ 12800 3900
 2.20000+ 3 1.17400+ 1 2.30000+ 3 1.17400+ 1 2.40000+ 3 1.17400+ 12800 3900
 2.55000+ 3 1.17400+ 1 2.70000+ 3 1.57600+ 1 2.80000+ 3 1.57600+ 12800 3900
 3.00000+ 3 1.57600+ 1 3.20000+ 3 1.57600+ 1 3.40000+ 3 2.97500+ 12800 3900
 3.60000+ 3 2.97500+ 1 3.80000+ 3 2.97500+ 1 4.00000+ 3 2.97500+ 12800 3900
 4.25000+ 3 2.97500+ 1 4.50000+ 3 4.07100+ 1 4.75000+ 3 4.07100+ 12800 3900
 5.00000+ 3 4.07100+ 1 5.25000+ 3 4.07100+ 1 5.50000+ 3 2.59000+ 12800 3900
 5.75000+ 3 2.59000+ 1 6.00000+ 3 2.59000+ 1 6.30000+ 3 2.59000+ 12800 3900
 6.60000+ 3 2.59000+ 1 6.90000+ 3 2.59000+ 1 7.20000+ 3 2.01200+ 12800 3900
 7.60000+ 3 2.01200+ 1 8.00000+ 3 2.01200+ 1 8.40000+ 3 2.01200+ 12800 3900
 8.80000+ 3 2.01200+ 1 9.20000+ 3 5.45400+ 1 9.60000+ 3 5.45400+ 12800 3900
 1.00000+ 4 5.45400+ 1 1.05000+ 4 5.45400+ 1 1.10000+ 4 5.45400+ 12800 3900
 1.15000+ 4 5.45400+ 1 1.20000+ 4 1.76200+ 2 1.27500+ 4 1.76200+ 22800 3900
 1.35000+ 4 1.76200+ 2 1.42500+ 4 1.76200+ 2 1.50000+ 4 3.05300+ 22800 3900
 1.60000+ 4 3.05300+ 2 1.70000+ 4 3.05300+ 2 1.80000+ 4 3.05300+ 22800 3900
 1.90000+ 4 1.01900+ 2 2.00000+ 4 1.01900+ 2 2.10000+ 4 1.01900+ 22800 3900
 2.20000+ 4 1.01900+ 2 2.30000+ 4 1.01900+ 2 2.40000+ 4 1.01900+ 22800 3900
 2.55000+ 4 1.03800+ 2 2.70000+ 4 1.03800+ 2 2.80000+ 4 1.03800+ 22800 3900
 3.00000+ 4 1.03800+ 2 3.20000+ 4 8.69000+ 1 3.40000+ 4 8.69000+ 12800 3900
 3.60000+ 4 8.69000+ 1 3.80000+ 4 8.69000+ 1 4.00000+ 4 6.99300+ 12800 3900
 4.25000+ 4 6.99300+ 1 4.50000+ 4 6.99300+ 1 4.75000+ 4 6.99300+ 12800 3900
 5.00000+ 4 6.99300+ 1 5.25000+ 4 1.46500+ 2 5.50000+ 4 1.46500+ 22800 3900
 5.75000+ 4 1.46500+ 2 6.00000+ 4 1.46500+ 2 6.30000+ 4 1.46500+ 22800 3900
 6.60000+ 4 1.68700+ 2 6.90000+ 4 1.68700+ 2 7.20000+ 4 1.68700+ 22800 3900
 7.60000+ 4 1.68700+ 2 8.00000+ 4 1.68700+ 2 8.40000+ 4 1.68700+ 22800 3900
 8.80000+ 4 1.39800+ 2 9.20000+ 4 1.39800+ 2 9.60000+ 4 1.39800+ 22800 3900
 1.00000+ 5 1.39800+ 2 1.05000+ 5 1.39800+ 2 1.10000+ 5 1.11000+ 22800 3900
 1.15000+ 5 1.11000+ 2 1.20000+ 5 1.11000+ 2 1.27500+ 5 1.11000+ 22800 3900
 1.35000+ 5 2.31700+ 2 1.42500+ 5 2.31700+ 2 1.50000+ 5 2.31700+ 22800 3900
 1.60000+ 5 2.69200+ 2 1.70000+ 5 2.69200+ 2 1.80000+ 5 2.69200+ 22800 3900
 1.90000+ 5 3.63200+ 2 2.00000+ 5 3.63200+ 2 2.10000+ 5 3.63200+ 22800 3900
 2.20000+ 5 3.36700+ 2 2.30000+ 5 3.36700+ 2 2.40000+ 5 3.36700+ 22800 3900
 2.55000+ 5 3.35700+ 2 2.70000+ 5 3.35700+ 2 2.80000+ 5 3.53500+ 22800 3900
 3.00000+ 5 3.71300+ 2 3.20000+ 5 3.93400+ 2 3.40000+ 5 3.93400+ 22800 3900
 3.60000+ 5 3.19600+ 2 3.80000+ 5 3.19600+ 2 4.00000+ 5 3.51000+ 22800 3900
 4.25000+ 5 3.51000+ 2 4.50000+ 5 4.07400+ 2 4.75000+ 5 4.07400+ 22800 3900
 5.00000+ 5 3.97800+ 2 5.25000+ 5 3.97800+ 2 5.50000+ 5 3.42000+ 22800 3900
 5.75000+ 5 3.42000+ 2 6.00000+ 5 4.88600+ 2 6.30000+ 5 4.88600+ 22800 3900
 6.60000+ 5 4.20100+ 2 6.90000+ 5 4.20100+ 2 7.20000+ 5 4.00200+ 22800 3900
 7.60000+ 5 4.62100+ 2 8.00000+ 5 5.24000+ 2 8.40000+ 5 5.48900+ 22800 3900
 8.80000+ 5 5.48900+ 2 9.20000+ 5 5.39000+ 2 9.60000+ 5 5.39000+ 22800 3900
 1.00000+ 6 5.96500+ 2 1.10000+ 6 5.96500+ 2 1.20000+ 6 8.03000+ 22800 3900
 1.30000+ 6 8.03000+ 2 1.40000+ 6 8.06300+ 2 1.50000+ 6 8.06300+ 22800 3900
 1.60000+ 6 7.74800+ 2 1.70000+ 6 7.74800+ 2 1.80000+ 6 8.19900+ 22800 3900
 1.90000+ 6 8.19900+ 2 2.00000+ 6 9.15000+ 2 2.10000+ 6 9.15000+ 22800 3900
 2.20000+ 6 9.15000+ 2 2.30000+ 6 1.02000+ 3 2.40000+ 6 1.02000+ 32800 3900
 2.50000+ 6 1.02000+ 3 2.60000+ 6 1.13800+ 3 2.70000+ 6 1.13800+ 32800 3900
 2.80000+ 6 1.13800+ 3 2.90000+ 6 1.13000+ 3 3.00000+ 6 1.13000+ 32800 3900
 3.10000+ 6 1.13000+ 3 3.20000+ 6 1.13000+ 3 3.30000+ 6 1.27300+ 32800 3900
 3.40000+ 6 1.27300+ 3 3.50000+ 6 1.27300+ 3 3.60000+ 6 1.27300+ 32800 3900
 3.70000+ 6 1.32100+ 3 3.80000+ 6 1.32100+ 3 3.90000+ 6 1.32100+ 32800 3900
 4.00000+ 6 1.32100+ 3 4.10000+ 6 1.60400+ 3 4.20000+ 6 1.60400+ 32800 3900
 4.30000+ 6 1.60400+ 3 4.40000+ 6 1.60400+ 3 4.50000+ 6 1.68600+ 32800 3900
 4.60000+ 6 1.68600+ 3 4.70000+ 6 1.68600+ 3 4.80000+ 6 1.68600+ 32800 3900
 4.90000+ 6 1.68600+ 3 5.00000+ 6 1.79800+ 3 5.10000+ 6 1.79800+ 32800 3900
 5.20000+ 6 1.79800+ 3 5.30000+ 6 1.79800+ 3 5.40000+ 6 1.79800+ 32800 3900
 5.50000+ 6 1.86900+ 3 5.60000+ 6 1.86900+ 3 5.70000+ 6 1.86900+ 32800 3900
 5.80000+ 6 1.86900+ 3 5.90000+ 6 1.86900+ 3 6.00000+ 6 2.07400+ 32800 3900
 6.10000+ 6 2.07400+ 3 6.20000+ 6 2.07400+ 3 6.30000+ 6 2.07400+ 32800 3900
 6.40000+ 6 2.07400+ 3 6.50000+ 6 2.07400+ 3 6.60000+ 6 2.07400+ 32800 3900
 6.70000+ 6 2.18800+ 3 6.80000+ 6 2.18800+ 3 6.90000+ 6 2.18800+ 32800 3900
 7.00000+ 6 2.18800+ 3 7.10000+ 6 2.18800+ 3 7.20000+ 6 2.18800+ 32800 3900
 7.30000+ 6 2.18800+ 3 7.40000+ 6 2.23600+ 3 7.50000+ 6 2.23600+ 32800 3900
 7.60000+ 6 2.23600+ 3 7.70000+ 6 2.23600+ 3 7.80000+ 6 2.23600+ 32800 3900
 7.90000+ 6 2.23600+ 3 8.00000+ 6 2.23600+ 3 8.10000+ 6 2.23600+ 32800 3900
 8.20000+ 6 2.31500+ 3 8.30000+ 6 2.31500+ 3 8.40000+ 6 2.31500+ 32800 3900
 8.50000+ 6 2.31500+ 3 8.60000+ 6 2.31500+ 3 8.70000+ 6 2.31500+ 32800 3900
 8.80000+ 6 2.31500+ 3 8.90000+ 6 2.31500+ 3 9.00000+ 6 2.56500+ 32800 3900
 9.10000+ 6 2.56500+ 3 9.20000+ 6 2.56500+ 3 9.30000+ 6 2.56500+ 32800 3900
 9.40000+ 6 2.56500+ 3 9.50000+ 6 2.56500+ 3 9.60000+ 6 2.56500+ 32800 3900
 9.70000+ 6 2.56500+ 3 9.80000+ 6 2.56500+ 3 9.90000+ 6 2.56500+ 32800 3900
 1.00000+ 7 2.69300+ 3 1.01000+ 7 2.69300+ 3 1.02000+ 7 2.69300+ 32800 3900
 1.03000+ 7 2.69300+ 3 1.04000+ 7 2.69300+ 3 1.05000+ 7 2.69300+ 32800 3900
 1.06000+ 7 2.69300+ 3 1.07000+ 7 2.69300+ 3 1.08000+ 7 2.69300+ 32800 3900
 1.09000+ 7 2.69300+ 3 1.10000+ 7 2.81800+ 3 1.11000+ 7 2.81800+ 32800 3900
 1.12000+ 7 2.81800+ 3 1.13000+ 7 2.81800+ 3 1.14000+ 7 2.81800+ 32800 3900
 1.15000+ 7 2.81800+ 3 1.16000+ 7 2.81800+ 3 1.17000+ 7 2.81800+ 32800 3900
 1.18000+ 7 2.81800+ 3 1.19000+ 7 2.81800+ 3 1.20000+ 7 2.89400+ 32800 3900
 1.21000+ 7 2.89400+ 3 1.22000+ 7 2.89400+ 3 1.23000+ 7 2.89400+ 32800 3900
 1.24000+ 7 2.89400+ 3 1.25000+ 7 2.89400+ 3 1.26000+ 7 2.89400+ 32800 3900
 1.27000+ 7 2.89400+ 3 1.28000+ 7 2.89400+ 3 1.29000+ 7 2.89400+ 32800 3900
 1.30000+ 7 2.96400+ 3 1.31000+ 7 2.96400+ 3 1.32000+ 7 2.96400+ 32800 3900
 1.33000+ 7 2.96400+ 3 1.34000+ 7 2.96400+ 3 1.35000+ 7 2.96400+ 32800 3900
 1.36000+ 7 2.96400+ 3 1.37000+ 7 2.96400+ 3 1.38000+ 7 2.96400+ 32800 3900
 1.39000+ 7 2.96400+ 3 1.40000+ 7 3.00100+ 3 1.41000+ 7 3.00100+ 32800 3900
 1.42000+ 7 3.00100+ 3 1.43000+ 7 3.00100+ 3 1.44000+ 7 3.00100+ 32800 3900
 1.45000+ 7 3.00100+ 3 1.46000+ 7 3.00100+ 3 1.47000+ 7 3.00100+ 32800 3900
 1.48000+ 7 3.00100+ 3 1.49000+ 7 3.00100+ 3 1.50000+ 7 3.03400+ 32800 3900
 1.51000+ 7 3.03400+ 3 1.52000+ 7 3.03400+ 3 1.53000+ 7 3.03400+ 32800 3900
 1.54000+ 7 3.03400+ 3 1.55000+ 7 3.03400+ 3 1.56000+ 7 3.03400+ 32800 3900
 1.57000+ 7 3.03400+ 3 1.58000+ 7 3.03400+ 3 1.59000+ 7 3.03400+ 32800 3900
 1.60000+ 7 3.09000+ 3 1.61000+ 7 3.09000+ 3 1.62000+ 7 3.09000+ 32800 3900
 1.63000+ 7 3.09000+ 3 1.64000+ 7 3.09000+ 3 1.65000+ 7 3.09000+ 32800 3900
 1.66000+ 7 3.09000+ 3 1.67000+ 7 3.09000+ 3 1.68000+ 7 3.09000+ 32800 3900
 1.69000+ 7 3.09000+ 3 1.70000+ 7 3.24600+ 3 1.71000+ 7 3.24600+ 32800 3900
 1.72000+ 7 3.24600+ 3 1.73000+ 7 3.24600+ 3 1.74000+ 7 3.24600+ 32800 3900
 1.75000+ 7 3.24600+ 3 1.76000+ 7 3.24600+ 3 1.77000+ 7 3.24600+ 32800 3900
 1.78000+ 7 3.24600+ 3 1.79000+ 7 3.24600+ 3 1.80000+ 7 3.33500+ 32800 3900
 1.81000+ 7 3.33500+ 3 1.82000+ 7 3.33500+ 3 1.83000+ 7 3.33500+ 32800 3900
 1.84000+ 7 3.33500+ 3 1.85000+ 7 3.33500+ 3 1.86000+ 7 3.33500+ 32800 3900
 1.87000+ 7 3.33500+ 3 1.88000+ 7 3.33500+ 3 1.89000+ 7 3.33500+ 32800 3900
 1.90000+ 7 3.37500+ 3 1.91000+ 7 3.37500+ 3 1.92000+ 7 3.37500+ 32800 3900
 1.93000+ 7 3.37500+ 3 1.94000+ 7 3.37500+ 3 1.95000+ 7 3.37500+ 32800 3900
 1.96000+ 7 3.37500+ 3 1.97000+ 7 3.37500+ 3 1.98000+ 7 3.37500+ 32800 3900
 1.99000+ 7 3.37500+ 3 2.00000+ 7 0.0    + 0                      2800 3900
                                                                  2800 3  0
                                                                  2800 0  0
                                                                     0 0  0
 2.80580E+4 5.74376E+1          0          0         34         102825 1451
 0.0        0.0                 0          0          0          62825 1451
 1.00000E+0 2.00000E+7          0          0         10       20022825 1451
 3.00000E+2 0.0                 1          0        378          52825 1451
 28-Ni- 58 IRK-IJS/FEIEVAL-AUG99 EUR.JOINT COLLAB. K.I.Zolotarev  2825 1451
                      DIST-Feb2004                                2825 1451
----IRDF-2002         MATERIAL 2825                               2825 1451
-----INCIDENT NEUTRON DATA                                        2825 1451
------ENDF-6 FORMAT                                               2825 1451
                                                                  2825 1451
*****************************   JEFF-3.0   ***********************2825 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2825 1451
DOSIMETRY                                                         2825 1451
******************************************************************2825 1451
 28-NI- 58 IRK-IJS    EVAL-AUG99 EUROPEAN JOINT COLLABORATION     2825 1451
                                                                  2825 1451
   DATA TAKEN FROM   :-   EFF-3.1 (DIST-AUG99 REV1-SEP00)         2825 1451
MF=3                                                              2825 1451
MT= 16 - (n,2n)                                                   2825 1451
                                                                  2825 1451
******************************************************************2825 1451
 28-Ni- 58 FEI        EVAL-Dec02 K.I.Zolotarev                    2825 1451
                      DIST-Jan03                                  2825 1451
----BROND-3           MATERIAL 2825                               2825 1451
-----INCIDENT NEUTRON DATA                                        2825 1451
------ENDF-6 FORMAT                                               2825 1451
***************************************************************** 2825 1451
------Russian Reactor Dosimetry File RRDF-2002                    2825 1451
***************************************************************** 2825 1451
Author  of evaluation: K.I.Zolotarev                              2825 1451
***************************************************************** 2825 1451
MF=3                                                              2825 1451
MT=103 - Ni58(n,p)Co58m+g reaction                                2825 1451
                                                                  2825 1451
***************************************************************** 2825 1451
******* Start of JEFF-3.0 (N,2N) bibliographical component ****** 2825 1451
***************************************************************** 2825 1451
*****************************   JEFF-3.0   ***********************2825 1451
 28-NI- 58 IRK-IJS    EVAL-AUG99 EUROPEAN JOINT COLLABORATION     2825 1451
                                                                  2825 1451
   DATA TAKEN FROM   :-   EFF-3.1 (DIST-AUG99 REV1-SEP00)         2825 1451
                                                                  2825 1451
******************************************************************2825 1451
Authors and Responsibilities:                                     2825 1451
S.Tagesen, H.Vonach and A.Wallner, I.R.K.:                        2825 1451
 - Complete evaluation of the cross sections including covariance 2825 1451
   matrices by generalized least squares cross section update     2825 1451
   code GLUCS (9, 10). The total cross section is evaluated in    2825 1451
   broad energy bins.                                             2825 1451
A.Trkov, I.J.S.:                                                  2825 1451
 - Final assembly of the file.                                    2825 1451
 - Consistency corrections (interp.law in MF6 starter file).      2825 1451
 - Implementation of the resonance fluctuations on the smooth     2825 1451
   newly evaluated cross sections.                                2825 1451
 - Preliminary data verification and benchmarking.                2825 1451
                                                                  2825 1451
Evaluation Details:                                               2825 1451
The Oak Ridge Ni-58 ENDF/B-VI Revision 1 evaluation (MAT 2825)    2825 1451
by Larson et.al.  was the chosen starter file. All neutron        2825 1451
cross sections above 810 keV were re-evaluated and include        2825 1451
covariance data. The following data sets were selected as         2825 1451
"priors":                                                         2825 1451
        MT16     IRDF-90  (evaluation Pavlik)(11)                 2825 1451
                                                                  2825 1451
Resonance Fluctuations in the Cross Sections:                     2825 1451
The cross sections below 810 keV are not affected by the re-      2825 1451
evaluation process. Above this energy, broad bin average total    2825 1451
cross section from the starter file was calculated. The bins      2825 1451
correspond to those used in the re-evaluation process. A smooth   2825 1451
cross section curve was generated, conserving the bin average     2825 1451
values. Fluctuations modulating function was defined as the       2825 1451
ratio of the original and the smoothed cross section. This        2825 1451
modulating function is applied on the re-evaluated smooth total   2825 1451
cross section and the inelastic cross sections.                   2825 1451
                                                                  2825 1451
The remaining comments are taken over from the starter file,      2825 1451
except for the sections referring to the data, which have been    2825 1451
superseeded. Modified sections are identified by the use of       2825 1451
lower case characters.                                            2825 1451
                                                                  2825 1451
******************************************************************2825 1451
 CAPTURE WIDTHS CORRECTED FOR 58.7 AND 439.52 KEV RESONANCES.     2825 1451
 THE ELASTIC TRANSFORMATION MATRIX WAS REMOVED.                   2825 1451
 FIXED TYPO IN MINIMUM ENERGY IN MF=6, MT=51                      2825 1451
******************************************************************2825 1451
                                                                  2825 1451
   THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE           2825 1451
DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1)        2825 1451
AND THE HAUSER-FESHBACH CODE TNG (2,3,4).  THE TNG CODE PROVIDES  2825 1451
ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE      2825 1451
COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2825 1451
REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2825 1451
CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF          2825 1451
CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).2825 1451
                                                                  2825 1451
----- DESCRIPTION OF FILES                                        2825 1451
(MF-MT)                                                           2825 1451
  1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS.         2825 1451
  3-16  (n,2n) cross sections were re-evaluated.                  2825 1451
--------------------------------------------------------------    2825 1451
                UNCERTAINTY FILES                                 2825 1451
        ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS       2825 1451
        REQUIRED BY ENDF/B-VI FORMATS                             2825 1451
        For all evaluated reactions full covariance matrices are  2825 1451
        given as calculated by the bayesian evaluation update code2825 1451
        GLUCS. This includes full inter-reaction covariances.     2825 1451
                                                                  2825 1451
 33-16  (N,2N) covariances from GLUCS.                            2825 1451
                                                                  2825 1451
 REFERENCES:                                                      2825 1451
(11) M. Wagner et al., Physics Data 13-5, Fachinformationszentrum 2825 1451
     Karlsruhe, 1990                                              2825 1451
***************************************************************** 2825 1451
*******  End of JEFF-3.0 (N,2N) bibliographical component ******* 2825 1451
***************************************************************** 2825 1451
***************************************************************** 2825 1451
*******   Start of RRDF (N,P) bibliographical component   ******* 2825 1451
***************************************************************** 2825 1451
 28-Ni- 58 FEI        EVAL-Dec02 K.I.Zolotarev                    2825 1451
                      DIST-Jan03                                  2825 1451
----BROND-3           MATERIAL 2825                               2825 1451
-----INCIDENT NEUTRON DATA                                        2825 1451
------ENDF-6 FORMAT                                               2825 1451
***************************************************************** 2825 1451
------Russian Reactor Dosimetry File RRDF-2002                    2825 1451
***************************************************************** 2825 1451
Author  of evaluation: K.I.Zolotarev                              2825 1451
***************************************************************** 2825 1451
MF=3                                                              2825 1451
MT=103 - Ni58(n,p)Co58m+g reaction                                2825 1451
-------------------------------------                             2825 1451
    Microscopic experimental data  [1-64]  were  analyzed in  the 2825 1451
process of preparation  of input data base  for the evaluation of 2825 1451
cross sections and their uncertainty  for the  Ni-58(n,p)Co-58m+g 2825 1451
reaction.  During this procedure all experimental data  if it was 2825 1451
possible were corrected to the new recommended cross section data 2825 1451
for monitor reactions used in the measurements and to the new re- 2825 1451
commended decay data.                                             2825 1451
    Excitation function for the  Ni-58(n,p)Co-58m+g  reaction  in 2825 1451
the energy region from threshold to 20 MeV was evaluated by means 2825 1451
of statistical analysis of experimental cross section data [1-41].2825 1451
    Special correction was done with experimental data [9], [12], 2825 1451
[13], [15], [16], [20], [23], [26], [34], [35]. Experimental data 2825 1451
of Decowski et al. [9] corresponding to the neutron energy region 2825 1451
12.42 - 17.08 MeV  were renormalized to the preliminary evaluated 2825 1451
cross section value at En=14.77 MeV. Original cross sections from 2825 1451
ref. [9] were multiplied to the correction factor Fc=0.53807 .    2825 1451
    Data of Paulsen and Widera [12] measured in the energy ranges 2825 1451
3.35 - 6.27 MeV and 12.8 - 16.4 MeV  were corrected to the preli- 2825 1451
minary evaluated cross section value at  6.3 MeV  and integral of 2825 1451
cross section  in the energy interval  14-15 MeV.  The correction 2825 1451
factors for the experimental data [12] were Fc=1.059 and Fc=1.058,2825 1451
respectively.                                                     2825 1451
    Data of  Smith and Meadows [13] measured  with using neutrons 2825 1451
from  D(d,n)He3 reaction were renormalized to the results of this 2825 1451
experiment obtained with  Li7(p,n)Be7 neutron source in the over- 2825 1451
lapping interval 5.398 - 5.870 MeV.  D(d,n)He3 data in the energy 2825 1451
range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 .   2825 1451
    Data of Wu and Chou [15]  and  Husain and Hunt [20] were also 2825 1451
renormalized to the  Smith and Meadows  measurements  carried out 2825 1451
with  Li7(p,n)Be7  neutron source. Correction factors for experi- 2825 1451
mental data were Fc=1.04592 [15] and Fc=1.03634 [20].             2825 1451
    Cross sections measured  by Hudson et al. [16]  in the energy 2825 1451
range 13.3 - 17.10 MeV were corrected to Filatenkov et al. experi-2825 1451
mental data at 14.1 MeV [40], Fc=0.95367.                         2825 1451
    Experimental data of  Rondio et al. [23,26] were corrected to 2825 1451
the preliminary evaluated  Ni-58(n,p)Co-58m+g reaction cross sec- 2825 1451
tion integral in the energy interval 17.3 - 18.5 MeV (Fc=0.1889). 2825 1451
    Data of Yuan Junqian et al. [34]  and  Li Tingyan et al. [35] 2825 1451
were multiplied to the factors Fc=0.9000 and Fc=0.9259 , respec-  2825 1451
tively, using preliminary evaluated cross section integral in the 2825 1451
energy interval 13.6 - 14.6 MeV.                                  2825 1451
    Experimental cross section data [42-64] were rejected  due to 2825 1451
their discrepancy with the main bulk of experimental data [1-41]. 2825 1451
Data of Decowski et al. [9]  obtained in the neutron energy range 2825 1451
from 1.95 to 4.79 MeV  were rejected for the same reason.  In the 2825 1451
rejected experiments [42-43], [51-55], [57-60], [62-64] the cross 2825 1451
section values  were measured only  in a one  energy point in the 2825 1451
interval 14 - 15 MeV.                                             2825 1451
    Statistical analysis of input cross section data was carried  2825 1451
out by means of PADE-2 code [65].  Rational function was used as  2825 1451
the model function [66].                                          2825 1451
     U-235 thermal fission [67] and  Cf-252 spontaneous  fission  2825 1451
neutron spectra [68] averaged cross sections calculated from the  2825 1451
the evaluated  Ni-58(n,p)Co-58m+g  excitation function  are  the  2825 1451
following:                                                        2825 1451
                                                                  2825 1451
  2825 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2825 1451
  2825 1451
 U-235 neutron fission     107.44         108.9 +- 5.2    [69]  2825 1451
                                          106.0 +- 7.0    [70]  2825 1451
                                          108.5 +- 1.4    [71]  2825 1451
  2825 1451
 Cf-252 spont. fission     117.36         117.6 +- 1.5    [72]  2825 1451
                                          117.5 +- 1.53   [73]  2825 1451
  2825 1451
                                                                  2825 1451
MT=33                                                             2825 1451
MT=103 -(n,p) cross section cov. matrix                           2825 1451
---------------------------------------                           2825 1451
     Uncertainties in the evaluated excitation function  for  the 2825 1451
reaction Ni-58(n,p)Co-58m+g are given in the form of relative co- 2825 1451
variance matrix for the  37-neutron energy groups (LB=5). Covari- 2825 1451
ance matrix of uncertainties  was calculated simultaneously  with 2825 1451
recommended cross section data by means of PADE-2 code.           2825 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2825 1451
given in the 33-file are the following:                           2825 1451
                                                                  2825 1451
    3.02525E-08    3.04415E-08    3.07179E-08    3.11702E-08      2825 1451
    3.17809E-08    3.24784E-08    3.36976E-08    3.53449E-08      2825 1451
    3.74027E-08    4.12908E-08    4.46423E-08    5.24310E-08      2825 1451
    5.88770E-08    7.22029E-08    8.98365E-08    1.06273E-07      2825 1451
    1.47057E-07    1.97171E-07    2.32636E-07    3.12653E-07      2825 1451
    4.31346E-07    5.83111E-07    7.69089E-07    9.84635E-07      2825 1451
    1.17764E-06    1.35691E-06    6.11687E-05    6.46368E-04      2825 1451
    7.13990E-04    1.20451E-03    1.41032E-03    1.54898E-03      2825 1451
    2.03298E-03    3.77118E-03    4.31020E-03    5.28282E-03      2825 1451
    3.31282E-02                                                   2825 1451
                                                                  2825 1451
References:                                                       2825 1451
 1. I.Kumabe, R.W.Fink  Nucl. Phys., v.15, p.316, February 1960   2825 1451
 2. R.N.Glover, E.Weigold  Nucl. Phys., v.29, p.309, 1962         2825 1451
 3. J.F.Barry  J. Nucl. Energy AB, v.16, p.467, 1962              2825 1451
 4. W.G.Cross et al.  Progress Report EANDC(CAN)-16, p.1,         2825 1451
    January 1963                                                  2825 1451
 5. J.W.Meadows, J.F.Whalen  Phys. Rev., v.130, p.2022, 1963      2825 1451
 6. K.Debertin, E.Rossle  Nucl. Phys., v.70, p.89, August 1965    2825 1451
 7. M.Bormann et al.  Zeitschrift f. Naturforschung, section A,   2825 1451
    v.21, p.988, 1966 ;                                           2825 1451
 8. A.Paulsen, H.Liskien  Proc. of the 1-st IAEA Conference on    2825 1451
    Nuclear Data for Reactors, Paris, 17-21 October 1966, p.217   2825 1451
 9. P.Decowski et al.  Nucl. Phys. A, v.112, p.513, May 1968      2825 1451
10. R.C.Barrall et al.  Report AFWL-TR-68-134, Albuquerque, NM,   2825 1451
    March 1969                                                    2825 1451
11. R.W.Fink, W.-D.Lu  Bull. Amer. Phys. Soc., v.15, p.1372(EH6), 2825 1451
    July 1970                                                     2825 1451
12. A.Paulsen, R.Widera  Proc. of Conference on Chemical Nuclear  2825 1451
    Data, Measurements and Applicat., Univ. of Kent, Canterbury,  2825 1451
    20-22 September 1971, p.129                                   2825 1451
13. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2825 1451
14. R.A.Sigg  Dissertation Abstracts section B, v.37, p.2237,     2825 1451
    November 1976                                                 2825 1451
15. M.W.Wu, J.C.Chou  Nucl. Sci. Eng., v.63, p.268, 1977          2825 1451
16. C.G.Hudson et al.  Annals of Nuclear Energy, v.5, p.589, 1978 2825 1451
17. M.T.Swinhoe, C.A.Uttley  Report AERE-R-9929, September 1980   2825 1451
18. P.N.Ngoc et al.  Nukleonika, v.29, p.87, 1984                 2825 1451
19. P.Raics et al.  Atomki Koezlemenyek, v.23, p.45, June 1981    2825 1451
20. H.A.Husain, S.E.Hunt  Internat. J. of Applied Radiation and   2825 1451
    Isotopes, v.34, n.4, p.731, 1983                              2825 1451
21. A.Pavlik et al.  Nucl. Sci. Eng., v.90, p.186, June 1985      2825 1451
22. Lu Hanlin et al.  Chinese J. of Nuclear Physics, v.7, n.3,    2825 1451
    p.242, August 1985;                                           2825 1451
    Lu Hanlin et al.  Report INDC(CPR)-16, September 1989         2825 1451
23. J.Rondio et al.  Journ. Phys. G: Nucl. Part. Phys., v.11,     2825 1451
    pp.549-564, March 1985                                        2825 1451
24. N.V.Kornilov et al. Atomnaya Energiya, v.58, n.2, p.128, 1985 2825 1451
25. L.R.Greenwood  Progress Report ASTM-STP-956, p.743, 1987      2825 1451
26. J.Rondio et al.  Acta Physica Polonica Section B, v.20,       2825 1451
    No.7, p.627, 1987 ;                                           2825 1451
    J.Rondio et al.  Acta Physica Polonica Section B, v.18,       2825 1451
    p.1065, November 1987                                         2825 1451
27. Y.Ikeda et al.  Report JAERI-1312, March 1988                 2825 1451
28. H.Vonach, M.Wagner, R.C.Haight  Report NEANDC-259, p.165,     2825 1451
    September 1989                                                2825 1451
29. S.Cabral, G.Boerker, H.Klein, W.Mannhart                      2825 1451
    Nucl. Sci. Eng., v.106, p.308, 1990                           2825 1451
30. Y.Ikeda et al. Prog. Report NEANDC(J)-155, p.11, August, 1990 2825 1451
31. D.L.Smith, J.W.Meadows, H.Vonach, M.Wagner, R.C.Haight,       2825 1451
    W.Mannhart  Proc. of Int. Conf. on Nuclear Data for Science   2825 1451
    and Technology, Juelich, Germany, 13 - 17 May 1991, p.282     2825 1451
32. M.Viennot et al.  Nucl. Sci. Eng., v.108, p.289, July 1991    2825 1451
33. P.M.Dighe et al.  Journ. Phys. G: Nucl. Part. Phys., v.17,    2825 1451
    p.169-172, 1991                                               2825 1451
34. Yuan Junqian et al.  High Energy Physics and Nuclear Physics, 2825 1451
    v.16(1), p.57, January 1992                                   2825 1451
35. Li Tingyan et.al.  Chinese J. High Energy Physics and Nuclear 2825 1451
    Physics, v.16, n.2, p.151, Feb.1992                           2825 1451
36. Cs.M.Buczko, J.Csikai et al.  Physical Review C, v.52, no.4,  2825 1451
    p.1940, 1995                                                  2825 1451
37. K.T.Osman, F.I.Habbani Report INDC(SUD)-001,Distr.L, IAEA,    2825 1451
    Vienna, October 1996                                          2825 1451
38. J.Cezar Suita et al.  Nucl. Sci. Eng., v.126, p.101, 1997     2825 1451
39. Lu Hanlin et al.  Report INDC(CPR)-045, IAEA, October 1998    2825 1451
40. A.A.Filatenkov et al.  Report RI-252, May 1999                2825 1451
41. T.Senga et al. Report JAERI-Conf 2000-005, p.208, March 2000  2825 1451
42. D.L.Allan  Proc. Phys. Soc., sec.A, v.70, p.195, March 1957   2825 1451
43. K.H.Purser, F.W.Titterton  Australian J. Physic, v.12,        2825 1451
    p.103, 1959                                                   2825 1451
44. L.Gonzalez et al.  Phys. Rev., v.120, p.1319, 1960            2825 1451
45. K.Nakai et al.  J. Phys. Soc. Japan, v.17, p.1215, July 1962  2825 1451
46. J.M.F.Jeronimo et al.  Nucl. Phys., v.47, p.157, July 1963    2825 1451
47. J.Konijn, A.Lauber  Nucl. Phys., v.48, p.191, November 1963   2825 1451
48. S.Okumura   Nucl. Phys., v.A93, p.74, March 1967              2825 1451
49. M.Bormann et al. Proc. of 1-st IAEA Conf. on Nuclear Data for 2825 1451
    Reactors, Paris, 17-21 October 1966, v.1, p.225, April 1967   2825 1451
50. J.K.Temperley  Nucl. Sci. Eng., v.32, p.195, 1968             2825 1451
51. V.N.Levkovskij et al.  Yadernaja Fizika (Sov.), v.10, no.1,   2825 1451
    p.44, July 1969                                               2825 1451
52. G.N.Maslov et al.  Yadernye Konstanty, v.9, p.50, 1972        2825 1451
53. J.D.Hemingway  J. Nucl. Energy, v.27, p.241, April 1973       2825 1451
54. J.Dresler et al.  Progress Report, INR-1464, p.12, May 1973   2825 1451
55. K.Fukuda et al.  Progress Report NEANDC(J)-56/U, p.44,        2825 1451
    September 1978                                                2825 1451
56. Ju.A.Nemilov, Ju.N.Trofimov   Progress Report YFI-26,p.25,    2825 1451
    November 1978                                                 2825 1451
57. I.Garlea et al. Rev. Roum. Phys., v.30, no.8, p.673, 1985     2825 1451
58. I.Ribansky et al.  Report INDC(CSR)-7, June 1985              2825 1451
59. N.I.Molla et al.  Report INDC(BAN)-003, September 1986 ;      2825 1451
    M.Rahman et al. Nuclear Science and Applications, v.1, no.1,  2825 1451
    p.32, January 1989                                            2825 1451
60. K.Kobayashi, I.Kimura  Proc. of an Intern. Conf. on Nuclear   2825 1451
    Data for Science and Technology, 30 May - 3 June 1988, Mito,  2825 1451
    Japan, Saikon Publishing Co., LTD, pp.261-265, 1989           2825 1451
61. N.I.Molla et al.  Proc. of Int. Conf. on Nuclear Data for     2825 1451
    Science and Technology, Juelich, FRG, 13-17 May 1991, p. 355  2825 1451
62. A.Ercan et.al.  Proc. of an Intern. Conf. on Nuclear Data for 2825 1451
    Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2825 1451
    Verlag, 1992                                                  2825 1451
63. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992  2825 1451
64. L.I.Klochkova et al.  Vop. At. Nauki i Tekhn., Ser. Yadernye  2825 1451
    Konstanty, v.1, p.27, 1992                                    2825 1451
65. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2825 1451
66. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 2825 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 2825 1451
    Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105                2825 1451
67. L.W.Weston et al.  Evaluated Neutron Data  for  Uranium-235,  2825 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    2825 1451
68. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2825 1451
69. W.Mannhart  Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry,  2825 1451
    Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985    2825 1451
70. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2825 1451
    Washington D.C., 25-28 April 1989, v.2, p.923                 2825 1451
71. W.Mannhart  Prog. Report INDC(Ger)-045, pp.40-43, June 1999   2825 1451
72. W.Mannhart  Handbook on Nuclear Activation Cross Sections ,   2825 1451
    IAEA Technical Report Ser. No.273, p.413, 1987                2825 1451
73. W.Mannhart  Validation of Differential Cross Sections with    2825 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2825 1451
    September 2002                                                2825 1451
***************************************************************** 2825 1451
*******    End of RRDF (N,P) bibliographical component    ******* 2825 1451
***************************************************************** 2825 1451
 The Q values and threshold energies were updated prior to pro-   2825 1451
 cessing through the codes to comply with the values obtained     2825 1451
 using the NNDC calculation program which is based on the 1995    2825 1451
 Update to the Atomic mass Evaluation.                            2825 1451
                                                                  2825 1451
 Psuedo Threshold of 1.0E+5 changed to 4.0E+5 ev in MF/MT=3/103   2825 1451
 in original input file before processing through codes.          2825 1451
                                                                  2825 1451
 File 2 added to the pointwise file containing only the effective 2825 1451
 scattering radius with no resonance parameters given.            2825 1451
 Taken from ENDF/B-VI                                             2825 1451
***************************************************************** 2825 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2825 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2825 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2825 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2825 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2825 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2825 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2825 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2825 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2825 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2825 1451
 ***************** Program FIXUP (Version 2002-1) ****************2825 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2825 1451
 Corrected Thresholds-----------------------------------------Yes 2825 1451
 Extended Cross Sections to 20 MeV----------------------------No  2825 1451
 Allow Cross Section Deletion---------------------------------No  2825 1451
 Allow Cross Section Reconstruction---------------------------No  2825 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2825 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2825 1451
 Deleted Duplicate Points-------------------------------------Yes 2825 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2825 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2825 1451
 Allow Creation of Missing Sections---------------------------No  2825 1451
 Allow Insertion of Energy Points-----------------------------No  2825 1451
 Create Uniform Energy Grid-----------------------------------No  2825 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2825 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2825 1451
 Unshielded Group Averages Using  640 Groups                      2825 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2825 1451
                                1        451        387          02825 1451
                                3         16         29          02825 1451
                                3        103         73          02825 1451
                               33         16         58          02825 1451
                               33        103        133          02825 1451
                                                                  2825 1  0
                                                                  2825 0  0
 2.80580E+4 5.74376E+1          0          0          0          02825 3 16
-1.22190E+7-1.22190E+7          0          0          1         772825 3 16
         77          1                                            2825 3 16
 12400000.0 .000186234 12500000.0 .000945261 12600000.0 .0017445372825 3 16
 12700000.0 .002543812 12800000.0 .003343087 12900000.0 .0041423622825 3 16
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 8.390600-5 1.038060-4 1.225950-4 1.333900-4 1.313770-4 1.150770-4282533103
 8.728570-5 5.472140-5 2.582750-5 7.650840-6 3.665640-6 1.362650-5282533103
 3.491240-5 6.412560-5 9.816150-5 1.346460-4 1.719590-4 2.090840-4282533103
 2.454320-4 2.806830-4 3.303550-4 5.509690-4 5.154950-4 4.391610-4282533103
 3.415510-4 2.425000-4 1.581420-4 9.888940-5 6.904830-5 6.730480-5282533103
 8.747470-5 1.193440-4 1.499790-4 1.661890-4 1.585030-4 1.256480-4282533103
 7.668980-5 2.811230-5-3.501070-6-8.510460-6 1.331180-5 5.508020-5282533103
 1.069860-4 1.597770-4 2.065590-4 2.431800-4 2.678620-4 2.805190-4282533103
 2.821410-4 2.667540-4 5.417870-4 5.136820-4 4.421840-4 3.463000-4282533103
 2.471180-4 1.628990-4 1.061840-4 8.253620-5 9.022130-5 1.204470-4282533103
 1.584500-4 1.863210-4 1.883670-4 1.580330-4 1.027010-4 4.182840-5282533103
-1.978330-6-1.327090-5 1.116700-5 6.402690-5 1.326760-4 2.041440-4282533103
 2.680810-4 3.177620-4 3.498570-4 3.636240-4 3.600870-4 3.260650-4282533103
 5.356030-4 5.047340-4 4.339530-4 3.427590-4 2.515430-4 1.772730-4282533103
 1.309670-4 1.163850-4 1.294420-4 1.584730-4 1.862670-4 1.949470-4282533103
 1.733930-4 1.238480-4 6.254270-5 1.232750-5-8.083160-6 8.890820-6282533103
 5.948470-5 1.328500-4 2.160810-4 2.977160-4 3.693570-4 4.259120-4282533103
 4.650730-4 4.866190-4 4.870360-4 5.190510-4 4.882560-4 4.249420-4282533103
 3.458300-4 2.676380-4 2.040560-4 1.637790-4 1.491530-4 1.553460-4282533103
 1.707500-4 1.798490-4 1.691530-4 1.342560-4 8.344380-5 3.432480-5282533103
 5.388940-6 8.354180-6 4.544650-5 1.112280-4 1.962950-4 2.905760-4282533103
 3.853630-4 4.741820-4 5.528270-4 6.190620-4 6.913430-4 5.024060-4282533103
 4.788270-4 4.267050-4 3.585150-4 2.874830-4 2.254110-4 1.808090-4282533103
 1.570670-4 1.510140-4 1.530260-4 1.501600-4 1.322530-4 9.798390-5282533103
 5.640470-5 2.258470-5 1.070060-5 2.911650-5 7.923250-5 1.570580-4282533103
 2.556770-4 3.673960-4 4.851380-4 6.031250-4 7.170790-4 8.712440-4282533103
 4.970580-4 4.795620-4 4.324380-4 3.657970-4 2.920140-4 2.237100-4282533103
 1.713170-4 1.402170-4 1.283050-4 1.259800-4 1.202720-4 1.019710-4282533103
 7.128330-5 3.787990-5 1.585420-5 1.766380-5 5.050340-5 1.157090-4282533103
 2.100860-4 3.278250-4 4.621730-4 6.066060-4 7.554590-4 9.737830-4282533103
 4.965710-4 4.768580-4 4.259810-4 3.542670-4 2.752360-4 2.031230-4282533103
 1.494550-4 1.191020-4 1.078110-4 1.038920-4 9.468250-5 7.415140-5282533103
 4.629310-5 2.249660-5 1.582900-5 3.625710-5 8.855090-5 1.725350-4282533103
 2.844970-4 4.187960-4 5.692020-4 7.297510-4 9.749350-4 4.857170-4282533103
 4.593680-4 4.041610-4 3.309560-4 2.534110-4 1.850300-4 1.352320-4282533103
 1.058290-4 9.053390-5 7.906960-5 6.388380-5 4.475090-5 2.842960-5282533103
 2.474100-5 4.238370-5 8.649170-5 1.581960-4 2.554100-4 3.740190-4282533103
 5.090310-4 6.553620-4 8.830400-4 4.622530-4 4.355670-4 3.846720-4282533103
 3.182150-4 2.468290-4 1.807690-4 1.271380-4 8.806000-5 6.133630-5282533103
 4.348530-5 3.294650-5 3.123000-5 4.181120-5 6.821470-5 1.125520-4282533103
 1.750150-4 2.540930-4 3.471210-4 4.509020-4 5.621480-4 7.341490-4282533103
 4.443900-4 4.285000-4 3.883860-4 3.274380-4 2.526150-4 1.740560-4282533103
 1.032440-4 5.000760-5 1.996620-5 1.392060-5 2.917560-5 6.150500-5282533103
 1.066340-4 1.608980-4 2.213640-4 2.857300-4 3.521810-4 4.192940-4282533103
 4.859520-4 5.816950-4 4.533230-4 4.499330-4 4.120140-4 3.396280-4282533103
 2.425640-4 1.399160-4 5.435630-5 3.576710-6-5.659410-6 2.242670-5282533103
 7.696180-5 1.454850-4 2.173370-4 2.849970-4 3.440510-4 3.925320-4282533103
 4.301430-4 4.575930-4 4.813510-4 4.857840-4 4.796520-4 4.229260-4282533103
 3.212120-4 1.967440-4 8.186660-5 5.950200-6-1.591330-5 1.429150-5282533103
 8.259810-5 1.707280-4 2.620060-4 3.441990-4 4.099680-4 4.560900-4282533103
 4.822470-4 4.899470-4 4.712900-4 5.075800-4 4.787570-4 3.916490-4282533103
 2.653020-4 1.347580-4 3.667320-5-5.566650-6 1.196800-5 7.748840-5282533103
 1.717390-4 2.753360-4 3.729060-4 4.543540-4 5.144220-4 5.515150-4282533103
 5.665130-4 5.513000-4 4.853190-4 4.324480-4 3.299840-4 2.066050-4282533103
 9.851580-5 3.426120-5 2.566870-5 6.833550-5 1.479640-4 2.473420-4282533103
 3.510700-4 4.476680-4 5.298930-4 5.940640-4 6.391730-4 6.707940-4282533103
 4.268960-4 3.717220-4 2.815670-4 1.833800-4 1.050470-4 6.517590-5282533103
 6.955450-5 1.135560-4 1.869020-4 2.778930-4 3.760410-4 4.732300-4282533103
 5.638770-4 6.446410-4 7.413690-4 3.756610-4 3.387300-4 2.724920-4282533103
 1.974450-4 1.345130-4 9.832300-5 9.510360-5 1.240620-4 1.800840-4282533103
 2.562310-4 3.454750-4 4.416530-4 5.398550-4 6.807700-4 3.609990-4282533103
 3.428820-4 2.928950-4 2.278710-4 1.658530-4 1.206730-4 1.000320-4282533103
 1.060900-4 1.371230-4 1.892260-4 2.576610-4 3.377130-4 4.689640-4282533103
 3.773120-4 3.706070-4 3.302750-4 2.703650-4 2.058530-4 1.486660-4282533103
 1.063320-4 8.229690-5 7.695740-5 8.879390-5 1.152710-4 1.761240-4282533103
 4.157570-4 4.230810-4 3.969360-4 3.463240-4 2.815830-4 2.118020-4282533103
 1.437810-4 8.194610-5 2.873500-5-1.490280-5-6.107470-5 4.907200-4282533103
 5.238500-4 5.211570-4 4.859640-4 4.245560-4 3.440740-4 2.512660-4282533103
 1.518550-4 5.033710-5-9.636020-5 6.307180-4 7.025830-4 7.333670-4282533103
 7.226360-4 6.745020-4 5.953670-4 4.923350-4 3.722470-4 1.753660-4282533103
 8.655530-4 9.914960-4 1.072120-3 1.105210-3 1.093810-3 1.043780-3282533103
 9.621740-4 7.967120-4 1.233200-3 1.438380-3 1.597930-3 1.708340-3282533103
 1.771340-3 1.791690-3 1.753320-3 1.793410-3 2.115770-3 2.395310-3282533103
 2.626540-3 2.809320-3 2.992600-3 2.629110-3 3.114210-3 3.559090-3282533103
 3.955790-3 4.443820-3 3.832780-3 4.524150-3 5.174450-3 6.033490-3282533103
 5.486660-3 6.417000-3 7.695010-3 7.646890-3 9.373270-3 1.179240-2282533103
                                                                  282533  0
                                                                  2825 0  0
                                                                     0 0  0
 2.80600E+4 5.94159E+1          0          0         34         102831 1451
 0.0        0.0                 0          0          0          62831 1451
 1.00000E+0 1.50000E+8          0          0         10       20022831 1451
 3.00000E+2 0.0                 1          0        395          32831 1451
 28-Ni- 60  LANL,ORNL EVAL-SEP97 S.CHIBA,M.B.CHADWICK,LARSON      2831 1451
                      DIST-Feb2004                                2831 1451
----IRDF-2002         MATERIAL 2831                               2831 1451
-----INCIDENT NEUTRON DATA                                        2831 1451
------ENDF-6 FORMAT                                               2831 1451
                                                                  2831 1451
******************************************************************2831 1451
----ENDF/B-VI         MATERIAL 2831         REVISION 3            2831 1451
 Ch97,Ch99            DIST-SEP 1 REV3-                 20010926   2831 1451
                                                                  2831 1451
 ENDF/B-VI MOD 4 Revision, June 2000, S.C. Frankle, R.C. Reedy,   2831 1451
                 P.G. Young (LANL)                                2831 1451
                                                                  2831 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   2831 1451
 MT 102) has been updated for new experimental data at incident   2831 1451
 neutron energies up to 1 keV.                                    2831 1451
 The Q-value for radiative capture was also updated in File 3.    2831 1451
 Details of these changes are described in Frankel et al. [Fr01]. 2831 1451
                                                                  2831 1451
******************************************************************2831 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2831 1451
DOSIMETRY                                                         2831 1451
******************************************************************2831 1451
                                                                  2831 1451
  ENDF/B-VI MOD 3 Evaluation, September 1997, S. Chiba,           2831 1451
                  M.B. Chadwick, P.G. Young (LANL), and           2831 1451
                  A.J. Koning (ECN)                               2831 1451
                                                                  2831 1451
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     2831 1451
   in cooperation with ECN Petten.                                2831 1451
                                                                  2831 1451
   This evaluation provides a complete representation of the      2831 1451
 nuclear data needed for transport, damage, heating,              2831 1451
 radioactivity, and shielding applications over the incident      2831 1451
 neutron energy range from 1.0E-11 to 150 MeV.  The discussion    2831 1451
 here is divided into the region below and above 20 MeV.          2831 1451
                                                                  2831 1451
INCIDENT NEUTRON ENERGIES < 20 MeV                                2831 1451
                                                                  2831 1451
  Below 20 MeV the evaluation is based completely on the ENDF/B-  2831 1451
VI.5(Rev1,Release 2) evaluation by Larson, C. Perey, Hetrich, and 2831 1451
Fu.                                                               2831 1451
                                                                  2831 1451
INCIDENT NEUTRON ENERGIES > 20 MeV                                2831 1451
                                                                  2831 1451
  The ENDF/B-VI Release 2 evaluation extends to 20 MeV and        2831 1451
includes cross sections and energy-angle data for all             2831 1451
significant reactions.  The present evaluation utilizes a more    2831 1451
compact composite reaction spectrum representation above 20 MeV   2831 1451
in order to reduce the length of the file. No essential data for  2831 1451
applications is lost with this representation.                    2831 1451
  The evaluation above 20 MeV utilizes MF=6, MT=5 to represent    2831 1451
all reaction data.  Production cross sections and emission        2831 1451
spectra are given for neutrons, protons, deuterons, tritons,      2831 1451
alpha particles, gamma rays, and all residual nuclides produced   2831 1451
(A>5) in the reaction chains. To summarize, the ENDF sections     2831 1451
with non-zero data above En = 20 MeV are:                         2831 1451
                                                                  2831 1451
  MF=3 MT=  1  Total Cross Section                                2831 1451
       MT=  2  Elastic Scattering Cross Section                   2831 1451
       MT=  3  Nonelastic Cross Section                           2831 1451
       MT=  5  Sum of Binary (n,n') and (n,x) Reactions           2831 1451
                                                                  2831 1451
  MF=4 MT=  2  Elastic Angular Distributions                      2831 1451
                                                                  2831 1451
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         2831 1451
               Distributions for Emission Neutrons, Protons,      2831 1451
               Deuterons, Tritons, and Alphas; and Angle-         2831 1451
               Integrated Spectra for Gamma Rays and Residual     2831 1451
               Nuclei That Are Stable Against Particle Emission   2831 1451
                                                                  2831 1451
  The evaluation is based on nuclear model calculations that      2831 1451
have been benchmarked to experimental data, especially for n +    2831 1451
Ni58 and p + Ni58 reactions [Ch97]. We use the GNASH code system  2831 1451
[Yo92], which utilizes Hauser-Feshbach statistical,               2831 1451
preequilibrium and direct-reaction theories. Spherical optical    2831 1451
model calculations are used to obtain particle transmission       2831 1451
coefficients for the Hauser-Feshbach calculations, as well as     2831 1451
for the elastic neutron angular distributions.                    2831 1451
  Cross sections and spectra for producing individual residual    2831 1451
nuclei are included for reactions. The energy-angle-correlations  2831 1451
for all outgoing particles are based on Kalbach systematics       2831 1451
[Ka88].                                                           2831 1451
  A model was developed to calculate the energy distributions of  2831 1451
all recoil nuclei in the GNASH calculations [Ch96]. The recoil    2831 1451
energy distributions are represented in the laboratory system in  2831 1451
MT=5, MF=6, and are given as isotropic in the lab system. All     2831 1451
other data in MT=5,MF=6 are given in the center-of-mass system.   2831 1451
This method of representation utilizes the LCT=3 option approved  2831 1451
at the November, 1996, CSEWG meeting.                             2831 1451
  Preequilibrium corrections were performed in the course of the  2831 1451
GNASH calculations using the exciton model of Kalbach [Ka77,      2831 1451
Ka85], validated by comparison with calculations using Feshbach,  2831 1451
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     2831 1451
nuclear data sheets were matched to continuum level densities     2831 1451
using the formulation of Gilbert and Cameron [Gi65] and pairing   2831 1451
and shell parameters from the Cook [Co67] analysis. Neutron and   2831 1451
charged- particle transmission coefficients were obtained from    2831 1451
the optical potentials, as discussed below.  Gamma-ray            2831 1451
transmission coefficients were calculated using the Kopecky-Uhl   2831 1451
model [Ko90].                                                     2831 1451
                                                                  2831 1451
SOME Ni-SPECIFIC INFORMATION CONCERNING THE EVAL.                 2831 1451
                                                                  2831 1451
  The neutron total cross section was evaluated based on the      2831 1451
least-squares method taking account of the experimental           2831 1451
data [Du67, Bo71, St71, Sm79, Fe80, Pe82, Ha82a, Di97].  The data 2831 1451
for natural Ni [Di97] was also used because there was not enough  2831 1451
data for Ni-60 above 20 MeV.  The data for natural Ni were        2831 1451
transformed to the Ni-60 cross section according to A*(2/3) law.  2831 1451
Result of this estimation  was used as the evaluated total cross  2831 1451
section data above 20 MeV.                                        2831 1451
                                                                  2831 1451
The evaluated total cross section data (1 to 250 MeV), s-wave     2831 1451
strength function [Mu81] and elastic scattering angular           2831 1451
distribution data [Bo71, Gu85, Tu73, Ya79] were used to obtain the2831 1451
neutron optical potential parameters. The parameter estimation was2831 1451
carried out based on Marquart-Bayesian approach [Sm91], where     2831 1451
ECIS95 code [Ra96] was used for the optical model calculation.  We2831 1451
have employed the energy dependence of the optical potential      2831 1451
similar to Delaroche's work [De89].  The initial potential        2831 1451
parameters were adopted from Koning and Delaroche [Ko97].  A total2831 1451
of 7 parameters concerning the central potential depth were       2831 1451
estimated with associated covariance matrix, while the geometrical2831 1451
parameters were fixed to the result of a similar search for n +   2831 1451
Ni-58. Presently obtained potential was used for the calculation  2831 1451
of neutron transmission coefficients and DWBA cross sections in   2831 1451
the energy region above 20 MeV.  Below 20 MeV, the Harper neutron 2831 1451
potential [Ha82b] was used for the calculation of transmission    2831 1451
coefficients.                                                     2831 1451
                                                                  2831 1451
The proton optical potential was also searched to obtain a good   2831 1451
description of proton-total reaction cross section as predicted by2831 1451
Wellisch-Axen systematic [We96] above 50 MeV.  The parameter      2831 1451
estimation was carried out by the Marquart-Bayesian approach      2831 1451
similar to the neutron OMP, but trying to seek the best parameter 2831 1451
to reproduce the reaction cross sections compiled by Carlson      2831 1451
[Ca96] and Wellisch values. The experimental data in Carlson      2831 1451
[Ca96] was scaled for Ni-60 according to A**(2/3) law.  In this   2831 1451
search, the geometrical parameters were fixed to be same as the   2831 1451
neutron potential.  The present potential gives a good description2831 1451
of the proton total reaction cross section from 10 MeV to 250 MeV.2831 1451
However, after some trial and error to reproduce both the elastic 2831 1451
scattering and reaction cross section data for Ni-58, we have     2831 1451
employed the following combination of proton potentials:          2831 1451
                                                                  2831 1451
 0 to   5 MeV  :  Harper potential [Ha82b]                        2831 1451
 6 to  47 MeV  :  Koning and Delaroche [Ko97]                     2831 1451
48 to 260 MeV  :  Present OMP                                     2831 1451
                                                                  2831 1451
For deuterons, the Lohr-Haeberli [Lo74] global potential was used;2831 1451
for alpha particles the McFadden-Satchler [Mc66] potential was    2831 1451
used; and for tritons the Becchetti-Greenlees [Be71] potential was2831 1451
used.  The He-3 channel was ignored.                              2831 1451
                                                                  2831 1451
The direct collective inelastic scattering to the following level 2831 1451
in Ni-60 was considered by the DWBA-mode calculation of ECIS95    2831 1451
[Ra96]:                                                           2831 1451
                                                                  2831 1451
Jpi Ex(MeV) Deformation length                                    2831 1451
 2+ 1.331   0.8535                                                2831 1451
                                                                  2831 1451
The deformation length was determined to much the ENDF/B-VI value 2831 1451
at 20 MeV.                                                        2831 1451
                                                                  2831 1451
Certain nuclear level densities were modified from the default    2831 1451
values in accordance with experimental level-density information  2831 1451
as follows. The level density of 57Fe was matched to the observed 2831 1451
D-0 value by modifying the pairing energy to 0.15 MeV; the level  2831 1451
density of the main competition channel, neutron emission to 60Ni,2831 1451
was adjusted to match Fischer et al.'s [Fi86] value of 3.85e3+-25%2831 1451
at 11 MeV excitation energy, but then increased slightly (a total 2831 1451
density increase of 11%) by using a pairing energy of 1.1 MeV,    2831 1451
since Haight's alpha production at 14 MeV is approx. 11% smaller  2831 1451
than Fischer et al.'s measurements. Production of residual nuclei 2831 1451
59Ni and 56Fe, through (n,2n) and (n,na) reactions, become        2831 1451
particularly important above 14 MeV. Pairing energies were        2831 1451
adjusted to match the level density for 59Ni at the neutron       2831 1451
binding energy (D0=12.9 keV), and Fischer et al.'s result for the 2831 1451
56Fe level density at 11 MeV.                                     2831 1451
                                                                  2831 1451
The new Haight et al. [Ha97] LANL/WNR 60Ni(n,x alpha) data up to  2831 1451
50 MeV was used to benchmark our model calculations [Ch97] -      2831 1451
agreement with experiment was good.                               2831 1451
                                                                  2831 1451
 **************************************************************** 2831 1451
                                                                  2831 1451
REFERENCES                                                        2831 1451
                                                                  2831 1451
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees in                2831 1451
   "Polarization Phenomena in Nuclear Reactions," (Ed: H.H.       2831 1451
   Barschall and W. Haeberli, The University of Wisconsin Press,  2831 1451
   1971) p.682                                                    2831 1451
[Bo71]  P. Boschung et al, Nucl.Phys. A161, 593 (1971)            2831 1451
[Ca96]  R.F. Carlson, Atomic Data and Nuclear Data Tables, 63,    2831 1451
   93 (1996)                                                      2831 1451
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2831 1451
[Ch96]  M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      2831 1451
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   2831 1451
   Driven Technologies:  Calculational Method for Heavy Recoils," 2831 1451
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    2831 1451
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996     2831 1451
[Ch97]  M.B. Chadwick and P.G. Young, "Model Calculations of      2831 1451
   n,p + 58,60,61,62,64Ni" in APT PROGRESS REPORT: 1 August - 1   2831 1451
   September 1997, internal Los Alamos National Laboratory memo T-2831 1451
   2-97/MS-51, 8 September 1997 from R.E. MacFarlane to L. Waters.2831 1451
[Ch99]  M.B. Chadwick, P.G. Young, G.M. Hale, et al., Los Alamos  2831 1451
   National Laboratory report, LA-UR-99-1222 (1999)               2831 1451
[Ci68]  S. Cierjack et al, report KFK-1000 (1968)                 2831 1451
[Co67]  J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J.    2831 1451
   Phys. 20, 477 (1967)                                           2831 1451
[De89]  J.P. Delaroche, Y. Wang and J. Rapaport, Phys.Rev.C 39,   2831 1451
   391 (1989)                                                     2831 1451
[Di97]  F. Dietrich et al., private communication (1997).         2831 1451
[Du67]  Yu.V. Dukarevich et al., Nucl.Phys. A92, 433 (1967)       2831 1451
[Fe80]  M.B. Fedorov et al., 80Kiev, 1, 309(1980)                 2831 1451
[Fi86]  R. Fischer, G. Traxler, M. Uhl at al., Phys Rev C34, 460  2831 1451
   (1986)                                                         2831 1451
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      2831 1451
   National Laboratory Report, LA-13812 (2001).                   2831 1451
[Fu96]   C.Y. Fu, "Effects of shape differences in the level      2831 1451
   densities of three formalisms on calculated cross sections",   2831 1451
   personal communication, 1996 (to be published by the IAEA).    2831 1451
[Gi65]  A. Gilbert and A.G.W. Cameron, Can.J.Phys. 43, 1446 (1965)2831 1451
[Gu85]  P.P. Guss et al, Nucl.Phys. A438, 187 (1985)              2831 1451
[Ha82a] J.A. Harvey et al, 82Anterp  (1982) p. 856                2831 1451
[Ha82b] R.C. Harper and W.L. Alford, J.Phys.G. 8, 153 (1982)      2831 1451
[Ha97]  R.C. Haight, F.B. Bateman, S.M. Sterbenz et al., Nuclear  2831 1451
   Data for Science and Technol., Proc. Conf. Trieste, Italy, May 2831 1451
   1997, Vol. I (Edit. Compositori, Bologna, Italy, 1997) p. 603  2831 1451
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     2831 1451
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         2831 1451
   LA-10248-MS (1985)                                             2831 1451
[Ka88]  C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also          2831 1451
   C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981)           2831 1451
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990)         2831 1451
[Ko97]  A. Koning and J.P. Delaroche, private communication.      2831 1451
[La83]  D.C. Larson et al, report ORNL-TM-8203 (1983)             2831 1451
[Lo74]  J.M.Lohr and W.Haeberli, Nucl.Phys. A232, 381 (1974)      2831 1451
[Mc66]  L. McFadden and G.R. Satchler, Nucl.Phys. 84, 177 (1966)  2831 1451
[Mu81]  S.F. Mughabghab, M. Divadeenam and N.E. Holden, "Neutron  2831 1451
   Cross Sections", Vol. 1, Part A (Academic Press, 1981)         2831 1451
[Pe73]  F.G. Perey, private communication (1973) [EXFOR 10342]    2831 1451
[Pe82]  C.M. Perey et al, report ORNL-5893 (1982)                 2831 1451
[Pe88]  Pedroni et al, Phys.Rev.C 38, 2052 (1988)                 2831 1451
[Po81]  W. Poenitz, Proc. Conf. on Nuclear Data Evaluation Methods2831 1451
   and Procedures, Brookhaven National Laboratory Report BNL-NCS- 2831 1451
   51363, p.249(1981).                                            2831 1451
[Ra96]   J. Raynal, "Notes on ECIS94", Service de Physique        2831 1451
   Theorique, Saclay, France (personal communication through A. J.2831 1451
   Koning, 1996).                                                 2831 1451
[Sc73]  W. Schimmerling et al., Phys.Rev.C 7, 248 (1973)          2831 1451
[Sm79]  A.B. Smith et al., Nucl.Sci.Eng., 72, 293 (1979)          2831 1451
[Sm91]  D.L. Smith, "Probability, Statistics, and Data Uncertainty2831 1451
   in Nuclear Science and Technology" (American Nuclear Society,  2831 1451
   1991)                                                          2831 1451
[Sm92]  A.B. Smith et al., J.Phys.G, 18, 629 (1992)               2831 1451
[St71]  P. Stoler et al,, 71Knox, 1, 311 (1971)                   2831 1451
[Tu73]  A.I. Tutubalin et al., Neutron Physics, Proc. Conf., Kiev,2831 1451
   Ukraine, 1973, Vol. 3 [EXFOR 40417] p. 62                      2831 1451
[Ul83]  J.J. Ullmann, F.P. Brady, C.M. Castaneda, et al., Nucl.   2831 1451
   Phys. A427, 493 (1984)                                         2831 1451
[We96]  H.P. Wellisch and D. Axen, Phys.Rev.C 54, 1329 (1996)     2831 1451
[Ya79]  Y. Yamanouti, J. Rapaport, S.M. Grimes et al, Nuclear     2831 1451
   Cross Sections for Technology, Proc. Conf., Knoxville, TN,     2831 1451
   1979, NBS Special Publication 594 (1980) [EXFOR 10953]         2831 1451
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos    2831 1451
   report LA-12343-MS (1992)                                      2831 1451
                                                                  2831 1451
 **************************************************************** 2831 1451
                                                                  2831 1451
 ENDF/B-VI MOD 2 Revision, July 1991, D.C. Larson, C.M. Perey,    2831 1451
                 D.M. Hetrick, and C.Y. Fu (ORNL)                 2831 1451
                                                                  2831 1451
 REVISION 1 CHANGES                                               2831 1451
   The resonance region from 1.E-5 eV to 450 keV has been changed,2831 1451
 based on a SAMMY analysis by F. G. Perey of new 60Ni ORELA data  2831 1451
 up to 100 keV of Harvey, Hill and Perey. This new work reproduces2831 1451
 the thermal cross sections and includes the small resonances     2831 1451
 inadvertently left out in the original file.  No background      2831 1451
 cross sections are required in file 3 in the resonance region.   2831 1451
   Removed elastic transformation matrix.                         2831 1451
                                                                  2831 1451
 **************************************************************** 2831 1451
                                                                  2831 1451
 ENDF/B-VI MOD 1 Evaluation, October 1989, D.C. Larson,           2831 1451
                 C.M. Perey, D.M. Hetrick, and C.Y. Fu (ORNL)     2831 1451
                                                                  2831 1451
   This work employed nuclear model codes including the           2831 1451
 Distorted Wave Born Approximation (DWBA) program DWUCK (1)       2831 1451
 and the Hauser-Feshbach code TNG (2,3,4).  The TNG code provides 2831 1451
 energy and angular distributions of particles emitted in the     2831 1451
 compound and pre-compound reactions, ensures consistency among   2831 1451
 all reactions, and maintains energy balance. Details pertinent   2831 1451
 to the contents of this evaluation and extensive comparisons of  2831 1451
 calculations with experimental data can be found in Hetrick (5). 2831 1451
                                                                  2831 1451
----- DESCRIPTION OF FILES                                        2831 1451
(MF-MT)                                                           2831 1451
  1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS.         2831 1451
  2-151 RESONANCE PARAMETERS -- TAKEN FROM REF (6).  THE NEGATIVE 2831 1451
        ENERGY RESONANCES WERE ADJUSTED BY THE AUTHOR, C. M. PEREY2831 1451
        TO GIVE THE PROPER THERMAL SCATTERING AND CAPTURE.  USING 2831 1451
        THESE MODIFIED RESULTS, NO BACKGROUND IS REQUIRED IN 3/1  2831 1451
        IE, THE TOTAL, SCATTERING AND CAPTURE IS GIVEN COMPLETELY 2831 1451
        BY THE RESONANCE PARAMETERS FROM 1.E-5 EV TO 450 KEV.     2831 1451
        RESONANCE PARAMETERS ARE FROM A REICH-MOORE ANALYSIS WITH 2831 1451
        THE CODE SAMMY (REF 7).                                   2831 1451
        THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE         2831 1451
        RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 3.9 B,        2831 1451
        ELASTIC SCATTERING 0.98 B, AND CAPTURE 2.92 B.            2831 1451
        NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2831 1451
        OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE       2831 1451
        PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON    2831 1451
        A FINER ENERGY GRID THAN GIVEN IN 4/2.                    2831 1451
  3-103 (N,P) CROSS SECTIONS FOR EN LESS THAN OR EQUAL TO 12 MEV  2831 1451
        WERE TAKEN FROM A CURVE DRAWN THROUGH THE DATA OF VONACH  2831 1451
        ET AL. (10). THIS CURVE WAS CONNECTED TO THE ANL DATA OF  2831 1451
        GREENWOOD (11) FROM 14.5 TO 15.0 MEV. FROM 15.0 TO 20.0   2831 1451
        MEV THE DATA OF PAULSEN AND LISKIEN (12) WERE USED        2831 1451
        AS A GUIDE.  THE NEW DATA BY VONACH ET AL. (10)           2831 1451
        RESULT IN AN INTEGRAL VALUE CLOSE TO THE 2.39 VALUE OF    2831 1451
        MANNHART (13).                                            2831 1451
--------------------------------------------------------------    2831 1451
            UNCERTAINTY FILES                                     2831 1451
        AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS  2831 1451
        REQUIRED BY ENDF/B-VI.                                    2831 1451
                                                                  2831 1451
 33-103 COVARIANCES - DATA OF VONACH ET AL. (10) AND CSISRS DATA  2831 1451
        USED AS A GUIDE.                                          2831 1451
 ---------------------------------------------------------------- 2831 1451
 REFERENCES:                                                      2831 1451
 (1) P.D. Kunz, "DISTORTED WAVE CODE DWUCK72," UNIV. OF           2831 1451
     COLORADO, UNPUBLISHED (1972).                                2831 1451
 (2) C.Y. Fu, Oak Ridge report ORNL/TM-7042 (1980)                2831 1451
 (3) C.Y Fu, Neutron Cross Sections from 10 to 50 MeV, Proc. Symp.2831 1451
     Upton, NY, May 1980, Brookhaven report BNL-NCS-51425 (1980)  2831 1451
     p. 675                                                       2831 1451
 (4) K. Shibata and C.Y. Fu, Oak Ridge National Laboratory report 2831 1451
     ORNL/TM-10093 (1986)                                         2831 1451
 (5) D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge National   2831 1451
     Laboratory report ORNL/TM-10219 [ENDF-344] (1987)            2831 1451
 (6) C.M. Perey, J.A. Harvey, R.L. Macklin, et al., Phys.Rev.C,   2831 1451
     27, 2556 (1983)                                              2831 1451
 (7) N.M. Larson and F.G. Perey, Oak Ridge National Laboratory    2831 1451
     reports ORNL/TM-7485 (1980); ORNL/TM-9179 (1984);            2831 1451
     ORNL/TM-9179/R1 (1985); ORNL/TM-9179/R2 (1988)               2831 1451
 (8) J.A. Harvey, ORELA, 80-M NE110 TA TARGET DATA, SENT TO NNDC  2831 1451
 (9) J.A. Harvey, ORELA, 80-M NE110 BE BLOCK DATA, SENT TO NNDC   2831 1451
(10) H. Vonach, M. Wagner, and R.C. Haight, "Neutron cross        2831 1451
     sections of 58Ni and 60Ni for 8-12 MeV neutrons," private    2831 1451
     communication, 1989                                          2831 1451
(11) L.R. Greenwood, report DOE-ER-0046-21 (1985) p. 15           2831 1451
(12) A. Paulsen, Nukleonik 10, 91 (1967)                          2831 1451
(13) W. Mannhart, Proc. Fourth ASTM-Euratom Symp. on Reactor      2831 1451
     Dosimetry, Vol. II, March 22-26, 1982 (1982) p. 637          2831 1451
(14) M. Divadeenam, Brookhaven National Laboratory report         2831 1451
     BNL-NCS-51346 [ENDF-294] (1979)                              2831 1451
(15) S.M. Grimes, R.C. Haight, K.R. Alvar, et al., Phys.Rev.C     2831 1451
     19, 2127 (1979)                                              2831 1451
  ****************************************************************2831 1451
 Psuedo Threshold of 0.0 added at 2.7E+6 ev to MF/MT=3/103 after  2831 1451
 processing thro. codes.                                          2831 1451
                                                                  2831 1451
                                                                  2831 1451
                                                                  2831 1451
 ************************ C O N T E N T S *********************** 2831 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2831 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2831 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2831 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2831 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2831 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2831 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2831 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2831 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2831 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2831 1451
 ***************** Program FIXUP (Version 2002-1) ****************2831 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2831 1451
 Corrected Thresholds-----------------------------------------Yes 2831 1451
 Extended Cross Sections to 20 MeV----------------------------No  2831 1451
 Allow Cross Section Deletion---------------------------------No  2831 1451
 Allow Cross Section Reconstruction---------------------------No  2831 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2831 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2831 1451
 Deleted Duplicate Points-------------------------------------Yes 2831 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2831 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2831 1451
 Allow Creation of Missing Sections---------------------------No  2831 1451
 Allow Insertion of Energy Points-----------------------------No  2831 1451
 Create Uniform Energy Grid-----------------------------------No  2831 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2831 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2831 1451
 Unshielded Group Averages Using  640 Groups                      2831 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2831 1451
                                1        451        402          42831 1451
                                3        103         61          12831 1451
                               33        103         14          12831 1451
                                                                  2831 1  0
                                                                  2831 0  0
 2.80600E+4 5.94159E+1          0          0          0          02831 3103
-2.04150E+6-2.04150E+6          0          0          1        1742831 3103
        174          1                                            2831 3103
 2700000.00 1.27290E-8 2800000.00 3.81870E-8 2900000.00 6.36450E-82831 3103
 3000000.00 2.39704E-6 3100000.00 7.03836E-6 3200000.00 1.16797E-52831 3103
 3300000.00 1.63210E-5 3400000.00 2.09623E-5 3500000.00 5.81607E-52831 3103
 3600000.00 .000127916 3700000.00 .000197672 3800000.00 .0002674272831 3103
 3900000.00 .000337182 4000000.00 .000553104 4100000.00 .0009151922831 3103
 4200000.00 .001277280 4300000.00 .001639368 4400000.00 .0020014562831 3103
 4500000.00 .002751840 4600000.00 .003890520 4700000.00 .0050292002831 3103
 4800000.00 .006167880 4900000.00 .007306560 5000000.00 .0085883102831 3103
 5100000.00 .010013130 5200000.00 .011437950 5300000.00 .0128627702831 3103
 5400000.00 .014287590 5500000.00 .015850000 5600000.00 .0175500002831 3103
 5700000.00 .019250000 5800000.00 .020950000 5900000.00 .0226500002831 3103
 6000000.00 .024550000 6100000.00 .026650000 6200000.00 .0287500002831 3103
 6300000.00 .030850000 6400000.00 .032950000 6500000.00 .0350000002831 3103
 6600000.00 .037000000 6700000.00 .039000000 6800000.00 .0410000002831 3103
 6900000.00 .043000000 7000000.00 .045200000 7100000.00 .0476000002831 3103
 7200000.00 .050000000 7300000.00 .052400000 7400000.00 .0548000002831 3103
 7500000.00 .056900000 7600000.00 .058700000 7700000.00 .0605000002831 3103
 7800000.00 .062300000 7900000.00 .064100000 8000000.00 .0660000002831 3103
 8100000.00 .068000000 8200000.00 .070000000 8300000.00 .0720000002831 3103
 8400000.00 .074000000 8500000.00 .075800000 8600000.00 .0774000002831 3103
 8700000.00 .079000000 8800000.00 .080600000 8900000.00 .0822000002831 3103
 9000000.00 .083950000 9100000.00 .085850000 9200000.00 .0877500002831 3103
 9300000.00 .089650000 9400000.00 .091550000 9500000.00 .0933500002831 3103
 9600000.00 .095050000 9700000.00 .096750000 9800000.00 .0984500002831 3103
 9900000.00 .100150000 10000000.0 .101850000 10100000.0 .1035500002831 3103
 10200000.0 .105250000 10300000.0 .106950000 10400000.0 .1086500002831 3103
 10500000.0 .110350000 10600000.0 .112050000 10700000.0 .1137500002831 3103
 10800000.0 .115450000 10900000.0 .117150000 11000000.0 .1188500002831 3103
 11100000.0 .120550000 11200000.0 .122250000 11300000.0 .1239500002831 3103
 11400000.0 .125650000 11500000.0 .127350000 11600000.0 .1290500002831 3103
 11700000.0 .130750000 11800000.0 .132450000 11900000.0 .1341500002831 3103
 12000000.0 .135550000 12100000.0 .136650000 12200000.0 .1377500002831 3103
 12300000.0 .138850000 12400000.0 .139950000 12500000.0 .1410500002831 3103
 12600000.0 .142150000 12700000.0 .143250000 12800000.0 .1443500002831 3103
 12900000.0 .145450000 13000000.0 .146200000 13100000.0 .1466000002831 3103
 13200000.0 .147000000 13300000.0 .147400000 13400000.0 .1478000002831 3103
 13500000.0 .147850000 13600000.0 .147550000 13700000.0 .1472500002831 3103
 13800000.0 .146950000 13900000.0 .146650000 14000000.0 .1460520002831 3103
 14100000.0 .145156000 14200000.0 .144260000 14300000.0 .1433640002831 3103
 14400000.0 .142468000 14500000.0 .140559000 14600000.0 .1376370002831 3103
 14700000.0 .134715000 14800000.0 .131793000 14900000.0 .1288710002831 3103
 15000000.0 .125949000 15100000.0 .123027000 15200000.0 .1201050002831 3103
 15300000.0 .117183000 15400000.0 .114261000 15500000.0 .1113390002831 3103
 15600000.0 .108417000 15700000.0 .105495000 15800000.0 .1025730002831 3103
 15900000.0 .099651000 16000000.0 .097483667 16100000.0 .0960710002831 3103
 16200000.0 .094658333 16300000.0 .093245667 16400000.0 .0918330002831 3103
 16500000.0 .090420333 16600000.0 .089007667 16700000.0 .0875950002831 3103
 16800000.0 .086182333 16900000.0 .084769667 17000000.0 .0833570002831 3103
 17100000.0 .081944333 17200000.0 .080531667 17300000.0 .0791190002831 3103
 17400000.0 .077706333 17500000.0 .076620000 17600000.0 .0758600002831 3103
 17700000.0 .075100000 17800000.0 .074340000 17900000.0 .0735800002831 3103
 18000000.0 .072820000 18100000.0 .072060000 18200000.0 .0713000002831 3103
 18300000.0 .070540000 18400000.0 .069780000 18500000.0 .0690200002831 3103
 18600000.0 .068260000 18700000.0 .067500000 18800000.0 .0667400002831 3103
 18900000.0 .065980000 19000000.0 .065220000 19100000.0 .0644600002831 3103
 19200000.0 .063700000 19300000.0 .062940000 19400000.0 .0621800002831 3103
 19500000.0 .061420000 19600000.0 .060660000 19700000.0 .0599000002831 3103
 19800000.0 .059140000 19900000.0 .058380000 20000000.0 0.0       2831 3103
                                                                  2831 3  0
                                                                  2831 0  0
 2.80600E+4 5.94160E+1          0          0          0          1283133103
 0.000000+0 0.000000+0          0        103          0          4283133103
 0.000000+0 0.000000+0          0          0          6          3283133103
 1.000000-5 0.000000+0 2.700000+6 1.00000-11 2.000000+7 0.000000+0283133103
 0.000000+0 0.000000+0          0          1          6          3283133103
 1.000000-5 0.000000+0 2.700000+6 5.000000-3 2.000000+7 0.000000+0283133103
 0.000000+0 0.000000+0          0          1         16          8283133103
 1.000000-5 0.000000+0 2.700000+6 7.200000-2 5.000000+6 1.800000-2283133103
 7.000000+6 1.012500-2 1.000000+7 1.012500-2 1.200000+7 4.500000-3283133103
 1.600000+7 1.800000-2 2.000000+7 0.000000+0                      283133103
 0.000000+0 0.000000+0          0          8         16          8283133103
 1.000000-5 0.000000+0 2.700000+6 4.66640-17 5.000000+6 1.240600-7283133103
 7.000000+6 2.178000-6 1.000000+7 1.147600-5 1.200000+7 9.112500-6283133103
 1.600000+7 1.928300-5 2.000000+7 0.000000+0                      283133103
                                                                  283133  0
                                                                  2831 0  0
                                                                     0 0  0
 2.90630E+4 6.23890E+1          0          0         34         102925 1451
 0.0        0.0                 0          0          0          62925 1451
 1.00000E+0 2.00000E+7          0          0         10       20022925 1451
 3.00000E+2 0.0                 1          0        550          72925 1451
 29-Cu- 63  LANL,ORNL,FEI EVAL-FEB98 A.KONING ET AL,K.Zolotarev   2925 1451
                      DIST-Feb2004                                2925 1451
----IRDF-2002         MATERIAL 2925                               2925 1451
-----INCIDENT NEUTRON DATA                                        2925 1451
------ENDF-6 FORMAT                                               2925 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2925 1451
DOSIMETRY                                                         2925 1451
******************************************************************2925 1451
  29-Cu- 63 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK      2925 1451
 Ch98,Ch99            DIST-SEP 1 REV4-                 20010926   2925 1451
----ENDF/B-VI         MATERIAL 2925         REVISION 3            2925 1451
 **************************************************************** 2925 1451
***************************************************************** 2925 1451
 29-Cu- 63 FEI        EVAL-Nov01 K.I.Zolotarev                    2925 1451
                      DIST-Jan02                       20020110   2925 1451
----BROND-2           MATERIAL 2925                               2925 1451
-----INCIDENT NEUTRON DATA                                        2925 1451
------ENDF-6 FORMAT                                               2925 1451
------Russian Reactor Dosimetry File RRDF-2002                    2925 1451
***************************************************************** 2925 1451
Author  of evaluation: K.I.Zolotarev                              2925 1451
                    ----- MF=3 MT=107 -----                       2925 1451
***************************************************************** 2925 1451
******** Start of (N,2N), (N.G) bibliographical component ******* 2925 1451
***************************************************************** 2925 1451
                                                                  2925 1451
 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy,    2925 1451
                 P.G. Young (LANL)                                2925 1451
                                                                  2925 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   2925 1451
 MT 102) has been updated for new experimental data at incident   2925 1451
 neutron energies up to 1 keV. The Q-value for radiative capture  2925 1451
 was also updated in File 3.                                      2925 1451
 Details of these changes are described in Frankel et al. [Fr01]. 2925 1451
                                                                  2925 1451
 **************************************************************** 2925 1451
                                                                  2925 1451
 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN),    2925 1451
                 M.B. Chadwick, P.G. Young (LANL)                 2925 1451
                                                                  2925 1451
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     2925 1451
   in cooperation with ECN Petten.                                2925 1451
                                                                  2925 1451
  This evaluation provides a complete representation of the       2925 1451
nuclear data needed for transport, damage, heating,               2925 1451
radioactivity, and shielding applications over the incident       2925 1451
neutron energy range from 1.0E-11 to 150 MeV.  The discussion     2925 1451
here is divided into the region below and above 20 MeV.           2925 1451
                                                                  2925 1451
INCIDENT NEUTRON ENERGIES < 20 MeV                                2925 1451
                                                                  2925 1451
  Below 20 MeV the evaluation is based completely on the ENDF/B-  2925 1451
VI (Mod 3) evaluation by D. Hetrick, C.Y. Fu, and D. Larson.      2925 1451
                                                                  2925 1451
INCIDENT NEUTRON ENERGIES > 20 MeV                                2925 1451
                                                                  2925 1451
  The ENDF/B-VI Release 2 evaluation extends to 20 MeV and        2925 1451
includes cross sections and energy-angle data for all             2925 1451
significant reactions.  The present evaluation utilizes a more    2925 1451
compact composite reaction spectrum representation above 20 MeV   2925 1451
in order to reduce the length of the file. No essential data for  2925 1451
applications is lost with this representation.                    2925 1451
  The evaluation above 20 MeV utilizes MF=6, MT=5 to represent    2925 1451
all reaction data.  Production cross sections and emission        2925 1451
spectra are given for neutrons, protons, deuterons, tritons,      2925 1451
alpha particles, gamma rays, and all residual nuclides produced   2925 1451
(A>5) in the reaction chains. To summarize, the ENDF sections     2925 1451
with non-zero data above En = 20 MeV are:                         2925 1451
                                                                  2925 1451
  MF=3 MT=  1  Total Cross Section                                2925 1451
       MT=  2  Elastic Scattering Cross Section                   2925 1451
       MT=  3  Nonelastic Cross Section                           2925 1451
       MT=  5  Sum of Binary (n,n') and (n,x) Reactions           2925 1451
                                                                  2925 1451
  MF=4 MT=  2  Elastic Angular Distributions                      2925 1451
                                                                  2925 1451
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         2925 1451
               Distributions for Emission Neutrons, Protons,      2925 1451
               Deuterons, Tritons, and Alphas; and Angle-         2925 1451
               Integrated Spectra for Gamma Rays and Residual     2925 1451
               Nuclei That Are Stable Against Particle Emission   2925 1451
                                                                  2925 1451
  The evaluation is based on nuclear model calculations that      2925 1451
have been benchmarked to experimental data, especially for n +    2925 1451
Cu65 and p + Cu65 reactions [Ch98]. We use the GNASH code system  2925 1451
[Yo92], which utilizes Hauser-Feshbach statistical, preequilib-   2925 1451
rium and direct-reaction theories. Spherical optical model        2925 1451
calculations are used to obtain particle transmission             2925 1451
coefficients for the Hauser-Feshbach calculations, as well as     2925 1451
for the elastic neutron angular distributions.                    2925 1451
  Cross sections and spectra for producing individual residual    2925 1451
nuclei are included for reactions. The energy-angle-correlations  2925 1451
for all outgoing particles are based on Kalbach systematics       2925 1451
[Ka88].                                                           2925 1451
  A model was developed to calculate the energy distributions of  2925 1451
all recoil nuclei in the GNASH calculations [Ch96a]. The recoil   2925 1451
energy distributions are represented in the laboratory system in  2925 1451
MT=5, MF=6, and are given as isotropic in the lab system. All     2925 1451
other data in MT=5,MF=6 are given in the center-of-mass system.   2925 1451
This method of representation utilizes the LCT=3 option approved  2925 1451
at the November, 1996, CSEWG meeting.                             2925 1451
  Preequilibrium corrections were performed in the course of the  2925 1451
GNASH calculations using the exciton model of Kalbach [Ka77,      2925 1451
Ka85], validated by comparison with calculations using Feshbach,  2925 1451
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     2925 1451
nuclear data sheets were matched to continuum level densities     2925 1451
using the formulation of Ignatyuk et al. [Ig75] and pairing and   2925 1451
shell parameters from the Cook [Co67] analysis. Neutron and       2925 1451
charged- particle transmission coefficients were obtained from    2925 1451
the optical potentials, as discussed below.  Gamma-ray            2925 1451
transmission coefficients were calculated using the Kopecky-Uhl   2925 1451
model [Ko90].                                                     2925 1451
                                                                  2925 1451
SPECIFIC INFORMATION CONCERNING THE 63Cu EVALUATION               2925 1451
                                                                  2925 1451
This evaluation is documented in some detail in Ref. [Ko98b].     2925 1451
                                                                  2925 1451
The neutron total cross section above 20 MeV was obtained by      2925 1451
evaluating experimental data, with a particular emphasis on the   2925 1451
Finlay [Fi93] elemental data. This resulted in an evaluated       2925 1451
elemental Cu total cross section; to obtain an isotopic 63Cu total2925 1451
cross section, it was assumed that 63Cu and 65Cu have total cross 2925 1451
sections in an A**2/3 ratio to one another. The total neutron     2925 1451
nonelastic cross section was obtained directly from an optical    2925 1451
model calculation (see below), after verifying that it was in good2925 1451
agreement with the experimental data [Ko98b].                     2925 1451
                                                                  2925 1451
To obtain the neutron optical potential we used total cross       2925 1451
section data from 1.2 to 4.5 MeV [Gu86] and from 5.3 to 600 MeV   2925 1451
[Fi93], and elastic scattering angular distribution data from 1.6 2925 1451
to 96 MeV [Br50, Sa60, Ki74, El82, Gu86]. The optical potential   2925 1451
parameters were obtained using a combination of a grid search code2925 1451
and the interactive optical model viewer ECISVIEW [Ko97], both    2925 1451
built around the coupled channels code ECIS96 [Ra94]. The energy  2925 1451
dependence of the optical model parameters is as described in     2925 1451
[Ko98]. This optical potential was used for the calculation, with 2925 1451
ECIS96, of neutron transmission coefficients and DWBA cross       2925 1451
sections for the entire energy region above 20 MeV.               2925 1451
                                                                  2925 1451
Due to the lack of proton elastic scattering data in numerical    2925 1451
form, we used a combination of global optical models for the      2925 1451
proton channel. The Becchetti-Greenlees potential [Be69]was       2925 1451
adopted below 47 MeV, and the non-relativistic version of the     2925 1451
Madland potential [Ma88] above 47 MeV. At this particular energy  2925 1451
point the two potentials join smoothly.                           2925 1451
                                                                  2925 1451
For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2925 1451
for alpha particles the Moyen potential (MacFadden-Satchler       2925 1451
[Ma66]) was used; and for tritons the Becchetti-Greenlees         2925 1451
potential [Be71] was used. The He-3 channel was ignored, due to   2925 1451
its small importance.                                             2925 1451
                                                                  2925 1451
Following Delaroche et al. [De82], we adopted the weak-coupling   2925 1451
model for direct collective inelastic scattering for Cu-63, using 2925 1451
Ni-64 as a basis. For the calculation of the cross sections,      2925 1451
ECIS96 was used in DWBA mode. We used the following direct        2925 1451
transitions for Cu-63  (ground state 3/2- ) :                     2925 1451
                                                                  2925 1451
Jpi Ex(MeV) Deformation lengths                                   2925 1451
 0.5- 0.669   Delta(2)=0.319                                      2925 1451
 2.5- 0.962   Delta(2)=0.552                                      2925 1451
 3.5- 1.327   Delta(2)=0.638                                      2925 1451
 1.5- 1.547   Delta(2)=0.451                                      2925 1451
 1.5- 3.382   Delta(3)=0.313                                      2925 1451
 2.5- 3.632   Delta(3)=0.384                                      2925 1451
 3.5- 3.882   Delta(3)=0.444                                      2925 1451
 4.5- 4.132   Delta(3)=0.496                                      2925 1451
                                                                  2925 1451
No measurements exist for neutron-induced emission spectra above  2925 1451
20 MeV for 63Cu. However, for Cu-65 there exists 25.7 MeV (n,xn)  2925 1451
data by Marcinkowski et al [Ma83]. This has been used to benchmark2925 1451
the Cu-65 data. Without adjusting any of the level density or pre-2925 1451
equilibrium parameters the GNASH calculation was in good agreement2925 1451
with these data. Hence we also adopted these parameters for the   2925 1451
whole energy region for Cu-63.                                    2925 1451
                                                                  2925 1451
****************************************************************  2925 1451
                                                                  2925 1451
REFERENCES                                                        2925 1451
                                                                  2925 1451
[Ab93]  W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra,   2925 1451
   Phys.Rev. C47, 1033 (1993)                                     2925 1451
[Al83]  R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986)   2925 1451
[Ar80]  E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross  2925 1451
   Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross   2925 1451
   Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2925 1451
   Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2925 1451
   1980] p. 731.                                                  2925 1451
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182,   2925 1451
   1190 (1969)                                                    2925 1451
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization  2925 1451
   Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W.    2925 1451
   Haeberli, The University of Wisconsin Press, 1971) p.682.      2925 1451
[Be92]  O. Bersillon, "SCAT2 - A Spherical Optical Model Code,"   2925 1451
   in Proc. ICTP Workshop on Computation and Analysis of Nuclear  2925 1451
   data Relevant to Nuclear Energy and Safety, February-March,    2925 1451
   1999 Trieste, Italy, to be published in World Scientific Press,2925 1451
   and Progress Report of the Nuclear Physics Division, Bruyeres- 2925 1451
   le-Chatel 1977, CEA-N-2037 (1978) p.111                        2925 1451
[Br50]  S. Bratenahl, S. Fernbach, R.H. Hildebrand et al.,        2925 1451
   Phys.Rev. 77, 597 (1950)                                       2925 1451
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2925 1451
[Ch96]  M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      2925 1451
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   2925 1451
   Driven Technologies:  Calculational Method for Heavy Recoils," 2925 1451
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    2925 1451
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996     2925 1451
[Ch98]   M. B. Chadwick and P. G. Young, "GNASH Calculations of   2925 1451
   n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2925 1451
   REPORT: 1 February - 1 March 1998, internal Los Alamos National2925 1451
   Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2925 1451
[Ch99]  M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2925 1451
   National Laboratory report, LA-UR-99-1222 (1999)               2925 1451
[Co67]  J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J.    2925 1451
   Phys. 20, 477 (1967)                                           2925 1451
[De82]  J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and   2925 1451
   R.L. Walter, Nucl. Phys. A390, 541 (1982).                     2925 1451
[El82]  S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2925 1451
   A390, 509 (1982)                                               2925 1451
[Fi93]  R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys.Rev. 2925 1451
   C 47, 237 (1993)                                               2925 1451
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      2925 1451
   National Laboratory Report, LA-13812 (2001).                   2925 1451
[Gu86]  P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl.   2925 1451
   Phys. A448, 280 (1986)                                         2925 1451
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    2925 1451
   Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485     2925 1451
   (1975)                                                         2925 1451
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     2925 1451
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         2925 1451
   LA-10248-MS (1985)                                             2925 1451
[Ka88]  C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also          2925 1451
   C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981)           2925 1451
[Ki74]  W.E. Kinney, F.G. Perey, Oak Ridge report ORNL-4908 (1974)2925 1451
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990)         2925 1451
[Ko97]  A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2925 1451
   A Graphical Interface for ECIS95", Proceedings of the NEA      2925 1451
   Specialists' Meeting on the Nucleon Nucleus Optical Model up to2925 1451
   200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2925 1451
   http://db.nea.fr/html/science/om200/.                          2925 1451
[Ko98]  A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear   2925 1451
   Data for Accelerator-Driven Systems: Nuclear models, Experiment2925 1451
   and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2925 1451
[Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI    2925 1451
   neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2925 1451
   Los Alamos National Laboratory report LAUR- (1998); ECN lab and2925 1451
   JEFF report (1998).                                            2925 1451
[Lo74]  J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974)    2925 1451
[Ma66]  Macfadden and Satchler, Nuc.Phys. 84, 177 (1966)          2925 1451
[Ma83]  A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al.,  2925 1451
   Nucl.Phys. A402, 220 (1983)                                    2925 1451
[Ma88]  D.G. Madland, "Recent Results in the Development of a     2925 1451
   Global Medium-Energy Nucleon-Nucleus Optical-Model Potential,  2925 1451
   "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2925 1451
   Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2925 1451
   (1988).                                                        2925 1451
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      2925 1451
[Ra94]  J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772  2925 1451
   (1994)                                                         2925 1451
[Sa60]  G.L. Salmon, Nucl.Phys. 21, 15 (1960)                     2925 1451
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, report        2925 1451
   LA-12343-MS (1992)                                             2925 1451
                                                                  2925 1451
 **************************************************************** 2925 1451
                                                                  2925 1451
 ENDF/B-VI MOD 3 Revision, July 1991 (ORNL)                       2925 1451
                                                                  2925 1451
 MOD 3 changes                                                    2925 1451
   1) Corrections to MF=6, MT=65 at 17.0 MeV to prevent negative  2925 1451
      values in the angular distribution.                         2925 1451
   2) Corrections to MF=33, MT=102                                2925 1451
                                                                  2925 1451
 **************************************************************** 2925 1451
                                                                  2925 1451
 * Note there was no MOD 2 released.                              2925 1451
                                                                  2925 1451
 **************************************************************** 2925 1451
                                                                  2925 1451
 ENDF/B-VI MOD 1 Evaluation, October 1989, D. Hetrick, F.Y. Fu,   2925 1451
           D. Larson (ORNL)                                       2925 1451
                                                                  2925 1451
   This work employed several nuclear model codes including the   2925 1451
 optical-model code GENOA [1], the Distorted Wave Born            2925 1451
 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach  2925 1451
 code TNG [3,4]. The TNG code provides energy and angular         2925 1451
 distributions of particles emitted in the compound and pre-      2925 1451
 compound reactions, ensures consistency among all reactions, and 2925 1451
 maintains energy balance. Details pertinent to the contents of   2925 1451
 this evaluation and extensive comparisons of calculations with   2925 1451
 experimental data can be found in reference [5].                 2925 1451
                                                                  2925 1451
----- DESCRIPTION OF FILES                                        2925 1451
(MF-MT)                                                           2925 1451
  1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS.         2925 1451
  2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB[6].  POINT2925 1451
        WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER  2925 1451
        FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2925 1451
        99.5 KEV.  REICH-MOORE PARAMETERS ARE GIVEN.  AGREEMENT   2925 1451
        WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2925 1451
        FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN       2925 1451
        AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND     2925 1451
        COMPARED WITH THE BINNED DATA.  THIS IS PROBABLY DUE TO   2925 1451
        TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES    2925 1451
        SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION.             2925 1451
        NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2925 1451
        OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE       2925 1451
        PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON    2925 1451
        A FINER ENERGY GRID THAN GIVEN IN 4/2.                    2925 1451
  3-1   THE TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS  2925 1451
        FROM 1.E-5 EV TO 99.5 KEV.  FROM 1.E-5 TO 1 EV, -0.9B IS  2925 1451
        GIVEN TO REDUCE THE ELASTIC AND GIVE THE CORRECT TOTAL    2925 1451
        THERMAL CROSS SECTION (9.6 B).  FROM 1 EV TO 170 EV THIS  2925 1451
        GOES LINEARLY TO ZERO.  A SMALL CONTRIBUTION FROM 3/102 IS2925 1451
        REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151   2925 1451
        REPRODUCES THE CAPTURE DATA.  FROM 99.5 KEV TO 1.12 MEV   2925 1451
        CU63 DATA FROM [7] IS USED, AFTER APPROPRIATE AVERAGING.  2925 1451
        ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE.  FROM 1.12 TO  2925 1451
        4.0 MEV,NAT CU  DATA OF PEREY [8] USED IN V5 IS RETAINED. 2925 1451
        FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL [9] IS    2925 1451
        AVERAGED AND USED.  COMPARISONS FROM 1.2 TO 4.5           2925 1451
        MEV WITH AVERAGED ARGONNE DATA FOR NAT CU [10] SHOW       2925 1451
        1% AGREEMENT.                                             2925 1451
  3-2   ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY        2925 1451
        SUBTRACTING THE NONELASTIC (3-3) FROM THE TOTAL.  THE     2925 1451
        THERMAL VALUE OF 5.1 B IS REPRODUCED.                     2925 1451
  3-102 (N,G) DATA TAKEN FROM RESONANCE PARAMETERS FROM 1.E-5 EVTO2925 1451
        99.5 KEV.  A SMALL BACKGROUND IS GIVEN HERE WHICH WHEN    2925 1451
        ADDED TO THE RESONACE CONTRIBUTION REPRODUCES EXPERIMENTAL2925 1451
        DATA,INCLUDING THE THERMAL VALUE OF 4.50B.  V5 DATA FROM  2925 1451
        99.5 KEV TO 20.0 MEV WERE REPLACED BY POINTS ON A CURVE   2925 1451
        DRAWN THROUGH DATA FROM THE CSISRS LIBRARY [5,13]; RESULTS2925 1451
        COMPARABLE TO EYE GUIDE IN REF [14].                      2925 1451
 ---------------------------------------------------------------- 2925 1451
 UNCERTAINTY FILES                                                2925 1451
        ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS       2925 1451
        REQUIRED BY ENDF/B-VI FORMATS                             2925 1451
                                                                  2925 1451
 33-1   TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV  2925 1451
        EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8.       2925 1451
 33-2   EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV2925 1451
 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2925 1451
        ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 2925 1451
        DATA [5,13,14] FROM 99.5 KEV TO 20 MEV.                   2925 1451
 **************************************************************** 2925 1451
                                                                  2925 1451
 REFERENCES:                                                      2925 1451
                                                                  2925 1451
 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967)    2925 1451
 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of           2925 1451
     Colorado, unpublished (1972)                                 2925 1451
 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu,           2925 1451
     Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2925 1451
     May 1980, Brookhaven National Lab. report BNL-NCS-51245      2925 1451
     (1980) p.675                                                 2925 1451
 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986)          2925 1451
 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report     2925 1451
     ORNL/TM-9083 [ENDF-337] (1984)                               2925 1451
 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron    2925 1451
     Cross Sections, Vol. 1, Neutron Resonance Parameters and     2925 1451
     Thermal Cross Sections, Part A, Z=1-60," (Academic Press,    2925 1451
     1981)                                                        2925 1451
 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600  2925 1451
     (1977), and private communication.                           2925 1451
 [8] F.G. Perey, private communication (1977)                     2925 1451
 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50   2925 1451
     MeV, Upton, NY, May 1980, Brookhaven National Lab. report    2925 1451
     BNL-NCS-51245 (1980) p.277                                   2925 1451
[10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl.   2925 1451
     Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and    2925 1451
     W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80      2925 1451
     (1983) [CSISRS data set 12853]                               2925 1451
[11] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341      2925 1451
     [ENDF-303] (1980)                                            2925 1451
[12] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp.    2925 1451
     on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26,   2925 1451
     1982 (U.S. National Bureau of Standards) p.877               2925 1451
[13] CSISRS Library, National Nuclear Data Center, Brookhaven     2925 1451
     National Laboratory, Upton, N.Y. 11973.                      2925 1451
[14] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross        2925 1451
     Sections, Vol. 2, Neutron Cross Section Curves" (Academic    2925 1451
     Press, 1988)                                                 2925 1451
[15] S.M. Qaim, Radiochimica Acta, 25, 13 (1978)                  2925 1451
[16] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys.     2925 1451
     A404, 250 (1983)                                             2925 1451
                                                                  2925 1451
***************************************************************** 2925 1451
********   End of (N,2N), (N.G) bibliographical component ******* 2925 1451
***************************************************************** 2925 1451
***************************************************************** 2925 1451
********       Start of (N,A) bibliographical component   ******* 2925 1451
***************************************************************** 2925 1451
------Russian Reactor Dosimetry File RRDF-2002                    2925 1451
***************************************************************** 2925 1451
Author  of evaluation: K.I.Zolotarev                              2925 1451
***************************************************************** 2925 1451
                                                                  2925 1451
                    ----- MF=3 MT=107 -----                       2925 1451
    Evaluation of Cu-63(n,a)Co-60m+g- excitation function was car-2925 1451
ried out by means of statistical analysis of cross sections from  2925 1451
data base prepared in the energy range 2 - 20 Mev. In the energy  2925 1451
range 3.56 - 19.55 MeV  input data base  was formed with using of 2925 1451
experimental data from ref. [1-19].  Cross section data  in  the  2925 1451
interval 2.0 - 3.5 MeV were taken from theoretical model calcula- 2925 1451
tion.  Experimental data  included  in  the input data base were  2925 1451
renormalized using new cross sections standards for monitor reac- 2925 1451
tions and new standards for decay data.                           2925 1451
    The special correction  was applied to the experimental data  2925 1451
[2,3,12,13,15,18]. Cross section data of A.Paulsen and H.Liskien  2925 1451
[2-3] measured in the energy region 12.09 - 19.55 MeV with using  2925 1451
T(d,n)He4 neutron source were multiplied to the factor 1.20805 .  2925 1451
Experimental data of Lu Hanlin et al. [12,18] , Wang Yongchag et  2925 1451
al. [13] and Konno et al. [15]  were multiplied  to  the factors  2925 1451
0.88110, 0.86000, 0.84377 and 1.060, respectively. The correction 2925 1451
factors were derived from preliminary evaluated  cross sections   2925 1451
integrals in the energy intervals 8.4 - 11.4 and 13 - 15 MeV .    2925 1451
    Data  of A.Paulsen and H.Liskien [2]  in  the energy  range   2925 1451
5.76 - 11.48 MeV  measured  with using  D(d,n)He3 , Be9(a,n)C12 , 2925 1451
C14(d,n)N15 , N15(d,n)O16  neutron sources  were rejected  due to 2925 1451
their inconsistency with precision measurements  of  G.Winkler et 2925 1451
al. [8]  and integral experimental data for U-235 fission neutron 2925 1451
spectrum [20-23]  and Cf-252 spontaneous fission neutron spectrum 2925 1451
[25-26]. Cross sections measured by J.Kantele and D.Gardner [27], 2925 1451
M.Bormann et al. [28], G.Maslov et al. [29] and K.Kayashima et al.2925 1451
[30]  were also rejected due to the big discrepancy with the main 2925 1451
bulk experimental data.                                           2925 1451
    The final procedure  evaluation of (n,a) excitation function  2925 1451
was carried out by means of Pade-2 code [31].                     2925 1451
    Evaluated excitation function for the reaction  Cu63(n,a)Co60 2925 1451
was tested  with using  integral experimental  data  [20-24]  for 2925 1451
U-235 thermal fission neutron spectrum and evaluated integral ex- 2925 1451
perimental data [32-33]  for Cf-252 spontaneous  fission  neutron 2925 1451
spectrum.  Calculated  and  measured average cross section values 2925 1451
for U-235 thermal fission neutron spectrum [34] and Cf-252 sponta-2925 1451
neous fission neutron spectrum [35] are given in the table 1.     2925 1451
                                                      Table 1     2925 1451
  2925 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  2925 1451
  2925 1451
 U-235 neutron fission    0.53294        0.5269 +- 0.0316 [20]  2925 1451
                                         0.520  +- 0.040  [21]  2925 1451
                                         0.519  +- 0.035  [22]  2925 1451
                                         0.5273 +- 0.0156 [23]  2925 1451
                                         0.538  +- 0.015  [23]  2925 1451
                                         0.4935 +- 0.0242 [24]  2925 1451
  2925 1451
 CF-252 spont. fission    0.69248        0.709  +- 0.017  [25]  2925 1451
                                         0.675  +- 0.018  [26]  2925 1451
                                         0.6897 +- 0.0130 [32]  2925 1451
                                         0.6887 +- 0.0135 [33]  2925 1451
  2925 1451
                                                                  2925 1451
                    ----- MF=33 MT=107 -----                      2925 1451
    Uncertainties in the evaluated  excitation function  for the  2925 1451
reaction Cu-63(n,a)Co-60m+g are given in the form of relative co- 2925 1451
variance matrix for the 18-neutron energy groups (LB=5).  Covari- 2925 1451
ance matrix of uncertainties was calculated simultaneously  with  2925 1451
recommended cross section data by means of PADE-2 code.           2925 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  2925 1451
given in the 33-file are the following:                           2925 1451
                                                                  2925 1451
    9.39333E-10    1.02943E-09    1.18681E-09    1.49652E-09      2925 1451
    2.03282E-09    2.85265E-09    3.69906E-09    6.46604E-09      2925 1451
    7.96491E-09    7.69999E-08    4.12497E-06    6.99817E-04      2925 1451
    1.07280E-03    1.33316E-03    1.89667E-03    4.15078E-03      2925 1451
    6.87556E-03    8.81528E-03                                    2925 1451
                                                                  2925 1451
References :                                                      2925 1451
 1. B.Czapp, H.Vonach  Oesterr. Akad. Wiss, Math, Naturw.         2925 1451
    Anzeiger, v.97, p.13, January 1960                            2925 1451
 2. A.Paulsen, H.Liskien  Nukleonik, v.10, p.91, July 1967        2925 1451
 3. A.Paulsen  Zeitschrift f. Phys., v.205, p.226, August 1967    2925 1451
 4. R.C.Barrall et al.  Report AFWL-TR-68-134, March 1969         2925 1451
 6. G.Winkler  Nucl. Sci. Eng., v.67, n.2, p.260, August 1978     2925 1451
 8. G.Winkler, D.L.Smith, J.W.Meadows  Nucl. Sci. Eng., v.76,     2925 1451
    p.30, October 1980                                            2925 1451
 9. U.Garuska et al.  Prog. Report INDC(POL)-11, p.15, July 1980  2925 1451
10. O.I.Artem'ev et al.  Atomnaja Energija (Sov.), v.49, n.3,     2925 1451
    p.195, September 1980                                         2925 1451
11. L.R.Greenwood  Progress Report ASTM-STP-956, p.743, 1987      2925 1451
12. Lu Hanlin et al.  China Journal of Nuclear Phys., v.12. n.4,  2925 1451
    p. 373, 1990                                                  2925 1451
13. Wang Yongchang et al. Chinese J. High Energy Phys. and Nucl.  2925 1451
    Phys., v.14, p.919, October 1990                              2925 1451
14. J.Csikai, C.M.Buczko, R.Pepelnik, H.M.Agrawal Annals of Nucl. 2925 1451
    Energy, v.18, n.1, p.1, 1991                                  2925 1451
15. C.Konno et al.  Report JAERI-1329, October 1993               2925 1451
16. J.W.Meadows et al.  Annals of Nucl. Energy, v.23, p.877, 1996 2925 1451
17. A.A.Filatenkov et al.  VANT, Ser.: Yadernye Konstanty, v.2,   2925 1451
    p.8, Moscow, 1996                                             2925 1451
18. Lu Hanlin et al.  Report INDC(CPR)-045, IAEA, October 1998    2925 1451
19. A.A.Filatenkov et al.  Report RI-252, St.Petersburg, May 1999 2925 1451
20. R.Lloret Progress Report EANDC(E)-57, p.172, February 1965    2925 1451
21. A.Fabry, J.P.Deworm  Progress Report EANDC(E)-66, p.125,      2925 1451
    February 1966                                                 2925 1451
22. K.Kobayashi et al.  Nucl. Sci. Techn., v.13, p.531, Oct. 1976 2925 1451
23. L.P.Geraldo et al.  Radiochimica Acta, v.57, pp.63-67, 1992   2925 1451
24. W.Mannhart  Progress Report INDC(Ger)-045, pp.40-43, 1999     2925 1451
25. G.Winkler et al.  Nucl. Sci. Eng., v.78, p.415, August 1981   2925 1451
26. W.Mannhart  Proc. of Int. Conf. Nuclear Data for Science and  2925 1451
    Technology, 6-10 September 1982, Antwerp, Holland, D.Reidel   2925 1451
    Publishing Company, p.429                                     2925 1451
27. J.Kantele, D.Gardner  Nucl. Phys., v.35, p.353, 1962          2925 1451
28. M.Bormann et al.  Nucl. Phys., v.A186, p.65, May 1972         2925 1451
29. G.N.Maslov, F.Nasyrov, N.F.Pashkin  Yadernye Konstanty, v.9,  2925 1451
    p.50, Obninsk, 1972                                           2925 1451
30. K.Kayashima et al.  Prog. Rep. NEANDC(J)-61U, p.94, Sep. 1979 2925 1451
31. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          2925 1451
32. W.Mannhart   Handbook on Nuclear Activation Cross Sections,   2925 1451
    IAEA Technical Report series No.273, p.413, 1987              2925 1451
33. W.Mannhart  Validation of Differential Cross Sections with    2925 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2925 1451
    September 2002                                                2925 1451
34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2925 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              2925 1451
35. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        2925 1451
***************************************************************** 2925 1451
********       End of (N,A)bibliographical component      ******* 2925 1451
***************************************************************** 2925 1451
                                                                  2925 1451
                                                                  2925 1451
                                                                  2925 1451
 ************************ C O N T E N T S *********************** 2925 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2925 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2925 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2925 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2925 1451
 ***************** Program FIXUP (Version 2002-1) ****************2925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2925 1451
 Corrected Thresholds-----------------------------------------Yes 2925 1451
 Extended Cross Sections to 20 MeV----------------------------No  2925 1451
 Allow Cross Section Deletion---------------------------------No  2925 1451
 Allow Cross Section Reconstruction---------------------------No  2925 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2925 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2925 1451
 Deleted Duplicate Points-------------------------------------Yes 2925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2925 1451
 Allow Creation of Missing Sections---------------------------No  2925 1451
 Allow Insertion of Energy Points-----------------------------No  2925 1451
 Create Uniform Energy Grid-----------------------------------No  2925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2925 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2925 1451
 Unshielded Group Averages Using  640 Groups                      2925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2925 1451
                                1        451        561          52925 1451
                                3         16         34          12925 1451
                                3        102        217          52925 1451
                                3        107         63          12925 1451
                               33         16         58          12925 1451
                               33        102         23          32925 1451
                               33        107         38          12925 1451
                                                                  2925 1  0
                                                                  2925 0  0
 2.90630E+4 6.23890E+1          0          0          0          02925 3 16
-1.08540E+7-1.08540E+7          0          0          1         912925 3 16
         91          1                                            2925 3 16
 11000000.0 .000429103 11100000.0 .002018629 11200000.0 .0036728762925 3 16
 11300000.0 .006750000 11400000.0 .011250000 11500000.0 .0182500002925 3 16
 11600000.0 .027750000 11700000.0 .039500000 11800000.0 .0535000002925 3 16
 11900000.0 .067948571 12000000.0 .082845714 12100000.0 .0977428572925 3 16
 12200000.0 .113333107 12300000.0 .133082000 12400000.0 .1535240002925 3 16
 12500000.0 .173966000 12600000.0 .194408000 12700000.0 .2150142502925 3 16
 12800000.0 .236606000 12900000.0 .258362000 13000000.0 .2801180002925 3 16
 13100000.0 .301874000 13200000.0 .322831375 13300000.0 .3389970002925 3 16
 13400000.0 .354364000 13500000.0 .369731000 13600000.0 .3850980002925 3 16
 13700000.0 .400465000 13800000.0 .415832000 13900000.0 .4311990002925 3 16
 14000000.0 .446566000 14100000.0 .461933000 14200000.0 .4772576252925 3 16
 14300000.0 .492328000 14400000.0 .507356000 14500000.0 .5223840002925 3 16
 14600000.0 .537412000 14700000.0 .551971000 14800000.0 .5637160002925 3 16
 14900000.0 .574992000 15000000.0 .586268000 15100000.0 .5975440002925 3 16
 15200000.0 .608696250 15300000.0 .619106000 15400000.0 .6293920002925 3 16
 15500000.0 .639678000 15600000.0 .649964000 15700000.0 .6598785002925 3 16
 15800000.0 .667564000 15900000.0 .674878000 16000000.0 .6821920002925 3 16
 16100000.0 .689506000 16200000.0 .696376250 16300000.0 .7005840002925 3 16
 16400000.0 .704348000 16500000.0 .708112000 16600000.0 .7118760002925 3 16
 16700000.0 .715640000 16800000.0 .719404000 16900000.0 .7231680002925 3 16
 17000000.0 .726932000 17100000.0 .730696000 17200000.0 .7344600002925 3 16
 17300000.0 .738224000 17400000.0 .741988000 17500000.0 .7456035002925 3 16
 17600000.0 .749070500 17700000.0 .752537500 17800000.0 .7560045002925 3 16
 17900000.0 .759471500 18000000.0 .762938500 18100000.0 .7664055002925 3 16
 18200000.0 .769872500 18300000.0 .773339500 18400000.0 .7768065002925 3 16
 18500000.0 .780356000 18600000.0 .783988000 18700000.0 .7876200002925 3 16
 18800000.0 .791252000 18900000.0 .794884000 19000000.0 .7985160002925 3 16
 19100000.0 .802148000 19200000.0 .805780000 19300000.0 .8094120002925 3 16
 19400000.0 .813044000 19500000.0 .814860000 19600000.0 .8148600002925 3 16
 19700000.0 .814860000 19800000.0 .814860000 19900000.0 .8148600002925 3 16
 20000000.0 0.0                                                   2925 3 16
                                                                  2925 3  0
 2.90630E+4 6.23890E+1          0          0          0          02925 3102
 7.91602E+6 7.91602E+6          0          0          1        6412925 3102
        641          1                                            2925 3102
 .000100000 70.2622035 .000105000 68.6237994 .000110000 67.06729662925 3102
 .000115000 65.6330228 .000120000 63.9563318 .000127500 62.09608432925 3102
 .000135000 60.3964753 .000142500 58.8345693 .000150000 57.14415432925 3102
 .000160000 55.3831827 .000170000 53.7809161 .000180000 52.31015672925 3102
 .000190000 50.9417396 .000200000 49.6923602 .000210000 48.51690432925 3102
 .000220000 47.4250938 .000230000 46.4106884 .000240000 45.21999192925 3102
 .000255000 43.9115387 .000270000 42.9018717 .000280000 41.77792572925 3102
 .000300000 40.4072412 .000320000 39.1661345 .000340000 38.02948142925 3102
 .000360000 36.9825557 .000380000 36.0268482 .000400000 35.03047112925 3102
 .000425000 34.0108822 .000450000 33.0793104 .000475000 32.22522862925 3102
 .000500000 31.4220240 .000525000 30.6879457 .000550000 29.99449532925 3102
 .000575000 29.3495653 .000600000 28.6885484 .000630000 28.00852932925 3102
 .000660000 27.3863025 .000690000 26.7894143 .000720000 26.15430472925 3102
 .000760000 25.4733201 .000800000 24.8410401 .000840000 24.26158862925 3102
 .000880000 23.7108520 .000920000 23.2059760 .000960000 22.72247112925 3102
 .001000000 22.2213219 .001050000 21.6981447 .001100000 21.20766632925 3102
 .001150000 20.7557256 .001200000 20.2226031 .001275000 19.63667432925 3102
 .001350000 19.1009798 .001425000 18.6010383 .001500000 18.06989802925 3102
 .001600000 17.5145304 .001700000 17.0075434 .001800000 16.53864652925 3102
 .001900000 16.1102769 .002000000 15.7131069 .002100000 15.34042142925 3102
 .002200000 14.9991178 .002300000 14.6727519 .002400000 14.30037732925 3102
 .002550000 13.8866075 .002700000 13.5629647 .002800000 13.21129882925 3102
 .003000000 12.7783972 .003200000 12.3844533 .003400000 12.02419892925 3102
 .003600000 11.6956128 .003800000 11.3922199 .004000000 11.07599662925 3102
 .004250000 10.7549209 .004500000 10.4618809 .004750000 10.18802182925 3102
 .005000000 9.93673050 .005250000 9.70381542 .005500000 9.483662792925 3102
 .005750000 9.28192760 .006000000 9.07007595 .006300000 8.858338982925 3102
 .006600000 8.65797260 .006900000 8.47214444 .007200000 8.270077472925 3102
 .007600000 8.05397897 .008000000 7.85633023 .008400000 7.670846742925 3102
 .008800000 7.49826745 .009200000 7.33790053 .009600000 7.185308302925 3102
 .010000000 7.02731331 .010500000 6.86027264 .011000000 6.707461282925 3102
 .011500000 6.56167121 .012000000 6.39456365 .012750000 6.209568252925 3102
 .013500000 6.03822804 .014250000 5.88109160 .015000000 5.713610162925 3102
 .016000001 5.53789841 .017000001 5.37685181 .017999999 5.229868612925 3102
 .018999999 5.09475963 .020000000 4.96798271 .021000000 4.851895632925 3102
 .022000000 4.74247553 .023000000 4.64047366 .024000000 4.520598702925 3102
 .025500000 4.39132258 .027000001 4.28886478 .028000001 4.177587382925 3102
 .029999999 4.04061874 .032000002 3.91584643 .034000002 3.801962922925 3102
 .035999998 3.69805476 .037999999 3.60225066 .039999999 3.502310602925 3102
 .042500000 3.40086338 .045000002 3.30811484 .047499999 3.221662622925 3102
 .050000001 3.14208839 .052499998 3.06842366 .055000000 2.998664342925 3102
 .057500001 2.93480171 .059999999 2.86765333 .063000001 2.800501852925 3102
 .066000000 2.73722878 .068999998 2.67830068 .071999997 2.614436882925 3102
 .075999998 2.54606707 .079999998 2.48366401 .083999999 2.425014432925 3102
 .088000000 2.37046825 .092000000 2.31988364 .096000001 2.271530822925 3102
 .100000001 2.22155040 .104999997 2.16862115 .109999999 2.120249322925 3102
 .115000002 2.07404477 .119999997 2.02112934 .127499998 1.962654412925 3102
 .135000005 1.90844599 .142499998 1.85872498 .150000006 1.805830022925 3102
 .159999996 1.75027369 .170000002 1.69942005 .180000007 1.652808602925 3102
 .189999998 1.60995002 .200000003 1.56967971 .209999993 1.532852322925 3102
 .219999999 1.49810051 .230000004 1.46573550 .239999995 1.428372152925 3102
 .254999995 1.38659312 .270000011 1.35462976 .280000001 1.319114872925 3102
 .300000012 1.27564619 .319999993 1.23640287 .340000004 1.201193502925 3102
 .360000014 1.16754401 .379999995 1.13707345 .400000006 1.105578112925 3102
 .425000012 1.07317407 .449999988 1.04354331 .474999994 1.016426382925 3102
 .500000000 .990997135 .524999976 .967568345 .550000012 .9457333122925 3102
 .574999988 .925091982 .600000024 .903971010 .629999995 .8831860302925 3102
 .660000026 .863367838 .689999998 .844306064 .720000029 .8240036922925 3102
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 .879999995 .746318045 .920000017 .730139243 .959999979 .7149109362925 3102
 1.00000000 .698831730 1.04999995 .682283042 1.10000002 .6665914032925 3102
 1.14999998 .652238681 1.20000005 .635245219 1.27499998 .6165585522925 3102
 1.35000002 .599628464 1.42499995 .583807325 1.50000000 .5668068512925 3102
 1.60000002 .549051099 1.70000005 .532869404 1.79999995 .5179035022925 3102
 1.89999998 .504165632 2.00000000 .491529322 2.09999990 .4795813572925 3102
 2.20000005 .468648836 2.29999995 .458213190 2.40000010 .4462346112925 3102
 2.54999995 .433010060 2.70000005 .422667338 2.79999995 .4114881472925 3102
 3.00000000 .397621803 3.20000005 .384942471 3.40000010 .3733207992925 3102
 3.59999990 .362690507 3.79999995 .352926310 4.00000000 .3426862962925 3102
 4.25000000 .332273009 4.50000000 .322781297 4.75000000 .3139566862925 3102
 5.00000000 .305775792 5.25000000 .298280761 5.50000000 .2911442622925 3102
 5.75000000 .284570789 6.00000000 .277665264 6.30000019 .2707298672925 3102
 6.59999990 .264325443 6.90000010 .258305768 7.19999981 .2515323122925 3102
 7.59999990 .244499479 8.00000000 .237997939 8.39999962 .2318562262925 3102
 8.80000019 .226218093 9.19999981 .220822288 9.60000038 .2158801322925 3102
 10.0000000 .210577911 10.5000000 .205101020 11.0000000 .2000608982925 3102
 11.5000000 .195304503 12.0000000 .189618666 12.7500000 .1834147062925 3102
 13.5000000 .177744563 14.2500000 .172462458 15.0000000 .1667994802925 3102
 16.0000000 .160887832 17.0000000 .155490509 18.0000000 .1504680722925 3102
 19.0000000 .145825409 20.0000000 .141569631 21.0000000 .1375452902925 3102
 22.0000000 .133835471 23.0000000 .130321643 24.0000000 .1263017862925 3102
 25.5000000 .121725563 27.0000000 .118265087 28.0000000 .1143610192925 3102
 30.0000000 .109605590 32.0000000 .105293926 34.0000000 .1013529082925 3102
 36.0000000 .097691052 38.0000000 .094306318 40.0000000 .0908373242925 3102
 42.5000000 .087259596 45.0000000 .083955886 47.5000000 .0809392262925 3102
 50.0000000 .078166135 52.5000000 .075545286 55.0000000 .0731605132925 3102
 57.5000000 .070846166 60.0000000 .068504381 63.0000000 .0661650282925 3102
 66.0000000 .063929306 69.0000000 .061851612 72.0000000 .0596063642925 3102
 76.0000000 .057237194 80.0000000 .055054440 84.0000000 .0530339982925 3102
 88.0000000 .051157235 92.0000000 .049420404 96.0000000 .0478088132925 3102
 100.000000 .046075291 105.000000 .044326491 110.000000 .0427115182925 3102
 115.000000 .041202203 120.000000 .039464769 127.500000 .0375586362925 3102
 135.000000 .035856054 142.500000 .034278756 150.000000 .0326370332925 3102
 160.000000 .030954016 170.000000 .029474509 180.000000 .0281204392925 3102
 190.000000 .026933840 200.000000 .025853772 210.000000 .0248895622925 3102
 220.000000 .024010332 230.000000 .023221487 240.000000 .0223683502925 3102
 255.000000 .021467091 270.000000 .020835647 280.000000 .0202159012925 3102
 300.000000 .019583405 320.000000 .019196792 340.000000 .0190620882925 3102
 360.000000 .019265403 380.000000 .021665902 400.000000 .0925170382925 3102
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 500.000000 .061705904 525.000000 .147182668 550.000000 1.868321322925 3102
 575.000000 46.0061716 600.000000 .200111344 630.000000 1.355487312925 3102
 660.000000 .035260491 690.000000 .020802007 720.000000 .0154297422925 3102
 760.000000 .015363512 800.000000 .330341068 840.000000 .0100129162925 3102
 880.000000 .009269458 920.000000 .009606887 960.000000 .8528706612925 3102
 1000.00000 .016553777 1050.00000 .008447028 1100.00000 .0082293162925 3102
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 1350.00000 .010605520 1425.00000 .012341874 1500.00000 .0156162212925 3102
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 1900.00000 .448941427 2000.00000 2.80238553 2100.00000 .2479769022925 3102
 2200.00000 .064002243 2300.00000 .033783096 2400.00000 .0370697492925 3102
 2550.00000 3.39830388 2700.00000 .058439117 2800.00000 .0129655402925 3102
 3000.00000 .005963652 3200.00000 .072973641 3400.00000 .0579893052925 3102
 3600.00000 .003689999 3800.00000 .013971609 4000.00000 .0071559712925 3102
 4250.00000 .925532977 4500.00000 .022111476 4750.00000 .8103107942925 3102
 5000.00000 .038217922 5250.00000 1.00817673 5500.00000 .0591121072925 3102
 5750.00000 .844108255 6000.00000 .009366183 6300.00000 .0052543062925 3102
 6600.00000 .124991723 6900.00000 .250886850 7200.00000 .2980666732925 3102
 7600.00000 .528928020 8000.00000 .381709799 8400.00000 .1918954152925 3102
 8800.00000 .113597176 9200.00000 .167841136 9600.00000 .4285984042925 3102
 10000.0000 .134282984 10500.0000 .187730609 11000.0000 .0108148172925 3102
 11500.0000 .003502296 12000.0000 .076775056 12750.0000 .2561983682925 3102
 13500.0000 .106843315 14250.0000 .064051633 15000.0000 .1442864042925 3102
 16000.0000 .149704487 17000.0000 .081769306 18000.0000 .0462818412925 3102
 19000.0000 .050830274 20000.0000 .104570720 21000.0000 .2042916832925 3102
 22000.0000 .085151443 23000.0000 .140274596 24000.0000 .0541256882925 3102
 25500.0000 .104722656 27000.0000 .049342813 28000.0000 .0393934112925 3102
 30000.0000 .065008048 32000.0000 .046721038 34000.0000 .0727883722925 3102
 36000.0000 .052703011 38000.0000 .053637289 40000.0000 .0781513932925 3102
 42500.0000 .043039656 45000.0000 .049125858 47500.0000 .0557586742925 3102
 50000.0000 .006468657 52500.0000 .014090851 55000.0000 .0412020222925 3102
 57500.0000 .054507662 60000.0000 .029851598 63000.0000 .0395868852925 3102
 66000.0000 .036396140 69000.0000 .031661174 72000.0000 .0418829762925 3102
 76000.0000 .029841708 80000.0000 .027799368 84000.0000 .0325954352925 3102
 88000.0000 .034971714 92000.0000 .034649996 96000.0000 .0290141752925 3102
 100000.000 .029880499 105000.000 .029681499 110000.000 .0294824992925 3102
 115000.000 .029283499 120000.000 .029034749 127500.000 .0287362492925 3102
 135000.000 .028437749 142500.000 .028139249 150000.000 .0277909992925 3102
 160000.000 .027392999 170000.000 .026995000 180000.000 .0265970002925 3102
 190000.000 .026198895 200000.000 .025694023 210000.000 .0250950702925 3102
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 9.034280-6-1.641020-5 1.661120-6 6.958950-5 1.707410-4 2.942290-4292533107
 1.430690-3 1.171940-3 7.971110-4 5.514220-4 4.345810-4 3.760900-4292533107
 3.055560-4 1.896340-4 8.544220-5 2.295550-5-3.706030-5-4.438810-5292533107
 4.268180-5 2.163080-4 4.676900-4 1.230320-3 1.067120-3 8.606320-4292533107
 6.657200-4 4.806270-4 3.023890-4 1.475950-4 6.104790-5 2.518420-5292533107
 3.893420-6 1.881610-5 7.051740-5 1.450370-4 2.353030-4 1.167910-3292533107
 1.098490-3 9.064420-4 6.368410-4 3.512730-4 1.303170-4 3.773650-5292533107
 1.762610-5 3.077580-5 8.134670-5 1.278940-4 1.424820-4 1.145190-4292533107
 1.153480-3 1.034720-3 7.810150-4 4.620260-4 1.832590-4 5.101270-5292533107
 1.413080-5 2.017000-5 8.668080-5 1.691200-4 2.283950-4 2.472330-4292533107
 1.019520-3 8.586540-4 5.832950-4 2.813990-4 1.006260-4 2.759680-5292533107
-6.493400-6 4.651990-5 1.727150-4 3.297290-4 4.943410-4 8.217660-4292533107
 6.464730-4 3.759330-4 1.718880-4 6.780650-5-1.364010-5 9.720900-7292533107
 1.407710-4 3.698800-4 6.635980-4 5.910990-4 4.124830-4 2.402880-4292533107
 1.360380-4 3.274150-5 2.902680-6 1.040010-4 3.151620-4 6.171420-4292533107
 3.612330-4 2.787220-4 2.169140-4 1.413740-4 9.211770-5 1.194900-4292533107
 2.202080-4 3.862950-4 2.811990-4 2.707110-4 2.463720-4 2.133570-4292533107
 1.939780-4 1.950960-4 2.180510-4 2.960220-4 3.144670-4 3.107020-4292533107
 2.817370-4 2.377330-4 1.849270-4 3.947940-4 4.559440-4 4.588000-4292533107
 4.126890-4 3.275050-4 6.282220-4 7.551020-4 8.366480-4 8.809310-4292533107
 1.092850-3 1.456080-3 1.843700-3 2.235280-3 3.158140-3 4.788580-3292533107
                                                                  292533  0
                                                                  2925 0  0
                                                                     0 0  0
 2.90650E+4 6.43700E+1          0          0         34         102931 1451
 0.0        0.0                 0          0          0          62931 1451
 1.00000E+0 1.50000E+8          0          0         10       20022931 1451
 3.00000E+2 0.0                 1          0        384          32931 1451
 29-Cu- 65  LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK      2931 1451
                      DIST-Feb2002                                2931 1451
----IRDF-2002         MATERIAL 2931                               2931 1451
-----INCIDENT NEUTRON DATA                                        2931 1451
------ENDF-6 FORMAT                                               2931 1451
******************************************************************2931 1451
  29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK      2931 1451
 Ch98,Ch99            DIST-SEP 1 REV4-                 20010926   2931 1451
----ENDF/B-VI         MATERIAL 2931         REVISION 4            2931 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2931 1451
DOSIMETRY                                                         2931 1451
******************************************************************2931 1451
                                                                  2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy,    2931 1451
                 P.G. Young (LANL)                                2931 1451
                                                                  2931 1451
 The secondary gamma-ray spectrum for radiative capture (MF 12,   2931 1451
 MT 102) has been updated for new experimental data at incident   2931 1451
 neutron energies up to 1 keV.                                    2931 1451
 The MF=12, MT=102 yields above 1 keV were adjusted slightly to   2931 1451
 force energy conservation.                                       2931 1451
 The Q-value for radiative capture was also updated in File 3.    2931 1451
 Details of these changes are described in Frankel et al. [Fr01]. 2931 1451
                                                                  2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN),    2931 1451
                 M.B. Chadwick, P.G. Young (LANL)                 2931 1451
                                                                  2931 1451
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     2931 1451
   in cooperation with ECN Petten.                                2931 1451
                                                                  2931 1451
   This evaluation provides a complete representation of the      2931 1451
 nuclear data needed for transport, damage, heating,              2931 1451
 radioactivity, and shielding applications over the incident      2931 1451
 neutron energy range from 1.0E-11 to 150 MeV.  The discussion    2931 1451
 here is divided into the region below and above 20 MeV.          2931 1451
                                                                  2931 1451
INCIDENT NEUTRON ENERGIES < 20 MeV                                2931 1451
                                                                  2931 1451
  Below 20 MeV the evaluation is based completely on the ENDF/B-  2931 1451
VI (Release 2) evaluation by D. Hetrick, C.Y. Fu, and D. Larson.  2931 1451
                                                                  2931 1451
INCIDENT NEUTRON ENERGIES > 20 MeV                                2931 1451
                                                                  2931 1451
  The ENDF/B-VI Release 2 evaluation extends to 20 MeV and        2931 1451
includes cross sections and energy-angle data for all             2931 1451
significant reactions.  The present evaluation utilizes a more    2931 1451
compact composite reaction spectrum representation above 20 MeV   2931 1451
in order to reduce the length of the file. No essential data for  2931 1451
applications is lost with this representation.                    2931 1451
  The evaluation above 20 MeV utilizes MF=6, MT=5 to represent    2931 1451
all reaction data.  Production cross sections and emission        2931 1451
spectra are given for neutrons, protons, deuterons, tritons,      2931 1451
alpha particles, gamma rays, and all residual nuclides produced   2931 1451
(A>5) in the reaction chains. To summarize, the ENDF sections     2931 1451
with non-zero data above En = 20 MeV are:                         2931 1451
                                                                  2931 1451
  MF=3 MT=  1  Total Cross Section                                2931 1451
       MT=  2  Elastic Scattering Cross Section                   2931 1451
       MT=  3  Nonelastic Cross Section                           2931 1451
       MT=  5  Sum of Binary (n,n') and (n,x) Reactions           2931 1451
                                                                  2931 1451
  MF=4 MT=  2  Elastic Angular Distributions                      2931 1451
                                                                  2931 1451
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         2931 1451
               Distributions for Emission Neutrons, Protons,      2931 1451
               Deuterons, Tritons, and Alphas; and Angle-         2931 1451
               Integrated Spectra for Gamma Rays and Residual     2931 1451
                                                                  2931 1451
               Nuclei That Are Stable Against Particle Emission   2931 1451
                                                                  2931 1451
  The evaluation is based on nuclear model calculations that      2931 1451
have been benchmarked to experimental data, especially for n +    2931 1451
Cu65 and p + Cu65 reactions (Ch98). We use the GNASH code system  2931 1451
(Yo92), which utilizes Hauser-Feshbach statistical, preequilib-   2931 1451
rium and direct-reaction theories. Spherical optical model        2931 1451
calculations are used to obtain particle transmission             2931 1451
coefficients for the Hauser-Feshbach calculations, as well as     2931 1451
for the elastic neutron angular distributions.                    2931 1451
  Cross sections and spectra for producing individual residual    2931 1451
nuclei are included for reactions. The energy-angle-correlations  2931 1451
for all outgoing particles are based on Kalbach systematics       2931 1451
(Ka88).                                                           2931 1451
  A model was developed to calculate the energy distributions of  2931 1451
all recoil nuclei in the GNASH calculations (Ch96a). The recoil   2931 1451
energy distributions are represented in the laboratory system in  2931 1451
MT=5, MF=6, and are given as isotropic in the lab system. All     2931 1451
other data in MT=5,MF=6 are given in the center-of-mass system.   2931 1451
This method of representation utilizes the LCT=3 option approved  2931 1451
at the November, 1996, CSEWG meeting.                             2931 1451
  Preequilibrium corrections were performed in the course of the  2931 1451
GNASH calculations using the exciton model of Kalbach (Ka77,      2931 1451
Ka85), validated by comparison with calculations using Feshbach,  2931 1451
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     2931 1451
nuclear data sheets were matched to continuum level densities     2931 1451
using the formulation of Ignatyuk et al. (Ig75) and pairing and   2931 1451
shell parameters from the Cook (Co67) analysis. Neutron and       2931 1451
charged- particle transmission coefficients were obtained from    2931 1451
the optical potentials, as discussed below.  Gamma-ray            2931 1451
transmission coefficients were calculated using the Kopecky-Uhl   2931 1451
model (Ko90).                                                     2931 1451
                                                                  2931 1451
SPECIFIC INFORMATION CONCERNING THE 65Cu EVALUATION               2931 1451
                                                                  2931 1451
This evaluation is documented in some detail in Ref. (Ko98b).     2931 1451
                                                                  2931 1451
The neutron total cross section above 20 MeV was obtained by      2931 1451
evaluating experimental data, with a particular emphasis on the   2931 1451
Finlay (Fi93) elemental data. This resulted in an evaluated       2931 1451
elemental Cu total cross section; to obtain an isotopic 65Cu total2931 1451
cross section, it was assumed that 63Cu and 65Cu have total cross 2931 1451
sections in an A**2/3 ratio to one another. The total neutron     2931 1451
nonelastic cross section was obtained directly from an optical    2931 1451
model calculation (see below), after verifying that it was in good2931 1451
agreement with the experimental data (Ko98b).                     2931 1451
                                                                  2931 1451
To obtain the neutron optical potential we used total cross       2931 1451
section data from 1.2 to 4.5 MeV (Gu86) and from 5.3 to 600 MeV   2931 1451
(Fi93), and elastic scattering angular distribution data from 1.6 2931 1451
to 96 MeV (Br50, Sa60, Ki74, El82, Gu86). The optical potential   2931 1451
parameters were obtained using a combination of a grid search code2931 1451
and the interactive optical model viewer ECISVIEW [Ko97], both    2931 1451
built around the coupled channels code ECIS96 [Ra94]. The energy  2931 1451
dependence of the optical model parameters is as described in     2931 1451
[Ko98]. This optical potential was used for the calculation, with 2931 1451
ECIS96, of neutron transmission coefficients and DWBA cross       2931 1451
sections for the entire energy region above 20 MeV.               2931 1451
                                                                  2931 1451
Due to the lack of proton elastic scattering data in numerical    2931 1451
form, we used a combination of global optical models for the      2931 1451
proton channel. The Becchetti-Greenlees potential [Be69]was       2931 1451
adopted below 47 MeV, and the non-relativistic version of the     2931 1451
Madland potential [Ma88] above 47 MeV. At this particular energy  2931 1451
point the two potentials join smoothly.                           2931 1451
                                                                  2931 1451
For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2931 1451
for alpha particles the Moyen potential (MacFadden-Satchler       2931 1451
[Ma66]) was used; and for tritons the Becchetti-Greenlees         2931 1451
potential [Be71] was used. The He-3 channel was ignored, due to   2931 1451
its small importance.                                             2931 1451
                                                                  2931 1451
Following Delaroche et al. [De82], we adopted the weak-coupling   2931 1451
model for direct collective inelastic scattering for Cu-65, using 2931 1451
Ni-64 as a basis. For the calculation of the cross sections,      2931 1451
ECIS96 was used in DWBA mode. We used the following direct        2931 1451
transitions for Cu-65  (ground state 3/2- ) :                     2931 1451
                                                                  2931 1451
Jpi Ex(MeV) Deformation length (Delta) or parameter (Beta)        2931 1451
 0.5- 0.771  Beta(2)=0.0566                                       2931 1451
 2.5- 1.116  Beta(2)=0.0980                                       2931 1451
 3.5- 1.481  Beta(2)=0.1132                                       2931 1451
 1.5- 1.743  Beta(2)=0.0800                                       2931 1451
 1.5- 3.185  Delta(3)=0.3167                                      2931 1451
 2.5- 3.435  Delta(3)=0.3879                                      2931 1451
 3.5- 3.685  Delta(3)=0.4479                                      2931 1451
 4.5- 3.935  Delta(3)=0.5008                                      2931 1451
                                                                  2931 1451
Only one measurement exists for neutron-induced emission spectra  2931 1451
above 20 MeV for 65Cu: the 25.7 MeV (n,xn) data by Marcinkowski et2931 1451
al (Ma83). Without adjusting any of the level density or pre-     2931 1451
equilibrium parameters the GNASH calculation was in good agreement2931 1451
with these data (Ko98b). Hence we adopted these parameters for the2931 1451
whole energy region.                                              2931 1451
                                                                  2931 1451
****************************************************************  2931 1451
                                                                  2931 1451
REFERENCES                                                        2931 1451
                                                                  2931 1451
[Ab93]  W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra,   2931 1451
   Phys.Rev. C47, 1033 (1993)                                     2931 1451
[Al83]  R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986)   2931 1451
[Ar80]  E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross  2931 1451
   Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross   2931 1451
   Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2931 1451
   Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2931 1451
   1980] p. 731.                                                  2931 1451
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182,   2931 1451
   1190 (1969)                                                    2931 1451
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization  2931 1451
   Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W.    2931 1451
   Haeberli, The University of Wisconsin Press, 1971) p.682.      2931 1451
[Be92]  O. Bersillon, "SCAT2 - A Spherical Optical Model Code,"   2931 1451
   in Proc. ICTP Workshop on Computation and Analysis of Nuclear  2931 1451
   data Relevant to Nuclear Energy and Safety, February-March,    2931 1451
   1999 Trieste, Italy, to be published in World Scientific Press,2931 1451
   and Progress Report of the Nuclear Physics Division, Bruyeres- 2931 1451
   le-Chatel 1977, CEA-N-2037 (1978) p.111                        2931 1451
[Br50]  S. Bratenahl, S. Fernbach, R.H. Hildebrand et al.,        2931 1451
   Phys.Rev. 77, 597 (1950)                                       2931 1451
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2931 1451
[Ch96]  M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      2931 1451
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   2931 1451
   Driven Technologies:  Calculational Method for Heavy Recoils," 2931 1451
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    2931 1451
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996     2931 1451
[Ch98]  M.B. Chadwick and P.G. Young, "GNASH Calculations of      2931 1451
   n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2931 1451
   REPORT: 1 February - 1 March 1998, internal Los Alamos National2931 1451
   Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2931 1451
[Ch99]  M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2931 1451
   National Laboratory report, LA-UR-99-1222 (1999)               2931 1451
[Co67]  J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J.    2931 1451
   Phys. 20, 477 (1967)                                           2931 1451
[De82]  J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and   2931 1451
   R.L. Walter, Nucl. Phys. A390, 541 (1982).                     2931 1451
[El82]  S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2931 1451
   A390, 509 (1982)                                               2931 1451
[Fi93]  R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 2931 1451
   C 47, 237 (1993)                                               2931 1451
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      2931 1451
   National Laboratory Report, LA-13812 (2001).                   2931 1451
[Gu86]  P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl.   2931 1451
   Phys. A448, 280 (1986)                                         2931 1451
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    2931 1451
   Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485     2931 1451
   (1975)                                                         2931 1451
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     2931 1451
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         2931 1451
   LA-10248-MS (1985)                                             2931 1451
[Ka88]  C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also          2931 1451
   C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981)           2931 1451
[Ki74]  W.E. Kinney, F.G. Perey, report ORNL-4908 (1974)          2931 1451
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990)         2931 1451
[Ko97]  A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2931 1451
   A Graphical Interface for ECIS95", Proceedings of the NEA      2931 1451
   Specialists' Meeting on the Nucleon Nucleus Optical Model up to2931 1451
   200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2931 1451
   http://db.nea.fr/html/science/om200/.                          2931 1451
[Ko98]  A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear   2931 1451
  Data for Accelerator-Driven Systems: Nuclear models, Experiments2931 1451
   and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2931 1451
[Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI    2931 1451
   neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2931 1451
   Los Alamos National Laboratory report LAUR- (1998); ECN lab and2931 1451
   JEFF report (1998).                                            2931 1451
[Lo74]  J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974)    2931 1451
[Ma66]  Macfadden and Satchler, Nuc.Phys. 84, 177 (1966)          2931 1451
[Ma83]  A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al.,  2931 1451
   Nucl.Phys. A402, 220 (1983)                                    2931 1451
[Ma88]   D.G. Madland, "Recent Results in the Development of a    2931 1451
   Global Medium-Energy Nucleon-Nucleus Optical-Model Potential,  2931 1451
   "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2931 1451
   Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2931 1451
   (1988).                                                        2931 1451
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      2931 1451
[Ra94]  J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772  2931 1451
   (1994)                                                         2931 1451
[Sa60]  G.L. Salmon, Nucl.Phys. 21, 15 (1960)                     2931 1451
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, report        2931 1451
   LA-12343-MS (1992)                                             2931 1451
                                                                  2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 ENDF/B-VI MOD 3 Revision, March 1991, ORNL                       2931 1451
                                                                  2931 1451
 MOD 3 changes                                                    2931 1451
   1) Corrections to MF=6, MT=63 at 17.5 MeV to prevent negative  2931 1451
      values in the angular distribution.                         2931 1451
                                                                  2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 * Note there was no MOD 2 released.                              2931 1451
                                                                  2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 ENDF/B-VI MOD 1 Evaluation, November 1989, D. Hetrick, F.Y. Fu,  2931 1451
           D. Larson (ORNL)                                       2931 1451
                                                                  2931 1451
    This work employed several nuclear model codes including the  2931 1451
 optical-model code GENOA [1], the Distorted Wave Born            2931 1451
 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach  2931 1451
 code TNG [3,4]. The TNG code provides energy and angular         2931 1451
 distributions of particles emitted in the compound and pre-      2931 1451
 compound reactions, ensures consistency among all reactions, and 2931 1451
 maintains energy balance. Details pertinent to the contents of   2931 1451
 this evaluation and extensive comparisons of calculations with   2931 1451
 experimental data can be found in reference [5].                 2931 1451
                                                                  2931 1451
----- DESCRIPTION OF FILES                                        2931 1451
(MF-MT)                                                           2931 1451
  1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS.         2931 1451
  2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB [6]. POINT2931 1451
        WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER  2931 1451
        FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2931 1451
        99.5 KEV.  REICH-MOORE PARAMETERS ARE GIVEN.  AGREEMENT   2931 1451
        WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2931 1451
        FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN       2931 1451
        AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND     2931 1451
        COMPARED WITH THE BINNED DATA.  THIS IS PROBABLY DUE TO   2931 1451
        TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES    2931 1451
        SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION.             2931 1451
        NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2931 1451
        OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE       2931 1451
        PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON    2931 1451
        A FINER ENERGY GRID THAN GIVEN IN 4/2.                    2931 1451
  3-16  (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS [12]      2931 1451
        CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL-     2931 1451
        TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS          2931 1451
        SECTIONS [13].                                            2931 1451
 33-16  (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS [12]         2931 1451
        CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL-     2931 1451
        TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS          2931 1451
        SECTIONS [13].                                            2931 1451
 **************************************************************** 2931 1451
                                                                  2931 1451
 REFERENCES:                                                      2931 1451
                                                                  2931 1451
 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967)    2931 1451
 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of           2931 1451
     Colorado, unpublished (1972)                                 2931 1451
 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu,           2931 1451
     Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2931 1451
     May 1980, Brookhaven National Lab. report BNL-NCS-51245      2931 1451
     (1980) p.675                                                 2931 1451
 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986)          2931 1451
 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report     2931 1451
     ORNL/TM-9083 [ENDF-337] (1984)                               2931 1451
 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron    2931 1451
     Cross Sections, Vol. 1, Neutron Resonance Parameters and     2931 1451
     Thermal Cross Sections, Part A, Z=1-60," (Academic Press,    2931 1451
     1981)                                                        2931 1451
 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600  2931 1451
     (1977), and private communication.                           2931 1451
 [8] F.G. Perey, private communication (1977)                     2931 1451
 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50   2931 1451
     MeV, Upton, NY, May 1980, Brookhaven National Lab. report    2931 1451
     BNL-NCS-51245 (1980) p.277                                   2931 1451
[10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl.   2931 1451
     Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and    2931 1451
     W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80      2931 1451
     (1983) [CSISRS data set 12853]                               2931 1451
[11] A.I. Dyumin, D.M. Kaminker, G.N. Popova, and V.A. Smolin,    2931 1451
     Izv.Akad.Nauk SSSR, Ser.Fiz. 36, 852 (1972)                  2931 1451
[12] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341      2931 1451
     [ENDF-303] (1980)                                            2931 1451
[13] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp.    2931 1451
     on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26,   2931 1451
     1982 (U.S. National Bureau of Standards) p.877               2931 1451
[14] CSISRS Library, National Nuclear Data Center, Brookhaven     2931 1451
     National Laboratory, Upton, N.Y. 11973.                      2931 1451
[15] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross        2931 1451
     Sections, Vol. 2, Neutron Cross Section Curves" (Academic    2931 1451
     Press, 1988)                                                 2931 1451
[16] S.M. Qaim and G. Stoecklin, Nucl.Phys. A257, 233 (1976)      2931 1451
[17] S.M. Qaim, Radiochimica Acta, 25, 13 (1978)                  2931 1451
[18] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys.     2931 1451
     A404, 250 (1983)                                             2931 1451
 **************************************************************** 2931 1451
  The Q values and threshold energies were updated prior to pro-  2931 1451
  cessing through the codes to comply with the values obtained    2931 1451
  using the NNDC calculation program which is based on the 1995   2931 1451
  Update to the Atomic mass Evaluation.                           2931 1451
 ************************ C O N T E N T S *********************** 2931 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************2931 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     2931 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     2931 1451
 ***************** Program RECENT (VERSION 2002-1) ***************2931 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2931 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     2931 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************2931 1451
 Data Doppler Broadened to 300.000000 Kelvin                      2931 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     2931 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     2931 1451
 ***************** Program FIXUP (Version 2002-1) ****************2931 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 2931 1451
 Corrected Thresholds-----------------------------------------Yes 2931 1451
 Extended Cross Sections to 20 MeV----------------------------No  2931 1451
 Allow Cross Section Deletion---------------------------------No  2931 1451
 Allow Cross Section Reconstruction---------------------------No  2931 1451
 Make All Cross Sections Non-Negative-------------------------Yes 2931 1451
 Delete Energies Not in Ascending Order-----------------------Yes 2931 1451
 Deleted Duplicate Points-------------------------------------Yes 2931 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 2931 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 2931 1451
 Allow Creation of Missing Sections---------------------------No  2931 1451
 Allow Insertion of Energy Points-----------------------------No  2931 1451
 Create Uniform Energy Grid-----------------------------------No  2931 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 2931 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************2931 1451
 Unshielded Group Averages Using  640 Groups                      2931 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     2931 1451
                                1        451        391          52931 1451
                                3         16         37          12931 1451
                               33         16         62          12931 1451
                                                                  2931 1  0
                                                                  2931 0  0
 2.90650E+4 6.43700E+1          0          0          0          02931 3 16
-9.91020E+6-9.91020E+6          0          0          1        1012931 3 16
        101          1                                            2931 3 16
 10000000.0 .000366266 10100000.0 .004894626 10200000.0 .0105964472931 3 16
 10300000.0 .016298268 10400000.0 .022000089 10500000.0 .0367329002931 3 16
 10600000.0 .060496700 10700000.0 .084260500 10800000.0 .1080243002931 3 16
 10900000.0 .131788100 11000000.0 .159085000 11100000.0 .1899150002931 3 16
 11200000.0 .220745000 11300000.0 .251575000 11400000.0 .2824050002931 3 16
 11500000.0 .314504000 11600000.0 .347872000 11700000.0 .3812400002931 3 16
 11800000.0 .414608000 11900000.0 .447976000 12000000.0 .4801010002931 3 16
 12100000.0 .510983000 12200000.0 .541865000 12300000.0 .5727470002931 3 16
 12400000.0 .603629000 12500000.0 .630458000 12600000.0 .6532340002931 3 16
 12700000.0 .676010000 12800000.0 .698786000 12900000.0 .7215620002931 3 16
 13000000.0 .741987000 13100000.0 .760061000 13200000.0 .7781350002931 3 16
 13300000.0 .796209000 13400000.0 .814283000 13500000.0 .8305190002931 3 16
 13600000.0 .844917000 13700000.0 .859315000 13800000.0 .8737130002931 3 16
 13900000.0 .888111000 14000000.0 .900663000 14100000.0 .9113690002931 3 16
 14200000.0 .922075000 14300000.0 .932781000 14400000.0 .9434870002931 3 16
 14500000.0 .952211000 14600000.0 .958953000 14700000.0 .9656950002931 3 16
 14800000.0 .972437000 14900000.0 .979179000 15000000.0 .9856050002931 3 16
 15100000.0 .991715000 15200000.0 .997825000 15300000.0 1.003935002931 3 16
 15400000.0 1.01004500 15500000.0 1.01484000 15600000.0 1.018320002931 3 16
 15700000.0 1.02180000 15800000.0 1.02528000 15900000.0 1.028760002931 3 16
 16000000.0 1.03373000 16100000.0 1.04019000 16200000.0 1.046650002931 3 16
 16300000.0 1.05311000 16400000.0 1.05957000 16500000.0 1.063630002931 3 16
 16600000.0 1.06529000 16700000.0 1.06695000 16800000.0 1.068610002931 3 16
 16900000.0 1.07027000 17000000.0 1.07157000 17100000.0 1.072510002931 3 16
 17200000.0 1.07345000 17300000.0 1.07439000 17400000.0 1.075330002931 3 16
 17500000.0 1.07734000 17600000.0 1.08042000 17700000.0 1.083500002931 3 16
 17800000.0 1.08658000 17900000.0 1.08966000 18000000.0 1.089750002931 3 16
 18100000.0 1.08685000 18200000.0 1.08395000 18300000.0 1.081050002931 3 16
 18400000.0 1.07815000 18500000.0 1.07277167 18600000.0 1.064915002931 3 16
 18700000.0 1.05705833 18800000.0 1.04920167 18900000.0 1.041345002931 3 16
 19000000.0 1.03348833 19100000.0 1.02563167 19200000.0 1.017775002931 3 16
 19300000.0 1.00991833 19400000.0 1.00206167 19500000.0 .9942050002931 3 16
 19600000.0 .986348333 19700000.0 .978491667 19800000.0 .9706350002931 3 16
 19900000.0 .962778333 20000000.0 0.0                             2931 3 16
                                                                  2931 3  0
                                                                  2931 0  0
 2.90650E+4 6.43700E+1          0          0          0          1293133 16
 0.0000E+00 0.0000E+00          0         16          0          2293133 16
 0.0000E+00 0.0000E+00          1          5        300         24293133 16
 1.0000E-05 1.0000E+07 1.0500E+07 1.1000E+07 1.1500E+07 1.2000E+07293133 16
 1.2500E+07 1.3000E+07 1.3500E+07 1.4000E+07 1.4500E+07 1.5000E+07293133 16
 1.5500E+07 1.6000E+07 1.6500E+07 1.7000E+07 1.7500E+07 1.8000E+07293133 16
 1.8500E+07 1.9000E+07 1.9200E+07 1.9400E+07 1.9600E+07 2.0000E+07293133 16
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16
 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 9.2849E-02293133 16
-3.9100E-03 9.8124E-04 3.2315E-04 2.5973E-04 1.6943E-04 1.0837E-04293133 16
 5.2219E-05 4.5076E-05 3.6396E-05 1.4418E-05 2.1146E-05 2.1805E-05293133 16
 2.1931E-05 2.1318E-05 2.1522E-05 2.2348E-05 2.9580E-05 2.2389E-05293133 16
 2.5489E-05 2.5389E-05 2.8747E-05 1.1582E-03 9.4407E-05 2.2976E-04293133 16
 1.6573E-04 1.1914E-04 8.3873E-05 4.3373E-05 3.6631E-05 3.1684E-05293133 16
 1.1438E-05 1.8007E-05 1.8412E-05 1.8521E-05 1.7912E-05 1.8306E-05293133 16
 1.8801E-05 2.5446E-05 1.8989E-05 2.1853E-05 2.1701E-05 2.4893E-05293133 16
 1.0426E-03 4.1291E-04 3.5993E-04 2.4206E-04 1.6476E-04 8.2857E-05293133 16
 7.0722E-05 6.0039E-05 2.2260E-05 3.4439E-05 3.5246E-05 3.5284E-05293133 16
 3.4112E-05 3.4653E-05 3.5725E-05 4.7927E-05 3.4885E-05 4.0981E-05293133 16
 4.0699E-05 4.6995E-05 1.6804E-03 5.1367E-04 3.7077E-04 2.3022E-04293133 16
 1.0876E-04 9.5181E-05 7.5902E-05 3.0612E-05 4.4708E-05 4.6020E-05293133 16
 4.5981E-05 4.4753E-05 4.5221E-05 4.6593E-05 6.1095E-05 4.5008E-05293133 16
 5.2412E-05 5.2045E-05 5.9698E-05 1.5662E-03 4.6015E-04 3.0034E-04293133 16
 1.3181E-04 1.1846E-04 8.7882E-05 3.8955E-05 5.3423E-05 5.5298E-05293133 16
 5.5174E-05 5.4093E-05 5.4391E-05 5.6078E-05 7.1784E-05 5.4017E-05293133 16
 6.1977E-05 6.1527E-05 6.9508E-05 1.6388E-03 3.8739E-04 1.8183E-04293133 16
 1.6833E-04 1.0963E-04 5.4807E-05 6.8444E-05 7.1184E-05 7.0989E-05293133 16
 6.9696E-05 7.0329E-05 7.2399E-05 9.1321E-05 7.1362E-05 7.9338E-05293133 16
 7.8939E-05 8.7688E-05 1.6381E-03 2.7820E-04 2.9569E-04 1.5183E-04293133 16
 8.4051E-05 9.2687E-05 9.9014E-05 9.7400E-05 9.5524E-05 9.6112E-05293133 16
 9.8898E-05 1.2500E-04 9.5565E-05 1.0835E-04 1.0739E-04 1.2076E-04293133 16
 7.2642E-04 1.3861E-04 2.1090E-04 1.3154E-04 1.2590E-04 1.3899E-04293133 16
 1.3473E-04 1.3248E-04 1.3243E-04 1.3654E-04 1.7246E-04 1.2855E-04293133 16
 1.4860E-04 1.4753E-04 1.6696E-04 7.4020E-04 2.2844E-04 1.3272E-04293133 16
 1.3480E-04 1.4740E-04 1.4311E-04 1.4019E-04 1.4023E-04 1.4475E-04293133 16
 1.8498E-04 1.3536E-04 1.5870E-04 1.5766E-04 1.7994E-04 2.5205E-04293133 16
 6.7831E-05 1.5774E-04 1.5403E-04 1.5480E-04 1.4692E-04 1.4845E-04293133 16
 1.5385E-04 2.1458E-04 1.4114E-04 1.7986E-04 1.7866E-04 2.1467E-04293133 16
 4.9848E-04 7.7213E-05 1.8584E-04 1.5113E-04 1.7339E-04 1.6481E-04293133 16
 1.6703E-04 1.1799E-04 1.6769E-04 1.2330E-04 1.2252E-04 8.4028E-05293133 16
 3.4942E-04 1.7432E-04 1.9975E-04 2.0482E-04 1.8271E-04 1.7572E-04293133 16
 1.6743E-04 1.7022E-04 1.5696E-04 1.5582E-04 1.4329E-04 5.2876E-04293133 16
 1.9487E-04 2.1525E-04 1.9401E-04 1.9702E-04 1.6642E-04 1.6184E-04293133 16
 1.5555E-04 1.5458E-04 1.4089E-04 2.9887E-04 1.8750E-04 2.0981E-04293133 16
 1.8656E-04 1.7332E-04 1.7942E-04 1.6365E-04 1.6255E-04 1.4448E-04293133 16
 3.3467E-04 2.2613E-04 2.0243E-04 1.7545E-04 1.8875E-04 1.6806E-04293133 16
 1.6691E-04 1.4225E-04 1.1229E-03-7.4862E-05 3.1309E-04 2.2013E-04293133 16
 2.7176E-04 2.6985E-04 1.8472E-04 4.9986E-04 1.6697E-04 2.0915E-04293133 16
 1.7615E-04 1.7495E-04 1.4909E-04 1.3873E-03 1.1714E-04 5.8852E-04293133 16
 5.8437E-04 3.4575E-04 4.7121E-04 2.5339E-04 2.5176E-04 1.7365E-04293133 16
 3.0426E-03 1.1867E-03 9.8857E-04 3.0015E-03 9.8197E-04 2.7731E-03293133 16
 0.0000E+00 0.0000E+00          0          8         48         24293133 16
 1.0000E-05 0.0000E+00 1.0000E+07 1.8167E-06 1.0500E+07 9.2309E-07293133 16
 1.1000E+07 5.7456E-06 1.1500E+07 2.8213E-05 1.2000E+07 5.2413E-05293133 16
 1.2500E+07 8.4423E-05 1.3000E+07 1.1260E-04 1.3500E+07 6.0382E-05293133 16
 1.4000E+07 7.0182E-05 1.4500E+07 2.6190E-05 1.5000E+07 5.5253E-05293133 16
 1.5500E+07 4.0563E-05 1.6000E+07 6.4379E-05 1.6500E+07 3.7846E-05293133 16
 1.7000E+07 4.2917E-05 1.7500E+07 1.4672E-04 1.8000E+07 6.5269E-05293133 16
 1.8500E+07 1.7207E-04 1.9000E+07 5.8043E-05 1.9200E+07 3.7985E-04293133 16
 1.9400E+07 3.6370E-04 1.9600E+07 3.0778E-04 2.0000E+07 2.6557E-04293133 16
                                                                  293133  0
                                                                  2931 0  0
                                                                     0 0  0
 3.00640E+4 6.33800E+1          0          0         34         103025 1451
 0.0        0.0                 0          0          0          63025 1451
 1.00000E+0 2.00000E+7          0          0         10       20023025 1451
 3.00000E+2 0.0                 1          0         42          33025 1451
 30-Zn- 64 IRK-VIENNA EVAL-APR90                                  3025 1451
                      DIST-Feb2004                                3025 1451
----IRDF-2002         MATERIAL 3025                               3025 1451
-----INCIDENT NEUTRON DATA                                        3025 1451
------ENDF-6 FORMAT                                               3025 1451
 *****************************************************************3025 1451
 30-ZN- 64 IRK-VIENNA EVAL-APR90                                  3025 1451
                      DIST-JUN90                                  3025 1451
    IRK-EVAL.NLIB 25  3025                                        3025 1451
 *****************************************************************3025 1451
  The Q value was updated prior to processing through the pre-    3025 1451
  processing codes to comply with the values obtained using the   3025 1451
  NNDC calculation program which is based on the 1995 Update to   3025 1451
  the Atomic mass Evaluation.                                     3025 1451
  Threshold value of 0.0 inserted at 1.0e-5 ev for MF/MT=3/103    3025 1451
  after processing using the pre-processing codes.                3025 1451
 *****************************************************************3025 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************3025 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     3025 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     3025 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************3025 1451
 Data Doppler Broadened to 300.000000 Kelvin                      3025 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     3025 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     3025 1451
 ***************** Program FIXUP (Version 2002-1) ****************3025 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 3025 1451
 Corrected Thresholds-----------------------------------------Yes 3025 1451
 Extended Cross Sections to 20 MeV----------------------------No  3025 1451
 Allow Cross Section Deletion---------------------------------No  3025 1451
 Allow Cross Section Reconstruction---------------------------No  3025 1451
 Make All Cross Sections Non-Negative-------------------------Yes 3025 1451
 Delete Energies Not in Ascending Order-----------------------Yes 3025 1451
 Deleted Duplicate Points-------------------------------------Yes 3025 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 3025 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 3025 1451
 Allow Creation of Missing Sections---------------------------No  3025 1451
 Allow Insertion of Energy Points-----------------------------No  3025 1451
 Create Uniform Energy Grid-----------------------------------No  3025 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 3025 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************3025 1451
 Unshielded Group Averages Using  640 Groups                      3025 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     3025 1451
                                1        451         49          13025 1451
                                3        103         72          13025 1451
                               33        103        103          13025 1451
                                                                  3025 1  0
                                                                  3025 0  0
 3.00640E+4 6.33800E+1          0          0          0          03025 3103
 2.03600E+5 2.03600E+5          0          0          1        2063025 3103
        206          1                                            3025 3103
 500000.000 2.82258E-6 525000.000 8.46774E-6 550000.000 1.41129E-53025 3103
 575000.000 1.97581E-5 600000.000 2.59677E-5 630000.000 3.27419E-53025 3103
 660000.000 3.95161E-5 690000.000 4.62903E-5 720000.000 5.41935E-53025 3103
 760000.000 6.32258E-5 800000.000 7.22581E-5 840000.000 8.12903E-53025 3103
 880000.000 9.03226E-5 920000.000 9.93548E-5 960000.000 .0001083873025 3103
 1000000.00 .000124194 1100000.00 .000182215 1200000.00 .0003133333025 3103
 1300000.00 .000456625 1400000.00 .000819000 1500000.00 .0012710003025 3103
 1600000.00 .001886125 1700000.00 .002900000 1800000.00 .0039791253025 3103
 1900000.00 .006095000 2000000.00 .008635000 2100000.00 .0116475003025 3103
 2200000.00 .015815000 2300000.00 .020140000 2400000.00 .0267750003025 3103
 2500000.00 .034355000 2600000.00 .043101625 2700000.00 .0547000003025 3103
 2800000.00 .066454875 2900000.00 .078801000 3000000.00 .0913890003025 3103
 3100000.00 .101665125 3200000.00 .106290000 3300000.00 .1107457503025 3103
 3400000.00 .117132000 3500000.00 .124308000 3600000.00 .1297177503025 3103
 3700000.00 .130810000 3800000.00 .131846375 3900000.00 .1359710003025 3103
 4000000.00 .141359000 4100000.00 .146190125 4200000.00 .1496600003025 3103
 4300000.00 .153109375 4400000.00 .157469000 4500000.00 .1622010003025 3103
 4600000.00 .166793500 4700000.00 .171045000 4800000.00 .1752502503025 3103
 4900000.00 .178779000 5000000.00 .182031000 5100000.00 .1846305003025 3103
 5200000.00 .185635000 5300000.00 .186509125 5400000.00 .1861100003025 3103
 5500000.00 .185190000 5600000.00 .184270000 5700000.00 .1837981673025 3103
 5800000.00 .186015333 5900000.00 .188680667 6000000.00 .1913460003025 3103
 6100000.00 .194011333 6200000.00 .196676667 6300000.00 .1993420003025 3103
 6400000.00 .202007333 6500000.00 .203560000 6600000.00 .2040000003025 3103
 6700000.00 .204440000 6800000.00 .204880000 6900000.00 .2053200003025 3103
 7000000.00 .205760000 7100000.00 .206200000 7200000.00 .2066400003025 3103
 7300000.00 .207080000 7400000.00 .207520000 7500000.00 .2085860003025 3103
 7600000.00 .210278000 7700000.00 .211970000 7800000.00 .2136620003025 3103
 7900000.00 .215354000 8000000.00 .217046000 8100000.00 .2187380003025 3103
 8200000.00 .220430000 8300000.00 .222122000 8400000.00 .2238140003025 3103
 8500000.00 .225047333 8600000.00 .225822000 8700000.00 .2265966673025 3103
 8800000.00 .227371333 8900000.00 .228146000 9000000.00 .2289206673025 3103
 9100000.00 .229695333 9200000.00 .230470000 9300000.00 .2312446673025 3103
 9400000.00 .232019333 9500000.00 .232794000 9600000.00 .2335686673025 3103
 9700000.00 .234343333 9800000.00 .235118000 9900000.00 .2358926673025 3103
 10000000.0 .237326750 10100000.0 .239420250 10200000.0 .2415137503025 3103
 10300000.0 .243607250 10400000.0 .245700750 10500000.0 .2477942503025 3103
 10600000.0 .249887750 10700000.0 .251981250 10800000.0 .2540747503025 3103
 10900000.0 .256168250 11000000.0 .258261750 11100000.0 .2603552503025 3103
 11200000.0 .262448750 11300000.0 .264542250 11400000.0 .2666357503025 3103
 11500000.0 .268729250 11600000.0 .270822750 11700000.0 .2729162503025 3103
 11800000.0 .275009750 11900000.0 .277103250 12000000.0 .2756823333025 3103
 12100000.0 .270747000 12200000.0 .265811667 12300000.0 .2608763333025 3103
 12400000.0 .255941000 12500000.0 .251005667 12600000.0 .2460703333025 3103
 12700000.0 .241135000 12800000.0 .236199667 12900000.0 .2312643333025 3103
 13000000.0 .226329000 13100000.0 .221393667 13200000.0 .2164583333025 3103
 13300000.0 .211523000 13400000.0 .206587667 13500000.0 .2014446673025 3103
 13600000.0 .196094000 13700000.0 .190743333 13800000.0 .1853926673025 3103
 13900000.0 .180042000 14000000.0 .174691333 14100000.0 .1693406673025 3103
 14200000.0 .164595167 14300000.0 .163480667 14400000.0 .1629713333025 3103
 14500000.0 .162462000 14600000.0 .161564917 14700000.0 .1597200003025 3103
 14800000.0 .157882450 14900000.0 .157154800 15000000.0 .1568812003025 3103
 15100000.0 .156607600 15200000.0 .156334000 15300000.0 .1560604003025 3103
 15400000.0 .155786800 15500000.0 .153846000 15600000.0 .1502380003025 3103
 15700000.0 .146630000 15800000.0 .143022000 15900000.0 .1394140003025 3103
 16000000.0 .135806000 16100000.0 .132198000 16200000.0 .1285900003025 3103
 16300000.0 .124982000 16400000.0 .121374000 16500000.0 .1190220003025 3103
 16600000.0 .117926000 16700000.0 .116830000 16800000.0 .1157340003025 3103
 16900000.0 .114638000 17000000.0 .113542000 17100000.0 .1124460003025 3103
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                                                                  302533  0
                                                                  3025 0  0
                                                                     0 0  0
 3.10000+ 4 0.0000 + 0         -1          0         34          13100 1451
 0.0    + 0 0.0    + 0          0          0          0          63100 1451
 1.0    + 0 0.0    + 0          0          0         10          23100 1451
 0.0    + 0 0.0    + 0          0          0          8          23100 1451
 31-GA-  0 ASTM       EVAL-   91 P.J.GRIFFIN                      3100 1451
 PRIVATE COM.         DIST-JAN03                       811211     3100 1451
----IRDF-02/NMF-G     MATERIAL 3100                               3100 1451
-----INCIDENT NEUTRON DATA                                        3100 1451
------ENDF-6 FORMAT                                               3100 1451
              DAMAGE CROSS SECTION OF GaAs (ELECTRONICS DAMAGE)   3100 1451
              ASTM E722           MT=900                          3100 1451
                                                                  3100 1451
                                1        451         14          03100 1451
                                3        900        217          03100 1451
                                                                  3100 1  0
                                                                  3100 0  0
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        641          1                                            3100 3900
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 1.9000E-04 1.3836E+02 2.0000E-04 1.3845E+02 2.1000E-04 1.3854E+023100 3900
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 2.5500E-04 1.3905E+02 2.7000E-04 1.3919E+02 2.8000E-04 1.3901E+023100 3900
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 1.0000E+01 3.4059E+01 1.0500E+01 3.2995E+01 1.1000E+01 3.1874E+013100 3900
 1.1500E+01 3.0694E+01 1.2000E+01 2.9514E+01 1.2750E+01 2.8316E+013100 3900
 1.3500E+01 2.7264E+01 1.4250E+01 2.6413E+01 1.5000E+01 2.5561E+013100 3900
 1.6000E+01 2.4701E+01 1.7000E+01 2.3855E+01 1.8000E+01 2.3024E+013100 3900
 1.9000E+01 2.2192E+01 2.0000E+01 2.1352E+01 2.1000E+01 2.0509E+013100 3900
 2.2000E+01 1.9659E+01 2.3000E+01 1.8800E+01 2.4000E+01 1.8044E+013100 3900
 2.5500E+01 1.7385E+01 2.7000E+01 1.6723E+01 2.8000E+01 1.6061E+013100 3900
 3.0000E+01 1.5445E+01 3.2000E+01 1.4875E+01 3.4000E+01 1.4305E+013100 3900
 3.6000E+01 1.3734E+01 3.8000E+01 1.3165E+01 4.0000E+01 1.2595E+013100 3900
 4.2500E+01 1.2024E+01 4.5000E+01 1.1449E+01 4.7500E+01 1.0988E+013100 3900
 5.0000E+01 1.0638E+01 5.2500E+01 1.0287E+01 5.5000E+01 9.9305E+003100 3900
 5.7500E+01 9.5438E+00 6.0000E+01 9.1394E+00 6.3000E+01 8.8213E+003100 3900
 6.6000E+01 8.5463E+00 6.9000E+01 8.2122E+00 7.2000E+01 7.9198E+003100 3900
 7.6000E+01 7.6836E+00 8.0000E+01 7.4465E+00 8.4000E+01 7.1973E+003100 3900
 8.8000E+01 6.9399E+00 9.2000E+01 6.6811E+00 9.6000E+01 6.4211E+003100 3900
 1.0000E+02 6.1490E+00 1.0500E+02 5.8714E+00 1.1000E+02 5.6203E+003100 3900
 1.1500E+02 5.3994E+00 1.2000E+02 5.1737E+00 1.2750E+02 4.9434E+003100 3900
 1.3500E+02 4.7500E+00 1.4250E+02 4.5899E+00 1.5000E+02 4.4266E+003100 3900
 1.6000E+02 4.2632E+00 1.7000E+02 4.3032E+00 1.8000E+02 4.4669E+003100 3900
 1.9000E+02 4.7471E+00 2.0000E+02 3.8537E+00 2.1000E+02 4.3474E+003100 3900
 2.2000E+02 3.6819E+00 2.3000E+02 3.5706E+00 2.4000E+02 4.0995E+003100 3900
 2.5500E+02 3.0083E+00 2.7000E+02 5.0913E+00 2.8000E+02 5.6135E+003100 3900
 3.0000E+02 4.5205E+00 3.2000E+02 6.3687E+00 3.4000E+02 2.6198E+003100 3900
 3.6000E+02 3.4985E+00 3.8000E+02 3.6612E+00 4.0000E+02 3.1876E+003100 3900
 4.2500E+02 2.1236E+00 4.5000E+02 2.1440E+00 4.7500E+02 4.5796E+003100 3900
 5.0000E+02 2.9168E+00 5.2500E+02 4.3518E+00 5.5000E+02 1.4714E+003100 3900
 5.7500E+02 2.0069E+00 6.0000E+02 2.2452E+00 6.3000E+02 4.2128E+003100 3900
 6.6000E+02 3.1071E+00 6.9000E+02 1.5217E+00 7.2000E+02 1.2373E+003100 3900
 7.6000E+02 1.1523E+00 8.0000E+02 1.1249E+00 8.4000E+02 1.4630E+003100 3900
 8.8000E+02 1.9094E+00 9.2000E+02 3.0359E+00 9.6000E+02 1.3138E+003100 3900
 1.0000E+03 3.6528E+00 1.0500E+03 9.7157E-01 1.1000E+03 2.9769E+003100 3900
 1.1500E+03 1.0585E+00 1.2000E+03 4.0391E+00 1.2750E+03 6.7298E-013100 3900
 1.3500E+03 6.7764E-01 1.4250E+03 1.3796E+00 1.5000E+03 6.7876E-013100 3900
 1.6000E+03 6.7518E-01 1.7000E+03 7.0528E-01 1.8000E+03 4.1174E+003100 3900
 1.9000E+03 7.0857E-01 2.0000E+03 7.4331E-01 2.1000E+03 1.9085E+003100 3900
 2.2000E+03 9.7233E-01 2.3000E+03 1.0859E+01 2.4000E+03 8.7289E-013100 3900
 2.5500E+03 3.9196E+00 2.7000E+03 7.9842E-01 2.8000E+03 7.5788E-013100 3900
 3.0000E+03 7.9912E-01 3.2000E+03 1.0212E+00 3.4000E+03 1.4056E+013100 3900
 3.6000E+03 1.0500E+00 3.8000E+03 9.8169E-01 4.0000E+03 1.8146E+003100 3900
 4.2500E+03 7.5901E-01 4.5000E+03 7.1954E-01 4.7500E+03 7.7158E-013100 3900
 5.0000E+03 8.2704E-01 5.2500E+03 9.7526E-01 5.5000E+03 1.0957E+013100 3900
 5.7500E+03 1.1932E+01 6.0000E+03 8.0398E-01 6.3000E+03 6.6710E-013100 3900
 6.6000E+03 7.2019E-01 6.9000E+03 5.1709E+00 7.2000E+03 5.2950E-013100 3900
 7.6000E+03 4.3053E-01 8.0000E+03 4.0029E-01 8.4000E+03 3.8638E-013100 3900
 8.8000E+03 3.9023E-01 9.2000E+03 4.0210E-01 9.6000E+03 4.3962E-013100 3900
 1.0000E+04 5.2598E-01 1.0500E+04 6.7015E-01 1.1000E+04 8.2229E-013100 3900
 1.1500E+04 9.4465E-01 1.2000E+04 1.1042E+00 1.2750E+04 1.3147E+003100 3900
 1.3500E+04 1.4648E+00 1.4250E+04 1.5236E+00 1.5000E+04 1.6630E+003100 3900
 1.6000E+04 1.8367E+00 1.7000E+04 1.9586E+00 1.8000E+04 1.4132E+013100 3900
 1.9000E+04 3.5726E+00 2.0000E+04 1.0644E+00 2.1000E+04 7.5453E-013100 3900
 2.2000E+04 5.9425E-01 2.3000E+04 4.9881E-01 2.4000E+04 4.4759E-013100 3900
 2.5500E+04 4.2479E-01 2.7000E+04 4.1753E-01 2.8000E+04 4.2775E-013100 3900
 3.0000E+04 4.6204E-01 3.2000E+04 5.5056E-01 3.4000E+04 8.5951E-013100 3900
 3.6000E+04 1.5997E+00 3.8000E+04 1.3238E+01 4.0000E+04 1.0191E+023100 3900
 4.2500E+04 2.1149E+00 4.5000E+04 8.6689E-01 4.7500E+04 3.9098E-013100 3900
 5.0000E+04 3.9491E-01 5.2500E+04 3.9884E-01 5.5000E+04 4.0281E-013100 3900
 5.7500E+04 4.0683E-01 6.0000E+04 4.1084E-01 6.3000E+04 4.1389E-013100 3900
 6.6000E+04 4.1647E-01 6.9000E+04 4.1956E-01 7.2000E+04 4.2216E-013100 3900
 7.6000E+04 4.2429E-01 8.0000E+04 4.2642E-01 8.4000E+04 4.2855E-013100 3900
 8.8000E+04 4.3085E-01 9.2000E+04 4.3332E-01 9.6000E+04 4.3585E-013100 3900
 1.0000E+05 4.3844E-01 1.0500E+05 4.4112E-01 1.1000E+05 4.4351E-013100 3900
 1.1500E+05 4.4564E-01 1.2000E+05 4.4783E-01 1.2750E+05 4.5009E-013100 3900
 1.3500E+05 4.5204E-01 1.4250E+05 4.5363E-01 1.5000E+05 4.5532E-013100 3900
 1.6000E+05 4.5702E-01 1.7000E+05 4.5940E-01 1.8000E+05 4.6246E-013100 3900
 1.9000E+05 4.6553E-01 2.0000E+05 4.6868E-01 2.1000E+05 4.7191E-013100 3900
 2.2000E+05 4.7515E-01 2.3000E+05 4.7853E-01 2.4000E+05 4.8152E-013100 3900
 2.5500E+05 4.8412E-01 2.7000E+05 4.8686E-01 2.8000E+05 4.8962E-013100 3900
 3.0000E+05 4.9219E-01 3.2000E+05 4.9467E-01 3.4000E+05 4.9715E-013100 3900
 3.6000E+05 4.9970E-01 3.8000E+05 5.0412E-01 4.0000E+05 5.1023E-013100 3900
 4.2500E+05 5.1650E-01 4.5000E+05 5.2290E-01 4.7500E+05 5.2871E-013100 3900
 5.0000E+05 5.3410E-01 5.2500E+05 5.3952E-01 5.5000E+05 5.4507E-013100 3900
 5.7500E+05 5.5083E-01 6.0000E+05 5.5678E-01 6.3000E+05 5.6135E-013100 3900
 6.6000E+05 5.6532E-01 6.9000E+05 5.7023E-01 7.2000E+05 5.7450E-013100 3900
 7.6000E+05 5.7802E-01 8.0000E+05 5.8169E-01 8.4000E+05 5.8544E-013100 3900
 8.8000E+05 5.8936E-01 9.2000E+05 5.9345E-01 9.6000E+05 5.9769E-013100 3900
 1.0000E+06 6.0216E-01 1.1000E+06 6.0685E-01 1.2000E+06 6.1120E-013100 3900
 1.3000E+06 6.1510E-01 1.4000E+06 6.1924E-01 1.5000E+06 6.2363E-013100 3900
 1.6000E+06 6.2743E-01 1.7000E+06 6.3069E-01 1.8000E+06 6.3409E-013100 3900
 1.9000E+06 6.3765E-01 2.0000E+06 6.4213E-01 2.1000E+06 6.4731E-013100 3900
 2.2000E+06 6.5271E-01 2.3000E+06 6.5828E-01 2.4000E+06 6.6416E-013100 3900
 2.5000E+06 6.7016E-01 2.6000E+06 6.7640E-01 2.7000E+06 6.8214E-013100 3900
 2.8000E+06 6.8713E-01 2.9000E+06 6.9241E-01 3.0000E+06 6.9792E-013100 3900
 3.1000E+06 7.0329E-01 3.2000E+06 7.0832E-01 3.3000E+06 7.1344E-013100 3900
 3.4000E+06 7.1892E-01 3.5000E+06 7.2678E-01 3.6000E+06 7.3700E-013100 3900
 3.7000E+06 7.4758E-01 3.8000E+06 7.5868E-01 3.9000E+06 7.6888E-013100 3900
 4.0000E+06 7.7855E-01 4.1000E+06 7.8855E-01 4.2000E+06 7.9881E-013100 3900
 4.3000E+06 8.0987E-01 4.4000E+06 8.2143E-01 4.5000E+06 8.3051E-013100 3900
 4.6000E+06 8.3849E-01 4.7000E+06 8.4870E-01 4.8000E+06 8.5739E-013100 3900
 4.9000E+06 8.6500E-01 5.0000E+06 8.7283E-01 5.1000E+06 8.8088E-013100 3900
 5.2000E+06 8.8981E-01 5.3000E+06 8.9897E-01 5.4000E+06 9.0855E-013100 3900
 5.5000E+06 9.1915E-01 5.6000E+06 9.3000E-01 5.7000E+06 9.4042E-013100 3900
 5.8000E+06 9.5009E-01 5.9000E+06 9.6033E-01 6.0000E+06 9.7162E-013100 3900
 6.1000E+06 9.8114E-01 6.2000E+06 9.8952E-01 6.3000E+06 9.9837E-013100 3900
 6.4000E+06 1.0078E+00 6.5000E+06 1.0189E+00 6.6000E+06 1.0315E+003100 3900
 6.7000E+06 1.0447E+00 6.8000E+06 1.0583E+00 6.9000E+06 1.0726E+003100 3900
 7.0000E+06 1.0877E+00 7.1000E+06 1.1034E+00 7.2000E+06 1.1178E+003100 3900
 7.3000E+06 1.1311E+00 7.4000E+06 1.1448E+00 7.5000E+06 1.1593E+003100 3900
 7.6000E+06 1.1731E+00 7.7000E+06 1.1864E+00 7.8000E+06 1.2006E+003100 3900
 7.9000E+06 1.2158E+00 8.0000E+06 1.2312E+00 8.1000E+06 1.2482E+003100 3900
 8.2000E+06 1.2663E+00 8.3000E+06 1.2848E+00 8.4000E+06 1.3038E+003100 3900
 8.5000E+06 1.3213E+00 8.6000E+06 1.3402E+00 8.7000E+06 1.3604E+003100 3900
 8.8000E+06 1.3826E+00 8.9000E+06 1.4067E+00 9.0000E+06 1.4263E+003100 3900
 9.1000E+06 1.4438E+00 9.2000E+06 1.4683E+00 9.3000E+06 1.4925E+003100 3900
 9.4000E+06 1.5131E+00 9.5000E+06 1.5349E+00 9.6000E+06 1.5580E+003100 3900
 9.7000E+06 1.5822E+00 9.8000E+06 1.6083E+00 9.9000E+06 1.6361E+003100 3900
 1.0000E+07 1.6661E+00 1.0100E+07 1.6983E+00 1.0200E+07 1.7287E+003100 3900
 1.0300E+07 1.7570E+00 1.0400E+07 1.7871E+00 1.0500E+07 1.8197E+003100 3900
 1.0600E+07 1.8492E+00 1.0700E+07 1.8738E+00 1.0800E+07 1.9004E+003100 3900
 1.0900E+07 1.9292E+00 1.1000E+07 1.9574E+00 1.1100E+07 1.9858E+003100 3900
 1.1200E+07 2.0158E+00 1.1300E+07 2.0480E+00 1.1400E+07 2.0812E+003100 3900
 1.1500E+07 2.1171E+00 1.1600E+07 2.1558E+00 1.1700E+07 2.1921E+003100 3900
 1.1800E+07 2.2245E+00 1.1900E+07 2.2593E+00 1.2000E+07 2.2965E+003100 3900
 1.2100E+07 2.3330E+00 1.2200E+07 2.3672E+00 1.2300E+07 2.4048E+003100 3900
 1.2400E+07 2.4446E+00 1.2500E+07 2.4905E+00 1.2600E+07 2.5429E+003100 3900
 1.2700E+07 2.5983E+00 1.2800E+07 2.6581E+00 1.2900E+07 2.7158E+003100 3900
 1.3000E+07 2.7717E+00 1.3100E+07 2.8309E+00 1.3200E+07 2.8945E+003100 3900
 1.3300E+07 2.9637E+00 1.3400E+07 3.0395E+00 1.3500E+07 3.1006E+003100 3900
 1.3600E+07 3.1571E+00 1.3700E+07 3.2270E+00 1.3800E+07 3.2906E+003100 3900
 1.3900E+07 3.3451E+00 1.4000E+07 3.4043E+00 1.4100E+07 3.4657E+003100 3900
 1.4200E+07 3.5320E+00 1.4300E+07 3.6035E+00 1.4400E+07 3.6814E+003100 3900
 1.4500E+07 3.7638E+00 1.4600E+07 3.8551E+00 1.4700E+07 3.9420E+003100 3900
 1.4800E+07 4.0228E+00 1.4900E+07 4.1110E+00 1.5000E+07 4.2070E+003100 3900
 1.5100E+07 4.2920E+00 1.5200E+07 4.3682E+00 1.5300E+07 4.4448E+003100 3900
 1.5400E+07 4.5282E+00 1.5500E+07 4.6114E+00 1.5600E+07 4.6932E+003100 3900
 1.5700E+07 4.7819E+00 1.5800E+07 4.8711E+00 1.5900E+07 4.9679E+003100 3900
 1.6000E+07 5.0727E+00 1.6100E+07 5.1861E+00 1.6200E+07 5.2913E+003100 3900
 1.6300E+07 5.3884E+00 1.6400E+07 5.4949E+00 1.6500E+07 5.6036E+003100 3900
 1.6600E+07 5.7124E+00 1.6700E+07 5.8213E+00 1.6800E+07 5.9305E+003100 3900
 1.6900E+07 6.0479E+00 1.7000E+07 6.1830E+00 1.7100E+07 6.3300E+003100 3900
 1.7200E+07 6.4924E+00 1.7300E+07 6.6592E+00 1.7400E+07 6.8311E+003100 3900
 1.7500E+07 6.9891E+00 1.7600E+07 7.1631E+00 1.7700E+07 7.3530E+003100 3900
 1.7800E+07 7.5532E+00 1.7900E+07 7.7820E+00 1.8000E+07 7.9627E+003100 3900
 1.8100E+07 8.1275E+00 1.8200E+07 8.3410E+00 1.8300E+07 8.5423E+003100 3900
 1.8400E+07 8.7103E+00 1.8500E+07 8.8786E+00 1.8600E+07 9.0687E+003100 3900
 1.8700E+07 9.2755E+00 1.8800E+07 9.4867E+00 1.8900E+07 9.7326E+003100 3900
 1.9000E+07 9.9862E+00 1.9100E+07 1.0262E+01 1.9200E+07 1.0532E+013100 3900
 1.9300E+07 1.0782E+01 1.9400E+07 1.1053E+01 1.9500E+07 1.1350E+013100 3900
 1.9600E+07 1.1615E+01 1.9700E+07 1.1845E+01 1.9800E+07 1.2091E+013100 3900
 1.9900E+07 1.2355E+01 2.0000E+07 0.0000E+00                      3100 3900
                                                                  3100 3  0
                                                                  3100 0  0
                                                                     0 0  0
 3.30750E+4 7.42780E+1          0          0         34         103325 1451
 0.0        0.0                 0          0          0          63325 1451
 1.00000E+0 2.00000E+7          0          0         10       20023325 1451
 3.00000E+2 0.0                 1          0        128          33325 1451
 33-As- 75 FEI        EVAL-Sep94  K.I.Zolotarev                   3325 1451
                      DIST-Feb2004                                3325 1451
----IRDF-2002         MATERIAL 3325                               3325 1451
-----INCIDENT NEUTRON DATA                                        3325 1451
------ENDF-6 FORMAT                                               3325 1451
***************************************************************** 3325 1451
 33-AS-75  FEI        EVAL-Sep94  K.I.Zolotarev                   3325 1451
                      DIST-May03                       20030504   3325 1451
----BROND-2           MATERIAL 3325                               3325 1451
------Russian Reactor Dosimetry File RRDF-2002                    3325 1451
***************************************************************** 3325 1451
Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko      3325 1451
***************************************************************** 3325 1451
MF=3                                                              3325 1451
MT= 16  -  As-75(n,2n)As-74 reaction                              3325 1451
-------------------------------------                             3325 1451
    Evaluation was made on the base  of experimental data [1-18]. 3325 1451
Systematics of  (n,2n) excitation functions [19]  was taken into  3325 1451
account in the energy region 10.5 - 13.0 MeV and 19.0 - 20.0 MeV. 3325 1451
Cross section data of  E.Paul  and  R.Clarke [20]  were rejected  3325 1451
due to  inconsistency  with  main bulk  of experimental data and  3325 1451
systematics of (n,2n) excitation functions.                       3325 1451
    Statistical analysis of input cross section data was carried  3325 1451
out by means of PADE-2 code [21].  Rational function was used as  3325 1451
the model function [22].                                          3325 1451
    Integral experimental data for  U-235 thermal fission neutron 3325 1451
spectrum [23-24]  was used for testing evaluated As-75(n,2n)As-74 3325 1451
excitation function. The results of testing are given in Table 1. 3325 1451
    Data for  U-235 thermal fission neutron spectrum  and  Cf-252 3325 1451
spontaneous fission neutron spectrum were taken from ref.[25] and 3325 1451
[26],  respectively.                                              3325 1451
                                                        Table 1.  3325 1451
  3325 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  3325 1451
  3325 1451
 U-235 neutron fission  0.30920           0.304 +- 0.036 [22]   3325 1451
                                          0.311 +- 0.023 [23]   3325 1451
  3325 1451
 Cf-252 spont. fission  0.61804                                 3325 1451
  3325 1451
                                                                  3325 1451
MT=33                                                             3325 1451
MT= 16 -(n,2n) cross section cov. matrix                          3325 1451
----------------------------------------                          3325 1451
    Uncertainties in the evaluated  excitation function  for the  3325 1451
reaction As-75(n,2n)As-74 are given in the form of relative cova- 3325 1451
riance matrix for the 19-neutron energy groups (LB=5). Covariance 3325 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  3325 1451
recommended cross section data by means of PADE-2 code.           3325 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  3325 1451
given in the 33-file are the following:                           3325 1451
                                                                  3325 1451
    2.79921E-09    3.33401E-09    3.76284E-09    3.88107E-09      3325 1451
    4.19122E-09    4.53852E-09    4.91657E-09    5.52980E-09      3325 1451
    7.06444E-09    1.01577E-08    1.66580E-08    2.33017E-08      3325 1451
    1.49576E-07    8.86325E-04    3.07310E-03    3.70656E-03      3325 1451
    9.37209E-03    1.38518E-02    8.84386E-02                     3325 1451
                                                                  3325 1451
References :                                                      3325 1451
 1. R.J.Prestwood, B.P.Bayhurst  Phys. Rev., v.121, p.1438, 1961  3325 1451
 2. B.Granger, M.Longueve  Report EANDC(E)-49L, p.83, Oct. 1963   3325 1451
 3. P.Hille, H.Muenzer Acta Phys. Austriaca, v.23, p.44, May 1966 3325 1451
 4. J.Csikai, G.Peto  Acta Physica Hung., v.23, p.87, May 1967    3325 1451
 5. M.Bormann et al.  Nucl. Phys., v.A115, p.309, July 1968       3325 1451
    M.Bormann et al.  Progress Report EANDC(E)-76, January 1967   3325 1451
 6. P.R.Prasad et al.  Nucl. Phys., v.A138, p.85, November 1969   3325 1451
 7. P.Venugopala Rao et al.  Phys. Rev., v.C3, p.629, Feb. 1971   3325 1451
 8. M.Wagner, M.Uhl  Oesterr. Akad. Wiss., Math.+Naturw. Anzeiger,3325 1451
    v.108, p.185, November 1971                                   3325 1451
 9. S.M.Qaim  Nucl. Phys., Sec.A, v.185, p.614, May 1972          3325 1451
10. D.V.Viktorov, V.L.Sjablin Jadernaja Fizika, v.15,p.1099, 1972 3325 1451
11. C.T.Simpson et al. Inorg. Nucl. Chem., v.35, p.2085,June 1973 3325 1451
12. J.Araminowicz,J.Dresler Peport INR-1464, Warsaw,p.14,May 1973 3325 1451
13. W.Grochulski et al.  Acta Physica Polonica, v.B6,p.139, 1975  3325 1451
14. R.A.Sigg  Dissertation Abstract, sec.B, v.37,p.2237, Nov.1976 3325 1451
15. J.L.Casanova,M.L.Sanchez  Anales de Fisica y Quimica (Spain), 3325 1451
    v.72, n.3, p.186, July 1976                                   3325 1451
16. S.Sothras  Dissertation Abstract, sec.B,v.38,p.280, July 1978 3325 1451
17. C.Konno, Y.Ikeda, K.Oishi e.a.  Report JAERI-1329, Oct. 1993  3325 1451
18. I.Birn, S.M.Qaim  Nucl. Sci. Eng., v.116, p.125-137, 1994     3325 1451
19. V.N.Manokhin   VANT, Ser.:Yadernye Konstanty, 1994, No. 1,    3325 1451
    pp.18-22                                                      3325 1451
20. E.B.Paul, R.L.Clarke  Can. J. Phys., v.31, p.267, 1953        3325 1451
21. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          3325 1451
22. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 3325 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 3325 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  3325 1451
23. F.Nasyrov, B.D.Sciborskij   Atomnaja Energija, v.25, no.5,    3325 1451
    p.437, November 1968                                          3325 1451
24. E.Steinnes  Radiochimica Acta, v.13, p.169, June 1970         3325 1451
25. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 3325 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              3325 1451
26. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        3325 1451
***************************************************************** 3325 1451
  The Q values and threshold energies were updated prior to pro-  3325 1451
  cessing through the codes to comply with the values obtained    3325 1451
  using the NNDC calculation program which is based on the 1995   3325 1451
  Update to the Atomic mass Evaluation.                           3325 1451
                                                                  3325 1451
 File 2 added to the pointwise file containing only the effective 3325 1451
 scattering radius with no resonance parameters given.            3325 1451
 Taken from ENDF/B-VI                                             3325 1451
***************************************************************** 3325 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************3325 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     3325 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     3325 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************3325 1451
 Data Doppler Broadened to 300.000000 Kelvin                      3325 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     3325 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     3325 1451
 ***************** Program FIXUP (Version 2002-1) ****************3325 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 3325 1451
 Corrected Thresholds-----------------------------------------Yes 3325 1451
 Extended Cross Sections to 20 MeV----------------------------No  3325 1451
 Allow Cross Section Deletion---------------------------------No  3325 1451
 Allow Cross Section Reconstruction---------------------------No  3325 1451
 Make All Cross Sections Non-Negative-------------------------Yes 3325 1451
 Delete Energies Not in Ascending Order-----------------------Yes 3325 1451
 Deleted Duplicate Points-------------------------------------Yes 3325 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 3325 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 3325 1451
 Allow Creation of Missing Sections---------------------------No  3325 1451
 Allow Insertion of Energy Points-----------------------------No  3325 1451
 Create Uniform Energy Grid-----------------------------------No  3325 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 3325 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************3325 1451
 Unshielded Group Averages Using  640 Groups                      3325 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     3325 1451
                                1        451        135          13325 1451
                                3         16         36          13325 1451
                               33         16         42          13325 1451
                                                                  3325 1  0
                                                                  3325 0  0
 3.30750E+4 7.42780E+1          0          0          0          03325 3 16
-1.02442E+7-1.02442E+7          0          0          1         983325 3 16
         98          1                                            3325 3 16
 10300000.0 6.04805E-6 10400000.0 .000608735 10500000.0 .0052550503325 3 16
 10600000.0 .014392210 10700000.0 .028074850 10800000.0 .0460172503325 3 16
 10900000.0 .067864400 11000000.0 .093207450 11100000.0 .1216010003325 3 16
 11200000.0 .152580000 11300000.0 .185675000 11400000.0 .2204260003325 3 16
 11500000.0 .256394000 11600000.0 .293214625 11700000.0 .3303858753325 3 16
 11800000.0 .367557125 11900000.0 .404728375 12000000.0 .4413397503325 3 16
 12100000.0 .477391250 12200000.0 .512559250 12300000.0 .5468437503325 3 16
 12400000.0 .580032500 12500000.0 .612125500 12600000.0 .6430125003325 3 16
 12700000.0 .672693500 12800000.0 .701134000 12900000.0 .7283340003325 3 16
 13000000.0 .754310750 13100000.0 .779064250 13200000.0 .8026452503325 3 16
 13300000.0 .825053750 13400000.0 .846360250 13500000.0 .8665647503325 3 16
 13600000.0 .885747750 13700000.0 .903909250 13800000.0 .9209150003325 3 16
 13900000.0 .936765000 14000000.0 .952615000 14100000.0 .9672400003325 3 16
 14200000.0 .980640000 14300000.0 .994040000 14400000.0 1.006381673325 3 16
 14500000.0 1.01766500 14600000.0 1.02894833 14700000.0 1.039320003325 3 16
 14800000.0 1.04878000 14900000.0 1.05824000 15000000.0 1.066797503325 3 16
 15100000.0 1.07445250 15200000.0 1.08210750 15300000.0 1.089762503325 3 16
 15400000.0 1.09655125 15500000.0 1.10247375 15600000.0 1.108396253325 3 16
 15700000.0 1.11431875 15800000.0 1.11943900 15900000.0 1.123757003325 3 16
 16000000.0 1.12807500 16100000.0 1.13239300 16200000.0 1.136711003325 3 16
 16300000.0 1.14029100 16400000.0 1.14313300 16500000.0 1.145975003325 3 16
 16600000.0 1.14881700 16700000.0 1.15165900 16800000.0 1.153825003325 3 16
 16900000.0 1.15531500 17000000.0 1.15680500 17100000.0 1.158295003325 3 16
 17200000.0 1.15978500 17300000.0 1.16127500 17400000.0 1.162227503325 3 16
 17500000.0 1.16264250 17600000.0 1.16305750 17700000.0 1.163472503325 3 16
 17800000.0 1.16335214 17900000.0 1.16269643 18000000.0 1.162040713325 3 16
 18100000.0 1.16138500 18200000.0 1.16072929 18300000.0 1.160073573325 3 16
 18400000.0 1.15941786 18500000.0 1.15810500 18600000.0 1.156135003325 3 16
 18700000.0 1.15416500 18800000.0 1.15219500 18900000.0 1.150225003325 3 16
 19000000.0 1.14825500 19100000.0 1.14628500 19200000.0 1.143680003325 3 16
 19300000.0 1.14044000 19400000.0 1.13720000 19500000.0 1.133960003325 3 16
 19600000.0 1.13072000 19700000.0 1.12748000 19800000.0 1.123815003325 3 16
 19900000.0 1.11972500 20000000.0 0.0                             3325 3 16
                                                                  3325 3  0
                                                                  3325 0  0
 3.30750E+4 7.42780E+1          0          0          0          1332533 16
 0.000000+0 0.000000+0          0         16          0          1332533 16
 0.000000+0 0.000000+0          1          5        231         21332533 16
 1.000000-5 1.030000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7332533 16
 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7332533 16
 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7332533 16
 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0332533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 4.254080-2332533 16
 3.081380-2 1.931400-2 1.096210-2 5.721690-3 2.776080-3 1.317510-3332533 16
 7.548290-4 7.048850-4 9.349630-4 1.310960-3 1.762040-3 2.258760-3332533 16
 2.801620-3 3.416040-3 4.153410-3 5.097380-3 6.377610-3 8.194880-3332533 16
 2.503660-2 1.710970-2 1.053310-2 5.933080-3 3.031140-3 1.367320-3332533 16
 5.445340-4 2.757120-4 3.679670-4 6.958200-4 1.178950-3 1.766310-3332533 16
 2.425790-3 3.137300-3 3.888340-3 4.670750-3 5.478140-3 6.303300-3332533 16
 1.310600-2 9.036090-3 5.666770-3 3.182410-3 1.508880-3 4.972090-4332533 16
 1.198130-6-9.989140-5 9.961100-5 5.258840-4 1.118440-3 1.823060-3332533 16
 2.584600-3 3.338870-3 4.001790-3 4.453280-3 4.511120-3 6.875300-3332533 16
 4.696000-3 2.853540-3 1.466050-3 5.292000-4-8.635390-6-2.104130-4332533 16
-1.367730-4 1.569210-4 6.190120-4 1.197530-3 1.834630-3 2.458670-3332533 16
 2.972620-3 3.235810-3 3.033720-3 3.453490-3 2.266810-3 1.294430-3332533 16
 5.856670-4 1.362000-4-8.005410-5-9.600310-5 5.413750-5 3.357960-4332533 16
 7.122850-4 1.141420-3 1.569950-3 1.925150-3 2.101170-3 1.936260-3332533 16
 1.633900-3 1.073760-3 6.359020-4 3.318380-4 1.553350-4 9.295540-5332533 16
 1.286100-4 2.450480-4 4.234780-4 6.421240-4 8.735030-4 1.080130-3332533 16
 1.207510-3 1.172240-3 8.558900-4 6.667390-4 5.164420-4 4.069720-4332533 16
 3.365790-4 3.019830-4 2.993640-4 3.246310-4 3.733290-4 4.402050-4332533 16
 5.184400-4 5.983440-4 6.651090-4 6.747560-4 6.610070-4 6.283810-4332533 16
 5.803990-4 5.209130-4 4.541900-4 3.852600-4 3.205060-4 2.685700-4332533 16
 2.417760-4 2.584150-4 3.465920-4 7.554750-4 7.992770-4 7.962510-4332533 16
 7.522190-4 6.741240-4 5.700760-4 4.499610-4 3.267030-4 2.183230-4332533 16
 1.513670-4 1.667090-4 9.133160-4 9.708800-4 9.754440-4 9.318050-4332533 16
 8.455750-4 7.232610-4 5.729600-4 4.056160-4 2.372370-4 9.272540-5332533 16
 1.098580-3 1.177030-3 1.204940-3 1.180740-3 1.101790-3 9.636420-4332533 16
 7.593550-4 4.784870-4 1.055990-4 1.346410-3 1.473240-3 1.545330-3332533 16
 1.546490-3 1.454020-3 1.234990-3 8.406180-4 1.969000-4 1.715590-3332533 16
 1.907030-3 2.015320-3 1.996100-3 1.785820-3 1.290460-3 3.668920-4332533 16
 2.227810-3 2.458970-3 2.534180-3 2.358840-3 1.793140-3 6.238740-4332533 16
 2.814850-3 2.998200-3 2.888540-3 2.307930-3 9.846970-4 3.293890-3332533 16
 3.287670-3 2.782370-3 1.476680-3 3.433180-3 3.143590-3 2.141300-3332533 16
 3.284480-3 3.040640-3 4.268410-3                                 332533 16
                                                                  332533  0
                                                                  3325 0  0
                                                                     0 0  0
 3.90890E+4 8.81421E+1          0          0         34         103925 1451
 0.0        0.0                 0          0          0          63925 1451
 1.00000E+0 2.00000E+7          0          0         10       20023925 1451
 3.00000E+2 0.0                 1          0         91          33925 1451
 39-Y - 89 SRI        EVAL-JAN96 N.ODANO (SHIP RES. INST.)        3925 1451
                      DIST-Feb2004                                3925 1451
----IRDF-2002         MATERIAL 3925                               3925 1451
-----INCIDENT NEUTRON DATA                                        3925 1451
------ENDF-6 FORMAT                                               3925 1451
***************************************************************   3925 1451
 39- Y- 89 SRI        EVAL-JAN96 N.ODANO (SHIP RES. INST.)        3925 1451
                      DIST-JUL98                                  3925 1451
----JENDL/D-99        MATERIAL 3925                               3925 1451
***************************************************************   3925 1451
HISTORY                                                           3925 1451
96-01   EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 3925 1451
        N.ODANO (SHIP RES. INST.).                                3925 1451
97-09   COMPILED TO JENDL DOSIMETRY FILE 99.                      3925 1451
                                                                  3925 1451
             ==== POINT-WISE DATA FILE ====                       3925 1451
                                                                  3925 1451
 <DOSIMETRY CROSS SECTION>                                        3925 1451
      Y-89 (N,2N) Y-88 (HALF-LIFE = 106.65D)                      3925 1451
                                                                  3925 1451
MF=1  GENERAL INFORMATION                                         3925 1451
 MT=451  DESCRIPTIVE DATA AND DICTIONARY                          3925 1451
                                                                  3925 1451
MF=2  RESONANCE PARAMETERS                                        3925 1451
 MT=151  PARAMETERS                                               3925 1451
     ONLY SPIN AND SCATTERING RADIUS ARE GIVEN.                   3925 1451
                                                                  3925 1451
MF=3 NEUTRON CROSS SECTIONS                                       3925 1451
 MT=16 (N,2N) CROSS SECTION                                       3925 1451
        EXPERIMENTAL DATA/1-17/ IN NESTOR-2/18/ WERE TAKEN FOR    3925 1451
        THE EVALUATION USING GMA CODE/19/.                        3925 1451
                                                                  3925 1451
MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS                       3925 1451
 MT=16  GENERATED USING THE GMA CODE.                             3925 1451
                                                                  3925 1451
REFERENCES                                                        3925 1451
 1) O.M.HUDSON, JR.+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 3925 1451
    6, 506(E9) (1961).                                            3925 1451
 2) R.RIEDER+ : EANDC(E)-49L, 83 (1965).                          3925 1451
 3) D.G.VALLIS : AWRE-O-76/66 (1966).                             3925 1451
 4) J.CSIKAI+ : ACTA PHYSICA HANGARIA, 23, 87 (1967).             3925 1451
 5) D.S.MATHER+ : AWRE-O-47/69 (1969).                            3925 1451
 6) D.R.NETHAWAY : NUCL. PHYS. A, 190, 635 (1972).                3925 1451
 7) S.M.QAIM+ : EUR-5182E, 939 (1974).                            3925 1451
 8) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975).                 3925 1451
 9) S.K.GHORAI+ : NUCL. PHYS. A, 266, 53 (1976).                  3925 1451
10) M.BORMANN+ : Z. PHYS. A, 277, 203 (1976).                     3925 1451
11) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977).                  3925 1451
12) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 3925 1451
    MEV, UPTON, USA, 12-14, 1980, P.399 (1980).                   3925 1451
13) J.LAUREC+ : CEA-R-5109 (1981).                                3925 1451
14) P.RAICS+ : ATOMKI KOZLEMENYEK, 23, 45 (1981).                 3925 1451
15) L.R.GREENWOOD : ASTM-STP-956, 743 (1987).                     3925 1451
16) HUANG JIAN-ZHOU+ : INDC(CPR)-16 (189).                        3925 1451
17) C.KONNO+ : JAERI 1329 (1993).                                 3925 1451
18) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED.      3925 1451
19) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS     3925 1451
    AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS-       3925 1451
    51363, P.249 (1981).                                          3925 1451
***************************************************************   3925 1451
  The Q values and threshold energies were updated prior to pro-  3925 1451
  cessing through the codes to comply with the values obtained    3925 1451
  using the NNDC calculation program which is based on the 1995   3925 1451
  Update to the Atomic mass Evaluation.                           3925 1451
***************************************************************   3925 1451
                                                                  3925 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************3925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     3925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     3925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************3925 1451
 Data Doppler Broadened to 300.000000 Kelvin                      3925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     3925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     3925 1451
 ***************** Program FIXUP (Version 2002-1) ****************3925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 3925 1451
 Corrected Thresholds-----------------------------------------Yes 3925 1451
 Extended Cross Sections to 20 MeV----------------------------No  3925 1451
 Allow Cross Section Deletion---------------------------------No  3925 1451
 Allow Cross Section Reconstruction---------------------------No  3925 1451
 Make All Cross Sections Non-Negative-------------------------Yes 3925 1451
 Delete Energies Not in Ascending Order-----------------------Yes 3925 1451
 Deleted Duplicate Points-------------------------------------Yes 3925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 3925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 3925 1451
 Allow Creation of Missing Sections---------------------------No  3925 1451
 Allow Insertion of Energy Points-----------------------------No  3925 1451
 Create Uniform Energy Grid-----------------------------------No  3925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 3925 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************3925 1451
 Unshielded Group Averages Using  640 Groups                      3925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     3925 1451
                                1        451         98          03925 1451
                                3         16         32          03925 1451
                               33         16         23          03925 1451
                                                                  3925 1  0
                                                                  3925 0  0
 3.90890E+4 8.81421E+1          0          0          0          03925 3 16
-1.14760E+7-1.14760E+7          0          0          1         853925 3 16
         85          1                                            3925 3 16
 11600000.0 .008985210 11700000.0 .030140246 11800000.0 .0559560003925 3 16
 11900000.0 .082542000 12000000.0 .109128000 12100000.0 .1357140003925 3 16
 12200000.0 .163801750 12300000.0 .200900000 12400000.0 .2395000003925 3 16
 12500000.0 .278100000 12600000.0 .316700000 12700000.0 .3553000003925 3 16
 12800000.0 .393900000 12900000.0 .432500000 13000000.0 .4743200003925 3 16
 13100000.0 .519360000 13200000.0 .564400000 13300000.0 .6094400003925 3 16
 13400000.0 .654480000 13500000.0 .692660000 13600000.0 .7239800003925 3 16
 13700000.0 .755300000 13800000.0 .786620000 13900000.0 .8179400003925 3 16
 14000000.0 .847270000 14100000.0 .874610000 14200000.0 .9019500003925 3 16
 14300000.0 .929290000 14400000.0 .956630000 14500000.0 .9775700003925 3 16
 14600000.0 .992110000 14700000.0 1.00665000 14800000.0 1.021190003925 3 16
 14900000.0 1.03573000 15000000.0 1.04830000 15100000.0 1.058900003925 3 16
 15200000.0 1.06950000 15300000.0 1.08010000 15400000.0 1.090700003925 3 16
 15500000.0 1.10190000 15600000.0 1.11370000 15700000.0 1.125500003925 3 16
 15800000.0 1.13730000 15900000.0 1.14910000 16000000.0 1.158350003925 3 16
 16100000.0 1.16505000 16200000.0 1.17175000 16300000.0 1.178450003925 3 16
 16400000.0 1.18515000 16500000.0 1.19185000 16600000.0 1.198550003925 3 16
 16700000.0 1.20525000 16800000.0 1.21195000 16900000.0 1.218650003925 3 16
 17000000.0 1.22130000 17100000.0 1.21990000 17200000.0 1.218500003925 3 16
 17300000.0 1.21710000 17400000.0 1.21570000 17500000.0 1.214300003925 3 16
 17600000.0 1.21290000 17700000.0 1.21150000 17800000.0 1.210100003925 3 16
 17900000.0 1.20870000 18000000.0 1.20790000 18100000.0 1.207700003925 3 16
 18200000.0 1.20750000 18300000.0 1.20730000 18400000.0 1.207100003925 3 16
 18500000.0 1.20690000 18600000.0 1.20670000 18700000.0 1.206500003925 3 16
 18800000.0 1.20630000 18900000.0 1.20610000 19000000.0 1.205900003925 3 16
 19100000.0 1.20570000 19200000.0 1.20550000 19300000.0 1.205300003925 3 16
 19400000.0 1.20510000 19500000.0 1.20490000 19600000.0 1.204700003925 3 16
 19700000.0 1.20450000 19800000.0 1.20430000 19900000.0 1.204100003925 3 16
 20000000.0 0.0                                                   3925 3 16
                                                                  3925 3  0
                                                                  3925 0  0
 3.90890E+4 8.81421E+1          0          0          0          1392533 16
 0.000000+0 0.000000+0          0         16          0          1392533 16
 0.000000+0 0.000000+0          1          5        120         15392533 16
 1.000000-5 1.160000+7 1.167980+7 1.200000+7 1.262500+7 1.325000+7392533 16
 1.375000+7 1.425000+7 1.475000+7 1.525000+7 1.575000+7 1.650000+7392533 16
 1.750000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0392533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 2.614690-2392533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16
 2.614690-2 2.351680-4 2.278630-4 9.042900-5 9.259900-5 8.525630-5392533 16
 8.299090-5 7.707100-5 4.918100-5 5.777220-5 5.017710-5 4.334530-5392533 16
 2.515020-3 1.482340-4 7.056710-5 6.952990-5 6.360270-5 6.802450-5392533 16
 5.867100-5 4.991930-5 5.263290-5 5.027740-5 3.826150-5 1.181300-3392533 16
 1.041660-4 9.116020-5 1.023620-4 8.883020-5 1.176510-4 7.032450-5392533 16
 8.058560-5 7.055540-5 5.527930-5 3.254420-4 1.149980-4 1.205660-4392533 16
 1.177200-4 1.281940-4 7.931010-5 9.144310-5 8.052500-5 6.472540-5392533 16
 2.271050-4 1.022200-4 1.102170-4 9.664150-5 6.583630-5 8.648390-5392533 16
 7.769970-5 7.830730-5 2.030630-4 1.065710-4 1.315170-4 8.313670-5392533 16
 9.450610-5 8.571720-5 6.963890-5 3.359890-4 1.032510-4 6.690090-5392533 16
 7.735940-5 6.913090-5 5.707270-5 1.877490-3 1.006380-4 1.132040-4392533 16
 9.722290-5 7.147720-5 7.645230-4 9.817050-5 1.108810-4 1.009150-4392533 16
 4.986290-4 1.343200-4 1.551700-4 5.697770-4 1.678400-4 4.251840-4392533 16
                                                                  392533  0
                                                                  3925 0  0
                                                                     0 0  0
 4.00900E+4 8.91324E+1          0          0         34         104025 1451
 0.0        0.0                 0          0          0          64025 1451
 1.00000E+0 2.00000E+7          0          0         10       20024025 1451
 3.00000E+2 0.0                 1          0         40          34025 1451
 40-Zr- 90 IRK-VIENNA EVAL-APR90                                  4025 1451
                      DIST-Feb2004                                4025 1451
----IRDF-2002         MATERIAL 4025                               4025 1451
-----INCIDENT NEUTRON DATA                                        4025 1451
------ENDF-6 FORMAT                                               4025 1451
 *****************************************************************4025 1451
 40-ZR- 90 IRK-VIENNA EVAL-APR90                                  4025 1451
                      DIST-JUN90                                  4025 1451
    IRK-EVAL.NLIB 25  4025                                        4025 1451
 *****************************************************************4025 1451
  The Q values and threshold energies were updated prior to pro-  4025 1451
  cessing through the codes to comply with the values obtained    4025 1451
  using the NNDC calculation program which is based on the 1995   4025 1451
  Update to the Atomic mass Evaluation.                           4025 1451
 *****************************************************************4025 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************4025 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     4025 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4025 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4025 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4025 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4025 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     4025 1451
 ***************** Program FIXUP (Version 2002-1) ****************4025 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4025 1451
 Corrected Thresholds-----------------------------------------Yes 4025 1451
 Extended Cross Sections to 20 MeV----------------------------No  4025 1451
 Allow Cross Section Deletion---------------------------------No  4025 1451
 Allow Cross Section Reconstruction---------------------------No  4025 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4025 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4025 1451
 Deleted Duplicate Points-------------------------------------Yes 4025 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4025 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4025 1451
 Allow Creation of Missing Sections---------------------------No  4025 1451
 Allow Insertion of Energy Points-----------------------------No  4025 1451
 Create Uniform Energy Grid-----------------------------------No  4025 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4025 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4025 1451
 Unshielded Group Averages Using  640 Groups                      4025 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4025 1451
                                1        451         47          14025 1451
                                3         16         30          14025 1451
                               33         16         46          14025 1451
                                                                  4025 1  0
                                                                  4025 0  0
 4.00900E+4 8.91324E+1          0          0          0          04025 3 16
-1.19700E+7-1.19700E+7          0          0          1         804025 3 16
         80          1                                            4025 3 16
 12100000.0 .005470824 12200000.0 .017333776 12300000.0 .0342925004025 3 16
 12400000.0 .056337500 12500000.0 .081985000 12600000.0 .1112350004025 3 16
 12700000.0 .144872500 12800000.0 .182897500 12900000.0 .2225200004025 3 16
 13000000.0 .263740000 13100000.0 .303092500 13200000.0 .3405775004025 3 16
 13300000.0 .376913750 13400000.0 .412101250 13500000.0 .4472887504025 3 16
 13600000.0 .482476250 13700000.0 .516977500 13800000.0 .5507925004025 3 16
 13900000.0 .582510000 14000000.0 .612130000 14100000.0 .6424475004025 3 16
 14200000.0 .673462500 14300000.0 .703267500 14400000.0 .7318625004025 3 16
 14500000.0 .757600000 14600000.0 .780480000 14700000.0 .8009275004025 3 16
 14800000.0 .818942500 14900000.0 .837701429 15000000.0 .8572042864025 3 16
 15100000.0 .876707143 15200000.0 .895960893 15300000.0 .9137200004025 3 16
 15400000.0 .931230000 15500000.0 .948740000 15600000.0 .9662500004025 3 16
 15700000.0 .983260583 15800000.0 .997274667 15900000.0 1.010789334025 3 16
 16000000.0 1.02430400 16100000.0 1.03781867 16200000.0 1.051333334025 3 16
 16300000.0 1.06484800 16400000.0 1.07836267 16500000.0 1.088679004025 3 16
 16600000.0 1.09579700 16700000.0 1.10291500 16800000.0 1.110033004025 3 16
 16900000.0 1.11715100 17000000.0 1.12426900 17100000.0 1.131387004025 3 16
 17200000.0 1.13850500 17300000.0 1.14562300 17400000.0 1.152741004025 3 16
 17500000.0 1.15859150 17600000.0 1.16317450 17700000.0 1.167757504025 3 16
 17800000.0 1.17234050 17900000.0 1.17692350 18000000.0 1.181506504025 3 16
 18100000.0 1.18608950 18200000.0 1.19067250 18300000.0 1.195255504025 3 16
 18400000.0 1.19983850 18500000.0 1.20193100 18600000.0 1.201533004025 3 16
 18700000.0 1.20113500 18800000.0 1.20073700 18900000.0 1.200339004025 3 16
 19000000.0 1.19994100 19100000.0 1.19954300 19200000.0 1.199145004025 3 16
 19300000.0 1.19874700 19400000.0 1.19834900 19500000.0 1.197951004025 3 16
 19600000.0 1.19755300 19700000.0 1.19715500 19800000.0 1.196757004025 3 16
 19900000.0 1.19635900 20000000.0 0.0                             4025 3 16
                                                                  4025 3  0
                                                                  4025 0  0
 4.00900E+4 8.91324E+1          0          0          0          1402533 16
 0.000000+0 0.000000+0          0         16          0          1402533 16
 0.000000+0 0.000000+0          1          5        253         22402533 16
 1.000000-5 1.210000+7 1.240000+7 1.260000+7 1.280000+7 1.300000+7402533 16
 1.320000+7 1.340000+7 1.360000+7 1.380000+7 1.400000+7 1.420000+7402533 16
 1.440000+7 1.460000+7 1.480000+7 1.500000+7 1.550000+7 1.600000+7402533 16
 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0402533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16
 0.000000+0 1.159000-2 4.915000-3 2.591000-3 2.281000-3 3.480000-3402533 16
 6.788000-4 1.381000-4 1.958000-4 5.803000-5 5.767000-5 2.137000-4402533 16
 6.431000-5 5.829000-5 5.247000-5 1.721000-3 1.796000-3 1.230000-3402533 16
 1.225000-3 1.489000-3 1.297000-3 4.873000-3 2.573000-3 2.266000-3402533 16
 1.893000-3 6.741000-4 1.371000-4 1.066000-4 5.763000-5 5.727000-5402533 16
 1.392000-4 6.387000-5 5.789000-5 5.211000-5 1.130000-3 1.177000-3402533 16
 7.987000-4 7.954000-4 9.667000-4 7.058000-4 8.925000-3 4.716000-3402533 16
 0.000000+0 1.403000-3 2.854000-4 0.000000+0 1.199000-4 1.192000-4402533 16
 1.055000-4 1.329000-4 1.205000-4 1.085000-4 8.939000-4 9.207000-4402533 16
 5.961000-4 5.941000-4 7.214000-4 0.000000+0 6.922000-3 0.000000+0402533 16
 1.236000-3 2.513000-4 0.000000+0 1.056000-4 1.050000-4 9.288000-5402533 16
 1.171000-4 1.061000-4 9.551000-5 7.873000-4 8.108000-4 5.250000-4402533 16
 5.232000-4 6.353000-4 0.000000+0 2.527000-3 0.000000+0 0.000000+0402533 16
 1.010000-4 0.000000+0 0.000000+0 8.392000-5 0.000000+0 0.000000+0402533 16
 0.000000+0 6.644000-4 6.964000-4 4.856000-4 4.834000-4 5.878000-4402533 16
 6.691000-4 1.279000-3 1.354000-4 3.023000-5 5.849000-5 3.123000-5402533 16
 5.003000-5 5.378000-5 5.446000-5 5.173000-5 2.342000-4 2.412000-4402533 16
 1.562000-4 1.557000-4 1.890000-4 0.000000+0 3.077000-4 8.119000-5402533 16
 7.922000-5 7.440000-5 7.199000-5 5.936000-5 7.322000-5 7.150000-5402533 16
 5.648000-5 5.854000-5 3.739000-5 3.784000-5 4.510000-5 0.000000+0402533 16
 1.171000-4 4.644000-5 4.378000-5 4.618000-5 3.271000-5 4.260000-5402533 16
 4.095000-5 3.739000-5 3.920000-5 2.733000-5 2.721000-5 3.308000-5402533 16
 3.766000-5 9.605000-5 4.489000-5 4.244000-5 3.238000-5 4.414000-5402533 16
 3.926000-5 2.002000-5 2.062000-5 1.335000-5 1.331000-5 1.616000-5402533 16
 0.000000+0 1.114000-4 4.042000-5 3.070000-5 4.347000-5 3.718000-5402533 16
 1.990000-5 2.050000-5 2.254000-5 2.203000-5 2.717000-5 1.259000-5402533 16
 8.344000-5 2.932000-5 3.968000-5 3.557000-5 5.268000-5 5.499000-5402533 16
 3.700000-5 3.711000-5 4.471000-5 3.129000-5 6.920000-5 3.915000-5402533 16
 3.852000-5 2.559000-5 2.649000-5 1.696000-5 1.712000-5 2.046000-5402533 16
 0.000000+0 1.115000-4 4.736000-5 2.011000-5 2.072000-5 1.341000-5402533 16
 1.337000-5 1.623000-5 0.000000+0 1.118000-4 2.243000-5 2.328000-5402533 16
 1.482000-5 1.505000-5 1.786000-5 0.000000+0 6.550000-4 4.446000-4402533 16
 2.989000-4 2.997000-4 3.613000-4 2.477000-4 6.887000-4 3.120000-4402533 16
 3.128000-4 3.770000-4 2.596000-4 4.461000-4 3.007000-4 2.902000-4402533 16
 3.123000-4 4.453000-4 2.888000-4 3.121000-4 5.414000-4 2.681000-4402533 16
 6.335000-4                                                       402533 16
                                                                  402533  0
                                                                  4025 0  0
                                                                     0 0  0
 4.10930E+4 9.21083E+1          0          0         34         104125 1451
 0.0        0.0                 0          0          0          64125 1451
 1.00000E+0 2.00000E+7          0          0         10       20024125 1451
 3.00000E+2 0.0                 1          0        408          94125 1451
 41-Nb- 93 FEI/LANL   EVAL-Aug96 Zolotarev et al.,Chadwick et al. 4125 1451
                      DIST-Feb2004                                4125 1451
----IRDF-2002         MATERIAL 4125                               4125 1451
-----INCIDENT NEUTRON DATA                                        4125 1451
------ENDF-6 FORMAT                                               4125 1451
***************************************************************** 4125 1451
-----RUSSIAN DOSIMETRY FILE RRDF-98                 Rev.2         4125 1451
                    ----- MF=3 MT=16  -----                       4125 1451
                    ----- MF=3 MT=51  -----                       4125 1451
  For the IRDF-2002 file these two reactions were converted at    4125 1451
  IAEA/NDS.                                                       4125 1451
  The reaction MF/MT=3/16 was converted to MF/MT=10/16            4125 1451
  The reaction MF/MT=3/51 was converted to MF/MT=10/ 4            4125 1451
  The corresponding co-variance files were also converted         4125 1451
  The reaction MF/MT=33/16 was converted to MF/MT=40/16           4125 1451
  The reaction MF/MT=33/51 was converted to MF/MT=40/ 4           4125 1451
***************************************************************** 4125 1451
 41-NB-93  FEI        EVAL-Aug96 K.Zolotarev and G.Badikov        4125 1451
                      DIST-Aug96                                  4125 1451
---BROND-3            MATERIAL 4112                               4125 1451
----INCIDENT NEUTRON DATA                                         4125 1451
-----ENDF-6 FORMAT                                                4125 1451
AUTHORS OF EVALUATION: K.Zolotarev and S.Badikov                  4125 1451
***************************************************************** 4125 1451
  41-NB- 93 LANL,ANL  EVAL-DEC97 M.CHADWICK,P.YOUNG,D.L.SMITH     4125 1451
 Ch97,Ch99            DIST-AUG99 REV2-                 19990803   4125 1451
----ENDF/B-VI         MATERIAL 4125         REVISION 2            4125 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4125 1451
DOSIMETRY                                                         4125 1451
******************************************************************4125 1451
****************************************************************  4125 1451
******** Start of (N,2N),(N,N') bibliographical component ******* 4125 1451
***************************************************************** 4125 1451
                                                                  4125 1451
                    ----- MF=3 MT=16  -----                       4125 1451
                                                                  4125 1451
    The  evaluation  Nb93(n,2n)Nb92m  excitation  function  from  4125 1451
9.20 MeV to 20 MeV  has been carried out within the framework of  4125 1451
generalized least squares method , rational function was used as  4125 1451
model function [1].The experimental data base covers the results  4125 1451
of measurements [2-37].                                           4125 1451
    Below  9.2 MeV  the evaluation  has been  carried out by      4125 1451
means of linear interpolation.  Experimental data [38-43] were    4125 1451
rejectected due to large discrepancy with the main bulk of the    4125 1451
experimental data.                                                4125 1451
    The evaluated  Nb93(n,2n)Nb92m  excitation function averaged  4125 1451
on  U-235  neutron fission spectrum [44] and  Cf-252 spontaneous  4125 1451
fission neutron spectrum [45] gives the next values :             4125 1451
                                                                  4125 1451
 ---------------------------------------------------------------  4125 1451
    TYPE OF SPECTRUM   I <CS>,MB (calc.) I   <CS>, MB (measured)  4125 1451
 ----------------------I-----------------I----------------------  4125 1451
                       I                 I                        4125 1451
 U-235 neutron fission I   0.4416        I  0.4796+-0.0293 [46]   4125 1451
                       I                 I                        4125 1451
 ----------------------I-----------------I----------------------  4125 1451
                       I                 I                        4125 1451
 CF-252 spont. fission I   0.7701        I  0.749 +- 0.038 [65]   4125 1451
                       I                 I                        4125 1451
                                                                  4125 1451
                                                                  4125 1451
                    ----- MF=3 MT=51  -----                       4125 1451
                                                                  4125 1451
    Evaluation  Nb93(n,n')Nb93m  excitation function from 30 keV  4125 1451
to  20 MeV  was performed  by  the  same method.  Evaluation  is  4125 1451
based on :                                                        4125 1451
from  170 keV to  680 keV on theoretical model calculations [47], 4125 1451
from  680 keV to  7.9 MeV on the  experimental  data  [48-51] ,   4125 1451
from 7.9 MeV to 20.0 MeV on theoretical model calculations [47]   4125 1451
and experimental data [52,53].  Below 170 keV the evaluation has  4125 1451
been carried out by means of linear interpolation.  Experimental  4125 1451
data [48] and [52] were renormalized , respectively , to the new  4125 1451
monitor cross-sections data  U235(n,f) [54] and  Nb93(n,2n)Nb92m  4125 1451
[55].                                                             4125 1451
Re-evaluated  experimental data  of M.Wagner e.a.  [49-50]  were  4125 1451
taken from ref. [56].                                             4125 1451
    The evaluated  Nb93(n,n')Nb93m  excitation function averaged  4125 1451
on  U-235  neutron fission spectrum [44] and  Cf-252 spontaneous  4125 1451
fission neutron spectrum [45] gives the next values :             4125 1451
                                                                  4125 1451
 ---------------------------------------------------------------  4125 1451
    TYPE OF SPECTRUM   I <CS>,MB (calc.) I   <CS>, MB (measured)  4125 1451
 ----------------------I-----------------I----------------------  4125 1451
                       I                 I  150.0 +- 13.0  [57]   4125 1451
 U-235 neutron fission I     143.47      I  147.0 +- 11.0  [58]   4125 1451
                       I                 I  146.2 +- 12.5  [59]   4125 1451
 ----------------------I-----------------I----------------------  4125 1451
                       I                 I  144.9 +-  4.8  [60]   4125 1451
 CF-252 spont. fission I     146.04      I  144.0 +-  5.0  [61]   4125 1451
                       I                 I  150.6 +-  3.6  [62]   4125 1451
                                                                  4125 1451
    The integal cross-section data [57 - 59], measured in  U-235  4125 1451
neutron fission spectrum, were renormalized to the  Nb93m  half-  4125 1451
life (16.1 +- 0.2) Year [63] , new values of  K X-rays branching  4125 1451
and  new monitor cross-section data.                              4125 1451
    The full description  of the cross sections data  evaluation  4125 1451
for the both reactions is given in [64].                          4125 1451
                                                                  4125 1451
References :                                                      4125 1451
 1. S.Badikov, K.Zolotarev  Proc. of NEANSC Specialist's. Meeting 4125 1451
    on Evaluation and Processing of Covariance Data, Oar Ridge ,  4125 1451
    USA, 1992, OECD, Paris, 1993, p.105                           4125 1451
 2. H.Vonach,H.Muenzer  J,OSA, v.96, p.120, Mar.1959              4125 1451
 3. H.A.Tewes e.a.  Report UCRL-6028-T, June 1960                 4125 1451
 4. R.J.Prestwood,B.P.Bayhurst  Phys.Rev., v.121, p.1438,Mar.1961 4125 1451
 5. V.L.Glagolev,P.A.Jampol'skij  J,JET, v.13, p.520, 1961        4125 1451
 6. D.G.Vallis  Report AWRE-O-76/66, Dec.1966                     4125 1451
 7. L.Husain e.a. Phys.Rev./C, v.1, p.1233, Apr.1970              4125 1451
12. A.Paulsen, R.Widera  Z.Physik, v.238, p.23, Sep.1970          4125 1451
13. M.Bormann e.a.  Progress Rep. EANDC(E)-127, p.38, Mar.1970    4125 1451
14. S.M.Qaim e.a.  Proc. of Conf.on Chemical Nuclear Data, Measur.4125 1451
    and Applicat., Univ.of Kent, Canterbury, 20-22 Sep.1971,p.121 4125 1451
15. D.R.Nethaway  Nucl.Phys./A, v.190, p.635, 1972                4125 1451
16. E.Holub,N.Cindro  J.of Phys./G, v.2, p.405, June 1976         4125 1451
17. D.R.Nethaway  J. of Inorg. and Nucl.Chem., v.40, p.1285, 1978 4125 1451
18. C.G.Hudson e.a.  J.Annals of Nucl.Energy, v.5, p.589, 1978    4125 1451
19. J.Laurec e.a.  Report CEA-R-5109, June 1981                   4125 1451
20. T.B.Ryves,P.Kolkowski  J.Phys.G, v.7, n.4, p.529, 1981        4125 1451
21. R.C.Harper,W.L.Alford  J,JP/G, v.8, p.153, Jan.1982           4125 1451
22. J.Csikai  Proc. of Int.Conf. Nuclear Data for Science  and    4125 1451
    Technology, Antverp., 6-10 Sept.1982, p.414                   4125 1451
23. A.Chiadli,G.Paic  Progress Report MOH-5, p.13, 1982           4125 1451
24. S.Daroczy e.a.  Conf.,83KIEV, v.3, p.191, Nov.1983            4125 1451
25. Y.Ikeda e.a  EXFOR 21945, 1984                                4125 1451
26. Lu Han-Lin e.a.  Chinese J.Nucl.Phys., v.6,n.1,p.76, Feb.1984 4125 1451
27. I.Garlea e.a.  J,RRP, v.29, p.421, 1984                       4125 1451
28. I.Garlea e.a.  Proc.of the 14-th Intern. Symp.on Nucl.Phys.,  4125 1451
    Gaussig,GDR,19-23 Nov.1984, ZFK-562, p.126, 1985              4125 1451
29. Y.S.Kim e.a.  J.of the Korean Nucl.Soc., v.18, n.2,p.92,1986  4125 1451
30. Y.Ikeda e.a.  Report  JAERI-1312, 1988                        4125 1451
31. K.Kobayashi,I.Kimura  Proc.of Int. Conf.on Nuclear Data for   4125 1451
    Science and Technology, Mito, Japan 30 May - 3 June 1988      4125 1451
32. R.Woelfle e.a. Appl.Radiat.and Isotopes, v.39, p.407, 1988    4125 1451
33. Lu Hanlin e.a.  Report INDC(CPR)-16, Aug.1989                 4125 1451
34. Wang Xiuyuan e.a.  EXFOR 30935, 1989                          4125 1451
35. D.C.Santry,R.D.Werner  Can.J.Phys., v.68, p.582, 1990         4125 1451
36. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992  4125 1451
37  D.L.Smith e.a. Proc.Int.Conf. Nuclear Data for Science and    4125 1451
    Technology, Julich, FRG, 13-17 May 1991. Springer Verlag,     4125 1451
    Berlin - Heidelberg, 1992, p.282-284                          4125 1451
38. D.R.Koehler,W.L.Alford  Report NP-11667, 1962                 4125 1451
39. F.Strohal e.a.  Nucl.Phys., v.30, p.49, Feb.1962              4125 1451
40. G.T.Western e.a.  Report AFWL-TR-65-216, p.2, June 1966       4125 1451
41. W.D.Lu e.a.  Phys.Rev.C, v.1, p.350, Jan.1970                 4125 1451
42. A.Mannan,S.M.Qaim  Phys.Rev./C, v.38, n.2, p.630, Aug.1988    4125 1451
43. S.I.Bhuiyan  EXFOR 30936, 1989                                4125 1451
44. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI   4125 1451
    Library, MAT=9228, MF=5, MT=18, eval.apr. 1989                4125 1451
45. W.Mannhart  IAEA-TECDOC-410, p.158, 1987                      4125 1451
46. K.Kobayashi, T.Kobayashi  Progress Report NEANDC(J)-155U,     4125 1451
    P.52, 1990                                                    4125 1451
47. B.Strohmaier  Ann. nucl. Energy, v.16, n.9, p.461, 1989       4125 1451
48. D.B.Gayther e.a. Rep. AERE-R-12612, MAY 1987                  4125 1451
49. M.Wagner e.a. J.Annals of Nucl.Energy, v.15, n.7, p.363, 1988 4125 1451
50. M.Wagner e.a. Proc.Int.Conf. Nucl.Data for Sci.Tec.,Mito,     4125 1451
    Japan,1988,p.1049                                             4125 1451
51. M.Wagner e.a. Measurement of the activation cross section for 4125 1451
    the reaction 93Nb(n,n')93mNb in the neutron energy range      4125 1451
    6 - 9 MeV. Progress Report 1991                               4125 1451
52. T.B.Ryves,P.Kolkowski  J. of Physics G, v.7, n.4, p.529, 1981 4125 1451
53. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992  4125 1451
54. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI   4125 1451
    Library, MAT=9228, MF=5, MT=18, eval.apr. 1989                4125 1451
55. M.Wagner e.a. Physics Data Nr.13-5, p.175, Karlsruhe 1990     4125 1451
56. M.Wagner e.a. Physics Data Nr.13-5, p.135, Karlsruhe 1990     4125 1451
57. K.Sakurai, I.Kondo  Nucl.Instr.and Meth., v.187, p.649, 1981  4125 1451
58. K.Kobayashi, I.Kimura  Rep. NEANDC(J)-61U, p.78, sep.1979     4125 1451
59. F.Hegedues  Proc. of the 1 st ASTM-EURATOM Symposium on Reac- 4125 1451
    tor Dosimetry, Petten 1975. EUR 5667e, Part 1, p.757, 1976    4125 1451
60. W.G.Alberts e.a. Proc. of the 6 th ASTM-EURATOM Symposium on  4125 1451
    Reactor Dosimetry, ASTM-STP-1OO1, p.223, 1989                 4125 1451
61. T.G.Williamson e.a. Proc. of the 6 th ASTM-EURATOM Symposium  4125 1451
    on Reactor Dosimetry, ASTM-STP-1OO1, p.229, 1989              4125 1451
62. J.G.Williams e.a. Proc.of the 6 th ASTM-EURATOM Symposium on  4125 1451
    Reactor Dosimetry, ASTM-STP-1OO1, p.235, 1989                 4125 1451
63. U.Schotzig,H.Schrader R.PTB-RA-16/2, 2 ed.,Braunschweig,1986  4125 1451
64. S.Badikov, K.Zolotarev  Proc.of the 8-th ASTM-EURATOM Sympos. 4125 1451
    on Reactor Dosimetry, Vail, Colorado, USA,29 Aug.-3 Sep. 1993 4125 1451
65. W.Mannhart Private communication, 1990                        4125 1451
******************************************************************4125 1451
********  End of (N,2N),(N,N') bibliographical component ******** 4125 1451
***************************************************************** 4125 1451
******************************************************************4125 1451
********     Start of (N,G) bibliographical component    ******** 4125 1451
***************************************************************** 4125 1451
                                                                  4125 1451
 ENDF/B-VI MOD 3 Evaluation, December 1997, M.B. Chadwick and     4125 1451
                 P.G. Young (LANL)                                4125 1451
                                                                  4125 1451
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     4125 1451
   in cooperation with ECN Petten.                                4125 1451
                                                                  4125 1451
  This evaluation provides a complete representation of the       4125 1451
nuclear data needed for transport, damage, heating, radioactivity,4125 1451
and shielding applications over the incident proton energy range  4125 1451
from 1 to 150 MeV.  The evaluation utilizes MF=6, MT=5 to         4125 1451
represent all reaction data.  Production cross sections and       4125 1451
emission spectra are given for neutrons, protons, deuterons,      4125 1451
tritons, alpha particles, gamma rays, and all residual nuclides   4125 1451
produced (A>5) in the reaction chains. To summarize, the ENDF     4125 1451
sections with non-zero data above are:                            4125 1451
                                                                  4125 1451
  MF=3 MT=  2  Integral of nuclear plus interference components   4125 1451
               of the elastic scattering cross section            4125 1451
                                                                  4125 1451
       MT=  5  Sum of binary (p,n') and (p,x) reactions           4125 1451
                                                                  4125 1451
  MF=6 MT=  2  Elastic (p,p) angular distributions given as       4125 1451
               ratios of the differential nuclear-plus-           4125 1451
               interference to the integrated value.              4125 1451
                                                                  4125 1451
       MT=  5  Production cross sections and energy-angle         4125 1451
               distributions for emission neutrons, protons,      4125 1451
               deuterons, and alphas; and angle-integrated        4125 1451
               spectra for gamma rays and residual nuclei that    4125 1451
               are stable against particle emission               4125 1451
                                                                  4125 1451
  The evaluation is based on nuclear model calculations that have 4125 1451
been benchmarked to experimental data, especially for n +93Nb and 4125 1451
n +93Nb reactions [Ch98]. We use the GNASH code system [Yo92],    4125 1451
which utilizes Hauser-Feshbach statistical, preequilibrium and    4125 1451
direct-reaction theories. Spherical optical model calculations are4125 1451
used to obtain particle transmission coefficients for the Hauser- 4125 1451
Feshbach calculations, as well as for the elastic proton angular  4125 1451
distributions.                                                    4125 1451
  Cross sections and spectra for producing individual residual    4125 1451
nuclei are included for reactions. The energy-angle-correlations  4125 1451
for all outgoing particles are based on Kalbach systematics       4125 1451
[Ka88].                                                           4125 1451
  A model was developed to calculate the energy distributions of  4125 1451
all recoil nuclei in the GNASH calculations [Ch96a]. The recoil   4125 1451
energy distributions are represented in the laboratory system in  4125 1451
MT=5, MF=6, and are given as isotropic in the lab system. All     4125 1451
other data in MT=5,MF=6 are given in the center-of-mass system.   4125 1451
This method of representation utilizes the LCT=3 option approved  4125 1451
at the November, 1996, CSEWG meeting.                             4125 1451
  Preequilibrium corrections were performed in the course of the  4125 1451
GNASH calculations using the exciton model of Kalbach [Ka77,      4125 1451
Ka85], validated by comparison with calculations using Feshbach,  4125 1451
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     4125 1451
nuclear data sheets were matched to continuum level densities     4125 1451
using the formulation of Ignatyuk et al. [Ig75] and pairing and   4125 1451
shell parameters from the Cook [Co67] analysis. Neutron and       4125 1451
charged- particle transmission coefficients were obtained from the4125 1451
optical potentials, as discussed below.  Gamma-ray transmission   4125 1451
coefficients were calculated using the Kopecky-Uhl model [Ko90].  4125 1451
                                                                  4125 1451
SPECIFIC INFORMATION CONCERNING THE 93Nb EVALUATION               4125 1451
                                                                  4125 1451
The total neutron cross section was obtained from the Finlay[Fi93]4125 1451
measurements.                                                     4125 1451
                                                                  4125 1451
The following optical potentials were used in the GNASH           4125 1451
calculations.  For incident neutrons, the Wilmore-Hodgson         4125 1451
potential was used below 15 MeV, and the Madland potential [Ma88] 4125 1451
was used at higher energies. For incident protons, the            4125 1451
Becchetti-Greenlees [Be69] potential was used up to 50 MeV, above 4125 1451
which the Madland potential [Ma88] was used. In both cases, the   4125 1451
matching energy between the potentials was chosen to result in    4125 1451
continuity of the reaction cross section. For protons at 50 MeV   4125 1451
the reaction cross section (and transmission coefficients) was    4125 1451
renormalized slightly to smoothen the transition between the      4125 1451
potentials. The Perey [Pe63] potential was used for indident      4125 1451
deuterons. For tritons, the Becchetti-Greenlees [Be71] was used   4125 1451
up to 80 MeV, above which the Watanabe potential was used. The    4125 1451
Moyen (McFadden Satchler) [Mc66] potential was used for alpha     4125 1451
particles over the whole energy range.                            4125 1451
                                                                  4125 1451
Direct inelastic scattering to low-lying states in Nb93 was       4125 1451
determined as follows. Coherent excitation of 2+ and 3-           4125 1451
vibrations were assumed to be fragmented over Nb93 states, after  4125 1451
coupling these excitations with the 4.5+ core. The magnitudes of  4125 1451
the deformation lengths of 2+ and 3- excitations was obtained by  4125 1451
fitting values of 34 and 46 mb respectively at 14 MeV, obtained   4125 1451
in ref. [Ch93] and accounting for measurements well. This         4125 1451
strength was then fragmented over Nb states.  For the 3-          4125 1451
excitation, the 7 states are in the "continuum" region of the     4125 1451
GNASH calculation at approximately 2.5 MeV, with spins 1.5-,2.5-, 4125 1451
..,7.5-.  For the 2+, the 5 states (2.5+,3.5+,...6.5+) near 1 MeV 4125 1451
were assumed to be those whose inelastic cross section in the     4125 1451
existing ENDF <20 MeV file are significant (note that the ENDF    4125 1451
file below 20 MeV appears to incorporate inelastic information    4125 1451
only up to 5 MeV for many states, after which a value of zero at  4125 1451
20 MeV was inserted).                                             4125 1451
                                                                  4125 1451
Experimental data is used to benchmark the calculations. For      4125 1451
incident neutrons, experimental neutron emission spectra data     4125 1451
exist at 20 and 26 MeV by Marcinkowski [Ma83]. For incident       4125 1451
protons, spectra data exist at 14 and 26 MeV by Watanabe et       4125 1451
al. [Wa97], and at 65 MeV by Sakai et al [Sa80]. Our evaluation   4125 1451
agrees reasonably well with these measurements.                   4125 1451
                                                                  4125 1451
****************************************************************  4125 1451
                                                                  4125 1451
REFERENCES                                                        4125 1451
                                                                  4125 1451
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182,   4125 1451
  1190 (1969)                                                     4125 1451
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees in Polarization   4125 1451
  Phenomena in Nuclear Reactions (Ed: H.H. Barschall and W.       4125 1451
  Haeberli, The University of Wisconsin Press, 1971) p.682        4125 1451
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 4125 1451
[Ch96a] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      4125 1451
  Koning, "High-Energy Nuclear Data Libraries for Accelerator-    4125 1451
  Driven Technologies:  Calculational Method for Heavy Recoils,"  4125 1451
  Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation     4125 1451
  Technology and Applications, Kalmar, Sweden, 3-7 June 1996.     4125 1451
[Ch98]  M.B. Chadwick and P.G. Young, "Model Calculations of      4125 1451
  n,p + 93Nb" in APT PROGRESS REPORT: 1 November 1997 - 1 January 4125 1451
  1998, internal Los Alamos National Laboratory memo              4125 1451
  January 1998 from R.E. MacFarlane to L. Waters.                 4125 1451
[Ch99]  M.B. Chadwick, P.G. Young, G.M. Hale et al., Los Alamos   4125 1451
  National Laboratory report, LA-UR-99-1222 (1999)                4125 1451
[Co67]  J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J.    4125 1451
  Phys. 20, 477 (1967)                                            4125 1451
[Fi93]  R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 4125 1451
  C 47, 237 (1993)                                                4125 1451
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    4125 1451
   Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485     4125 1451
   (1975)                                                         4125 1451
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     4125 1451
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         4125 1451
  LA-10248-MS (1985)                                              4125 1451
[Ka88]  C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also          4125 1451
  C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981)            4125 1451
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990)         4125 1451
[Lo74]  J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974)    4125 1451
[Ma88]  D.G. Madland, "Recent Results in the Development of a     4125 1451
  Global Medium-Energy Nucleon-Nucleus Optical-Model Potential,"  4125 1451
  Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear   4125 1451
  Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U'  4125 1451
  (1988).                                                         4125 1451
[Ma83]  A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al.,  4125 1451
  Nucl.Phys. A402, 220 (1983)                                     4125 1451
[Mc66]  L. McFadden and G. R. Satchler, Nucl. Phys. 84, 177       4125 1451
  (1966).                                                         4125 1451
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      4125 1451
[Sa80]  H. Sakai, K. Hosono, N. Matsuoka et al., Nucl.Phys. A344, 4125 1451
  41 (1980)                                                       4125 1451
[We96]  H.P. Wellisch and D. Axen, Phys.Rev. C 54, 1329(1996)     4125 1451
[Wa97]  Y. Watanabe, S. Yoshioka, M. Harada et al, Nuclear Data   4125 1451
  for Science and Technology, Proc. Conf. Trieste, May, 1997,     4125 1451
  G. Reffo, Ed. (Editrice Compositori, 1997) p.580                4125 1451
[Wi64]  D. Wilmore and P.E. Hodgson, Nucl.Phys. 55, 673 (1964)    4125 1451
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, report        4125 1451
  LA-12343-MS (1992)                                              4125 1451
                                                                  4125 1451
***************************************************************** 4125 1451
                                                                  4125 1451
 ENDF/B-VI MOD 2 Revision, August 1991, NNDC                      4125 1451
                                                                  4125 1451
 Only the section MOD numbers have been corrected in the          4125 1451
 directory.                                                       4125 1451
                                                                  4125 1451
***************************************************************** 4125 1451
                                                                  4125 1451
 ENDF/B-VI MOD 1 Evaluation, March 1990, A.B. Smith, D.L. Smith   4125 1451
                 L.P. Geraldo (ANL), and R. Howerton (LLNL)       4125 1451
                                                                  4125 1451
 Original evaluation fully documented in Smith et al. [1]         4125 1451
                                                                  4125 1451
 ---------------------------------------------------------------- 4125 1451
REFERENCES                                                        4125 1451
                                                                  4125 1451
[1.] A.B. Smith, D.L. Smith and R.J. Howerton, Argonne report     4125 1451
     ANL/NDM-88 (1985)                                            4125 1451
[2.] D.L. Smith and L.P. Geraldo, Argonne report ANL/NDM-117      4125 1451
     (1990)                                                       4125 1451
                                                                  4125 1451
******************************************************************4125 1451
********      End of (N,G) bibliographical component     ******** 4125 1451
***************************************************************** 4125 1451
 The Q values and threshold energies were updated prior to pro-   4125 1451
 cessing through the codes to comply with the values obtained     4125 1451
 using the NNDC calculation program which is based on the 1995    4125 1451
 Update to the Atomic mass Evaluation.                            4125 1451
                                                                  4125 1451
 File 2 taken from ENDF/B-VI                                      4125 1451
 ************************ C O N T E N T S *********************** 4125 1451
                                                                  4125 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************4125 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     4125 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4125 1451
 ***************** Program RECENT (VERSION 2002-1) ***************4125 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4125 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     4125 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4125 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4125 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4125 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     4125 1451
 ***************** Program FIXUP (Version 2002-1) ****************4125 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4125 1451
 Corrected Thresholds-----------------------------------------Yes 4125 1451
 Extended Cross Sections to 20 MeV----------------------------No  4125 1451
 Allow Cross Section Deletion---------------------------------No  4125 1451
 Allow Cross Section Reconstruction---------------------------No  4125 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4125 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4125 1451
 Deleted Duplicate Points-------------------------------------Yes 4125 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4125 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4125 1451
 Allow Creation of Missing Sections---------------------------No  4125 1451
 Allow Insertion of Energy Points-----------------------------No  4125 1451
 Create Uniform Energy Grid-----------------------------------No  4125 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4125 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4125 1451
 Unshielded Group Averages Using  640 Groups                      4125 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4125 1451
                                1        451        421          14125 1451
                                3        102        217          14125 1451
                                8          4          2          14125 1451
                                8         16          2          14125 1451
                               10          4         90          14125 1451
                               10         16         41          14125 1451
                               33        102          9          14125 1451
                               40          4        104          14125 1451
                               40         16         39          14125 1451
                                                                  4125 1  0
                                                                  4125 0  0
 4.10930E+4 9.21083E+1          0          0          0          04125 3102
 7.22747E+6 7.22747E+6          0          0          1        6414125 3102
        641          1                                            4125 3102
 .000100000 18.0213007 .000105000 17.6011498 .000110000 17.20200384125 3102
 .000115000 16.8342057 .000120000 16.4042406 .000127500 15.92720694125 3102
 .000135000 15.4913843 .000142500 15.0908610 .000150000 14.65737964125 3102
 .000160000 14.2058190 .000170000 13.7949650 .000180000 13.41782814125 3102
 .000190000 13.0669303 .000200000 12.7465657 .000210000 12.44515734125 3102
 .000220000 12.1651963 .000230000 11.9050847 .000240000 11.59977434125 3102
 .000255000 11.2642942 .000270000 11.0054074 .000280000 10.71720284125 3102
 .000300000 10.3657565 .000320000 10.0475333 .000340000 9.756094094125 3102
 .000360000 9.48766816 .000380000 9.24263675 .000400000 8.987176444125 3102
 .000425000 8.72576587 .000450000 8.48693672 .000475000 8.267962614125 3102
 .000500000 8.06202911 .000525000 7.87383007 .000550000 7.696048704125 3102
 .000575000 7.53070746 .000600000 7.36124432 .000630000 7.186911624125 3102
 .000660000 7.02739729 .000690000 6.87437954 .000720000 6.711564314125 3102
 .000760000 6.53699045 .000800000 6.37490446 .000840000 6.226363634125 3102
 .000880000 6.08518481 .000920000 5.95576328 .000960000 5.831819854125 3102
 .001000000 5.70335266 .001050000 5.56924720 .001100000 5.443531344125 3102
 .001150000 5.32768145 .001200000 5.19101925 .001275000 5.040835304125 3102
 .001350000 4.90352207 .001425000 4.77537192 .001500000 4.639231264125 3102
 .001600000 4.49688317 .001700000 4.36693655 .001800000 4.246754304125 3102
 .001900000 4.13696432 .002000000 4.03517330 .002100000 3.939658444125 3102
 .002200000 3.85218040 .002300000 3.76853368 .002400000 3.673114994125 3102
 .002550000 3.56708007 .002700000 3.48413693 .002800000 3.394028254125 3102
 .003000000 3.28311047 .003200000 3.18218089 .003400000 3.089889654125 3102
 .003600000 3.00571857 .003800000 2.92800045 .004000000 2.846995734125 3102
 .004250000 2.76475415 .004500000 2.68968605 .004750000 2.619526324125 3102
 .005000000 2.55514441 .005250000 2.49546942 .005500000 2.439062994125 3102
 .005750000 2.38736888 .006000000 2.33308297 .006300000 2.278836274125 3102
 .006600000 2.22751075 .006900000 2.17992067 .007200000 2.128184914125 3102
 .007600000 2.07287015 .008000000 2.02229362 .008400000 1.974836684125 3102
 .008800000 1.93068724 .009200000 1.88965925 .009600000 1.850618324125 3102
 .010000000 1.81018797 .010500000 1.76743310 .011000000 1.728284414125 3102
 .011500000 1.69092729 .012000000 1.64810317 .012750000 1.600693434125 3102
 .013500000 1.55678814 .014250000 1.51653646 .015000000 1.473662134125 3102
 .016000001 1.42869772 .017000001 1.38751152 .017999999 1.349987664125 3102
 .018999999 1.31541292 .020000000 1.28292398 .021000000 1.253192614125 3102
 .022000000 1.22518032 .023000000 1.19908065 .024000000 1.168416244125 3102
 .025500000 1.13534352 .027000001 1.10913958 .028000001 1.080676714125 3102
 .029999999 1.04563566 .032000002 1.01370832 .034000002 .9845640514125 3102
 .035999998 .957974892 .037999999 .933469317 .039999999 .9079185714125 3102
 .042500000 .882001933 .045000002 .858314863 .047499999 .8362348154125 3102
 .050000001 .815903693 .052499998 .797076973 .055000000 .7792451944125 3102
 .057500001 .762920918 .059999999 .745757234 .063000001 .7285945024125 3102
 .066000000 .712422526 .068999998 .697360082 .071999997 .6810306544125 3102
 .075999998 .663545193 .079999998 .647586836 .083999999 .6326033124125 3102
 .088000000 .618689959 .092000000 .605799476 .096000001 .5934841284125 3102
 .100000001 .580740692 .104999997 .567238550 .109999999 .5548929024125 3102
 .115000002 .543099907 .119999997 .529595519 .127499998 .5146728974125 3102
 .135000005 .500837179 .142499998 .488141351 .150000006 .4746142334125 3102
 .159999996 .460419968 .170000002 .447512045 .180000007 .4358671434125 3102
 .189999998 .424682446 .200000003 .414537148 .209999993 .4051062834125 3102
 .219999999 .396226892 .230000004 .388071824 .239999995 .3784338824125 3102
 .254999995 .367847380 .270000011 .359748630 .280000001 .3506513334125 3102
 .300000012 .339586358 .319999993 .329572955 .340000004 .3204548804125 3102
 .360000014 .312085870 .379999995 .304512822 .400000006 .2963389924125 3102
 .425000012 .288129965 .449999988 .280647711 .474999994 .2736699674125 3102
 .500000000 .267249196 .524999976 .261324591 .550000012 .2557032224125 3102
 .574999988 .250569355 .600000024 .245164713 .629999995 .2397563244125 3102
 .660000026 .234629738 .689999998 .229956081 .720000029 .2249389214125 3102
 .759999990 .219377641 .800000012 .214412077 .839999974 .2095743734125 3102
 .879999995 .205287449 .920000017 .201117331 .959999979 .1972194084125 3102
 1.00000000 .193144171 1.04999995 .188952517 1.10000002 .1849860024125 3102
 1.14999998 .181411664 1.20000005 .177069478 1.27499998 .1722936074125 3102
 1.35000002 .168067132 1.42499995 .163992345 1.50000000 .1598027644125 3102
 1.60000002 .155279789 1.70000005 .151227126 1.79999995 .1473994314125 3102
 1.89999998 .143982571 2.00000000 .140708632 2.09999990 .1377903334125 3102
 2.20000005 .134930277 2.29999995 .132383342 2.40000010 .1293041244125 3102
 2.54999995 .125954380 2.70000005 .123359271 2.79999995 .1205590374125 3102
 3.00000000 .117003541 3.20000005 .113818356 3.40000010 .1109472904125 3102
 3.59999990 .108223278 3.79999995 .105805012 4.00000000 .1031868124125 3102
 4.25000000 .100605679 4.50000000 .098236392 4.75000000 .0960118974125 3102
 5.00000000 .093988893 5.25000000 .092041920 5.50000000 .0902770754125 3102
 5.75000000 .088613386 6.00000000 .086909289 6.30000019 .0851493054125 3102
 6.59999990 .083560185 6.90000010 .082063272 7.19999981 .0803861634125 3102
 7.59999990 .078663891 8.00000000 .077001855 8.39999962 .0755136994125 3102
 8.80000019 .074118084 9.19999981 .072831750 9.60000038 .0715604124125 3102
 10.0000000 .070273917 10.5000000 .068923387 11.0000000 .0677183804125 3102
 11.5000000 .066530440 12.0000000 .065212677 12.7500000 .0637066964125 3102
 13.5000000 .062376173 14.2500000 .061106755 15.0000000 .0598156094125 3102
 16.0000000 .058434376 17.0000000 .057233258 18.0000000 .0561099024125 3102
 19.0000000 .055122657 20.0000000 .054279032 21.0000000 .0534859204125 3102
 22.0000000 .052806123 23.0000000 .052221292 24.0000000 .0516809544125 3102
 25.5000000 .051321727 27.0000000 .051340932 28.0000000 .0521454644125 3102
 30.0000000 .056070388 32.0000000 .075305021 34.0000000 2.145136594125 3102
 36.0000000 1.09955184 38.0000000 .084352271 40.0000000 1.343405374125 3102
 42.5000000 .413498725 45.0000000 .054923719 47.5000000 .0464178864125 3102
 50.0000000 .043601210 52.5000000 .042150748 55.0000000 .0412323614125 3102
 57.5000000 .040589832 60.0000000 .040061649 63.0000000 .0396397084125 3102
 66.0000000 .039357775 69.0000000 .039188537 72.0000000 .0391362604125 3102
 76.0000000 .039334796 80.0000000 .040032723 84.0000000 .0421793534125 3102
 88.0000000 .054697646 92.0000000 1.90564800 96.0000000 .0679057414125 3102
 100.000000 .105279263 105.000000 1.74244681 110.000000 .0804239034125 3102
 115.000000 13.1862451 120.000000 .290231261 127.500000 .0478872294125 3102
 135.000000 .041695504 142.500000 .040903630 150.000000 .0422966434125 3102
 160.000000 .047405648 170.000000 .065198158 180.000000 .2623448834125 3102
 190.000000 31.7843412 200.000000 .106103911 210.000000 .0538213784125 3102
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                                                                  4125 0  0
 4.10930E+4 9.21083E+1          0          0          0          1412533291
 0.0000E+00 0.0000E+00          0        291          0          1412533291
 0.0000E+00 0.0000E+00          1          5        595         34412533291
 1.000000-5 3.000000+4 2.200000+5 3.300000+5 4.300000+5 5.300000+5412533291
 6.300000+5 7.300000+5 8.700000+5 1.010000+6 1.250000+6 1.500000+6412533291
 1.750000+6 2.000000+6 2.510000+6 3.000000+6 4.000000+6 5.000000+6412533291
 6.000000+6 6.850000+6 7.860000+6 8.870000+6 9.880000+6 1.089000+7412533291
 1.190000+7 1.291000+7 1.392000+7 1.493000+7 1.595000+7 1.696000+7412533291
 1.797000+7 1.898000+7 1.999000+7 2.100000+7 1.056270-2 0.000000+0412533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533291
 0.000000+0 1.056270-2-4.712830-4 1.966910-3 1.806770-3 6.741760-4412533291
-7.698140-5-2.691050-4-6.637780-5 2.508250-4 3.745940-4 2.962000-4412533291
 1.733040-4 5.996550-5 5.103040-5 1.537080-4 2.570930-4 3.252750-4412533291
 4.301880-4 6.094740-4 8.649130-4 1.238930-3 1.849700-3 2.245100-3412533291
 1.994000-3 1.491610-3 1.155480-3 1.083730-3 1.239220-3 1.566990-3412533291
 2.029290-3 2.603240-3 3.276500-3 6.639730-3 2.160160-3 8.345460-4412533291
 8.651430-4 1.122380-3 1.110790-3 7.625850-4 1.679710-4-3.139040-4412533291
-4.698960-4-4.177790-4-1.866020-4 9.985080-5 3.203390-4 3.975540-4412533291
 4.239300-4 4.924280-4 6.247460-4 8.775000-4 1.392050-3 2.073540-3412533291
 2.269550-3 1.878770-3 1.441630-3 1.242290-3 1.275590-3 1.471790-3412533291
 1.775120-3 2.154770-3 2.594280-3 3.085800-3 4.299820-3 2.742730-3412533291
 1.180930-3 1.280130-4-3.168090-4-2.543180-4 7.409000-5 3.345940-4412533291
 3.825920-4 3.142290-4 1.609680-4 2.449690-5-3.732750-7 1.189920-4412533291
 2.866310-4 4.309180-4 5.567860-4 7.578590-4 1.276650-3 1.993510-3412533291
 2.136500-3 1.701400-3 1.293840-3 1.147460-3 1.216140-3 1.418870-3412533291
 1.701720-3 2.038410-3 2.416280-3 2.830230-3 3.739540-3 2.931830-3412533291
 1.975360-3 9.708180-4 2.607860-4-1.289430-4-1.991930-4-1.253310-4412533291
-4.579140-5 2.155070-5 4.919190-5 7.744570-5 1.831260-4 3.457120-4412533291
 4.943660-4 6.188340-4 8.343450-4 1.445800-3 2.255780-3 2.340310-3412533291
 1.812270-3 1.393220-3 1.288600-3 1.406360-3 1.643830-3 1.941020-3412533291
 2.271680-3 2.624370-3 2.995240-3 4.189860-3 4.024500-3 3.001730-3412533291
 1.587020-3 1.163030-4-7.682270-4-9.377610-4-7.640760-4-3.258930-4412533291
 1.361950-4 4.330750-4 4.873220-4 4.737090-4 5.292500-4 6.730000-4412533291
 9.466770-4 1.469850-3 2.166400-3 2.392460-3 2.003120-3 1.541370-3412533291
 1.319370-3 1.343650-3 1.544520-3 1.863190-3 2.266260-3 2.735490-3412533291
 3.261940-3 4.832140-3 4.170460-3 2.692670-3 7.597480-4-6.585320-4412533291
-1.121180-3-1.023370-3-4.735440-4 2.135970-4 6.904070-4 7.019770-4412533291
 5.269830-4 4.840870-4 6.322520-4 9.308430-4 1.273820-3 1.747660-3412533291
 2.124950-3 1.958180-3 1.500440-3 1.156890-3 1.057590-3 1.184700-3412533291
 1.489730-3 1.935410-3 2.497770-3 3.163020-3 4.224790-3 3.275700-3412533291
 1.671830-3 2.421390-4-4.249880-4-5.514110-4-2.532550-4 2.647080-4412533291
 6.744820-4 6.617330-4 4.482270-4 3.630620-4 4.795200-4 7.274500-4412533291
 9.033060-4 1.160090-3 1.543880-3 1.533820-3 1.169730-3 8.257270-4412533291
 6.754560-4 7.348610-4 9.717540-4 1.354640-3 1.861410-3 2.478470-3412533291
 3.208690-3 2.497910-3 1.568480-3 8.938470-4 5.088300-4 2.969030-4412533291
 2.890110-4 3.621250-4 3.379780-4 2.478700-4 2.125770-4 2.695270-4412533291
 3.943180-4 5.024480-4 6.618280-4 8.588960-4 8.344860-4 6.329920-4412533291
 4.514480-4 3.763480-4 4.135950-4 5.455180-4 7.555710-4 1.032330-3412533291
 1.368870-3 3.050000-3 3.030270-3 2.581900-3 1.987010-3 1.123700-3412533291
 3.185650-4-1.967760-4-2.476790-4-6.654280-5 5.095430-5 4.159240-5412533291
-1.919820-5 1.143880-4 3.203390-4 1.768150-4-3.885620-5-3.936770-5412533291
 8.313390-5 2.047000-4 2.735460-4 2.799000-4 2.281380-4 1.243990-4412533291
-2.614910-5 3.847210-3 3.737200-3 3.105740-3 1.820990-3 3.908990-4412533291
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 1.510730-4 2.694290-4 1.937640-4-5.807680-5-4.670250-4-1.018470-3412533291
 3.879860-3 3.364430-3 2.071040-3 5.003560-4-7.499150-4-9.504900-4412533291
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 1.713330-4-1.194240-4-5.940800-4-1.236840-3 3.024700-3 1.979160-3412533291
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 7.028490-4 1.145510-3 1.437010-3 6.735220-4 3.739970-5 1.022170-4412533291
 4.508550-4 7.565960-4 9.219840-4 9.481760-4 8.597130-4 6.784300-4412533291
 4.202460-4 1.724780-3 1.753300-3 1.451070-3 1.021880-3 1.000000-3412533291
 1.083020-3 8.507040-4 5.878650-4 4.902150-4 5.328860-4 6.521400-4412533291
 8.050520-4 9.709980-4 1.140600-3 1.310290-3 2.544470-3 2.659500-3412533291
 1.351230-3 4.060560-4 1.165450-3 1.653720-3 1.245440-3 6.684260-4412533291
 3.383700-4 3.185210-4 5.577470-4 9.980400-4 1.597830-3 2.330860-3412533291
 3.877570-3 3.510400-3 1.743520-3 1.068090-3 1.293590-3 1.293280-3412533291
 1.085130-3 9.278740-4 9.323140-4 1.111090-3 1.447670-3 1.923320-3412533291
 2.523750-3 7.414390-3 7.906620-3 3.667750-3 6.525600-4 4.905720-4412533291
 1.438910-3 2.364500-3 2.948770-3 3.187050-3 3.146430-3 2.886070-3412533291
 2.448850-3 1.358160-2 1.079440-2 4.940620-3 1.974820-3 1.624540-3412533291
 2.340620-3 3.263720-3 4.071640-3 4.684760-3 5.098500-3 5.330850-3412533291
 1.467980-2 1.129750-2 6.032760-3 2.540470-3 1.127810-3 1.190650-3412533291
 2.185420-3 3.781260-3 5.796800-3 8.136660-3 1.172610-2 7.841660-3412533291
 3.961550-3 1.607330-3 7.717680-4 1.097320-3 2.271930-3 4.092850-3412533291
 6.436120-3 6.765760-3 4.842440-3 3.296060-3 2.373560-3 2.015210-3412533291
 2.102290-3 2.542480-3 3.271550-3 5.031110-3 4.728170-3 4.240230-3412533291
 3.664830-3 3.050690-3 2.416420-3 1.766290-3 5.484480-3 5.547470-3412533291
 5.182140-3 4.486210-3 3.539640-3 2.388760-3 6.241640-3 6.330080-3412533291
 6.059800-3 5.483720-3 4.660630-3 7.169080-3 7.640690-3 7.944700-3412533291
 8.083190-3 9.263750-3 1.077760-2 1.235910-2 1.397320-2 1.732040-2412533291
 2.295720-2                                                       412533291
                                                                  412533  0
 4.10930E+4 9.21083E+1          0          0          0          1412533292
 0.0000E+00 0.0000E+00          0        292          0          1412533292
 0.0000E+00 0.0000E+00          1          5        210         20412533292
 1.000000-5 8.900000+6 9.250000+6 9.500000+6 9.750000+6 1.000000+7412533292
 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7 1.340000+7412533292
 1.400000+7 1.450000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7412533292
 1.900000+7 2.000000+7 1.841580-2 0.000000+0 0.000000+0 0.000000+0412533292
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533292
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412533292
 0.000000+0 0.000000+0 0.000000+0 1.841580-2-5.728770-4 1.657170-3412533292
 1.455050-3 7.578240-4 1.487120-4-1.526500-4-2.597930-4-2.584590-4412533292
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 1.316680-4-1.410550-5-2.449450-4 6.395990-4 2.205000-4 1.633780-4412533292
 1.415700-4 1.287930-4 1.149910-4 9.899660-5 8.228950-5 6.007320-5412533292
 3.972790-5 2.775870-5 1.927900-5 1.116450-5 7.840190-6 1.330400-5412533292
 2.737810-5 5.011810-5 3.976950-4 3.423270-4 2.498580-4 1.613170-4412533292
 1.040950-4 6.832150-5 4.629340-5 2.983180-5 2.172710-5 1.964290-5412533292
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 4.395540-5 2.694600-5 1.961410-5 1.563060-5 1.332480-5 1.463850-5412533292
 1.974150-5 2.763930-5 3.782410-5 2.460430-4 2.058710-4 1.660420-4412533292
 1.300130-4 9.897580-5 6.427630-5 3.669500-5 2.240070-5 1.326570-5412533292
 5.943540-6 5.400590-6 1.459300-5 3.254010-5 5.881020-5 1.976580-4412533292
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 1.464250-5 4.062820-6 1.535500-6 1.188640-5 3.429400-5 6.845130-5412533292
 1.599410-4 1.373640-4 1.124330-4 7.886480-5 4.844590-5 3.090580-5412533292
 1.874320-5 7.548880-6 3.762390-6 1.263450-5 3.354330-5 6.628720-5412533292
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 1.360900-5 9.229420-6 1.558390-5 3.223920-5 5.907260-5 8.921230-5412533292
 6.886480-5 4.947110-5 3.769610-5 2.916110-5 2.062300-5 1.620220-5412533292
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 5.761910-5 5.296130-5 6.734270-5 7.336690-5 7.759760-5 7.999730-5412533292
 8.807950-5 1.031920-4 1.196930-4 1.348650-4 1.722210-4 2.385330-4412533292
                                                                  412533  0
                                                                  4125 0  0
                                                                     0 0  0
 4.51030E+4 1.02021E+2          0          0         34         104525 1451
 0.0        0.0                 0          0          0          64525 1451
 1.00000E+0 2.00000E+7          0          0         10       20024525 1451
 3.00000E+2 0.0                 1          0        179          44525 1451
 45-Rh-103 FEI        EVAL-Mar03 K.I.Zolotarev                    4525 1451
                      DIST-Feb2004                                4525 1451
----IRDF-2002         MATERIAL 4525                               4525 1451
-----INCIDENT NEUTRON DATA                                        4525 1451
------ENDF-6 FORMAT                                               4525 1451
***************************************************************** 4525 1451
 45-Rh-103 FEI        EVAL-Mar03 K.I.Zolotarev                    4525 1451
                      DIST-Apr03                                  4525 1451
----BROND-2           MATERIAL 4525                               4525 1451
-----INCIDENT NEUTRON DATA                                        4525 1451
------ENDF-6 FORMAT                                               4525 1451
------Russian Reactor Dosimetry File RRDF-2002                    4525 1451
***************************************************************** 4525 1451
                    ----- MF=3 MT=51  -----                       4525 1451
  For the IRDF-2002 file this reaction was converted at IAEA/NDS  4525 1451
  The reaction MF/MT=3/51 was converted to MF/MT=10/ 4            4525 1451
  The corresponding co-variance files were also converted         4525 1451
  The reaction MF/MT=33/51 was converted to MF/MT=40/ 4           4525 1451
***************************************************************** 4525 1451
Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev            4525 1451
***************************************************************** 4525 1451
MF=3                                                              4525 1451
MT= 51 - Rh103(n,n')Rh103m  reaction                              4525 1451
-------------------------------------                             4525 1451
    In the section MT= 51 are given evaluated  cross section data 4525 1451
of neutron inelastic excitation of the  56.12-min  isomeric state 4525 1451
in the rhodium-103  in the energy range from threshold to 20 MeV. 4525 1451
The isomeric level has energy  39.756 keV  with spin  and  parity 4525 1451
7/2+ [1].                                                         4525 1451
    Microscopic experimental data  [4-13]  were  analyzed in  the 4525 1451
process of preparation  of input data base  for the evaluation of 4525 1451
cross sections and their uncertainty  for the Rh-103(n,n')Rh-103m 4525 1451
reaction.  During this procedure all experimental data  if it was 4525 1451
possible were corrected to the new recommended cross section data 4525 1451
for monitor reactions used in the measurements and to the new re- 4525 1451
commended decay data.  Decay data for the rhodium-103m were taken 4525 1451
from  ref. [2].   Recommended  cross sections  for  the  reaction 4525 1451
In-115(n,n')In-115m  used as a monitor  in  the measurements [4], 4525 1451
[5] and [7] were taken from ref. [3].                             4525 1451
    Special correction was done with experimental data [4], [6].  4525 1451
    Data  of Cross [4]  for the neutron energies  2.20 - 2.86 MeV 4525 1451
were renormalized to the measured by Paulsen et al. cross section 4525 1451
value (994.0+-45.7) mb at 2.60 MeV [8].                           4525 1451
    Data of  Santry and Butler [6]  obtained  in the measurements 4525 1451
with D(d,n)He3 source in the energy range 5.00-13.58 MeV were re- 4525 1451
normalized to the integral of Miach et al. experimental data [10] 4525 1451
in the energy interval 6.0-12.0 MeV, Fc= 0.90190 . Cross sections 4525 1451
measured by Santry and Butler  with using  Li7(p,n)Be7, T(p,n)He3 4525 1451
and T(d,n)He4 neutron sources were not corrected.                 4525 1451
    Excitation function for the Rh-103(n,n')Rh-103m  reaction  in 4525 1451
the energy region from threshold to 20 MeV was evaluated by means 4525 1451
of statistical analysis of experimental cross section data [4-10].4525 1451
    Experimental cross section data [11-13] were rejected  due to 4525 1451
their discrepancy with the main bulk of experimental data [4-10]. 4525 1451
In the rejected experiment [11] the cross section value was deter-4525 1451
mined only in a one energy point 14.20 MeV.                       4525 1451
    Statistical analysis of input cross section data was carried  4525 1451
out by means of PADE-2 code [14].  Rational function was used as  4525 1451
the model function [15].                                          4525 1451
    U-235 thermal fission [16]  and  Cf-252 spontaneous  fission  4525 1451
neutron spectra [17] averaged cross sections calculated from the  4525 1451
the evaluated Rh-103(n,n')Rh-103m  excitation function  are  the  4525 1451
following:                                                        4525 1451
  4525 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  4525 1451
  4525 1451
 U-235 neutron fission     715.85         702.16+-28.08   [18]  4525 1451
                                          721.16+-38.65   [19]  4525 1451
                                          670.73+-52.05   [20]  4525 1451
  4525 1451
 Cf-252 spont. fission     724.83         620.82+-67.17   [21]  4525 1451
                                          813.29+-24.24   [22]  4525 1451
                                          813.22+-24.15   [22]  4525 1451
  4525 1451
    Integral experimental data [18-22]  were corrected to the new 4525 1451
recommended cross sections  for monitor reactions  from ref. [23] 4525 1451
and [24].                                                         4525 1451
                                                                  4525 1451
MT=33                                                             4525 1451
MT= 51 -(n,n') cross section cov. matrix                          4525 1451
----------------------------------------                          4525 1451
    Uncertainties in the evaluated  excitation function  for  the 4525 1451
reaction  Rh-103(n,n')Rh-103m  are given  in the form of relative 4525 1451
covariance matrix for the 35-neutron energy groups (LB=5).  Cova- 4525 1451
riance matrix of uncertainties was calculated simultaneously with 4525 1451
recommended cross section data by means of PADE-2 code.           4525 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  4525 1451
given in the 33-file are the following:                           4525 1451
                                                                  4525 1451
    4.63337E-06    4.70345E-06    4.81607E-06    4.97060E-06      4525 1451
    5.13303E-06    5.31401E-06    5.53768E-06    5.79915E-06      4525 1451
    6.08110E-06    6.37403E-06    6.83267E-06    7.59079E-06      4525 1451
    8.42883E-06    9.54383E-06    1.23066E-05    1.44930E-05      4525 1451
    2.21073E-05    4.79767E-05    9.57950E-05    1.42110E-04      4525 1451
    1.77925E-04    2.06906E-04    5.62715E-04    1.57661E-03      4525 1451
    1.64672E-03    1.73445E-03    2.10901E-03    2.41391E-03      4525 1451
    2.81682E-03    8.17990E-03    1.45397E-02    1.82324E-02      4525 1451
    5.39402E-02    1.07397E-01    1.22586E-01                     4525 1451
                                                                  4525 1451
References:                                                       4525 1451
 1. R.B.Firestone   Table of Isotopes, Eighth edition, Vol. 1,    4525 1451
    John Wiley & Sons, Inc., New York, 1995                       4525 1451
 2. H.Vonach et al.    Proc. Inter. Conf. on Nuclear Data for     4525 1451
    Science and Technology Gatlinburg, Tennessee, USA May 9-13,   4525 1451
    1994, v.1, pp.278-280 ;                                       4525 1451
    U.Schotzig   PTB, Braunschweig, FRG, 1994                     4525 1451
 3. K.I.Zolotarev, F.K.Zolotarev  RRDF-2002, MAT=4921, IPPE,      4525 1451
    Obninsk, evaluated March 2003                                 4525 1451
 4. W.G.Cross     Physics in Canada, v.19, p.39, June 1963        4525 1451
 5. A.Pazsit, J.Csikai  Sov. Journal of Nuclear Physics, v.15,    4525 1451
    p.232, 1972.                                                  4525 1451
 6. D.C.Santry, J.P.Butler  Canadian Journal of Physics, v.52,    4525 1451
    p.1421, August 1974                                           4525 1451
 7. A.Pazsit, G.Peto, J.Csikai, I.Jozsa, J.Bacso                  4525 1451
    J. of Applied Radiation and Isotopes, v.26, p.621, Oct. 1975  4525 1451
 8. A.Paulsen et al.   Nucl. Sci. Eng., v.76, p.331, 1980         4525 1451
 9. Li Jianwei et al.  Proc. of an Int. Conf. on Nuclear Data for 4525 1451
    Science and Technology, 30 May - 3 June 1988, Mito, Japan,    4525 1451
    Saikon Publishing Co., LTD, p.315, 1988                       4525 1451
10. M.M.H.Miah, B.Strohmaier, H.Vonach, W.Mannhart, D.Schmidt     4525 1451
    Phys. Rev., pt. C, v.54, No.1, p.222, July 1996               4525 1451
11. W.Nagel    J. Nucl. Energ., v.20, p.475, June 1966            4525 1451
12. I.Kimura et al.    Journal of Nucl. Science and Technology,   4525 1451
    v.6, p.485, September 1969                                    4525 1451
13. E.Barnard, D.Reitmann  Nucl. Phys., v.A303, p.27, June 1978   4525 1451
14. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          4525 1451
15. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 4525 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 4525 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  4525 1451
16. L.W.Weston et al.  Evaluated Neutron Data  for  Uranium-235,  4525 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    4525 1451
17. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        4525 1451
18. E.I.Grigor'ev et al.  6th All Union Conference on Neutron     4525 1451
    Physics, Kiev, 2-6 October 1983, v.3, p.187, Moscow 1984      4525 1451
19. O.Horibe et al. Proc. of Conf.: 50 Years with Nuclear Fission,4525 1451
    Washington D.C., 25-28 April 1989, v.2, p.923 ;               4525 1451
    O.Horibe, H.Chatani  Proc. of Int. Conf. Nuclear Data for Sci.4525 1451
    and Technology, Julich, FRG, 13-17 May 1991, Springer-Verlag, 4525 1451
    Berlin Heidelberg, 1992, p.68                                 4525 1451
20. K.Kobayashi, T.Kobayashi   Progress Report NEANDC(J)-155/U,   4525 1451
    p.52, August 1990                                             4525 1451
21. G.J.Kirouac et al.  Report - 4005, Knolls Atomic Power Lab.,  4525 1451
    January 1974                                                  4525 1451
22. G.P.Lamaze et al.  Nucl. Sci. Eng., v.100, p.43, Sept. 1988   4525 1451
23. W.Mannhart  Prog. Report INDC(Ger)-045, pp.40-43, June 1999   4525 1451
24. W.Mannhart  Validation of Differential Cross Sections with    4525 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4525 1451
    September 2002                                                4525 1451
***************************************************************** 4525 1451
 File 2 added to the pointwise file containing only the effective 4525 1451
 scattering radius with no resonance parameters given.            4525 1451
 Taken from ENDF/B-VI                                             4525 1451
                                                                  4525 1451
***************************************************************** 4525 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************4525 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     4525 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4525 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4525 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4525 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4525 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     4525 1451
 ***************** Program FIXUP (Version 2002-1) ****************4525 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4525 1451
 Corrected Thresholds-----------------------------------------Yes 4525 1451
 Extended Cross Sections to 20 MeV----------------------------No  4525 1451
 Allow Cross Section Deletion---------------------------------No  4525 1451
 Allow Cross Section Reconstruction---------------------------No  4525 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4525 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4525 1451
 Deleted Duplicate Points-------------------------------------Yes 4525 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4525 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4525 1451
 Allow Creation of Missing Sections---------------------------No  4525 1451
 Allow Insertion of Energy Points-----------------------------No  4525 1451
 Create Uniform Energy Grid-----------------------------------No  4525 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4525 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4525 1451
 Unshielded Group Averages Using  640 Groups                      4525 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4525 1451
                                1        451        187          14525 1451
                                8          4          2          14525 1451
                               10          4         88          14525 1451
                               40          4        122          14525 1451
                                                                  4525 1  0
                                                                  4525 0  0
 4.51030E+4 1.02021E+2          0          0          1          04525 3291
 0.0       -3.97560E+4      45103          0          1        2554525 3291
        255          1                                            4525 3291
 40000.0000 .000131876 42500.0000 .000427871 45000.0000 .0007243454525 3291
 47500.0000 .001020819 50000.0000 .001317294 52500.0000 .0016137684525 3291
 55000.0000 .001910242 57500.0000 .002206716 60000.0000 .0025328374525 3291
 63000.0000 .002888606 66000.0000 .003244375 69000.0000 .0036001444525 3291
 72000.0000 .004015207 76000.0000 .004489566 80000.0000 .0049639244525 3291
 84000.0000 .005438283 88000.0000 .005912641 92000.0000 .0063870004525 3291
 96000.0000 .006861358 100000.000 .007395012 105000.000 .0079879604525 3291
 110000.000 .008580908 115000.000 .009173856 120000.000 .0113795894525 3291
 127500.000 .015198106 135000.000 .019016624 142500.000 .0228351414525 3291
 150000.000 .027155800 160000.000 .031978600 170000.000 .0368014004525 3291
 180000.000 .041624200 190000.000 .046447000 200000.000 .0510906904525 3291
 210000.000 .055555270 220000.000 .060019850 230000.000 .0644844304525 3291
 240000.000 .070033483 255000.000 .076286950 270000.000 .0813926004525 3291
 280000.000 .087519380 300000.000 .095305320 320000.000 .1027081604525 3291
 340000.000 .110028295 360000.000 .116852200 380000.000 .1235934004525 3291
 400000.000 .131003000 425000.000 .139081000 450000.000 .1476172504525 3291
 475000.000 .156611750 500000.000 .167446250 525000.000 .1801207504525 3291
 550000.000 .196934000 575000.000 .217886000 600000.000 .2482181004525 3291
 630000.000 .288511267 660000.000 .336357000 690000.000 .3858751334525 3291
 720000.000 .440636700 760000.000 .491837800 800000.000 .5305162004525 3291
 840000.000 .558433425 880000.000 .578712300 920000.000 .5945503004525 3291
 960000.000 .607908400 1000000.00 .628120500 1100000.00 .6549508334525 3291
 1200000.00 .680806500 1300000.00 .706662167 1400000.00 .7317365004525 3291
 1500000.00 .756029500 1600000.00 .779401000 1700000.00 .8018510004525 3291
 1800000.00 .823285500 1900000.00 .843704500 2000000.00 .8631160004525 3291
 2100000.00 .881520000 2200000.00 .898753833 2300000.00 .9148175004525 3291
 2400000.00 .930881167 2500000.00 .945705167 2600000.00 .9592895004525 3291
 2700000.00 .972873833 2800000.00 .985413333 2900000.00 .9969080004525 3291
 3000000.00 1.00840267 3100000.00 1.01891625 3200000.00 1.028448754525 3291
 3300000.00 1.03798125 3400000.00 1.04751375 3500000.00 1.056186254525 3291
 3600000.00 1.06399875 3700000.00 1.07181125 3800000.00 1.079623754525 3291
 3900000.00 1.08670417 4000000.00 1.09305250 4100000.00 1.099400834525 3291
 4200000.00 1.10574917 4300000.00 1.11209750 4400000.00 1.118445834525 3291
 4500000.00 1.12419500 4600000.00 1.12934500 4700000.00 1.134495004525 3291
 4800000.00 1.13964500 4900000.00 1.14479500 5000000.00 1.149945004525 3291
 5100000.00 1.15509500 5200000.00 1.16024500 5300000.00 1.165395004525 3291
 5400000.00 1.17054500 5500000.00 1.17544306 5600000.00 1.180089174525 3291
 5700000.00 1.18473528 5800000.00 1.18938139 5900000.00 1.194027504525 3291
 6000000.00 1.19867361 6100000.00 1.20331972 6200000.00 1.207965834525 3291
 6300000.00 1.21261194 6400000.00 1.21725806 6500000.00 1.221904174525 3291
 6600000.00 1.22655028 6700000.00 1.23119639 6800000.00 1.235842504525 3291
 6900000.00 1.24048861 7000000.00 1.24513472 7100000.00 1.249780834525 3291
 7200000.00 1.25442694 7300000.00 1.25850604 7400000.00 1.262018114525 3291
 7500000.00 1.26553018 7600000.00 1.26904225 7700000.00 1.272554324525 3291
 7800000.00 1.27521141 7900000.00 1.27701386 8000000.00 1.278816324525 3291
 8100000.00 1.28061877 8200000.00 1.28169500 8300000.00 1.281263334525 3291
 8400000.00 1.28005000 8500000.00 1.27883667 8600000.00 1.276161674525 3291
 8700000.00 1.27202500 8800000.00 1.26788833 8900000.00 1.262252504525 3291
 9000000.00 1.25511750 9100000.00 1.24661250 9200000.00 1.236737504525 3291
 9300000.00 1.22537250 9400000.00 1.21251750 9500000.00 1.198110004525 3291
 9600000.00 1.18215000 9700000.00 1.16464500 9800000.00 1.145595004525 3291
 9900000.00 1.12508000 10000000.0 1.10310000 10100000.0 1.079822504525 3291
 10200000.0 1.05524750 10300000.0 1.02941133 10400000.0 1.002314004525 3291
 10500000.0 .975216667 10600000.0 .947189214 10700000.0 .9182316434525 3291
 10800000.0 .889274071 10900000.0 .860316500 11000000.0 .8313589294525 3291
 11100000.0 .802401357 11200000.0 .773443786 11300000.0 .7452375004525 3291
 11400000.0 .717782500 11500000.0 .690327500 11600000.0 .6638342504525 3291
 11700000.0 .638302750 11800000.0 .613694250 11900000.0 .5900087504525 3291
 12000000.0 .567314500 12100000.0 .545611500 12200000.0 .5249195004525 3291
 12300000.0 .505238500 12400000.0 .486307000 12500000.0 .4686130004525 3291
 12600000.0 .451881500 12700000.0 .436085000 12800000.0 .4211920004525 3291
 12900000.0 .407167500 13000000.0 .394168750 13100000.0 .3817702504525 3291
 13200000.0 .370108000 13300000.0 .359182000 13400000.0 .3489147504525 3291
 13500000.0 .339306250 13600000.0 .330282250 13700000.0 .3218427504525 3291
 13800000.0 .313918750 13900000.0 .306510250 14000000.0 .2995535004525 3291
 14100000.0 .293048500 14200000.0 .286937500 14300000.0 .2812205004525 3291
 14400000.0 .275846750 14500000.0 .270816250 14600000.0 .2660832504525 3291
 14700000.0 .261647750 14800000.0 .257527167 14900000.0 .2537215004525 3291
 15000000.0 .249915833 15100000.0 .246422000 15200000.0 .2432400004525 3291
 15300000.0 .240058000 15400000.0 .237126500 15500000.0 .2344455004525 3291
 15600000.0 .231764500 15700000.0 .229284167 15800000.0 .2270045004525 3291
 15900000.0 .224724833 16000000.0 .222629125 16100000.0 .2207173754525 3291
 16200000.0 .218805625 16300000.0 .216893875 16400000.0 .2151590004525 3291
 16500000.0 .213601000 16600000.0 .212043000 16700000.0 .2104850004525 3291
 16800000.0 .208927000 16900000.0 .207520000 17000000.0 .2062640004525 3291
 17100000.0 .205008000 17200000.0 .203752000 17300000.0 .2024960004525 3291
 17400000.0 .201240000 17500000.0 .200100429 17600000.0 .1990772864525 3291
 17700000.0 .198054143 17800000.0 .197031000 17900000.0 .1960078574525 3291
 18000000.0 .194984714 18100000.0 .193961571 18200000.0 .1930241674525 3291
 18300000.0 .192172500 18400000.0 .191320833 18500000.0 .1904691674525 3291
 18600000.0 .189617500 18700000.0 .188765833 18800000.0 .1879141674525 3291
 18900000.0 .187062500 19000000.0 .186210833 19100000.0 .1854136674525 3291
 19200000.0 .184671000 19300000.0 .183928333 19400000.0 .1831856674525 3291
 19500000.0 .182443000 19600000.0 .181700333 19700000.0 .1809576674525 3291
 19800000.0 .180215000 19900000.0 .179472333 20000000.0 0.0       4525 3291
                                                                  4525 3  0
                                                                  4525 0  0
 4.51030E+4 1.02021E+2          0          0          0          1452533291
 0.0000E+00 0.0000E+00          0        291          0          1452533291
 0.0000E+00 0.0000E+00          1          5        703         37452533291
 1.000000-5 4.000000+4 2.000000+5 4.000000+5 6.000000+5 8.000000+5452533291
 1.000000+6 1.500000+6 2.000000+6 2.500000+6 3.000000+6 3.500000+6452533291
 4.000000+6 4.500000+6 5.000000+6 5.500000+6 6.000000+6 6.500000+6452533291
 7.000000+6 7.500000+6 8.000000+6 8.500000+6 9.000000+6 9.500000+6452533291
 1.000000+7 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.300000+7452533291
 1.400000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7452533291
 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291
 0.000000+0 1.201660-1 3.626120-3 5.830250-3 2.406510-3 3.829450-3452533291
 9.421320-4 1.744590-3 2.795830-3 3.312800-3 3.411860-3 3.250220-3452533291
 2.951990-3 2.607590-3 2.280380-3 2.013300-3 1.832210-3 1.747890-3452533291
 1.754810-3 1.828740-3 1.924420-3 1.979780-3 1.934180-3 1.764500-3452533291
 1.522800-3 1.330280-3 1.305230-3 1.474590-3 1.853920-3 1.977210-3452533291
 1.349570-3 4.168760-4-3.002750-4-5.661320-4-3.496380-4 3.015310-4452533291
 4.977370-3 2.720970-3 2.068350-3 1.881520-3 1.983850-3 1.918150-3452533291
 1.869060-3 1.891870-3 1.958280-3 2.036100-3 2.099860-3 2.130840-3452533291
 2.116050-3 2.047420-3 1.922130-3 1.743840-3 1.525610-3 1.293820-3452533291
 1.091910-3 9.788930-4 1.014690-3 1.226060-3 1.564980-3 1.899910-3452533291
 2.072300-3 1.986420-3 1.447110-3 5.347710-4 3.983250-6-8.023330-5452533291
 2.977010-5 1.266280-4 1.141860-4-3.476890-5 3.978040-3 1.824830-3452533291
 1.835480-3 1.643210-3 1.509900-3 1.548560-3 1.631550-3 1.715330-3452533291
 1.779860-3 1.814290-3 1.812430-3 1.770950-3 1.688910-3 1.568090-3452533291
 1.414200-3 1.238960-3 1.062830-3 9.173610-4 8.435190-4 8.807400-4452533291
 1.042860-3 1.290290-3 1.526610-3 1.640340-3 1.567330-3 1.166990-3452533291
 4.944710-4 8.494070-5-1.223270-5 3.449800-5 8.571080-5 7.622970-5452533291
-1.302230-5 3.335450-3 1.865850-3 1.320780-3 1.334180-3 1.390290-3452533291
 1.451910-3 1.509870-3 1.554570-3 1.578240-3 1.575150-3 1.541460-3452533291
 1.475190-3 1.376710-3 1.249730-3 1.103140-3 9.534420-4 8.269430-4452533291
 7.587760-4 7.840130-4 9.172220-4 1.127730-3 1.334760-3 1.441390-3452533291
 1.388440-3 1.050250-3 4.598470-4 8.317440-5-1.962000-5 1.285200-5452533291
 5.886810-5 5.874010-5-6.850400-6 2.935500-3 1.948970-3 1.404000-3452533291
 1.306000-3 1.344960-3 1.428310-3 1.513750-3 1.577960-3 1.607040-3452533291
 1.592890-3 1.531610-3 1.423490-3 1.273950-3 1.095770-3 9.119690-4452533291
 7.582380-4 6.808130-4 7.237980-4 9.020240-4 1.170360-3 1.421600-3452533291
 1.534730-3 1.444670-3 1.011040-3 3.314790-4-2.048830-5-3.774030-5452533291
 6.704390-5 1.361820-4 1.034210-4-4.532930-5 2.185350-3 1.745000-3452533291
 1.484750-3 1.385400-3 1.380060-3 1.419750-3 1.470620-3 1.509200-3452533291
 1.519370-3 1.490590-3 1.417430-3 1.300150-3 1.146670-3 9.756850-4452533291
 8.197340-4 7.246800-4 7.391190-4 8.883700-4 1.141090-3 1.399790-3452533291
 1.541910-3 1.489490-3 1.091420-3 4.000170-4-6.305970-6-6.918640-5452533291
 1.719700-5 8.992070-5 7.228520-5-5.799610-5 1.705580-3 1.595370-3452533291
 1.530820-3 1.492930-3 1.468820-3 1.447610-3 1.420890-3 1.382350-3452533291
 1.327560-3 1.254000-3 1.161690-3 1.054450-3 9.418340-4 8.412030-4452533291
 7.780520-4 7.807980-4 8.663420-4 1.019870-3 1.186120-3 1.289640-3452533291
 1.274940-3 1.030670-3 5.329380-4 1.499900-4-1.318120-5-3.406270-5452533291
-5.586790-6 1.728610-5 1.067630-5 1.634360-3 1.625160-3 1.607390-3452533291
 1.574910-3 1.528640-3 1.468740-3 1.395220-3 1.308330-3 1.209170-3452533291
 1.100440-3 9.876960-4 8.808630-4 7.955860-4 7.526840-4 7.727340-4452533291
 8.635330-4 1.004710-3 1.144140-3 1.218310-3 1.185790-3 9.517920-4452533291
 5.073870-4 1.702880-4 1.966640-5-9.743300-6 7.648470-6 2.860580-5452533291
 3.445390-5 1.686590-3 1.698690-3 1.689110-3 1.652020-3 1.589090-3452533291
 1.501910-3 1.392580-3 1.264400-3 1.123050-3 9.780430-4 8.445670-4452533291
 7.445890-4 7.049900-4 7.491080-4 8.799410-4 1.062630-3 1.225610-3452533291
 1.292150-3 1.222690-3 9.173280-4 4.205910-4 1.117470-4 2.198870-5452533291
 3.810570-5 6.754520-5 6.709010-5 2.377900-5 1.767300-3 1.787600-3452533291
 1.778600-3 1.732110-3 1.648380-3 1.529190-3 1.378700-3 1.204750-3452533291
 1.020930-3 8.488400-4 7.195540-4 6.705770-4 7.337420-4 9.118860-4452533291
 1.155540-3 1.366660-3 1.444020-3 1.340990-3 9.399110-4 3.401500-4452533291
 3.115010-5 3.701460-6 7.870690-5 1.295250-4 1.052400-4-3.667290-6452533291
 1.861870-3 1.882680-3 1.864080-3 1.796830-3 1.681110-3 1.520660-3452533291
 1.324290-3 1.108380-3 8.998800-4 7.381710-4 6.716550-4 7.426720-4452533291
 9.579980-4 1.258590-3 1.523090-3 1.624410-3 1.503210-3 1.018640-3452533291
 3.048970-4-3.417320-5-2.491860-5 9.680240-5 1.687500-4 1.281170-4452533291
-3.466650-5 1.953700-3 1.962510-3 1.921520-3 1.822400-3 1.667480-3452533291
 1.465140-3 1.232610-3 9.995790-4 8.107960-4 7.228870-4 7.879580-4452533291
 1.019900-3 1.357890-3 1.666730-3 1.798970-3 1.680060-3 1.145900-3452533291
 3.346320-4-6.084060-5-5.413060-5 8.742170-5 1.740050-4 1.296000-4452533291
-5.921130-5 2.018400-3 2.004860-3 1.934740-3 1.802360-3 1.614240-3452533291
 1.386200-3 1.147330-3 9.433540-4 8.345250-4 8.798450-4 1.102280-3452533291
 1.448040-3 1.781360-3 1.944550-3 1.848350-3 1.309570-3 4.359200-4452533291
-3.516650-5-7.449720-5 5.289760-5 1.437700-4 1.099710-4-6.921970-5452533291
 2.040740-3 2.005170-3 1.912560-3 1.760710-3 1.561740-3 1.340470-3452533291
 1.138540-3 1.013920-3 1.027960-3 1.213110-3 1.531140-3 1.860530-3452533291
 2.046820-3 1.989450-3 1.493570-3 6.047000-4 4.891560-5-7.673680-5452533291
 1.371240-6 8.420380-5 7.466270-5-5.785160-5 2.030060-3 1.988070-3452533291
 1.894760-3 1.751080-3 1.574090-3 1.396040-3 1.266060-3 1.241910-3452533291
 1.364030-3 1.616560-3 1.907010-3 2.099450-3 2.088640-3 1.678040-3452533291
 8.271910-4 1.904990-4-5.195030-5-5.417730-5 7.952040-6 3.346140-5452533291
-1.934510-5 2.026430-3 2.007020-3 1.944500-3 1.840030-3 1.711650-3452533291
 1.592980-3 1.531150-3 1.570710-3 1.721980-3 1.934050-3 2.105380-3452533291
 2.136050-3 1.839400-3 1.079100-3 3.806240-4 7.422110-6-9.667630-5452533291
-6.646090-5-7.827400-8 5.032560-5 2.093200-3 2.129750-3 2.125450-3452533291
 2.075490-3 1.991990-3 1.903760-3 1.852790-3 1.874580-3 1.967840-3452533291
 2.079970-3 2.129570-3 1.951850-3 1.324730-3 6.001410-4 1.053960-4452533291
-1.067670-4-1.161000-4-1.004330-5 1.517680-4 2.296170-3 2.410830-3452533291
 2.468920-3 2.452110-3 2.363350-3 2.232360-3 2.112150-3 2.051350-3452533291
 2.056640-3 2.080990-3 1.991540-3 1.516680-3 8.163850-4 2.383610-4452533291
-6.646880-5-1.167200-4 1.891140-5 2.788530-4 2.672770-3 2.860640-3452533291
 2.948500-3 2.902960-3 2.729070-3 2.481360-3 2.246960-3 2.093980-3452533291
 2.025970-3 1.946320-3 1.600080-3 9.808020-4 3.887410-4 3.325860-5452533291
-4.995920-5 9.539570-5 4.136770-4 3.207960-3 3.435520-3 3.495000-3452533291
 3.349050-3 3.028830-3 2.633510-3 2.280710-3 2.036640-3 1.834010-3452533291
 1.526700-3 1.031610-3 5.177390-4 1.802330-4 8.386130-5 2.111010-4452533291
 5.209780-4 3.846890-3 4.078820-3 4.065070-3 3.778340-3 3.287360-3452533291
 2.728170-3 2.229490-3 1.728800-3 1.286970-3 9.106850-4 5.629970-4452533291
 3.263650-4 2.481130-4 3.284120-4 5.472190-4 4.558340-3 4.791720-3452533291
 4.690670-3 4.237160-3 3.533920-3 2.750150-3 1.782710-3 9.594800-4452533291
 6.041830-4 4.544830-4 3.836850-4 3.591780-4 3.767060-4 4.355290-4452533291
 5.379550-3 5.621120-3 5.396270-3 4.706480-3 3.711170-3 2.207740-3452533291
 7.515130-4 2.027880-4 1.638560-4 2.563050-4 3.091530-4 2.787680-4452533291
 1.688620-4 6.326250-3 6.518650-3 6.081040-3 5.085380-3 3.181080-3452533291
 9.695790-4-6.047840-5-2.210550-4-7.551840-5 4.549650-5 2.181970-5452533291
-1.665760-4 7.262800-3 7.312700-3 6.627790-3 4.695680-3 1.873840-3452533291
 1.345200-4-4.562920-4-4.594530-4-3.217850-4-2.598870-4-3.491170-4452533291
 8.021830-3 7.929730-3 6.477870-3 3.489070-3 1.036030-3-2.102110-4452533291
-5.656340-4-4.887160-4-2.780690-4-8.214510-5 8.624160-3 8.079560-3452533291
 5.538600-3 2.666940-3 7.744250-4-2.478930-5-7.501780-5 2.929000-4452533291
 8.666250-4 9.222250-3 8.299370-3 5.838530-3 3.614610-3 2.386360-3452533291
 2.132820-3 2.604510-3 3.570530-3 1.032610-2 9.947210-3 8.723800-3452533291
 7.726450-3 7.373060-3 7.680650-3 8.542170-3 1.246280-2 1.317360-2452533291
 1.318840-2 1.302470-2 1.298030-2 1.315830-2 1.597610-2 1.695770-2452533291
 1.713040-2 1.683090-2 1.633520-2 1.899860-2 1.933620-2 1.899850-2452533291
 1.818580-2 2.027880-2 1.992020-2 1.919880-2 2.027840-2 1.978060-2452533291
 2.040580-2                                                       452533291
                                                                  452533  0
                                                                  4525 0  0
                                                                     0 0  0
 4.71090E+4 1.07969E+2          0          0         34         104731 1451
 0.0        0.0                 0          0          0          64731 1451
 1.00000E+0 2.00000E+7          0          0         10       20024731 1451
 3.00000E+2 0.0                 1          0        138          44731 1451
 47-Ag-109 CNDC       EVAL-JUN91 Z.X.ZHAO                         4731 1451
                      DIST-FEB2004                                4731 1451
----IRDF-2002         MATERIAL 4731                               4731 1451
-----INCIDENT NEUTRON DATA                                        4731 1451
------ENDF-6 FORMAT                                               4731 1451
 ================================================================ 4731 1451
 47-AG-109 CNDC       EVAL-JUN91 Z.X.ZHAO                         4731 1451
                      DIST-                                       4731 1451
----ENDF/A-6          MATERIAL 4731                               4731 1451
-----INCIDENT NEUTRON DATA                                        4731 1451
------ENDF-6 FORMAT                                               4731 1451
 *         *          *          *          *          *         *4731 1451
  THIS  EVALUATION WAS PERFORMED BASED ON THE MEASURED DATA       4731 1451
  AVAILABLE AND THE CALCULATIONS OF THE SYSTEMATICS AND THEORY.   4731 1451
     FILE 2    RESONANCE PARAMETER                                4731 1451
       ONLY POTENTIAL SCATTERING RADIUS R' OF REF.[1] IS GIVEN.   4731 1451
     FILE 10   PRODUCTION CROSS SECTION OF RADIOACTIVE NUCLIDE    4731 1451
       MT=102   109-AG(N,GAMMA)110M-AG REACTION                   4731 1451
          THE PRODUCTION CROSS SECTION OF ISOMER STATE OF 110-AG  4731 1451
       ARE OBTAINED BY                                            4731 1451
                    C.S.(M) = C.S.(M+G)*R                         4731 1451
       WHERE C.S.(M+G) REPRESENTS THE TOTAL CAPTURE CROSS SECTION 4731 1451
       AND R IS THE RATIO OF THE CAPTURE CROSS SECTION TO ISOMER  4731 1451
       STATE AND THE TOTAL CAPTURE CROSS SECTION. THE VALUES OF   4731 1451
       R WERE CALCULATED BY USING UNIFY2[2] AND NORMALIZED TO     4731 1451
       THE LIMITED DATA FOR C.S.(M) MEASURED AT THERMAL CROSS     4731 1451
       SECTION [3-5] AND AROUND 25 KEV [6-7].                     4731 1451
          C.S.(M+G) WAS RE-EVALUATED IN THE FOLLOWING WAY:        4731 1451
          BELOW 2.5 KEV, C.S.(M+G) WAS CALCULATED FROM A SET      4731 1451
       OF RESONANCE PARAMETERS OF JENDL-3 AND AVERAGED IN         4731 1451
       SMALL ENERGY INTERVALS.                                    4731 1451
          FROM 2.5 KEV TO 20 MEV, THE SYSTEMATICS OF ZHAO [8]     4731 1451
       WAS USED TO FIT THE MEASURED DATA OF REFS.[9-10] IN THE    4731 1451
       ENERGY RANGE FROM 2.5 KEV TO 2 MEV.                        4731 1451
     FILE 8  RADIOACTIVE DECAY DATA                               4731 1451
       THE RADIOACTIVE DECAY DATA FOR 110M-AG ARE TAKEN FROM      4731 1451
       REFS.[11] AND [12].                                        4731 1451
     FILE 40  COVARIANCE DATA FOR FILE 10                         4731 1451
       COVARIANCE DATA OF FILE 10 CONSISTES OF TWO COMPONENTS.    4731 1451
       ONE OF THEM COME FROM THE UNCERTAINTY OF TOTAL CAPTURE     4731 1451
       CROSS SECTION. ANOTHER COME FROM THE NORMALIZATION ERROR   4731 1451
       OF R=C.S.(M)/C.S.(M+G).                                    4731 1451
 ================================================================ 4731 1451
 =========== Processing done at IAEA/NDS for IRDF-2002 ========== 4731 1451
 ================================================================ 4731 1451
 The original file was calculated from a set of resonance         4731 1451
 parameters taken from JENDL-3.                                   4731 1451
 The JENDL-3.2 file for Ag109 was processed as follows to produce 4731 1451
 a pointwise file.                                                4731 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************4731 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     4731 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4731 1451
 ***************** Program RECENT (VERSION 2002-1) ***************4731 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4731 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     4731 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4731 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4731 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4731 1451
 Data Linearized to Within an Accuracy pf 1.00000000 per-cent     4731 1451
 ***************** Program FIXUP (Version 2002-1) ****************4731 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451
 Corrected Thresholds-----------------------------------------No  4731 1451
 Extended Cross Sections to 20 MeV----------------------------No  4731 1451
 Allow Cross Section Deletion---------------------------------No  4731 1451
 Allow Cross Section Reconstruction---------------------------No  4731 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451
 Deleted Duplicate Points-------------------------------------Yes 4731 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451
 Allow Creation of Missing Sections---------------------------No  4731 1451
 Allow Insertion of Energy Points-----------------------------No  4731 1451
 Create Uniform Energy Grid-----------------------------------No  4731 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451
                                                                  4731 1451
 The branching ratio of 0.0465 was applied to the Ag109 total     4731 1451
 capture cross section data of the JENDL-3.2 file in the energy   4731 1451
 range 1.0E-5 to 7.1E+3. The scaled data with the above energy    4731 1451
 range was then edited into the original data file as used for    4731 1451
 IRDF-90                                                          4731 1451
 ================================================================ 4731 1451
                                                                  4731 1451
       REFERENCES                                                 4731 1451
    [1]  S. F. Mughabghab et al., BNL-325, 4th ed., Vol.(1981)    4731 1451
                                                                  4731 1451
    [2]  Zhang Jinshang, " Semi-classical Theory Code UNIFY of    4731 1451
                 Multi-step Nuclear Reaction", to be published    4731 1451
                                                                  4731 1451
    [3]  T. B. Ryves et al., J. Nucl. Energy, 25(1971)129         4731 1451
                                                                  4731 1451
    [4]  J. K. Aaldijk et al., RCN-176,1972                       4731 1451
                                                                  4731 1451
    [5]  A. Simonits et al., JRN, 81(1984)369                     4731 1451
                                                                  4731 1451
    [6]  M. Sriramachandra et al., JP/A,5(1972)877                4731 1451
                                                                  4731 1451
    [7]  W. Poenitz, EANDC(E0-66,6(1966)                          4731 1451
                                                                  4731 1451
    [8]  Zhao Zhixiang et al., Chinese Nucl. Phys., 11(1989)71    4731 1451
                                                                  4731 1451
    [9]  R. L. Macklin, Nucl. Sci. and Eng., 82(1982)400          4731 1451
                                                                  4731 1451
    [10] M. Mizumoto et al., "Neutron Radiative Capture and Transm4731 1451
         Measurements of 107-Ag and 109-Ag", Proc. Int. Conf. Nucl4731 1451
         for Sci. and Tech., Antwerp, 1982, p.226(1983)           4731 1451
                                                                  4731 1451
    [11] J. K. Tuli, Nuclear Wallet Cards ,1990                   4731 1451
                                                                  4731 1451
    [12] C. M. Lederer and V. S. shirley, Table of Isotopes,      4731 1451
                              Seventh Edition(1978)               4731 1451
                                                                  4731 1451
 *****************************************************************4731 1451
                                                                  4731 1451
                                                                  4731 1451
                                                                  4731 1451
                                                                  4731 1451
 ***************** PROGRAM LINEAR (VERSION 96-1) *************    4731 1451
 FOR ALL DATA GREATER THAN 1.00000-10 BARNS IN ABSOLUTE VALUE     4731 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.10000000 PER-CENT     4731 1451
 ***************** Program FIXUP (Version 2002-1) ****************4731 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451
 Corrected Thresholds-----------------------------------------Yes 4731 1451
 Extended Cross Sections to 20 MeV----------------------------No  4731 1451
 Allow Cross Section Deletion---------------------------------No  4731 1451
 Allow Cross Section Reconstruction---------------------------No  4731 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451
 Deleted Duplicate Points-------------------------------------Yes 4731 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451
 Allow Creation of Missing Sections---------------------------No  4731 1451
 Allow Insertion of Energy Points-----------------------------No  4731 1451
 Create Uniform Energy Grid-----------------------------------No  4731 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451
 *****************************************************************4731 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4731 1451
 Unshielded Group Averages Using  640 Groups                      4731 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4731 1451
                                1        451        146          14731 1451
                                8        102          2          14731 1451
                               10        102        217          14731 1451
                               40        102         10          14731 1451
                                                                  4731 1  0
                                                                  4731 0  0
 4.71090E+4 1.07969E+2          0          0          1          04731 3293
 6.80570E+6 6.68810E+6      47110          2          1        6414731 3293
        641          1                                            4731 3293
 .000100000 65.8829137 .000105000 64.3718842 .000110000 62.86085474731 3293
 .000115000 61.3498263 .000120000 59.7673282 .000127500 58.20321084731 3293
 .000135000 56.6406830 .000142500 55.0781536 .000150000 53.37191724731 3293
 .000160000 51.8716541 .000170000 50.4199466 .000180000 48.96823914731 3293
 .000190000 47.5717224 .000200000 46.5123774 .000210000 45.51024454731 3293
 .000220000 44.5081115 .000230000 43.5059794 .000240000 42.26130974731 3293
 .000255000 41.0758325 .000270000 40.2204986 .000280000 39.19409774731 3293
 .000300000 37.8255648 .000320000 36.5868623 .000340000 35.63353004731 3293
 .000360000 34.6903705 .000380000 33.7472116 .000400000 32.71870274731 3293
 .000425000 31.8382357 .000450000 31.0164125 .000475000 30.19458894731 3293
 .000500000 29.3727662 .000525000 28.6779698 .000550000 28.11072844731 3293
 .000575000 27.5434871 .000600000 26.9195219 .000630000 26.23883294731 3293
 .000660000 25.5726206 .000690000 25.0645516 .000720000 24.52244004731 3293
 .000760000 23.9028841 .000800000 23.2833283 .000840000 22.66658764731 3293
 .000880000 22.1733049 .000920000 21.7461036 .000960000 21.31890204731 3293
 .001000000 20.8383005 .001050000 20.3042989 .001100000 19.80933414731 3293
 .001150000 19.4297664 .001200000 18.9647759 .001275000 18.40678754731 3293
 .001350000 17.8509108 .001425000 17.4043156 .001500000 16.95511074731 3293
 .001600000 16.4417334 .001700000 15.9283562 .001800000 15.45742554731 3293
 .001900000 15.1005702 .002000000 14.7500021 .002100000 14.39943394731 3293
 .002200000 14.0488654 .002300000 13.7117119 .002400000 13.39584104731 3293
 .002550000 13.0333777 .002700000 12.7313251 .002800000 12.36886184731 3293
 .003000000 11.9573379 .003200000 11.6207172 .003400000 11.28410504731 3293
 .003600000 10.9474928 .003800000 10.6608505 .004000000 10.39952984731 3293
 .004250000 10.1092003 .004500000 9.81887067 .004750000 9.533245344731 3293
 .005000000 9.31366335 .005250000 9.11538011 .005500000 8.917096864731 3293
 .005750000 8.71881361 .006000000 8.50070207 .006300000 8.293501074731 3293
 .006600000 8.12938064 .006900000 7.96555528 .007200000 7.774425614731 3293
 .007600000 7.55599162 .008000000 7.35868335 .008400000 7.205416584731 3293
 .008800000 7.05352267 .009200000 6.90162877 .009600000 6.749735044731 3293
 .010000000 6.58673552 .010500000 6.44858051 .011000000 6.317183784731 3293
 .011500000 6.18578706 .012000000 6.02154121 .012750000 5.828134244731 3293
 .013500000 5.68058330 .014250000 5.54674958 .015000000 5.390610104731 3293
 .016000001 5.21216495 .017000001 5.06131471 .017999999 4.941558864731 3293
 .018999999 4.82183021 .020000000 4.70210156 .021000000 4.582372914731 3293
 .022000000 4.46564013 .023000000 4.37180633 .024000000 4.260776584731 3293
 .025500000 4.14784660 .027000001 4.06409209 .028000001 3.963586784731 3293
 .029999999 3.82957962 .032000002 3.70540709 .034000002 3.616339924731 3293
 .035999998 3.53039126 .037999999 3.44444251 .039999999 3.347750154731 3293
 .042500000 3.24031417 .045000002 3.14491372 .047499999 3.074900804731 3293
 .050000001 3.00570398 .052499998 2.93650715 .055000000 2.867310284731 3293
 .057500001 2.80242435 .059999999 2.75162507 .063000001 2.698517274731 3293
 .066000000 2.64540951 .068999998 2.59230175 .071999997 2.530342684731 3293
 .075999998 2.45953229 .079999998 2.39456404 .083999999 2.350058724731 3293
 .088000000 2.30732082 .092000000 2.26458293 .096000001 2.221845034731 3293
 .100000001 2.17376493 .104999997 2.12034258 .109999999 2.068221254731 3293
 .115000002 2.03171845 .119999997 1.99179873 .127499998 1.943895024731 3293
 .135000005 1.89599133 .142499998 1.84808765 .150000006 1.792981224731 3293
 .159999996 1.74774667 .170000002 1.70994680 .180000007 1.672146964731 3293
 .189999998 1.63434712 .200000003 1.59654728 .209999993 1.558802044731 3293
 .219999999 1.53010801 .230000004 1.50800998 .239999995 1.480387454731 3293
 .254999995 1.44724039 .270000011 1.41961785 .280000001 1.386470804731 3293
 .300000012 1.34778961 .319999993 1.32230623 .340000004 1.297395754731 3293
 .360000014 1.27248529 .379999995 1.24757483 .400000006 1.219559984731 3293
 .425000012 1.19624620 .449999988 1.17972545 .474999994 1.163204694731 3293
 .500000000 1.14668395 .524999976 1.13016319 .550000012 1.113642434731 3293
 .574999988 1.09786761 .600000024 1.08820482 .629999995 1.079466694731 3293
 .660000026 1.07072857 .689999998 1.06199044 .720000029 1.051795954731 3293
 .759999990 1.04014512 .800000012 1.02949480 .839999974 1.025266604731 3293
 .879999995 1.02218176 .920000017 1.01909693 .959999979 1.016012094731 3293
 1.00000000 1.01258668 1.04999995 1.01406959 1.10000002 1.019183864731 3293
 1.14999998 1.02429814 1.20000005 1.03069099 1.27499998 1.038362404731 3293
 1.35000002 1.04603381 1.42499995 1.05634426 1.50000000 1.080408894731 3293
 1.60000002 1.10906368 1.70000005 1.13771845 1.79999995 1.167983984731 3293
 1.89999998 1.21190394 2.00000000 1.25892606 2.09999990 1.306404274731 3293
 2.20000005 1.36744349 2.29999995 1.43479879 2.40000010 1.526611894731 3293
 2.54999995 1.66486967 2.70000005 1.79066902 2.79999995 1.989481744731 3293
 3.00000000 2.30330077 3.20000005 2.72493186 3.40000010 3.316873394731 3293
 3.59999990 4.15243584 3.79999995 5.40661480 4.00000000 7.767223064731 3293
 4.25000000 13.0174447 4.50000000 27.1024049 4.75000000 96.88064274731 3293
 5.00000000 725.872816 5.25000000 308.558951 5.50000000 43.47225764731 3293
 5.75000000 16.1345931 6.00000000 7.94958909 6.30000019 4.455170674731 3293
 6.59999990 2.82372470 6.90000010 1.94246590 7.19999981 1.347709334731 3293
 7.59999990 .939412905 8.00000000 .689954724 8.39999962 .5248344064731 3293
 8.80000019 .412945385 9.19999981 .331773722 9.60000038 .2725911774731 3293
 10.0000000 .222468516 10.5000000 .181194952 11.0000000 .1502761294731 3293
 11.5000000 .126562773 12.0000000 .104511803 12.7500000 .0847258484731 3293
 13.5000000 .070440364 14.2500000 .059786029 15.0000000 .0505011334731 3293
 16.0000000 .042824923 17.0000000 .037420544 18.0000000 .0336902044731 3293
 19.0000000 .031168806 20.0000000 .029716841 21.0000000 .0294777274731 3293
 22.0000000 .030441135 23.0000000 .032962909 24.0000000 .0398415514731 3293
 25.5000000 .059888636 27.0000000 .105620350 28.0000000 .8699901974731 3293
 30.0000000 15.2751085 32.0000000 .138692160 34.0000000 .0607068904731 3293
 36.0000000 .079835222 38.0000000 3.23543961 40.0000000 5.838554654731 3293
 42.5000000 .052164789 45.0000000 .027849228 47.5000000 .0285981794731 3293
 50.0000000 .050652422 52.5000000 .354832624 55.0000000 9.537713244731 3293
 57.5000000 .102404270 60.0000000 .037831276 63.0000000 .0401934224731 3293
 66.0000000 .129269498 69.0000000 14.2435132 72.0000000 .1701945014731 3293
 76.0000000 .025944830 80.0000000 .018358104 84.0000000 2.090237064731 3293
 88.0000000 .354015421 92.0000000 .010773369 96.0000000 .0059440804731 3293
 100.000000 .005223153 105.000000 .031651345 110.000000 .0243531074731 3293
 115.000000 .008734952 120.000000 .021522423 127.500000 6.794720544731 3293
 135.000000 .412661156 142.500000 .015144510 150.000000 .0076705254731 3293
 160.000000 .042359786 170.000000 2.99459687 180.000000 .0089370114731 3293
 190.000000 .004496820 200.000000 1.18874814 210.000000 .2167235434731 3293
 220.000000 .002685375 230.000000 .002016628 240.000000 .4141073594731 3293
 255.000000 .074072058 270.000000 .070369378 280.000000 .2364871814731 3293
 300.000000 1.59099734 320.000000 .123750237 340.000000 .0040743924731 3293
 360.000000 .003670073 380.000000 .980494669 400.000000 .3900433674731 3293
 425.000000 .133332857 450.000000 .288156933 475.000000 .2733210544731 3293
 500.000000 .769591133 525.000000 .010301960 550.000000 .9797457704731 3293
 575.000000 .003933997 600.000000 .409036749 630.000000 .0067767814731 3293
 660.000000 .147393373 690.000000 .007641197 720.000000 .3092908884731 3293
 760.000000 .194711821 800.000000 .062692150 840.000000 .0593751734731 3293
 880.000000 .158772525 920.000000 .154912223 960.000000 .1738306654731 3293
 1000.00000 .137231276 1050.00000 .196444584 1100.00000 .1220674564731 3293
 1150.00000 .020582762 1200.00000 .350942078 1275.00000 .1661356294731 3293
 1350.00000 .257040498 1425.00000 .180871517 1500.00000 .2057077464731 3293
 1600.00000 .088162424 1700.00000 .139195318 1800.00000 .1025837144731 3293
 1900.00000 .136826314 2000.00000 .102542574 2100.00000 .1501113054731 3293
 2200.00000 .065128230 2300.00000 .199188779 2400.00000 .0959195764731 3293
 2550.00000 .126391554 2700.00000 .136546956 2800.00000 .0980007664731 3293
 3000.00000 .130069272 3200.00000 .121252881 3400.00000 .0652732104731 3293
 3600.00000 .075219147 3800.00000 .106302348 4000.00000 .0967330254731 3293
 4250.00000 .055626501 4500.00000 .070037029 4750.00000 .0773165294731 3293
 5000.00000 .074751017 5250.00000 .085971376 5500.00000 .0744562554731 3293
 5750.00000 .058042135 6000.00000 .070744369 6300.00000 .0870494064731 3293
 6600.00000 .067582619 6900.00000 .069959010 7200.00000 .0747916524731 3293
 7600.00000 .074231182 8000.00000 .073148724 8400.00000 .0721299794731 3293
 8800.00000 .071140112 9200.00000 .070250536 9600.00000 .0693695124731 3293
 10000.0000 .068723633 10500.0000 .068312899 11000.0000 .0679021654731 3293
 11500.0000 .067492934 12000.0000 .067040371 12750.0000 .0665211194731 3293
 13500.0000 .066040832 14250.0000 .065609035 15000.0000 .0651054374731 3293
 16000.0000 .064546652 17000.0000 .064073090 18000.0000 .0636114544731 3293
 19000.0000 .063149818 20000.0000 .062644853 21000.0000 .0620965594731 3293
 22000.0000 .061548265 23000.0000 .061002080 24000.0000 .0604020404731 3293
 25500.0000 .059715384 27000.0000 .059167431 28000.0000 .0585677864731 3293
 30000.0000 .057829890 32000.0000 .057138859 34000.0000 .0564765414731 3293
 36000.0000 .055891812 38000.0000 .055311404 40000.0000 .0547100904731 3293
 42500.0000 .054088351 45000.0000 .053513049 47500.0000 .0529683504731 3293
 50000.0000 .052104461 52500.0000 .050922003 55000.0000 .0498265864731 3293
 57500.0000 .048792862 60000.0000 .047730372 63000.0000 .0466484084731 3293
 66000.0000 .045653395 69000.0000 .044709909 72000.0000 .0437292994731 3293
 76000.0000 .042689774 80000.0000 .041756772 84000.0000 .0409255174731 3293
 88000.0000 .040159895 92000.0000 .039515200 96000.0000 .0388904004731 3293
 100000.000 .038201172 105000.000 .037447516 110000.000 .0367273044731 3293
 115000.000 .036062543 120000.000 .035304204 127500.000 .0344349894731 3293
 135000.000 .033647943 142500.000 .032907314 150000.000 .0321714544731 3293
 160000.000 .031472932 170000.000 .030825191 180000.000 .0302258504731 3293
 190000.000 .029666617 200000.000 .029075555 210000.000 .0284526644731 3293
 220000.000 .027858101 230000.000 .027310364 240000.000 .0266703324731 3293
 255000.000 .025971561 270000.000 .025417209 280000.000 .0248164134731 3293
 300000.000 .024019543 320000.000 .023245088 340000.000 .0225496894731 3293
 360000.000 .021902956 380000.000 .021313132 400000.000 .0206723904731 3293
 425000.000 .020014329 450000.000 .019400632 475000.000 .0188435564731 3293
 500000.000 .018260961 525000.000 .017678470 550000.000 .0171309374731 3293
 575000.000 .016628575 600000.000 .016110634 630000.000 .0155771424731 3293
 660000.000 .015092685 690000.000 .014631422 720000.000 .0141386414731 3293
 760000.000 .013621142 800000.000 .013143615 840000.000 .0127042984731 3293
 880000.000 .012298597 920000.000 .011922757 960000.000 .0115741584731 3293
 1000000.00 .010950648 1100000.00 .010136679 1200000.00 .0094413984731 3293
 1300000.00 .008843603 1400000.00 .008322346 1500000.00 .0078414004731 3293
 1600000.00 .007396664 1700000.00 .006998287 1800000.00 .0066447454731 3293
 1900000.00 .006325335 2000000.00 .005946862 2100000.00 .0055292094731 3293
 2200000.00 .005155062 2300000.00 .004821228 2400000.00 .0045213674731 3293
 2500000.00 .004251876 2600000.00 .004008317 2700000.00 .0037865304731 3293
 2800000.00 .003584598 2900000.00 .003399992 3000000.00 .0031123784731 3293
 3100000.00 .002752170 3200000.00 .002443123 3300000.00 .0021767274731 3293
 3400000.00 .001946699 3500000.00 .001745744 3600000.00 .0015703064731 3293
 3700000.00 .001417183 3800000.00 .001281677 3900000.00 .0011626124731 3293
 4000000.00 .001026912 4100000.00 .000883232 4200000.00 .0007621754731 3293
 4300000.00 .000660025 4400000.00 .000573415 4500000.00 .0004999414731 3293
 4600000.00 .000436975 4700000.00 .000383144 4800000.00 .0003367624731 3293
 4900000.00 .000296892 5000000.00 .000267552 5100000.00 .0002466744731 3293
 5200000.00 .000227835 5300000.00 .000210658 5400000.00 .0001950504731 3293
 5500000.00 .000180905 5600000.00 .000168025 5700000.00 .0001561924731 3293
 5800000.00 .000145436 5900000.00 .000135566 6000000.00 .0001316554731 3293
 6100000.00 .000133265 6200000.00 .000134875 6300000.00 .0001364854731 3293
 6400000.00 .000138095 6500000.00 .000139705 6600000.00 .0001413154731 3293
 6700000.00 .000142925 6800000.00 .000144535 6900000.00 .0001461454731 3293
 7000000.00 .000149313 7100000.00 .000154038 7200000.00 .0001588124731 3293
 7300000.00 .000163765 7400000.00 .000168735 7500000.00 .0001737054731 3293
 7600000.00 .000178803 7700000.00 .000184081 7800000.00 .0001893614731 3293
 7900000.00 .000194640 8000000.00 .000199311 8100000.00 .0002033724731 3293
 8200000.00 .000207434 8300000.00 .000211496 8400000.00 .0002155594731 3293
 8500000.00 .000219716 8600000.00 .000223928 8700000.00 .0002281404731 3293
 8800000.00 .000232352 8900000.00 .000236564 9000000.00 .0002396334731 3293
 9100000.00 .000241558 9200000.00 .000243483 9300000.00 .0002454084731 3293
 9400000.00 .000247333 9500000.00 .000249258 9600000.00 .0002511834731 3293
 9700000.00 .000253108 9800000.00 .000255033 9900000.00 .0002569584731 3293
 10000000.0 .000257529 10100000.0 .000256747 10200000.0 .0002559654731 3293
 10300000.0 .000255183 10400000.0 .000254401 10500000.0 .0002536194731 3293
 10600000.0 .000252837 10700000.0 .000252055 10800000.0 .0002512734731 3293
 10900000.0 .000250491 11000000.0 .000249258 11100000.0 .0002475744731 3293
 11200000.0 .000245890 11300000.0 .000244206 11400000.0 .0002425234731 3293
 11500000.0 .000240915 11600000.0 .000239356 11700000.0 .0002377974731 3293
 11800000.0 .000236238 11900000.0 .000234679 12000000.0 .0002328994731 3293
 12100000.0 .000230898 12200000.0 .000228897 12300000.0 .0002268964731 3293
 12400000.0 .000224896 12500000.0 .000222990 12600000.0 .0002211484731 3293
 12700000.0 .000219305 12800000.0 .000217463 12900000.0 .0002156214731 3293
 13000000.0 .000213685 13100000.0 .000211655 13200000.0 .0002096254731 3293
 13300000.0 .000207595 13400000.0 .000205566 13500000.0 .0002036334731 3293
 13600000.0 .000201768 13700000.0 .000199903 13800000.0 .0001980384731 3293
 13900000.0 .000196173 14000000.0 .000194206 14100000.0 .0001921394731 3293
 14200000.0 .000190072 14300000.0 .000188004 14400000.0 .0001859384731 3293
 14500000.0 .000183971 14600000.0 .000182078 14700000.0 .0001801844731 3293
 14800000.0 .000178290 14900000.0 .000176397 15000000.0 .0001745074731 3293
 15100000.0 .000172620 15200000.0 .000170734 15300000.0 .0001688474731 3293
 15400000.0 .000166961 15500000.0 .000165165 15600000.0 .0001634354731 3293
 15700000.0 .000161705 15800000.0 .000159975 15900000.0 .0001582454731 3293
 16000000.0 .000156534 16100000.0 .000154842 16200000.0 .0001531514731 3293
 16300000.0 .000151459 16400000.0 .000149768 16500000.0 .0001481554731 3293
 16600000.0 .000146600 16700000.0 .000145046 16800000.0 .0001434924731 3293
 16900000.0 .000141937 17000000.0 .000140408 17100000.0 .0001389034731 3293
 17200000.0 .000137399 17300000.0 .000135895 17400000.0 .0001343914731 3293
 17500000.0 .000132954 17600000.0 .000131569 17700000.0 .0001301834731 3293
 17800000.0 .000128798 17900000.0 .000127413 18000000.0 .0001260534731 3293
 18100000.0 .000124720 18200000.0 .000123387 18300000.0 .0001220534731 3293
 18400000.0 .000120720 18500000.0 .000119445 18600000.0 .0001182154731 3293
 18700000.0 .000116985 18800000.0 .000115755 18900000.0 .0001145254731 3293
 19000000.0 .000113321 19100000.0 .000112142 19200000.0 .0001109634731 3293
 19300000.0 .000109784 19400000.0 .000108605 19500000.0 .0001074764731 3293
 19600000.0 .000106386 19700000.0 .000105296 19800000.0 .0001042054731 3293
 19900000.0 .000103115 20000000.0 0.0                             4731 3293
                                                                  4731 3  0
                                                                  4731 0  0
 4.71090E+4 1.07969E+2          0          0          0          1473133293
 0.0000E+00 0.0000E+00          0        293          0          2473133293
 0.0000E+00 0.0000E+00          0          1         22         11473133293
 1.000000-5 1.000000-4 1.000000+0 2.500000-3 8.000000+1 1.000000-2473133293
 1.500000+2 2.250000-2 3.000000+2 2.250000-2 4.000000+2 2.250000-2473133293
 5.000000+2 9.000000-2 6.500000+2 1.000000-2 2.500000+4 2.500000-3473133293
 2.000000+6 9.000000-2 2.000000+7 0.000000+0                      473133293
 0.000000+0 0.000000+0          0          1          4          2473133293
 1.000000-5 2.500000-3 2.000000+7 0.000000+0                      473133293
                                                                  473133  0
                                                                  4731 0  0
                                                                     0 0  0
 4.80000E+4 1.11445E+2          0          0         34         104800 1451
 0.0        0.0                 0          0          0          64800 1451
 1.00000E+0 2.00000E+7          0          0         10       20024800 1451
 3.00000E+2 0.0                 1          0         77          24800 1451
 48-Cd-  0 BNL        EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.)    4800 1451
                      DIST-Feb2004                                4800 1451
----IRDF-2002         MATERIAL 4800                               4800 1451
-----INCIDENT NEUTRON DATA                                        4800 1451
------ENDF-6 FORMAT                                               4800 1451
 *****************************************************************4800 1451
 48-CD-  0 BNL        EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.)    4800 1451
                      DIST-JAN90                       19900130   4800 1451
----ENDF/B-VI         MATERIAL 4800                               4800 1451
-----INCIDENT NEUTRON DATA                                        4800 1451
------ENDF-6 FORMAT                                               4800 1451
 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC             4800 1451
 *         *          *          *          *          *         *4800 1451
 UKNDL TO ENDF/B TRANSLATION FROM                                 4800 1451
 UKNDL DATA FILE  70  RECEIVED JULY 1973 FROM CCDN                4800 1451
                                                                  4800 1451
  TRANSLATION OF UKNDL DFN-70 INTO ENDF FORMAT                    4800 1451
                                                                  4800 1451
  RESONANCE REGION EVALUATED BY MF JAMES IN 1969                  4800 1451
  FICTITIOUS CD-111 RESONANCE AT -0.5 EV TO ADJUST CROSS SECTION  4800 1451
  IN EV RANGE AND ACHIEVE CONSISTENCY WITH 20 BARN CD-111 THERMAL 4800 1451
  CAPTURE CROSS SECTION (BALDOCK ETAL 1966 ORNL-3994)             4800 1451
  ABOVE THE RESOLVED RESONANCE RANGE MAINLY DRAKE GA-6997 WAS USED4800 1451
                                                                  4800 1451
 PLEASE REFER COMMENTS OR QUESTIONS ABOUT THE TRANSLATION         4800 1451
 CODE (UKE*** ORNL-TM-2880 OR ENDF-134) TO                        4800 1451
 R. Q. WRIGHT, OAK RIDGE NATIONAL LABORATORY,                     4800 1451
 P. O. BOX X, OAK RIDGE, TENNESSEE 37830.                         4800 1451
                                                                  4800 1451
 RANGE EXTENDED FROM 1.0E-5 EV TO 20 MEV AND ELASTIC SCATTERING   4800 1451
 TRANSFER AND TRANSFORMATION DATA ADDED BY NNCSC AT BNL           4800 1451
                                0          0          0           4800 1451
 ***************** PROGRAM LINEAR (VERSION 89-1) *****************4800 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.300 PER-CENT      4800 1451
 ***************** PROGRAM FIXUP (VERSION 89-2) ******************4800 1451
 *RECONSTRUCTED MT NUMBERS                                        4800 1451
   4 =+( 50, 91)                                                  4800 1451
 103 =+(600,649)                                                  4800 1451
 104 =+(650,699)                                                  4800 1451
 105 =+(700,749)                                                  4800 1451
 106 =+(750,799)                                                  4800 1451
 107 =+(800,849)                                                  4800 1451
 101 =+( 18, 18)+(102,116)                                        4800 1451
  27 =+(101,101)                                                  4800 1451
   3 =+(  4,  5)+( 16, 17)+( 22, 37)+( 41, 42)                    4800 1451
  19 =+( 18, 18)-( 20, 21)-( 38, 38)                              4800 1451
   1 =+(  2,  3)                                                  4800 1451
 *****************************************************************4800 1451
                                                                  4800 1451
                                                                  4800 1451
                                                                  4800 1451
                                                                  4800 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************4800 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     4800 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4800 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4800 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4800 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4800 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     4800 1451
 ***************** Program FIXUP (Version 2002-1) ****************4800 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4800 1451
 Corrected Thresholds-----------------------------------------Yes 4800 1451
 Extended Cross Sections to 20 MeV----------------------------No  4800 1451
 Allow Cross Section Deletion---------------------------------No  4800 1451
 Allow Cross Section Reconstruction---------------------------No  4800 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4800 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4800 1451
 Deleted Duplicate Points-------------------------------------Yes 4800 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4800 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4800 1451
 Allow Creation of Missing Sections---------------------------No  4800 1451
 Allow Insertion of Energy Points-----------------------------No  4800 1451
 Create Uniform Energy Grid-----------------------------------No  4800 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4800 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4800 1451
 Unshielded Group Averages Using  640 Groups                      4800 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4800 1451
                                1        451         83          04800 1451
                                3          1        217          04800 1451
                                                                  4800 1  0
                                                                  4800 0  0
 4.80000E+4 1.11445E+2          0          0          0          04800 3  1
 0.0        0.0                 0          0          1        6414800 3  1
        641          1                                            4800 3  1
 .000100000 29407.0831 .000105000 28722.2715 .000110000 28071.69934800 3  1
 .000115000 27472.1873 .000120000 26771.4577 .000127500 25994.10844800 3  1
 .000135000 25283.8161 .000142500 24631.1861 .000150000 23924.94924800 3  1
 .000160000 23189.2717 .000170000 22519.9646 .000180000 21905.64114800 3  1
 .000190000 21334.1233 .000200000 20812.4749 .000210000 20321.70614800 3  1
 .000220000 19865.8497 .000230000 19442.4426 .000240000 18945.56044800 3  1
 .000255000 18399.5789 .000270000 17978.3029 .000280000 17509.39754800 3  1
 .000300000 16937.7319 .000320000 16420.3040 .000340000 15946.50474800 3  1
 .000360000 15510.1590 .000380000 15112.0336 .000400000 14697.04614800 3  1
 .000425000 14272.5001 .000450000 13884.8413 .000475000 13529.48074800 3  1
 .000500000 13195.3428 .000525000 12890.1169 .000550000 12601.86864800 3  1
 .000575000 12333.8926 .000600000 12059.2969 .000630000 11776.87624800 3  1
 .000660000 11518.6337 .000690000 11270.9767 .000720000 11007.59514800 3  1
 .000760000 10725.2895 .000800000 10463.2947 .000840000 10223.37704800 3  1
 .000880000 9995.42393 .000920000 9786.52876 .000960000 9586.559444800 3  1
 .001000000 9379.48279 .001050000 9163.38923 .001100000 8960.890154800 3  1
 .001150000 8774.47714 .001200000 8554.69557 .001275000 8313.263874800 3  1
 .001350000 8092.81498 .001425000 7887.24877 .001500000 7668.998014800 3  1
 .001600000 7441.12967 .001700000 7233.33162 .001800000 7041.324024800 3  1
 .001900000 6866.22705 .002000000 6704.03241 .002100000 6551.963094800 3  1
 .002200000 6412.97843 .002300000 6280.18478 .002400000 6128.854704800 3  1
 .002550000 5961.02976 .002700000 5829.97365 .002800000 5687.805354800 3  1
 .003000000 5513.21311 .003200000 5354.73926 .003400000 5210.126324800 3  1
 .003600000 5078.63908 .003800000 4957.52520 .004000000 4831.590394800 3  1
 .004250000 4704.11601 .004500000 4588.10821 .004750000 4479.985594800 3  1
 .005000000 4381.18708 .005250000 4288.67770 .005500000 4205.248774800 3  1
 .005750000 4127.02585 .006000000 4043.37604 .006300000 3961.529854800 3  1
 .006600000 3884.53340 .006900000 3813.54589 .007200000 3737.821294800 3  1
 .007600000 3656.81397 .008000000 3582.25830 .008400000 3514.463134800 3  1
 .008800000 3449.63480 .009200000 3392.57902 .009600000 3336.147454800 3  1
 .010000000 3280.01050 .010500000 3220.80780 .011000000 3165.827464800 3  1
 .011500000 3116.26697 .012000000 3056.75349 .012750000 2993.775724800 3  1
 .013500000 2934.84335 .014250000 2882.95374 .015000000 2826.684314800 3  1
 .016000001 2770.20627 .017000001 2720.23756 .017999999 2675.037884800 3  1
 .018999999 2635.72908 .020000000 2599.61340 .021000000 2567.624014800 3  1
 .022000000 2539.13231 .023000000 2511.86739 .024000000 2482.958104800 3  1
 .025500000 2454.26312 .027000001 2432.51438 .028000001 2411.048314800 3  1
 .029999999 2387.64170 .032000002 2371.17108 .034000002 2358.183324800 3  1
 .035999998 2351.54253 .037999999 2346.60570 .039999999 2348.374314800 3  1
 .042500000 2353.69662 .045000002 2365.02466 .047499999 2379.458304800 3  1
 .050000001 2399.04366 .052499998 2422.16151 .055000000 2450.519444800 3  1
 .057500001 2481.52148 .059999999 2521.31213 .063000001 2569.398904800 3  1
 .066000000 2622.60408 .068999998 2682.53395 .071999997 2757.488904800 3  1
 .075999998 2853.88658 .079999998 2958.39422 .083999999 3076.093264800 3  1
 .088000000 3206.13870 .092000000 3350.11632 .096000001 3508.935544800 3  1
 .100000001 3705.58156 .104999997 3945.41328 .109999999 4209.616384800 3  1
 .115000002 4496.69776 .119999997 4881.48369 .127499998 5373.986404800 3  1
 .135000005 5876.81399 .142499998 6366.94877 .150000006 6853.068364800 3  1
 .159999996 7200.83169 .170000002 7239.14753 .180000007 6937.009264800 3  1
 .189999998 6352.50044 .200000003 5598.87078 .209999993 4800.688484800 3  1
 .219999999 4034.74925 .230000004 3356.45914 .239999995 2658.838644800 3  1
 .254999995 2016.93859 .270000011 1617.35818 .280000001 1275.170164800 3  1
 .300000012 956.926116 .319999993 738.196276 .340000004 579.5011124800 3  1
 .360000014 463.202190 .379999995 378.444865 .400000006 306.8630344800 3  1
 .425000012 246.386840 .449999988 200.535076 .474999994 165.1785714800 3  1
 .500000000 137.745030 .524999976 116.343794 .550000012 99.81595164800 3  1
 .574999988 87.2726063 .600000024 76.2313859 .629999995 66.32663664800 3  1
 .660000026 58.1385929 .689999998 51.3339442 .720000029 44.83185604800 3  1
 .759999990 38.9150369 .800000012 34.1811769 .839999974 30.26465974800 3  1
 .879999995 27.0299071 .920000017 24.3467777 .959999979 22.15907454800 3  1
 1.00000000 20.1132763 1.04999995 18.1891628 1.10000002 16.58395784800 3  1
 1.14999998 15.2326021 1.20000005 13.8636836 1.27499998 12.53649874800 3  1
 1.35000002 11.4658666 1.42499995 10.6000374 1.50000000 9.809550904800 3  1
 1.60000002 9.07523028 1.70000005 8.48518897 1.79999995 8.011156894800 3  1
 1.89999998 7.63533454 2.00000000 7.31824312 2.09999990 7.052632074800 3  1
 2.20000005 6.82792281 2.29999995 6.63194712 2.40000010 6.422780654800 3  1
 2.54999995 6.21564102 2.70000005 6.07775650 2.79999995 5.947554254800 3  1
 3.00000000 5.79948658 3.20000005 5.68080063 3.40000010 5.581048504800 3  1
 3.59999990 5.49524048 3.79999995 5.42388368 4.00000000 5.368349114800 3  1
 4.25000000 5.31239383 4.50000000 5.26703134 4.75000000 5.229646654800 3  1
 5.00000000 5.19226197 5.25000000 5.16042262 5.50000000 5.134456754800 3  1
 5.75000000 5.11071758 6.00000000 5.09427758 6.30000019 5.077305734800 3  1
 6.59999990 5.06033387 6.90000010 5.04336201 7.19999981 5.026856804800 3  1
 7.59999990 5.01236071 8.00000000 4.99790729 8.39999962 4.985242924800 3  1
 8.80000019 4.97745744 9.19999981 4.96994846 9.60000038 4.962439494800 3  1
 10.0000000 4.95399189 10.5000000 4.94460568 11.0000000 4.935219464800 3  1
 11.5000000 4.92656246 12.0000000 4.92066554 12.7500000 4.914255084800 3  1
 13.5000000 4.90784462 14.2500000 4.90143416 15.0000000 4.893959944800 3  1
 16.0000000 4.89933288 17.0000000 5.07749352 18.0000000 8.933653324800 3  1
 19.0000000 4.98804073 20.0000000 4.87941088 21.0000000 4.863028544800 3  1
 22.0000000 4.85746135 23.0000000 4.87063574 24.0000000 4.917284354800 3  1
 25.5000000 5.94445738 27.0000000 75.1737718 28.0000000 7.668585214800 3  1
 30.0000000 5.03958656 32.0000000 4.91129536 34.0000000 4.855363764800 3  1
 36.0000000 4.82451761 38.0000000 4.79704762 40.0000000 4.766143894800 3  1
 42.5000000 4.73180641 45.0000000 4.69746893 47.5000000 4.663131454800 3  1
 50.0000000 4.63278288 52.5000000 4.63140732 55.0000000 5.446688354800 3  1
 57.5000000 106.491363 60.0000000 6.28119653 63.0000000 13.84048894800 3  1
 66.0000000 36.6060943 69.0000000 5.20468098 72.0000000 4.804301954800 3  1
 76.0000000 4.64240995 80.0000000 7.53506247 84.0000000 42.95883634800 3  1
 88.0000000 170.588009 92.0000000 8.17224460 96.0000000 19.04447974800 3  1
 100.000000 8.02683261 105.000000 14.1968044 110.000000 5.315013094800 3  1
 115.000000 51.2187320 120.000000 57.0460427 127.500000 5.747503794800 3  1
 135.000000 9.72056905 142.500000 7.02178081 150.000000 8.330220964800 3  1
 160.000000 21.2493702 170.000000 5.27579426 180.000000 5.100755834800 3  1
 190.000000 21.1613051 200.000000 5.14300643 210.000000 11.68636414800 3  1
 220.000000 25.3065812 230.000000 19.5187065 240.000000 5.280161004800 3  1
 255.000000 11.1828494 270.000000 10.7891875 280.000000 10.79748984800 3  1
 300.000000 5.25674537 320.000000 5.33989774 340.000000 7.764123144800 3  1
 360.000000 6.09741115 380.000000 123.642035 400.000000 15.55537324800 3  1
 425.000000 15.4433666 450.000000 6.10070927 475.000000 6.039831004800 3  1
 500.000000 8.49148003 525.000000 10.0394565 550.000000 9.905870144800 3  1
 575.000000 8.42107229 600.000000 10.0872010 630.000000 5.136112944800 3  1
 660.000000 22.1294461 690.000000 5.51734153 720.000000 19.58187534800 3  1
 760.000000 6.00485425 800.000000 5.32265104 840.000000 13.50076114800 3  1
 880.000000 5.39921715 920.000000 12.2365042 960.000000 5.361230104800 3  1
 1000.00000 4.92355848 1050.00000 4.36700737 1100.00000 53.51459984800 3  1
 1150.00000 11.6025416 1200.00000 9.45657189 1275.00000 7.632092934800 3  1
 1350.00000 6.68594731 1425.00000 6.37515821 1500.00000 6.249648624800 3  1
 1600.00000 6.21277708 1700.00000 6.23799452 1800.00000 6.278176974800 3  1
 1900.00000 6.32503067 2000.00000 6.34389500 2100.00000 6.348446104800 3  1
 2200.00000 6.35299720 2300.00000 6.35779053 2400.00000 6.364660344800 3  1
 2550.00000 6.37295581 2700.00000 6.37986870 2800.00000 6.388164174800 3  1
 3000.00000 6.39922480 3200.00000 6.41028543 3400.00000 6.421346064800 3  1
 3600.00000 6.43240669 3800.00000 6.44346732 4000.00000 6.455863114800 3  1
 4250.00000 6.46855054 4500.00000 6.48078238 4750.00000 6.493014214800 3  1
 5000.00000 6.50524605 5250.00000 6.51747789 5500.00000 6.529709724800 3  1
 5750.00000 6.54194156 6000.00000 6.55539658 6300.00000 6.570074794800 3  1
 6600.00000 6.58475299 6900.00000 6.59943119 7200.00000 6.616555774800 3  1
 7600.00000 6.63612671 8000.00000 6.65569764 8400.00000 6.675268584800 3  1
 8800.00000 6.69429303 9200.00000 6.70000000 9600.00000 6.699972834800 3  1
 10000.0000 6.69866480 10500.0000 6.69659439 11000.0000 6.697671274800 3  1
 11500.0000 6.70391228 12000.0000 6.71180714 12750.0000 6.721280994800 3  1
 13500.0000 6.73075483 14250.0000 6.74238943 15000.0000 6.748304154800 3  1
 16000.0000 6.74496246 17000.0000 6.74163236 18000.0000 6.744436604800 3  1
 19000.0000 6.75238667 20000.0000 6.76033674 21000.0000 6.768286824800 3  1
 22000.0000 6.77623689 23000.0000 6.78418697 24000.0000 6.794516984800 3  1
 25500.0000 6.80916542 27000.0000 6.82169292 28000.0000 6.836725934800 3  1
 30000.0000 6.85447352 32000.0000 6.86500000 34000.0000 6.875000004800 3  1
 36000.0000 6.88500000 38000.0000 6.89500000 40000.0000 6.906250004800 3  1
 42500.0000 6.91875000 45000.0000 6.93125000 47500.0000 6.943750004800 3  1
 50000.0000 6.95625000 52500.0000 6.96875000 55000.0000 6.981250004800 3  1
 57500.0000 6.99375000 60000.0000 7.00750000 63000.0000 7.022500004800 3  1
 66000.0000 7.03750000 69000.0000 7.05250000 72000.0000 7.070000004800 3  1
 76000.0000 7.09000000 80000.0000 7.11000000 84000.0000 7.130000004800 3  1
 88000.0000 7.15000000 92000.0000 7.17000000 96000.0000 7.189784494800 3  1
 100000.000 7.19872065 105000.000 7.20167611 110000.000 7.204631574800 3  1
 115000.000 7.20758703 120000.000 7.21128135 127500.000 7.215714544800 3  1
 135000.000 7.22014773 142500.000 7.22458092 150000.000 7.229752974800 3  1
 160000.000 7.23566389 170000.000 7.24157481 180000.000 7.247485734800 3  1
 190000.000 7.25339665 200000.000 7.25930757 210000.000 7.265218494800 3  1
 220000.000 7.27112941 230000.000 7.27704033 240000.000 7.276246634800 3  1
 255000.000 7.26874831 270000.000 7.26249972 280000.000 7.255001404800 3  1
 300000.000 7.25034797 320000.000 7.25068470 340000.000 7.250161134800 3  1
 360000.000 7.25053822 380000.000 7.25041933 400000.000 7.250249054800 3  1
 425000.000 7.24655811 450000.000 7.23604347 475000.000 7.212445804800 3  1
 500000.000 7.16191677 525000.000 7.09024850 550000.000 7.025511384800 3  1
 575000.000 6.97525235 600000.000 6.92046648 630000.000 6.862116574800 3  1
 660000.000 6.82499939 690000.000 6.79499864 720000.000 6.759997764800 3  1
 760000.000 6.71999675 800000.000 6.67999575 840000.000 6.639994744800 3  1
 880000.000 6.59999374 920000.000 6.55999273 960000.000 6.519991724800 3  1
 1000000.00 6.40005502 1100000.00 6.20018262 1200000.00 6.002167184800 3  1
 1300000.00 5.84242028 1400000.00 5.81242153 1500000.00 5.810578904800 3  1
 1600000.00 5.72230744 1700000.00 5.60712633 1800000.00 5.499628614800 3  1
 1900000.00 5.39988226 2000000.00 5.33750283 2100000.00 5.312490324800 3  1
 2200000.00 5.22498922 2300000.00 5.07499951 2400000.00 4.942505054800 3  1
 2500000.00 4.82750584 2600000.00 4.73417186 2700000.00 4.662503124800 3  1
 2800000.00 4.59083437 2900000.00 4.51750134 3000000.00 4.451488054800 3  1
 3100000.00 4.39445882 3200000.00 4.33915155 3300000.00 4.291018634800 3  1
 3400000.00 4.24367495 3500000.00 4.20600307 3600000.00 4.178003004800 3  1
 3700000.00 4.15000293 3800000.00 4.12200286 3900000.00 4.094002794800 3  1
 4000000.00 4.07400264 4100000.00 4.06200241 4200000.00 4.050002184800 3  1
 4300000.00 4.03800195 4400000.00 4.02600172 4500000.00 4.018390094800 3  1
 4600000.00 4.01516708 4700000.00 4.01194407 4800000.00 4.008721064800 3  1
 4900000.00 4.00549805 5000000.00 4.00227504 5100000.00 4.000068704800 3  1
 5200000.00 4.00000020 5300000.00 4.00070831 5400000.00 4.002459014800 3  1
 5500000.00 4.00421901 5600000.00 4.00597901 5700000.00 4.007739014800 3  1
 5800000.00 4.00955363 5900000.00 4.01237828 6000000.00 4.015582474800 3  1
 6100000.00 4.01878665 6200000.00 4.02199084 6300000.00 4.025195034800 3  1
 6400000.00 4.02839922 6500000.00 4.03500109 6600000.00 4.045000644800 3  1
 6700000.00 4.05500020 6800000.00 4.06499976 6900000.00 4.074999314800 3  1
 7000000.00 4.08299935 7100000.00 4.08899988 7200000.00 4.095000414800 3  1
 7300000.00 4.10100094 7400000.00 4.10700147 7500000.00 4.116682354800 3  1
 7600000.00 4.13004358 7700000.00 4.14340481 7800000.00 4.156766044800 3  1
 7900000.00 4.17012728 8000000.00 4.18315975 8100000.00 4.193024524800 3  1
 8200000.00 4.20209475 8300000.00 4.21116498 8400000.00 4.220235204800 3  1
 8500000.00 4.22930543 8600000.00 4.23837566 8700000.00 4.247445884800 3  1
 8800000.00 4.25651611 8900000.00 4.26558634 9000000.00 4.274656564800 3  1
 9100000.00 4.28372679 9200000.00 4.29279702 9300000.00 4.301867244800 3  1
 9400000.00 4.31093747 9500000.00 4.32000770 9600000.00 4.329077924800 3  1
 9700000.00 4.33814815 9800000.00 4.34721838 9900000.00 4.356288824800 3  1
 10000000.0 4.36559952 10100000.0 4.37512278 10200000.0 4.384646054800 3  1
 10300000.0 4.39416931 10400000.0 4.40369257 10500000.0 4.413973604800 3  1
 10600000.0 4.42633104 10700000.0 4.43880724 10800000.0 4.451283434800 3  1
 10900000.0 4.46375962 11000000.0 4.47338016 11100000.0 4.480145044800 3  1
 11200000.0 4.48780115 11300000.0 4.50025129 11400000.0 4.513414784800 3  1
 11500000.0 4.52154448 11600000.0 4.52464039 11700000.0 4.527736294800 3  1
 11800000.0 4.53086905 11900000.0 4.53473951 12000000.0 4.538898434800 3  1
 12100000.0 4.54305735 12200000.0 4.54721628 12300000.0 4.551375204800 3  1
 12400000.0 4.55553412 12500000.0 4.55969304 12600000.0 4.563851974800 3  1
 12700000.0 4.56801089 12800000.0 4.57216981 12900000.0 4.576328744800 3  1
 13000000.0 4.58048766 13100000.0 4.58464658 13200000.0 4.588805514800 3  1
 13300000.0 4.59296443 13400000.0 4.59712335 13500000.0 4.601282284800 3  1
 13600000.0 4.60544120 13700000.0 4.60960012 13800000.0 4.613759054800 3  1
 13900000.0 4.61791797 14000000.0 4.61799749 14100000.0 4.613997614800 3  1
 14200000.0 4.60999774 14300000.0 4.60599786 14400000.0 4.601997984800 3  1
 14500000.0 4.59499857 14600000.0 4.58499963 14700000.0 4.575000694800 3  1
 14800000.0 4.56500174 14900000.0 4.55500280 15000000.0 4.550003294800 3  1
 15100000.0 4.55000321 15200000.0 4.55000314 15300000.0 4.550003064800 3  1
 15400000.0 4.55000298 15500000.0 4.55000290 15600000.0 4.550002834800 3  1
 15700000.0 4.55000275 15800000.0 4.55000267 15900000.0 4.550002594800 3  1
 16000000.0 4.55000251 16100000.0 4.55000244 16200000.0 4.550002364800 3  1
 16300000.0 4.55000228 16400000.0 4.55000220 16500000.0 4.550002134800 3  1
 16600000.0 4.55000205 16700000.0 4.55000197 16800000.0 4.550001894800 3  1
 16900000.0 4.55000182 17000000.0 4.55000174 17100000.0 4.550001664800 3  1
 17200000.0 4.55000158 17300000.0 4.55000151 17400000.0 4.550001434800 3  1
 17500000.0 4.55000135 17600000.0 4.55000127 17700000.0 4.550001194800 3  1
 17800000.0 4.55000112 17900000.0 4.55000104 18000000.0 4.550000964800 3  1
 18100000.0 4.55000088 18200000.0 4.55000081 18300000.0 4.550000734800 3  1
 18400000.0 4.55000065 18500000.0 4.55000057 18600000.0 4.550000504800 3  1
 18700000.0 4.55000042 18800000.0 4.55000034 18900000.0 4.550000264800 3  1
 19000000.0 4.55000019 19100000.0 4.55000011 19200000.0 4.550000044800 3  1
 19300000.0 4.55000075 19400000.0 4.55000198 19500000.0 4.550003214800 3  1
 19600000.0 4.55000444 19700000.0 4.55000567 19800000.0 4.550006904800 3  1
 19900000.0 4.55000813 20000000.0 0.0                             4800 3  1
                                                                  4800 3  0
                                                                  4800 0  0
                                                                     0 0  0
 4.91150E+4 1.13917E+2          0          0         34         104931 1451
 0.0        0.0                 0          0          0          64931 1451
 1.00000E+0 2.00000E+7          0          0         10       20024931 1451
 3.00000E+2 0.0                 1          0        403         104931 1451
 49-In-115 FEI/CNDC   EVAL-Mar03 Zolotarev,Dunjiu,Schmittroth     4931 1451
                      DIST-Feb2004                                4931 1451
----IRDF-2002         MATERIAL 4931                               4931 1451
-----INCIDENT NEUTRON DATA                                        4931 1451
------ENDF-6 FORMAT                                               4931 1451
***************************************************************** 4931 1451
 49-In-115 FEI        EVAL-Mar03 K.I.Zolotarev                    4931 1451
                      DIST-Apr03                                  4931 1451
----BROND-2           MATERIAL 4931                               4931 1451
-----INCIDENT NEUTRON DATA                                        4931 1451
------ENDF-6 FORMAT                                               4931 1451
------Russian Reactor Dosimetry File RRDF-2002                    4931 1451
***************************************************************** 4931 1451
Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev            4931 1451
***************************************************************** 4931 1451
MF=3                                                              4931 1451
MT= 51 - In115(n,n')In115m  reaction                              4931 1451
-------------------------------------                             4931 1451
 *****************************************************************4931 1451
 CHINESE EVALUATION OF (N,2N) REACTION ADDED AT NDS               4931 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451
                                                                  4931 1451
THE FOLLOWING FILE SECTIONS ARE INCLUDED -                        4931 1451
     MF= 1    GENERAL INFORMATION (MT=451)                        4931 1451
     MF= 8    RADIOACTIVITY DATA (MT=51)                          4931 1451
     MF=10    CROSS SECTION FOR RADIOACTIVE NUCLIDE               4931 1451
                PRODUCTION (MT=51)                                4931 1451
     MF=40    DATA COVARIANCES FOR RADIOACTIVE NUCLIDE            4931 1451
                PRODUCTION (MT=51)                                4931 1451
49-IN-115  CNDC/CIAE  EVAL-JUN91 CAI DUNJIU AND WANG ZISHENG      4931 1451
---ENDF-6 FORMAT      4931                  REVISION 0            4931 1451
                                                                  4931 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451
 *****************************************************************4931 1451
 49-IN-115 HEDL,ANL   EVAL-MAR90 F.SCHMITTROTH,D.L.SMITH,S.CHIBA  4931 1451
                      DIST-JUN90                       19900329   4931 1451
----ENDF/B-VI         MATERIAL 4931                               4931 1451
 MF=2 MT=151   EVALUATION OF RESOLVED RESONANCE PARAMETERS        4931 1451
              BASED ON NEW BNL-325, REF(2).                       4931 1451
 MF=3 MT=102  VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA  4931 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4931 1451
DOSIMETRY                                                         4931 1451
******************************************************************4931 1451
***************************************************************** 4931 1451
**********   Start of (N,2N) bibliographical component ********** 4931 1451
***************************************************************** 4931 1451
                                                                  4931 1451
      Evaluation of the 115-In(n,2n) 114m-In Reaction             4931 1451
                 Cross-section                                    4931 1451
           Cai Dunjiu Yu Baosheng Wang Zisheng                    4931 1451
                Lu Hanlin Zhao Wenrong                            4931 1451
          Institute of Atomic Energy, China                       4931 1451
The excitation function of 115-In(n, 2n) 114m-In reaction         4931 1451
covers neutron energy range from threshold of 9.228 MeV           4931 1451
( including the energy of isomeric state which has                4931 1451
0.190 MeV ) to 20 MeV. Experimental knowledge of the              4931 1451
115-In(n,2n) reaction cross sections is entirely based            4931 1451
on the results of activation measurements. There are              4931 1451
still no experimental results obtained by                         4931 1451
direct-neutron-detection methods (such as scintillation           4931 1451
tank). For both the indium isotopes (115-In,0.957,                4931 1451
113-In,0.043), the primary activity                               4931 1451
resulting from the (n,2n) process is due to the decay of          4931 1451
metastable state. The interaction with the primary isotope,       4931 1451
115-In, results in a metastable state in 114-In which             4931 1451
decays with 49.5d half-life by means of an E4 transition          4931 1451
to the ground state (1+) with a half-life of 71.9s.               4931 1451
The (n,2n) process also directly populates the ground state.      4931 1451
Seventeen sets of data available (1-18) were collected.           4931 1451
For the sake of comparison, all the experimental results were     4931 1451
adjusted by the unified nuclear parameters (including             4931 1451
intensity of r-line, half-life, branching ratio),                 4931 1451
the reference cross section and the dependence of cross           4931 1451
section on energy which were taken from (Ref.1). In several       4931 1451
cases, adjustments were made only for the dependence of cross     4931 1451
section on energy since some necessary parameters used in data    4931 1451
analysis were not given by the authors.                           4931 1451
  The evaluation was made as the following ways:                  4931 1451
 (1).At 14.7MeV neutron energy .                                  4931 1451
  Based on the results measured by Ge(Li) detector,the            4931 1451
recommended cross section at 14.7 MeV is 1290 mb.                 4931 1451
 (2). For the neutron energy range from threshold to 20 MeV.      4931 1451
  The recommended excitation function was obtained                4931 1451
by fitting the measured data with least-square method in the      4931 1451
neutron energy range of 10.4-20MeV and by extrapolating from      4931 1451
10.4MeV to 9.228MeV based on the physical trend and systematics   4931 1451
calculation.                                                      4931 1451
Covariance matrix is given for the evaluated data considering all 4931 1451
adjusted factors.                                                 4931 1451
                      References                                  4931 1451
  1.Zhao Wenrong et al., CNDC-89014;INDC(CPR)-16, 1989            4931 1451
  2.Ke Wei et al., Chin.J.Nucl.Phys.11, 3, 11(1989)               4931 1451
  Lu Hanlin et al., INDC(CPR)-011/GI, P67(1988)                   4931 1451
  3.J.Csikai et al., INDC(NDS)-232/L, P45(1989)                   4931 1451
  4.Li Jiangwei et al.,  Chin.J.Nucl.Phys., 10, 1, 52(1988)       4931 1451
  Wu Zhihua et al., 88MITO, P315(1988)                            4931 1451
  5.T.B.Ryves et al., J.Phys., G9, 1549(1983)                     4931 1451
  6.T.B.Ryves et al., J.Phys., G7, 115(1981)                      4931 1451
  7.A.Reggoug et al.,  82Antwerp, P873(1982);NIMA, 227, 249(1984) 4931 1451
  8.K.Kayashima et al.,  NEANDC(J)-61U, P94(1979)                 4931 1451
  9.D.C.Santry et al.,  Can.J.Phys., 54, 757(1976)                4931 1451
  10.Lu Hanlin et al.,  At.Energy Sci.Technol., 2, 113(1975)      4931 1451
  11.A.Paulsen et al.,  Atomkernenergie, 26, 34(1975)             4931 1451
  12.J.K.Temperley et al.,  BRL-1491(1970)                        4931 1451
  13.R.C.Barrall et al.,  Nucl.Phys., A138, 387(1969)             4931 1451
  14.R.C.Barrall et al., AFWL-TR-68-134(1969)                     4931 1451
  15.H.Rotzer, Nucl.Phys., A109, 694(1968);DWA176, 289(1968)      4931 1451
  16.B.Minetti et al.,  Z.Physik., A217, 83(1968)                 4931 1451
  17.H.O.Menlove et al., Phys.Rev., 163, 1308(1967)               4931 1451
  18.R.J.Prestwood et al.,  Phys.Rev., 121, 1438(1961)            4931 1451
  19.T.Kozlowski et al.,  Acta.Phys.Pol., 33, 409(1968)           4931 1451
  20.R.Prasad et al.,Nucl.Phys., A94, 476(1967)                   4931 1451
  and 88,349(1966).                                               4931 1451
  21.W.Nagel, JNE, 20, 475(1966)                                  4931 1451
  22.K.C.Garg et al.,IPA17, 525(1979)                             4931 1451
  23.G.N.Salaita et al.,  ANS, 16, 59(1973)                       4931 1451
  24.J.Janczyszyn et al.,  J.Radiational Chem., 14, 201(1973)     4931 1451
  25.M.Bormann et al.,  EANDC(E)76u, 51(1967)                     4931 1451
  26.W.Grochulski et al., INR-1172(1970);INR-1197, 10(1970)       4931 1451
  Acta.Phys.Pol., BI, 271(1970)                                   4931 1451
  27.H.K.Vonach et al.,  68WASH, 2, P885(1968)                    4931 1451
  28.Wang Zisheng et al., Research on the Covariances Matrix      4931 1451
  of the Evaluation Data(to be published)                         4931 1451
                                                                  4931 1451
***************************************************************** 4931 1451
**********   End of (N,2N) bibliographical component   ********** 4931 1451
================================================================= 4931 1451
                                                                  4931 1451
***************************************************************** 4931 1451
**********   Start of (N,G) bibliographical component  ********** 4931 1451
***************************************************************** 4931 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451
 *         *          *          *          *          *         *4931 1451
 CAPTURE TO 54 MIN. ISOMERIC STATE DESCRIPTION MODIFIED BY        4931 1451
 R. SCHENTER ON 2/20/84.  THIS WAS DONE BY ADDING FILE9 WITH      4931 1451
 Y=.79 AND PUTTING THE TOTAL CAPTURE WIDTH IN FILE2 AND THE TOTAL 4931 1451
 CAPTURE CROSS SECTION IN FILE3.                                  4931 1451
 P.C.TO NORMALIZATION+STANDARDS SUBCOMITTEE (N,G)F.SCHMITTROTH    4931 1451
                                                 MF=2 MT=151      4931 1451
                                                 MF=2 MT=102      4931 1451
         *          *          *          *         *             4931 1451
                                                                  4931 1451
FILE INFORMATION                                                  4931 1451
                                                                  4931 1451
MF=1 MT=451   ATOMIC MASS FROM REF(1).                            4931 1451
                                                                  4931 1451
MF=2 MT=151   EVALUATION OF RESOLVED RESONANCE PARAMETERS         4931 1451
              BASED ON NEW BNL-325, REF(2).                       4931 1451
 MF=3 MT=102  VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA  4931 1451
              SEE DOCUMENTATION                                   4931 1451
              RADIATIVE NEUTRON CAPTURE TO THE IN 116M(54 MIN.)   4931 1451
              STATE.                                              4931 1451
              E  GREATER THAN  EH,                                4931 1451
              EVALUATION BASED ON EXPERIMENTAL DATA, REF.(3-7)    4931 1451
              AND THEORETICAL CALCULATIONS, REF.(8,9).            4931 1451
              E  LESS THAT  EH,                                   4931 1451
              A   1/V   COMPONENT WAS ADDED TO GIVE THE CORRECT   4931 1451
              2200M/S CROSS SECTION TO THE  54  MIN. STATE(THE    4931 1451
              2.2  SEC. STATE CROSS SECTION WAS INCLUDED).        4931 1451
     RADIATIVE CAPTURE TO THE  2.2  SEC. STATE OF IN 116 WAS      4931 1451
     INCLUDED AS PART OF THE CAPTURE TO THE  54  MIN. STATE       4931 1451
     FOR BOTH THERMAL AND FAST ENERGIES.                          4931 1451
             RESULTS WERE DIVIDED BY .79 TO GIVE THE TOTAL CAP-   4931 1451
             TURE CROSS SECTION IN FILE3.  FILE9 COMBINED WITH    4931 1451
             FILE3 IS REQUIRED TO GIVE CAPTURE TO 54 MIN.         4931 1451
             ISOMERIC STATE.                                      4931 1451
                                                                  4931 1451
2200M/S CAPTURE CROSS SECTION,BARNS TO THE 54 MIN. STATE          4931 1451
  (FROM RESONANCE PARAMETERS) = 166.413                           4931 1451
                                                                  4931 1451
COMPUTED RESONANCE INTEGRAL   =2587.345                           4931 1451
                                                                  4931 1451
REFERENCES.                                                       4931 1451
1.  A.H.WAPSTRA AND N.B.GOVE,NUCL. DATA TABLES,VOL.9,PART 1(1971).4931 1451
2.  S.F.MUGHABGHAB AND D.I.GARBER,BNL-325,3RD ED.,VOL.1(1973)     4931 1451
3.  H.A.GRENCH AND H.O.MENLOVE,PHYS.REV.,VOL.165(1968)1298.       4931 1451
4.  H.O.MENLOVE,ET AL.,PHYS.REV.,VOL.163(1967)1299.               4931 1451
5.  S.A.COX, PHYS.REV.,VOL.133(1964)B378.                         4931 1451
6.  A.E.JOHNSRUD,ET AL.,PHYS.REV.,VOL.116(1959)927.               4931 1451
7.  G.PETO ET AL., J.NUCL.EN.,VOL.21(1967)797.                    4931 1451
8.  F.SCHMITTROTH, HEDL-TME 71-106(AUGUST 1971).                  4931 1451
9.  F.SCHMITTROTH, HEDL-TME 73-79,ENDF-195(NOVEMBER 1973).        4931 1451
***************************************************************** 4931 1451
**********             (N,G) processing details        ********** 4931 1451
***************************************************************** 4931 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     4931 1451
 ***************** Program RECENT (VERSION 2002-1) ***************4931 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4931 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     4931 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************4931 1451
 Data Doppler Broadened to 300.000000 Kelvin                      4931 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     4931 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     4931 1451
 ***************** Program FIXUP (Version 2002-1) ****************4931 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 4931 1451
 Corrected Thresholds-----------------------------------------Yes 4931 1451
 Extended Cross Sections to 20 MeV----------------------------No  4931 1451
 Allow Cross Section Deletion---------------------------------No  4931 1451
 Allow Cross Section Reconstruction---------------------------No  4931 1451
 Make All Cross Sections Non-Negative-------------------------Yes 4931 1451
 Delete Energies Not in Ascending Order-----------------------Yes 4931 1451
 Deleted Duplicate Points-------------------------------------Yes 4931 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 4931 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 4931 1451
 Allow Creation of Missing Sections---------------------------No  4931 1451
 Allow Insertion of Energy Points-----------------------------No  4931 1451
 Create Uniform Energy Grid-----------------------------------No  4931 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 4931 1451
 *************** PROGRAM ACTIVATE (VERSION 2002-1) ***************4931 1451
 MF=10 Activation Cross Sections Defined by Combining MF=3        4931 1451
 Cross Sections and MF=9 Multipliers. MF=9 Deleted.               4931 1451
***************************************************************** 4931 1451
 The metastable state component (MF/MT=10/102) was extracted from 4931 1451
 the ACTIVATE output and merged with the other reactions in this  4931 1451
 In-115 evaluation. MF/MT=33/102 was adapted as MF/MT=40/102.     4931 1451
***************************************************************** 4931 1451
**********    End of (N,G) bibliographical component   ********** 4931 1451
================================================================= 4931 1451
**********  Start of (N,N') bibliographical component  ********** 4931 1451
***************************************************************** 4931 1451
------Russian Reactor Dosimetry File RRDF-2002                    4931 1451
***************************************************************** 4931 1451
Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev            4931 1451
***************************************************************** 4931 1451
MF=3                                                              4931 1451
MT= 51 - In115(n,n')In115m  reaction                              4931 1451
-------------------------------------                             4931 1451
    In the section MT= 51 are given evaluated  cross section data 4931 1451
of neutron inelastic excitation of the  4.486-hour isomeric state 4931 1451
in the  indium-115  in the energy range from threshold to 20 MeV. 4931 1451
The isomeric level has energy  336.24 keV  with spin  and  parity 4931 1451
1/2- [1].                                                         4931 1451
    Microscopic experimental data  [2-40]  were  analyzed in  the 4931 1451
process of preparation  of input data base  for the evaluation of 4931 1451
cross sections and their uncertainty  for the In-115(n,n')In-115m 4931 1451
reaction.  During this procedure all experimental data  if it was 4931 1451
possible were corrected to the new recommended cross section data 4931 1451
for monitor reactions used in the measurements and to the new re- 4931 1451
commended decay data.                                             4931 1451
    Special correction was done with experimental data [5], [6],  4931 1451
[12], [24], [25] and [27].                                        4931 1451
    Grench and Menlove data [5]  and  Kimura et al. data [6] bet- 4931 1451
ween 0.5 - 0.9 MeV obtained  in  the measurements with  T(p,n)He3 4931 1451
neutron source were renormalized to the integral of Liskien et al.4931 1451
experimental data [16] in the overlapping energy ranges. Original 4931 1451
cross section data  from ref. [5] and [6]  were multiplied to the 4931 1451
correction factors Fc=1.15268 and Fc=0.92486 , respectively.      4931 1451
    Data of Santry and Butler [12] obtained in the absolute measu-4931 1451
rements with T(p,n)He3 source in the energy range 0.348-5.130 MeV 4931 1451
were renormalized  to the integral of Liskien et al. experimental 4931 1451
data [16] in the energy interval 1.85-4.07 MeV, Fc= 1.0516        4931 1451
    Cross section data measured by Zhao Wenrong et al. [24]  were 4931 1451
renormalized also to the experimental data of Liskien et al. [16] 4931 1451
in the overlapping energy ranges.  All original experimental data 4931 1451
from ref. [24]  except cross section value  at  the  energy point 4931 1451
En=8.5 MeV were multiplied to the factor Fc=1.065 .               4931 1451
    Experimental data of Kimura et al. [6] in the energy interval 4931 1451
1.8-4.6 MeV and experimental data of Santry and Butler [12] obta- 4931 1451
ined relative to the S-32(n,p)P-32 standard were not corrected.   4931 1451
    Data of Zhao Wenrong et al. [24] for the energy point 8.5 MeV 4931 1451
was corrected  to the recommended cross section value for monitor 4931 1451
tor reaction Al-27(n,a)Na-24 from ref. [41].                      4931 1451
    Cross section data for the In-115(n,n')In-115m reaction measu-4931 1451
red by Lu Hanlin et al. [25] and by Csikai et al. [27] were renor-4931 1451
malized to the integral of Konno et al. [30] and Filatenkov et al.4931 1451
[31] experimental data in the overlapping energy interval 13.54 - 4931 1451
14.78 MeV . Original data from ref. [25] and [27] were multiplied 4931 1451
to the correction factors Fc=1.13622 and Fc=1.08624, respectively.4931 1451
    Excitation function for the In-115(n,n')In-115m  reaction  in 4931 1451
the energy region from threshold to 20 MeV was evaluated by means 4931 1451
of statistical analysis of experimental cross section data [2-31].4931 1451
    Experimental cross section data [32-40] were rejected  due to 4931 1451
their discrepancy with the main bulk of experimental data [2-31]. 4931 1451
Data of Martin et al. [2] in the energy interval  4.55 - 5.26 MeV 4931 1451
were not taken into account in the evaluation due to their syste- 4931 1451
matic underestimation In-115(n,n')In-115m reaction cross section. 4931 1451
In the rejected experiments [33-38] the cross section values were 4931 1451
measured only in a one energy point in the interval 14 - 15 MeV.  4931 1451
    Statistical analysis of input cross section data was carried  4931 1451
out by means of PADE-2 code [42].  Rational function was used as  4931 1451
the model function [43].                                          4931 1451
     U-235 thermal fission [44] and  Cf-252 spontaneous  fission  4931 1451
neutron spectra [45] averaged cross sections calculated from the  4931 1451
the evaluated In-115(n,n')In-115m  excitation function  are  the  4931 1451
following:                                                        4931 1451
                                                                  4931 1451
  4931 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  4931 1451
  4931 1451
 U-235 neutron fission     188.40         188.2 +- 2.3    [46]  4931 1451
  4931 1451
 Cf-252 spont. fission     191.66         197.4 +- 2.704  [47]  4931 1451
  4931 1451
                                                                  4931 1451
MT=33                                                             4931 1451
MT= 51 -(n,n') cross section cov. matrix                          4931 1451
----------------------------------------                          4931 1451
     Uncertainties in the evaluated excitation function  for  the 4931 1451
reaction  In-115(n,n')In-115m  are given  in the form of relative 4931 1451
covariance matrix for the 39-neutron energy groups (LB=5).  Cova- 4931 1451
riance matrix of uncertainties was calculated simultaneously with 4931 1451
recommended cross section data by means of PADE-2 code.           4931 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  4931 1451
given in the 33-file are the following:                           4931 1451
                                                                  4931 1451
    7.55059E-08    8.01529E-08    8.14328E-08    8.29955E-08      4931 1451
    8.51539E-08    8.79248E-08    8.88394E-08    9.29091E-08      4931 1451
    9.80564E-08    1.06991E-07    1.19646E-07    1.34061E-07      4931 1451
    1.46033E-07    1.71437E-07    2.22942E-07    2.66525E-07      4931 1451
    4.04740E-07    7.05147E-07    1.52953E-06    2.70059E-06      4931 1451
    2.95312E-06    3.08776E-06    6.35038E-06    4.26321E-04      4931 1451
    6.81158E-04    9.21024E-04    9.54240E-04    1.07694E-03      4931 1451
    1.11466E-03    1.28566E-03    1.45684E-03    1.62008E-03      4931 1451
    1.97248E-03    4.10930E-03    5.75040E-03    7.29359E-03      4931 1451
    1.00269E-02    1.64825E-02    2.00540E-02                     4931 1451
                                                                  4931 1451
References:                                                       4931 1451
 1. R.B.Firestone   Table of Isotopes, Eighth edition, Vol. 1,    4931 1451
    John Wiley & Sons, Inc., New York, 1995                       4931 1451
 2. H.C.Martin et al.  Phys. Rev., v.93, p.199, January 1954      4931 1451
 3. H.O.Menlove, K.L.Coop, H.A.Grench, R.Sher  Phys. Rev., v.163, 4931 1451
    p.1299, 1967                                                  4931 1451
 4. P.Bornemisza-Pauspertl et al. Atomki Koezlemenyek, v.10, No.2,4931 1451
    p.112, July 1968                                              4931 1451
 5. H.A.Grench, H.O.Menlove  Physical Review, v.165, p.1298, 1968 4931 1451
 6. I.Kimura et al.  Journal of Nucl. Science and Technology, v.6,4931 1451
    p.485, September 1969                                         4931 1451
 7. R.C.Barrall et al. Report AFWL-TR-68-134, Albuquerque, NM,    4931 1451
    March 1969 ;                                                  4931 1451
    R.C.Barrall et al. Nucl. Phys. A, v.138, p.387, December 1969 4931 1451
 8. J.K.Temperley, D.E.Barnes  Report BRL-R-1491, August 1970     4931 1451
 9. L.Lakosi, A.Veres  2nd Nat'l Soviet Conf. on Neutron Physics, 4931 1451
    Kiev, v.4, p.312, May 1973                                    4931 1451
10. K.Kobayashi et al. J. Nucl. Energ., v.27, p.741, October 1973 4931 1451
11. D.L.Smith, J.W.Meadows Report ANL-NDM-14, Argonne, July 1975; 4931 1451
    D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.60, p.319,July 1976 4931 1451
12. D.C.Santry, J.P.Butler  Can. J. Phys., v.54, p.757, 1976      4931 1451
13. C.G.Hudson, W.L.Alford  Bull. Amer. Phys. Soc., v.21,         4931 1451
    p.188(DB5), February 1976                                     4931 1451
14. G.Magnusson, I.Bergqvist  Nuclear Technology, v.34, p.114,    4931 1451
    June 1977                                                     4931 1451
15. S.Yamamoto et al.  Report NEANDC(J)-56, p.25, September 1978  4931 1451
16. H.Liskien et al. Nucl. Sci. Eng., v.67, p.334, September 1978 4931 1451
17. P.Andersson et al.  Report LUNF-D6-3021, University of Lund,  4931 1451
    November 1978                                                 4931 1451
18. C.F.Ai, J.C.Chou  Nuclear Science, Taiwan, v.16, no.3, p.157, 4931 1451
    September 1979                                                4931 1451
19. L.Adamski et al.  Annals of Nucl. Energy, v.7, p.397, 1980    4931 1451
20. Fan Pei-Guo et al.  Chinese J. of Nuclear Physics, v.2, No.4, 4931 1451
    p.337, November 1980                                          4931 1451
21. V.L.Demekhin et al.    6th All Union Conference on Neutron    4931 1451
    Physics, Kiev, 2 - 6 October, v.3, p.195, 1983                4931 1451
22. T.B.Ryves et al.  J. of Physics, pt.G, v.9, p.1549, Dec. 1983 4931 1451
23. K.Kudo et al.  Report NEANDC(J)-106/U, p.1, September 1984    4931 1451
24. Zhao Wenrong, Lu Hanlin et al. Report INDC(CPR)-16, Sep. 1989 4931 1451
25. Lu Hanlin et al.  Chinese J. of Nuclear Physics, v.11, n.2,   4931 1451
    p.53, 1989                                                    4931 1451
26. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990    4931 1451
27. J.Csikai et al.  Zeitschrift fuer Physik, v.A337, p.39, 1990  4931 1451
28. N.N.Moiseev et al. Atomnaja Energija, v.68, February 1990     4931 1451
29. Y.Ikeda et al.  Proc. of an Int. Conf. Nuclear Data for Sci-  4931 1451
    ence and Technology, Julich, FRG, 13-17 May 1991, Springer-   4931 1451
    Verlag, 1992, pp.294-296                                      4931 1451
30. C.Konno et al.  Report JAERI-1329, October 1993               4931 1451
31. A.A.Filatenkov et al. Report RI-252, St. Petesburg, May 1999; 4931 1451
    A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,4931 1451
    Moscow, 1996                                                  4931 1451
32. S.G.Cohen   Nature, v.161, p.475, March 1948                  4931 1451
33. W.Nagel  EXFOR 20198.016 , December 1966                      4931 1451
34. H.Roetzer  Nucl. Phys., v.A109, p.694, March 1968             4931 1451
35. B.Minetti, A.Pasquarelli  Zeitschrift f. Physik, v.217, p.83, 4931 1451
    October 1968                                                  4931 1451
36. I.Garlea et al. Rev. Roum. Phys., v.29, p.421, 1984           4931 1451
37. R.Pepelnik et al. Report NEANDC(E)-262U, No.5,p.32, June 1985 4931 1451
38. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992  4931 1451
39. J.Csikai, P.Raics  EXFOR 30983.002 July 1992                  4931 1451
40. N.I.Molla et al.  EXFOR 30984.002                             4931 1451
41. H.Vonach in "Nuclear Data Standards for Nuclear Measurements",4931 1451
    Report NEANDC-311 U, pp.75-77, OECD, Paris, 1992. ;           4931 1451
    M.Wagner et al. Evaluation of cross sections for 14 important 4931 1451
    neutron-dosimetry reactions. Physics Data No.13-5, 1990       4931 1451
42. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          4931 1451
43. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 4931 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 4931 1451
    Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105                4931 1451
44. L.W.Weston et al.  Evaluated Neutron Data  for  Uranium-235,  4931 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    4931 1451
45. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        4931 1451
46. W.Mannhart  Prog. Report INDC(Ger)-045, pp.40-43, June 1999   4931 1451
47. W.Mannhart  Validation of Differential Cross Sections with    4931 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4931 1451
    September 2002                                                4931 1451
***************************************************************** 4931 1451
**********   End of (N,N') bibliographical component   ********** 4931 1451
================================================================= 4931 1451
                                                                  4931 1451
                                                                  4931 1451
                                                                  4931 1451
                                                                  4931 1451
***************************************************************** 4931 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************4931 1451
 Unshielded Group Averages Using  640 Groups                      4931 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     4931 1451
                                1        451        417          14931 1451
                                8          4          2          14931 1451
                                8         16          2          14931 1451
                                8        102          2          14931 1451
                               10          4         75          14931 1451
                               10         16         40          14931 1451
                               10        102        217          14931 1451
                               40          4        148          14931 1451
                               40         16         24          14931 1451
                               40        102          7          04931 1451
                                                                  4931 1  0
                                                                  4931 0  0
 4.91150E+4 1.13917E+2          0          0          1          04931 3291
 0.0       -3.36240E+5      49115          0          1        2144931 3291
        214          1                                            4931 3291
 320000.000 4.26808E-7 340000.000 .000287403 360000.000 .0008195824931 3291
 380000.000 .001351761 400000.000 .001832778 425000.000 .0022626334931 3291
 450000.000 .002833150 475000.000 .003544330 500000.000 .0044273854931 3291
 525000.000 .005482315 550000.000 .006739573 575000.000 .0081991584931 3291
 600000.000 .010079555 630000.000 .012413913 660000.000 .0151792004931 3291
 690000.000 .018214627 720000.000 .022308693 760000.000 .0275836804931 3291
 800000.000 .033469320 840000.000 .039925608 880000.000 .0469139004931 3291
 920000.000 .054352860 960000.000 .062152500 1000000.00 .0764851004931 3291
 1100000.00 .097441650 1200000.00 .118527000 1300000.00 .1393800004931 3291
 1400000.00 .160001000 1500000.00 .180578500 1600000.00 .2013295004931 3291
 1700000.00 .222376500 1800000.00 .243662000 1900000.00 .2648950004931 3291
 2000000.00 .285530500 2100000.00 .304807500 2200000.00 .3218550004931 3291
 2300000.00 .335859000 2400000.00 .346250000 2500000.00 .3528395004931 3291
 2600000.00 .355853500 2700000.00 .355855500 2800000.00 .3535995004931 3291
 2900000.00 .349874500 3000000.00 .345387500 3100000.00 .3407005004931 3291
 3200000.00 .336218000 3300000.00 .332201500 3400000.00 .3287965004931 3291
 3500000.00 .326066000 3600000.00 .324015500 3700000.00 .3226115004931 3291
 3800000.00 .321799500 3900000.00 .321514500 4000000.00 .3216850004931 3291
 4100000.00 .322240000 4200000.00 .323110500 4300000.00 .3242315004931 3291
 4400000.00 .325543000 4500000.00 .326989000 4600000.00 .3285195004931 3291
 4700000.00 .330089000 4800000.00 .331656000 4900000.00 .3331850004931 3291
 5000000.00 .334645000 5100000.00 .336008000 5200000.00 .3372505004931 3291
 5300000.00 .338353000 5400000.00 .339299500 5500000.00 .3400775004931 3291
 5600000.00 .340676500 5700000.00 .341089500 5800000.00 .3413125004931 3291
 5900000.00 .341344000 6000000.00 .341183500 6100000.00 .3408330004931 3291
 6200000.00 .340296500 6300000.00 .339579000 6400000.00 .3386870004931 3291
 6500000.00 .337627500 6600000.00 .336408500 6700000.00 .3350390004931 3291
 6800000.00 .333529000 6900000.00 .331887000 7000000.00 .3301225004931 3291
 7100000.00 .328245500 7200000.00 .326265500 7300000.00 .3241925004931 3291
 7400000.00 .322034500 7500000.00 .319800000 7600000.00 .3174980004931 3291
 7700000.00 .315135880 7800000.00 .312720756 7900000.00 .3102588764931 3291
 8000000.00 .307756374 8100000.00 .305218372 8200000.00 .3026490004931 3291
 8300000.00 .300052000 8400000.00 .297430500 8500000.00 .2947870004931 3291
 8600000.00 .292123000 8700000.00 .289438500 8800000.00 .2867330004931 3291
 8900000.00 .284005500 9000000.00 .281254000 9100000.00 .2784760004931 3291
 9200000.00 .275667000 9300000.00 .272821500 9400000.00 .2699335004931 3291
 9500000.00 .266996500 9600000.00 .264002000 9700000.00 .2609410004931 3291
 9800000.00 .257803000 9900000.00 .254576500 10000000.0 .2512495004931 3291
 10100000.0 .247808500 10200000.0 .244240000 10300000.0 .2405295004931 3291
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 0.0000E+00 0.0000E+00          1          5        120         15493133292
 1.000000-5 9.100000+6 1.050000+7 1.150000+7 1.220000+7 1.270000+7493133292
 1.320000+7 1.380000+7 1.420000+7 1.470000+7 1.520000+7 1.600000+7493133292
 1.720000+7 1.800000+7 2.000000+7 4.000000-1 1.154500-2 2.886700-3493133292
 5.207500-4 6.268700-4 2.634000-4 7.576400-5 0.000000+0 0.000000+0493133292
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.360500-3493133292
 2.329300-4 4.379300-5 3.313900-5 2.375700-5 6.429400-6 0.000000+0493133292
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0493133292
 4.273200-4 5.149000-5 5.671700-5 3.970200-5 1.611800-5 0.000000+0493133292
 5.134000-6 5.330100-6 0.000000+0 3.850900-6 2.755800-7 0.000000+0493133292
 1.326100-4 7.026400-5 5.057000-5 1.814000-5 1.025800-5 1.091900-5493133292
 1.151900-5 6.054300-6 4.032200-6 3.183000-6 2.577000-7 1.061600-4493133292
 7.193900-5 1.827700-5 7.806100-6 1.063900-5 1.195100-5 9.623500-6493133292
 6.619700-6 5.063800-6 1.300400-6 9.380800-5 2.002300-5 8.788300-6493133292
 1.212000-5 1.349000-5 1.062800-5 7.704500-6 6.688400-6 1.473800-6493133292
 5.145600-5 2.270400-5 1.436100-5 1.686000-5 4.715200-6 4.452700-6493133292
 4.233200-6 1.200700-6 4.804900-5 1.232200-5 1.347800-5 3.765900-6493133292
 2.757000-6 3.388400-6 4.405400-7 4.707100-5 2.134000-5 8.565900-6493133292
 1.578500-5 9.719500-6 9.119700-6 4.580100-5 1.385300-5 1.154500-5493133292
 1.754500-5 3.156100-6 4.554700-5 1.336600-5 2.469400-5 2.724500-6493133292
 4.622500-5 2.217600-5 1.374800-5 7.898800-5 5.926800-6 8.012400-5493133292
                                                                  493133  0
 4.91150E+4 1.13917E+2          0          0          0          1493133293
 0.0000E+00 0.0000E+00          0        293          0          1493133293
 0.0000E+00 0.0000E+00          0          1         14          7493133293
 1.000000-5 3.600000-3 1.000000+3 2.250000-2 1.000000+4 8.100000-3493133293
 1.000000+5 2.500000-3 1.000000+6 4.900000-3 4.000000+6 1.690000-2493133293
 2.000000+7 0.000000+0                                            493133293
                                                                  493133  0
                                                                  4931 0  0
                                                                     0 0  0
 5.31270E+4 1.25814E+2          0          0         34         105325 1451
 0.0        0.0                 0          0          0          65325 1451
 1.00000E+0 2.00000E+7          0          0         10       20025325 1451
 3.00000E+2 0.0                 1          0         60          35325 1451
 53-I -127 CNDC/CIAE  EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451
                      DIST-Feb2004                                5325 1451
----IRDF-2002         MATERIAL 5325                               5325 1451
-----INCIDENT NEUTRON DATA                                        5325 1451
------ENDF-6 FORMAT                                               5325 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451
53-I -127  CNDC/CIAE  EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451
---ENDF-6 FORMAT      5325                  REVISION 0            5325 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451
                                                                  5325 1451
Results of the evaluation in ENDF/B Format                        5325 1451
53-I-127                                                          5325 1451
I-127(n,2n)I-126  Dosimetry reaction                              5325 1451
Evaluated by Zhao Wenrong Lu Hanlin Yu Weixiang Cai Dunjiu        5325 1451
Wang Zisheng Huang Xiaolong                                       5325 1451
                                                                  5325 1451
The reaction of I-127(n,2n)I-126 is used as a threshold detector  5325 1451
 for neutron dosimetry applications. Its threshold is 9.137       5325 1451
MeV. The half-life of I-126 is 13.023d. Iodine is a monoisotopic  5325 1451
element. The main gamma ray energies of I-126 are 388.6keV        5325 1451
and 666.4keV with branch ratios of 34.1 percent and 33.1 percent  5325 1451
respectively. In order to get reasonable results, the necessary   5325 1451
adjustment was made first for the collected data. Then the        5325 1451
evaluation was performed by spline fitting procedure. The         5325 1451
uncertainties were derived from experimental errors and           5325 1451
the consideration of systematics.                                 5325 1451
                                                                  5325 1451
Evaluation documentation: Zhao Wenrong.                           5325 1451
Report CIAE 1991                                                  5325 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451
  The Q values and threshold energies were updated prior to pro-  5325 1451
  cessing through the codes to comply with the values obtained    5325 1451
  using the NNDC calculation program which is based on the 1995   5325 1451
  Update to the Atomic mass Evaluation.                           5325 1451
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************5325 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     5325 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     5325 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************5325 1451
 Data Doppler Broadened to 300.000000 Kelvin                      5325 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     5325 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     5325 1451
 ***************** Program FIXUP (Version 2002-1) ****************5325 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 5325 1451
 Corrected Thresholds-----------------------------------------Yes 5325 1451
 Extended Cross Sections to 20 MeV----------------------------No  5325 1451
 Allow Cross Section Deletion---------------------------------No  5325 1451
 Allow Cross Section Reconstruction---------------------------No  5325 1451
 Make All Cross Sections Non-Negative-------------------------Yes 5325 1451
 Delete Energies Not in Ascending Order-----------------------Yes 5325 1451
 Deleted Duplicate Points-------------------------------------Yes 5325 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 5325 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 5325 1451
 Allow Creation of Missing Sections---------------------------No  5325 1451
 Allow Insertion of Energy Points-----------------------------No  5325 1451
 Create Uniform Energy Grid-----------------------------------No  5325 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 5325 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************5325 1451
 Unshielded Group Averages Using  640 Groups                      5325 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     5325 1451
                                1        451         67          15325 1451
                                3         16         40          15325 1451
                               33         16         19          15325 1451
                                                                  5325 1  0
                                                                  5325 0  0
 5.31270E+4 1.25814E+2          0          0          1          05325 3 16
-9.14300E+6-9.14300E+6          0          0          1        1095325 3 16
        109          1                                            5325 3 16
 9200000.00 .002125030 9300000.00 .003744262 9400000.00 .0058736845325 3 16
 9500000.00 .008981316 9600000.00 .031000000 9700000.00 .0660600005325 3 16
 9800000.00 .108200000 9900000.00 .155240000 10000000.0 .2070000005325 3 16
 10100000.0 .262032500 10200000.0 .320900000 10300000.0 .3824372375325 3 16
 10400000.0 .448368421 10500000.0 .517276842 10600000.0 .5849000005325 3 16
 10700000.0 .651900000 10800000.0 .718900000 10900000.0 .7858500005325 3 16
 11000000.0 .851150000 11100000.0 .915570000 11200000.0 .9773500005325 3 16
 11300000.0 1.03776632 11400000.0 1.09542105 11500000.0 1.151272635325 3 16
 11600000.0 1.19980000 11700000.0 1.24544000 11800000.0 1.285400005325 3 16
 11900000.0 1.32310750 12000000.0 1.35620000 12100000.0 1.387452505325 3 16
 12200000.0 1.41480000 12300000.0 1.44060105 12400000.0 1.463263165325 3 16
 12500000.0 1.48443579 12600000.0 1.50185000 12700000.0 1.518110005325 3 16
 12800000.0 1.53215000 12900000.0 1.54537000 13000000.0 1.556000005325 3 16
 13100000.0 1.56600000 13200000.0 1.57600000 13300000.0 1.585794875325 3 16
 13400000.0 1.59369231 13500000.0 1.60112821 13600000.0 1.608564105325 3 16
 13700000.0 1.61572676 13800000.0 1.62125000 13900000.0 1.626500005325 3 16
 14000000.0 1.63175000 14100000.0 1.63679830 14200000.0 1.640636365325 3 16
 14300000.0 1.64427273 14400000.0 1.64790909 14500000.0 1.651545455325 3 16
 14600000.0 1.65518182 14700000.0 1.65826506 14800000.0 1.660795185325 3 16
 14900000.0 1.66332530 15000000.0 1.66585542 15100000.0 1.668385545325 3 16
 15200000.0 1.67091566 15300000.0 1.67344578 15400000.0 1.675975905325 3 16
 15500000.0 1.67817198 15600000.0 1.67940000 15700000.0 1.680566675325 3 16
 15800000.0 1.68173333 15900000.0 1.68290000 16000000.0 1.684066675325 3 16
 16100000.0 1.68494750 16200000.0 1.68500000 16300000.0 1.684748725325 3 16
 16400000.0 1.68376923 16500000.0 1.68274359 16600000.0 1.681717955325 3 16
 16700000.0 1.68006051 16800000.0 1.67710000 16900000.0 1.674100005325 3 16
 17000000.0 1.67110000 17100000.0 1.66762000 17200000.0 1.663150005325 3 16
 17300000.0 1.65795667 17400000.0 1.65133333 17500000.0 1.644666675325 3 16
 17600000.0 1.63800000 17700000.0 1.63028708 17800000.0 1.621050005325 3 16
 17900000.0 1.61180000 18000000.0 1.60255000 18100000.0 1.592388755325 3 16
 18200000.0 1.58090000 18300000.0 1.56824212 18400000.0 1.553897445325 3 16
 18500000.0 1.53953846 18600000.0 1.52517949 18700000.0 1.509750005325 3 16
 18800000.0 1.49325000 18900000.0 1.47550000 19000000.0 1.456500005325 3 16
 19100000.0 1.43675000 19200000.0 1.41625000 19300000.0 1.393692315325 3 16
 19400000.0 1.36907692 19500000.0 1.34446154 19600000.0 1.319880525325 3 16
 19700000.0 1.29935484 19800000.0 1.28161290 19900000.0 1.263870975325 3 16
 20000000.0 0.0                                                   5325 3 16
                                                                  5325 3  0
                                                                  5325 0  0
 5.31270E+4 1.25814E+2          0          0          1          1532533 16
 0.000000+0 0.000000+0          0         16          0          1532533 16
 0.000000+0 0.000000+0          1          5         91         13532533 16
 1.000000-5 9.200000+6 1.058000+7 1.118000+7 1.157000+7 1.237000+7532533 16
 1.355000+7 1.415000+7 1.454000+7 1.514000+7 1.712000+7 1.811000+7532533 16
 1.989000+7 1.000000-1 2.189400-3 1.465900-3 8.864000-4 6.104300-4532533 16
 5.784600-4 5.628800-4 4.006100-4 5.706500-4 2.236400-4 3.653400-4532533 16
 3.326600-4 2.851600-4 6.491500-5 3.944500-5 2.716400-5 2.613800-5532533 16
 2.527600-5 1.805000-5 2.583600-5 9.952100-6 1.661900-5 1.522700-5532533 16
 1.278300-4 2.641000-5 1.818700-5 1.750000-5 1.692300-5 1.208500-5532533 16
 1.729800-5 6.663300-6 1.112700-5 1.019500-5 6.062500-5 1.953900-5532533 16
 2.045100-5 8.819600-6 9.864000-6 7.941700-6 3.976900-6 1.965700-5532533 16
 1.911200-5 4.140300-5 1.991900-5 1.119300-5 9.425200-6 1.161100-5532533 16
 1.011200-5 1.294000-5 1.031500-5 2.200000-5 9.677000-6 7.798900-6532533 16
 1.012000-5 7.187100-6 7.827200-6 6.343900-6 1.636500-5 1.249200-5532533 16
 1.206100-5 9.854400-6 9.520400-6 8.511300-6 1.567300-5 1.273500-5532533 16
 7.466100-6 8.300400-6 6.147800-6 1.539000-5 1.140600-5 1.040700-5532533 16
 8.057200-6 1.543700-5 1.378500-5 8.856800-6 1.830900-5 1.393200-5532533 16
 2.509400-5                                                       532533 16
                                                                  532533  0
                                                                  5325 0  0
                                                                     0 0  0
 5.71390E+4 1.37713E+2          0          0         34         105728 1451
 0.0        0.0                 0          0          0          65728 1451
 1.00000E+0 2.00000E+7          0          0         10       20025728 1451
 3.00000E+2 0.0                 1          0        200          35728 1451
 57-La-139 FEI        EVAL-SEP01 K.I.Zolotarev                    5728 1451
                      DIST-Feb2004                                5728 1451
----IRDF-2002         MATERIAL 5728                               5728 1451
-----INCIDENT NEUTRON DATA                                        5728 1451
------ENDF-6 FORMAT                                               5728 1451
***************************************************************** 5728 1451
 57-La-139 FEI        EVAL-Sep01 K.I.Zolotarev                    5728 1451
                      DIST-Jan04                                  5728 1451
----BROND-3           MATERIAL 5728         Revision 1, Jan. 2004 5728 1451
-----INCIDENT NEUTRON DATA                                        5728 1451
------ENDF-6 FORMAT                                               5728 1451
***************************************************************** 5728 1451
------Russian Reactor Dosimetry File RRDF-2002                    5728 1451
***************************************************************** 5728 1451
Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev             5728 1451
***************************************************************** 5728 1451
                                                                  5728 1451
                    ----- MF=2 MT=151 -----                       5728 1451
                                                                  5728 1451
RESONANCE PARAMETERS                                              5728 1451
1. Resolved Resonance Region (RRR)                                5728 1451
   Resolved MLBW resonance parameters  up to  20  keV were based  5728 1451
on data from ref.[1-2]. Spins of resonances, which were not iden- 5728 1451
tified, have been assigned by taking into account 2*J+1 law  and  5728 1451
approximate linear growth of the number of levels with given spin 5728 1451
and parity from energy.  Average gamma-widths  0.055 eV (L=0) and 5728 1451
0.04 eV (L=1)  were assigned  to  those resonances , widths  for  5728 1451
which were absent.  Fictitious  P-resonances  were  added  in the 5728 1451
energy region  (3 - 20) keV  to obtain the average capture cross  5728 1451
section values observed in the experiments [3],[4],[5].           5728 1451
                                                                  5728 1451
2. Unresolved Resonance Region (URR)                              5728 1451
   Unresolved resonance parameters in the region 20 keV - 160 keV 5728 1451
were evaluated  by  using  'EVPAR' code (Hauser-Feshbach-Moldauer 5728 1451
statistical model) [6]. URR parameters were calculated from opti- 5728 1451
mized fit to the measured capture [4,5]  and  total cross section 5728 1451
[7,8,9] data.  Average resolved resonance  parameters  have been  5728 1451
used as zero approximation.  Contribution of F-wave in the energy 5728 1451
interval  60 - 160 keV given as smoothed background cross section 5728 1451
in File 3.                                                        5728 1451
                                                                  5728 1451
                    ----- MF=3 MT=102 -----                       5728 1451
                                                                  5728 1451
   Capture cross sections from 1.0E-5 eV to 160 keV  are reconst- 5728 1451
ructed from evaluated  MLBW resolved and unresolved resonance pa- 5728 1451
rameters.  Small background  was added  in  RRR  for taking into  5728 1451
account non-statistical reaction mechanism contribution. Contri-  5728 1451
bution of F-wave in the energy interval (60 - 160) keV was added  5728 1451
as described above.                                               5728 1451
   Data base for the evaluation La-139(n,g)La-140 excitation func-5728 1451
tion in the neutron energy range 160 keV - 20 MeV was formed from 5728 1451
microscopic experimental  data [4-5], [10-28] and  data  received 5728 1451
from theoretical model calculations. Experimental data [4], [10], 5728 1451
[12-18], [20-22], [27-28] were  corrected  to  the  new standards.5728 1451
Uncertainties  for cross section data  measured by Johnsrud et al.5728 1451
[13] were evaluated between +-(16-19)%. Theoretical model calcula-5728 1451
tions was done by means of GNASH code [29].  New modified version 5728 1451
of GNASH was utilized for calculations [30].   Data from theoreti-5728 1451
cal model  calculation were dominant  in the neutron energy range 5728 1451
3.5 - 20 MeV.                                                     5728 1451
    The evaluation  La-139(n,g)La-140  excitation function in the 5728 1451
energy range above 160 keV has been carried out within the frame- 5728 1451
work  of generalized least squares method , rational function was 5728 1451
used as model function [31]. Procedure of calculation recommended 5728 1451
cross section data was performed by means of PADE-2 code [32].    5728 1451
   Capture cross section at En=0.0253 eV and Resonance Integral   5728 1451
(0.5 to 2.0E+7 eV) from present evaluation are given below in the 5728 1451
Table 1. in a comparison with data from compilations [1], [33-34].5728 1451
                                                      Table 1.    5728 1451
        -------------------------------------------------         5728 1451
          Parameters            Values, barn   References         5728 1451
        -------------------------------------------------         5728 1451
        Capture cross section   9.04 +- 0.23   This eval.         5728 1451
        at En=0.0253 eV         8.93 +- 0.04     [ 1]             5728 1451
                                9.04 +- 0.04     [33]             5728 1451
                                9.2  +- 0.2      [34]             5728 1451
        -------------------------------------------------         5728 1451
        Resonance Integral      12.1           This eval.         5728 1451
                                11.8 +- 0.8      [ 1]             5728 1451
                                12.1 +- 0.6      [33]             5728 1451
                                11.8 +- 0.8      [34]             5728 1451
        -------------------------------------------------         5728 1451
                                                                  5728 1451
                    ----- MF=33 MT=102-----                       5728 1451
    Uncertainties in  the evaluated excitation function  for the  5728 1451
reaction  La-139(n,g)La-140  are given  in the three independent  5728 1451
matrixes.                                                         5728 1451
    In the energy range 1.000E-05 - 200 eV uncertainties are pre- 5728 1451
sented in the form of relative covariance matrix for the 16-neut- 5728 1451
ron energy groups (LB=5). Eigenvalues  of this 6-th digits relat- 5728 1451
ive covariance matrix are the following:                          5728 1451
                                                                  5728 1451
    1.29984E-04    1.80010E-04    2.41627E-04    2.69103E-04      5728 1451
    3.03943E-04    3.53897E-04    4.83948E-04    5.72815E-04      5728 1451
    6.17194E-04    6.90136E-04    9.46937E-04    1.01148E-03      5728 1451
    1.61882E-03    1.76694E-03    1.98803E-03    2.89526E-02      5728 1451
                                                                  5728 1451
    In the energy range 0.2 - 20  keV  uncertainties are given in 5728 1451
the form  of  diagonal matrix of uncertainties  for  7-th neutron 5728 1451
energy intervals (LB=1)                                           5728 1451
    In the energy range 20 keV - 20 MeV uncertainties are presen- 5728 1451
ted in the form of relative covariance matrix for the 23-neutron  5728 1451
energy groups (LB=5).  Eigenvalues  of this 6-th digits relative  5728 1451
covariance matrix are the following:                              5728 1451
                                                                  5728 1451
    1.96223E-10    5.86297E-08    3.57093E-07    1.00052E-05      5728 1451
    3.95061E-05    1.50141E-04    2.46548E-04    3.34259E-04      5728 1451
    4.59370E-04    6.23219E-04    7.85798E-04    8.75929E-04      5728 1451
    1.31621E-03    1.62913E-03    1.81439E-03    1.87643E-03      5728 1451
    2.55332E-03    2.70062E-03    3.40954E-03    4.32177E-03      5728 1451
    1.13231E-02    3.04915E-02    1.18873E-01                     5728 1451
                                                                  5728 1451
References :                                                      5728 1451
 1. S.F.Mughabghab et al.  Neutron Cross Sections ,Vol.1,Part A,  5728 1451
    New York, Academic Press,1981                                 5728 1451
 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,5728 1451
    ed. H.Schopper, Springer, 1998, pp.w19-w23                    5728 1451
 3. V.A.Konks, Ju.P.Popov, F.L.Shapiro  Zurnal Eksperim. i Teor.  5728 1451
    Fiziki, (USSR), v.46,(1), p.80, 1963                          5728 1451
 4. D.C.Stupegia et al.  J. Nucl. Energ., v.22, p.267, May 1968   5728 1451
 5. B.J.Allen, J.W.Boldeman, R.L.Macklin  Nucl. Sci. Eng., v.82,  5728 1451
    n.2, p.230, October 1982                                      5728 1451
 6. G.N.Manturov et al. Voprosy Atomnoy Nauki i Tekhniki, Ser.:   5728 1451
    Jadernye Konstanty, v.1, p.50, 1983                           5728 1451
 7. M.Divadeenam et al. Diss. Abstr. B , 1968, v.28, p.3834, 1968 5728 1451
 8. E.Islam et al.  Nucl. Phys. A, v.209, p.189, 1973             5728 1451
 9. K.Nishimura et al. Report JAERI-M-6883, January 1977          5728 1451
10. R.L.Macklin et al.  Phys. Rev., v.107, p.504, 1957            5728 1451
11. J.L.Perkin et al.  Proc. Phys. Soc., v.72, p.505, 1958        5728 1451
12. W.S.Lyon, R.L.Macklin  Phys. Rev., v.114, p.1619, 1959        5728 1451
13. A.E.Johnsrud et al.  Phys. Rev., v.116, p.927, 1959           5728 1451
14. R.G.Wille, R.W.Fink  Phys. Rev., v.118, p.242, 1960           5728 1451
15. J.H.Gibbons et al.  Phys. Rev., v.122, p.182, 1961            5728 1451
16. J.Csikai et al.  J. Atomki Kozlemenyek, v.8, p.79, June 1966  5728 1451
    J.Csikai et al.  Nucl. Phys. A, v.95, p.229, March 1967       5728 1451
17. A.K.Chaubey, M.L.Sehgal  Phys. Rev., v.152, p.1055, Dec. 1966 5728 1451
18. G.Peto et al.  J. Nucl. Energ., v.21, p.797, October 1967     5728 1451
19. F.Rigaud et al.  Nucl. Phys., v.A176, p.545, December 1971    5728 1451
20. G.G.Zaikin et al.  Ukrainskij Fizichnij Zhurnal, v.16(7),     5728 1451
    p.1205, July 1971                                             5728 1451
21. F.Rigaud et al.  Nucl. Sci. Eng., v.55, p.17, September 1974  5728 1451
22. O.Schwerer et al.  Nucl. Phys. A, v.264, p.105, June 1976     5728 1451
23. A.R.Del.Musgrove, B.J.Allen, R.L.Macklin  Proc. of Int. Conf. 5728 1451
    on Neutron Physics and Nuclear Data for Reactors and Other    5728 1451
    Applied Purposes, AERE Harwell, 25-29 September 1978, p.449   5728 1451
24. W.P.Poenitz  Progress Report ANL-83-4, p.239, April 1982      5728 1451
25. J.Voignier et al.  Nucl. Sci. Eng., v.93, p.43, 1986          5728 1451
26. H.Beer  Progress Report NEANDC(E)-272U,(5), p.8, June 1986    5728 1451
27. Y.N.Trofimov   Voprosy Atomnoy Nauki i Tekhniki, Serija:      5728 1451
    Jadernye Konstanty, v.4, p.10, 1987                           5728 1451
28. Y.N.Trofimov Proc. of the 1-st International Conf. on Neutron 5728 1451
    Physics, Kiev, 14-18 September 1987, v.3, p.331, Moscow 1988  5728 1451
29. P.G.Young, E.D.Arthur  A Preequilibrium Statistical Nuclear   5728 1451
    Model Code for Calculation of Cross Section and Emission      5728 1451
    Spectra. Report LA-6947, Los Alamos, 1977                     5728 1451
30. E.L.Trykov, G.Ya.Tertychnyi  Private communication, IPPE,     5728 1451
    Obninsk, May 1999                                             5728 1451
31. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 5728 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 5728 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  5728 1451
32. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          5728 1451
33. S.F.Mughabghab et al.  Termal Neutron Capture Cross Sections  5728 1451
    Resonance Integrals and G-factors, Report INDC(NDS)-440,      5728 1451
    IAEA, Vienna, February 2003                                   5728 1451
34. A.V.Ignatyuk et al.  Landolt Bornstein New Series, v.I/16A,   5728 1451
    Part 1, ed. H.Schopper, Springer, 1998, p.8-18                5728 1451
***************************************************************** 5728 1451
 The total and elastic cross-section files gave double counting   5728 1451
 of the cross-sections in the resonance region, these two files   5728 1451
 were modified for this library. The total and elastic cross-     5728 1451
 section files were taken from ENDF/B-VI Release 1.               5728 1451
***************************************************************** 5728 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************5728 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     5728 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     5728 1451
 ***************** Program RECENT (VERSION 2002-1) ***************5728 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     5728 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     5728 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************5728 1451
 Data Doppler Broadened to 300.000000 Kelvin                      5728 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     5728 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     5728 1451
 ***************** Program FIXUP (Version 2002-1) ****************5728 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 5728 1451
 Corrected Thresholds-----------------------------------------Yes 5728 1451
 Extended Cross Sections to 20 MeV----------------------------No  5728 1451
 Allow Cross Section Deletion---------------------------------No  5728 1451
 Allow Cross Section Reconstruction---------------------------No  5728 1451
 Make All Cross Sections Non-Negative-------------------------Yes 5728 1451
 Delete Energies Not in Ascending Order-----------------------Yes 5728 1451
 Deleted Duplicate Points-------------------------------------Yes 5728 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 5728 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 5728 1451
 Allow Creation of Missing Sections---------------------------No  5728 1451
 Allow Insertion of Energy Points-----------------------------No  5728 1451
 Create Uniform Energy Grid-----------------------------------No  5728 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 5728 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************5728 1451
 Unshielded Group Averages Using  640 Groups                      5728 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     5728 1451
                                1        451        207          15728 1451
                                3        102        217          15728 1451
                               33        102         93          15728 1451
                                                                  5728 1  0
                                                                  5728 0  0
 5.71390E+4 1.37713E+2          0          0          0          05728 3102
 5.16110E+6 5.16110E+6          0          0          1        6415728 3102
        641          1                                            5728 3102
 .000100000 140.916130 .000105000 137.631355 .000110000 134.5107985728 3102
 .000115000 131.635331 .000120000 128.273838 .000127500 124.5443475728 3102
 .000135000 121.137008 .000142500 118.005771 .000150000 114.6169365728 3102
 .000160000 111.086689 .000170000 107.874636 .000180000 104.9262495728 3102
 .000190000 102.183048 .000200000 99.6784848 .000210000 97.32212995728 3102
 .000220000 95.1334787 .000230000 93.0999779 .000240000 90.71308485728 3102
 .000255000 88.0902320 .000270000 86.0663388 .000280000 83.81339225728 3102
 .000300000 81.0658839 .000320000 78.5781377 .000340000 76.29978725728 3102
 .000360000 74.2013117 .000380000 72.2857027 .000400000 70.28859095728 3102
 .000425000 68.2449527 .000450000 66.3776940 .000475000 64.66584655728 3102
 .000500000 63.0560178 .000525000 61.5847291 .000550000 60.19486285728 3102
 .000575000 58.9022361 .000600000 57.5773841 .000630000 56.21446155728 3102
 .000660000 54.9673858 .000690000 53.7710963 .000720000 52.49819395728 3102
 .000760000 51.1333567 .000800000 49.8661477 .000840000 48.70483885728 3102
 .000880000 47.6010855 .000920000 46.5892455 .000960000 45.62022485728 3102
 .001000000 44.6158143 .001050000 43.5673071 .001100000 42.58438595728 3102
 .001150000 41.6787045 .001200000 40.6103452 .001275000 39.43618005728 3102
 .001350000 38.3627031 .001425000 37.3608905 .001500000 36.29659065728 3102
 .001600000 35.1837725 .001700000 34.1678926 .001800000 33.22832805728 3102
 .001900000 32.3699820 .002000000 31.5741079 .002100000 30.82725945728 3102
 .002200000 30.1432970 .002300000 29.4892640 .002400000 28.74300675728 3102
 .002550000 27.9137684 .002700000 27.2651423 .002800000 26.56036255728 3102
 .003000000 25.6927903 .003200000 24.9033544 .003400000 24.18153895728 3102
 .003600000 23.5233241 .003800000 22.9156450 .004000000 22.28233845728 3102
 .004250000 21.6394275 .004500000 21.0525874 .004750000 20.50405645728 3102
 .005000000 20.0005519 .005250000 19.5336671 .005500000 19.09221685728 3102
 .005750000 18.6875024 .006000000 18.2624172 .006300000 17.83749075728 3102
 .006600000 17.4354503 .006900000 17.0627040 .007200000 16.65762145728 3102
 .007600000 16.2246636 .008000000 15.8289941 .008400000 15.45792825728 3102
 .008800000 15.1129611 .009200000 14.7923717 .009600000 14.48729375728 3102
 .010000000 14.1711393 .010500000 13.8366769 .011000000 13.53026115728 3102
 .011500000 13.2378410 .012000000 12.9026382 .012750000 12.53202545728 3102
 .013500000 12.1890031 .014250000 11.8743787 .015000000 11.53881425728 3102
 .016000001 11.1865003 .017000001 10.8633443 .017999999 10.56820955728 3102
 .018999999 10.2967876 .020000000 10.0420336 .021000000 9.808730435728 3102
 .022000000 9.58877507 .023000000 9.38366445 .024000000 9.142476865728 3102
 .025500000 8.88195969 .027000001 8.67540086 .028000001 8.450706555728 3102
 .029999999 8.17372671 .032000002 7.92099010 .034000002 7.689985845728 3102
 .035999998 7.47901504 .037999999 7.28440094 .039999999 7.081330265728 3102
 .042500000 6.87522787 .045000002 6.68700537 .047499999 6.511801765728 3102
 .050000001 6.35088582 .052499998 6.20203607 .055000000 6.061114085728 3102
 .057500001 5.93184964 .059999999 5.79580688 .063000001 5.659533175728 3102
 .066000000 5.53106507 .068999998 5.41136298 .071999997 5.281554185728 3102
 .075999998 5.14250084 .079999998 5.01545054 .083999999 4.896032725728 3102
 .088000000 4.78499161 .092000000 4.68226134 .096000001 4.584216205728 3102
 .100000001 4.48324276 .104999997 4.37641057 .109999999 4.278466245728 3102
 .115000002 4.18477460 .119999997 4.07719395 .127499998 3.958265675728 3102
 .135000005 3.84797071 .142499998 3.74673540 .150000006 3.638811005728 3102
 .159999996 3.52534063 .170000002 3.42151836 .180000007 3.326642955728 3102
 .189999998 3.23955461 .200000003 3.15757224 .209999993 3.082848935728 3102
 .219999999 3.01500584 .230000004 2.94815628 .239999995 2.871692795728 3102
 .254999995 2.78666076 .270000011 2.72163373 .280000001 2.649409865728 3102
 .300000012 2.56098805 .319999993 2.48088377 .340000004 2.408141305728 3102
 .360000014 2.34167369 .379999995 2.27990170 .400000006 2.216861425728 3102
 .425000012 2.15320686 .449999988 2.09521825 .474999994 2.043085085728 3102
 .500000000 1.99748613 .524999976 1.95603723 .550000012 1.922719395728 3102
 .574999988 1.89911055 .600000024 1.89200884 .629999995 1.936167425728 3102
 .660000026 2.16687669 .689999998 2.97714428 .720000029 3.585621135728 3102
 .759999990 2.36696288 .800000012 1.77176305 .839999974 1.606425365728 3102
 .879999995 1.52575725 .920000017 1.47129338 .959999979 1.427912665728 3102
 1.00000000 1.38690059 1.04999995 1.34698658 1.10000002 1.311277615728 3102
 1.14999998 1.27807038 1.20000005 1.24089857 1.27499998 1.199740075728 3102
 1.35000002 1.16237991 1.42499995 1.12842344 1.50000000 1.091833595728 3102
 1.60000002 1.05355434 1.70000005 1.01871964 1.79999995 .9868971405728 3102
 1.89999998 .957195969 2.00000000 .929820007 2.09999990 .9045599205728 3102
 2.20000005 .880519178 2.29999995 .858515706 2.40000010 .8324045995728 3102
 2.54999995 .803586799 2.70000005 .781502901 2.79999995 .7566564095728 3102
 3.00000000 .726321306 3.20000005 .698717025 3.40000010 .6735123945728 3102
 3.59999990 .650113662 3.79999995 .628541218 4.00000000 .6062990895728 3102
 4.25000000 .583326845 4.50000000 .562202761 4.75000000 .5427964895728 3102
 5.00000000 .524548155 5.25000000 .507547920 5.50000000 .4919011885728 3102
 5.75000000 .476953842 6.00000000 .461844036 6.30000019 .4460978025728 3102
 6.59999990 .431653795 6.90000010 .417874545 7.19999981 .4031649715728 3102
 7.59999990 .387421500 8.00000000 .372904078 8.39999962 .3591470755728 3102
 8.80000019 .346393405 9.19999981 .334728830 9.60000038 .3234366485728 3102
 10.0000000 .311655490 10.5000000 .299489725 11.0000000 .2881600605728 3102
 11.5000000 .277598486 12.0000000 .265340267 12.7500000 .2518482985728 3102
 13.5000000 .239496295 14.2500000 .228159008 15.0000000 .2160590055728 3102
 16.0000000 .203554474 17.0000000 .192244019 18.0000000 .1819503825728 3102
 19.0000000 .172599573 20.0000000 .164032558 21.0000000 .1561493165728 3102
 22.0000000 .148866331 23.0000000 .142145255 24.0000000 .1345352775728 3102
 25.5000000 .126274331 27.0000000 .120040369 28.0000000 .1132982805728 3102
 30.0000000 .105381402 32.0000000 .098493094 34.0000000 .0925493295728 3102
 36.0000000 .087415552 38.0000000 .083031106 40.0000000 .0789771265728 3102
 42.5000000 .075518280 45.0000000 .073173495 47.5000000 .0721248275728 3102
 50.0000000 .072673156 52.5000000 .075468582 55.0000000 .0816682855728 3102
 57.5000000 .093733114 60.0000000 .121278817 63.0000000 .1964426155728 3102
 66.0000000 .490808755 69.0000000 38.8167148 72.0000000 111.1919955728 3102
 76.0000000 .304919017 80.0000000 .111430450 84.0000000 .0641543615728 3102
 88.0000000 .044870529 92.0000000 .034798476 96.0000000 .0286624405728 3102
 100.000000 .024124733 105.000000 .020618111 110.000000 .0180647125728 3102
 115.000000 .016096787 120.000000 .014192109 127.500000 .0124010315728 3102
 135.000000 .010992289 142.500000 .009849928 150.000000 .0087673255728 3102
 160.000000 .007756086 170.000000 .006928042 180.000000 .0062457545728 3102
 190.000000 .005669140 200.000000 .005180843 210.000000 .0047636395728 3102
 220.000000 .004408720 230.000000 .004138393 240.000000 .1080888905728 3102
 255.000000 .003454009 270.000000 .003153382 280.000000 .0028934845728 3102
 300.000000 .002615505 320.000000 .065215765 340.000000 .0061120735728 3102
 360.000000 .002028551 380.000000 .001885322 400.000000 .0017555255728 3102
 425.000000 .001636953 450.000000 .001541719 475.000000 .0014665725728 3102
 500.000000 .001412820 525.000000 .001392752 550.000000 .0014553985728 3102
 575.000000 .002027849 600.000000 2.07704896 630.000000 .0022773975728 3102
 660.000000 .001569452 690.000000 .640595066 720.000000 .0012771835728 3102
 760.000000 .001170377 800.000000 .001296779 840.000000 .9758432225728 3102
 880.000000 .307883942 920.000000 .004246978 960.000000 .7127309865728 3102
 1000.00000 .002136035 1050.00000 .003638613 1100.00000 .0134021445728 3102
 1150.00000 1.60569954 1200.00000 .516360996 1275.00000 .0024041835728 3102
 1350.00000 .003761734 1425.00000 .136001840 1500.00000 .0009020785728 3102
 1600.00000 .374778034 1700.00000 .000940109 1800.00000 .0632213855728 3102
 1900.00000 .161078382 2000.00000 .017721336 2100.00000 1.140632905728 3102
 2200.00000 .003805450 2300.00000 .250131212 2400.00000 .3266767325728 3102
 2550.00000 .105724204 2700.00000 .001786788 2800.00000 .2871656745728 3102
 3000.00000 .078072363 3200.00000 .211893608 3400.00000 .2297869445728 3102
 3600.00000 .362415424 3800.00000 .091087950 4000.00000 .0710547765728 3102
 4250.00000 .185614814 4500.00000 .274804507 4750.00000 .1063477185728 3102
 5000.00000 .090699012 5250.00000 .166010721 5500.00000 .0889003525728 3102
 5750.00000 .263216372 6000.00000 .062719930 6300.00000 .1870963565728 3102
 6600.00000 .094966934 6900.00000 .231505410 7200.00000 .1428706705728 3102
 7600.00000 .083189902 8000.00000 .135855675 8400.00000 .1308038955728 3102
 8800.00000 .079543336 9200.00000 .057390732 9600.00000 .1698050635728 3102
 10000.0000 .095655986 10500.0000 .076190286 11000.0000 .0605745495728 3102
 11500.0000 .099569645 12000.0000 .091689585 12750.0000 .0743117505728 3102
 13500.0000 .059266584 14250.0000 .075003631 15000.0000 .0496516195728 3102
 16000.0000 .065352560 17000.0000 .060008600 18000.0000 .0492534465728 3102
 19000.0000 .046632142 20000.0000 .053979643 21000.0000 .0523238465728 3102
 22000.0000 .050783408 23000.0000 .049344663 24000.0000 .0476748405728 3102
 25500.0000 .045824588 27000.0000 .044399469 28000.0000 .0428424855728 3102
 30000.0000 .040906568 32000.0000 .039149431 34000.0000 .0375525075728 3102
 36000.0000 .036092256 38000.0000 .034751876 40000.0000 .0333649775728 3102
 42500.0000 .031957755 45000.0000 .030675222 47500.0000 .0295016015728 3102
 50000.0000 .028434482 52500.0000 .027462854 55000.0000 .0265665975728 3102
 57500.0000 .025737611 60000.0000 .024878936 63000.0000 .0240020325728 3102
 66000.0000 .023205451 69000.0000 .022446899 72000.0000 .0216415645728 3102
 76000.0000 .020797765 80000.0000 .020041074 84000.0000 .0193688115728 3102
 88000.0000 .018758350 92000.0000 .018163677 96000.0000 .0176087655728 3102
 100000.000 .017053673 105000.000 .016509385 110000.000 .0160154425728 3102
 115000.000 .015566041 120000.000 .015066526 127500.000 .0145490835728 3102
 135000.000 .014099105 142500.000 .013666508 150000.000 .0132054995728 3102
 160000.000 .012550055 170000.000 .011703463 180000.000 .0109078935728 3102
 190000.000 .010272784 200000.000 .009794410 210000.000 .0094180525728 3102
 220000.000 .009107929 230000.000 .008844160 240000.000 .0085619245728 3102
 255000.000 .008270502 270000.000 .008061967 280000.000 .0078451465728 3102
 300000.000 .007596099 320000.000 .007385542 340000.000 .0072139015728 3102
 360000.000 .007070845 380000.000 .006951513 400000.000 .0068385595728 3102
 425000.000 .006742198 450000.000 .006672964 475000.000 .0066163595728 3102
 500000.000 .006579159 525000.000 .006561226 550000.000 .0065507515728 3102
 575000.000 .006556875 600000.000 .006572463 630000.000 .0065982085728 3102
 660000.000 .006638844 690000.000 .006681843 720000.000 .0067492445728 3102
 760000.000 .006830836 800000.000 .006928983 840000.000 .0070320165728 3102
 880000.000 .007149792 920000.000 .007279764 960000.000 .0074257055728 3102
 1000000.00 .007753469 1100000.00 .008363678 1200000.00 .0090764805728 3102
 1300000.00 .009508900 1400000.00 .009378977 1500000.00 .0088884055728 3102
 1600000.00 .008338516 1700000.00 .007841272 1800000.00 .0074041655728 3102
 1900000.00 .007013814 2000000.00 .006658618 2100000.00 .0063311835728 3102
 2200000.00 .006027052 2300000.00 .005743452 2400000.00 .0054785075728 3102
 2500000.00 .005230797 2600000.00 .005001541 2700000.00 .0047847695728 3102
 2800000.00 .004581968 2900000.00 .004391955 3000000.00 .0042142745728 3102
 3100000.00 .004047834 3200000.00 .003891622 3300000.00 .0037452535728 3102
 3400000.00 .003607809 3500000.00 .003478471 3600000.00 .0033563605728 3102
 3700000.00 .003242869 3800000.00 .003134944 3900000.00 .0030314815728 3102
 4000000.00 .002936409 4100000.00 .002845264 4200000.00 .0027587345728 3102
 4300000.00 .002676819 4400000.00 .002598874 4500000.00 .0025248995728 3102
 4600000.00 .002454351 4700000.00 .002387231 4800000.00 .0023230815728 3102
 4900000.00 .002261901 5000000.00 .002203299 5100000.00 .0021472745728 3102
 5200000.00 .002093494 5300000.00 .002041959 5400000.00 .0019923895728 3102
 5500000.00 .001944784 5600000.00 .001898896 5700000.00 .0018547265728 3102
 5800000.00 .001812066 5900000.00 .001770916 6000000.00 .0017314885728 3102
 6100000.00 .001693783 6200000.00 .001656078 6300000.00 .0016198545728 3102
 6400000.00 .001585113 6500000.00 .001550371 6600000.00 .0015171935728 3102
 6700000.00 .001485580 6800000.00 .001453967 6900000.00 .0014234275728 3102
 7000000.00 .001393960 7100000.00 .001364493 7200000.00 .0013362315728 3102
 7300000.00 .001309173 7400000.00 .001282114 7500000.00 .0012559635728 3102
 7600000.00 .001230718 7700000.00 .001205473 7800000.00 .0011813835728 3102
 7900000.00 .001158448 8000000.00 .001135513 8100000.00 .0011135945728 3102
 8200000.00 .001092693 8300000.00 .001071791 8400000.00 .0010520455728 3102
 8500000.00 .001033455 8600000.00 .001015878 8700000.00 .0009993135728 3102
 8800000.00 .000983908 8900000.00 .000969661 9000000.00 .0009567295728 3102
 9100000.00 .000945110 9200000.00 .000934958 9300000.00 .0009262715728 3102
 9400000.00 .000919177 9500000.00 .000913676 9600000.00 .0009098475728 3102
 9700000.00 .000907690 9800000.00 .000907412 9900000.00 .0009090135728 3102
 10000000.0 .000911507 10100000.0 .000916185 10200000.0 .0009213895728 3102
 10300000.0 .000929159 10400000.0 .000937356 10500000.0 .0009468245728 3102
 10600000.0 .000957565 10700000.0 .000968504 10800000.0 .0009806325728 3102
 10900000.0 .000992958 11000000.0 .001005284 11100000.0 .0010176095728 3102
 11200000.0 .001029935 11300000.0 .001042261 11400000.0 .0010545875728 3102
 11500000.0 .001066156 11600000.0 .001076968 11700000.0 .0010875535728 3102
 11800000.0 .001096776 11900000.0 .001105771 12000000.0 .0011136955728 3102
 12100000.0 .001120547 12200000.0 .001127112 12300000.0 .0011319515728 3102
 12400000.0 .001136503 12500000.0 .001139873 12600000.0 .0011420615728 3102
 12700000.0 .001143954 12800000.0 .001144073 12900000.0 .0011438975728 3102
 13000000.0 .001142553 13100000.0 .001140041 13200000.0 .0011372435728 3102
 13300000.0 .001132729 13400000.0 .001127929 13500000.0 .0011220115728 3102
 13600000.0 .001114975 13700000.0 .001107668 13800000.0 .0010987315728 3102
 13900000.0 .001089523 14000000.0 .001079273 14100000.0 .0010679815728 3102
 14200000.0 .001056443 14300000.0 .001043429 14400000.0 .0010301695728 3102
 14500000.0 .001016003 14600000.0 .001000932 14700000.0 .0009856595728 3102
 14800000.0 .000969183 14900000.0 .000952506 15000000.0 .0009349815728 3102
 15100000.0 .000916607 15200000.0 .000898233 15300000.0 .0008798375728 3102
 15400000.0 .000860962 15500000.0 .000841450 15600000.0 .0008214985728 3102
 15700000.0 .000801545 15800000.0 .000781593 15900000.0 .0007616415728 3102
 16000000.0 .000741688 16100000.0 .000721736 16200000.0 .0007017845728 3102
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 16600000.0 .000623201 16700000.0 .000604179 16800000.0 .0005858265728 3102
 16900000.0 .000567585 17000000.0 .000549887 17100000.0 .0005327315728 3102
 17200000.0 .000515727 17300000.0 .000499636 17400000.0 .0004836975728 3102
 17500000.0 .000468404 17600000.0 .000453757 17700000.0 .0004392755728 3102
 17800000.0 .000425780 17900000.0 .000412451 18000000.0 .0003997725728 3102
 18100000.0 .000387745 18200000.0 .000375874 18300000.0 .0003649435728 3102
 18400000.0 .000354170 18500000.0 .000343988 18600000.0 .0003343975728 3102
 18700000.0 .000324944 18800000.0 .000316315 18900000.0 .0003078235728 3102
 19000000.0 .000299834 19100000.0 .000292347 19200000.0 .0002849755728 3102
 19300000.0 .000278285 19400000.0 .000271708 19500000.0 .0002655395728 3102
 19600000.0 .000259778 19700000.0 .000254108 19800000.0 .0002489825728 3102
 19900000.0 .000243947 20000000.0 0.0                             5728 3102
                                                                  5728 3  0
                                                                  5728 0  0
 5.71390E+4 1.37713E+2          0          0          0          1572833102
 0.000000+0 0.000000+0          0        102          0          3572833102
 0.000000+0 0.000000+0          1          5        171         18572833102
 1.000000-5 1.000000-4 1.000000-3 1.000000-2 2.000000-2 3.000000-2572833102
 1.000000-1 3.000000-1 5.000000-1 1.000000+0 3.000000+0 1.000000+1572833102
 3.000000+1 6.500000+1 8.000000+1 1.275000+2 2.000000+2 2.000000+7572833102
 3.767710-3 2.067300-3 2.139120-3 1.031770-3 6.641270-4 1.917110-3572833102
 1.953300-3 1.918890-3 2.400260-3 2.000230-3 1.928820-3 1.972660-3572833102
 1.644780-3 2.502550-3 1.626730-3 1.707750-3 0.000000+0 3.561890-3572833102
 1.793850-3 1.122410-3 6.875010-4 1.767050-3 1.826970-3 1.726430-3572833102
 2.252850-3 1.886500-3 1.788400-3 1.855740-3 1.533740-3 2.341410-3572833102
 1.524870-3 1.586460-3 0.000000+0 3.473920-3 1.593070-3 7.020780-4572833102
 1.948170-3 1.970670-3 1.931570-3 2.440230-3 2.033130-3 1.953630-3572833102
 2.004810-3 1.668060-3 2.540030-3 1.651850-3 1.730200-3 0.000000+0572833102
 2.132550-3 7.478990-4 9.323270-4 1.081060-3 9.104130-4 1.310740-3572833102
 1.112720-3 1.016470-3 1.087210-3 8.825080-4 1.356500-3 8.868590-4572833102
 9.059200-4 0.000000+0 6.664840-4 6.696790-4 6.745820-4 6.524970-4572833102
 8.165560-4 6.827530-4 6.571870-4 6.725670-4 5.606550-4 8.530610-4572833102
 5.544760-4 5.821580-4 0.000000+0 2.389520-3 1.894290-3 1.762430-3572833102
 2.284250-3 1.909940-3 1.818580-3 1.879830-3 1.557320-3 2.375260-3572833102
 1.546140-3 1.612410-3 0.000000+0 2.425000-3 2.081500-3 2.332430-3572833102
 1.919730-3 1.889840-3 1.905170-3 1.601770-3 2.429640-3 1.576590-3572833102
 1.668390-3 0.000000+0 2.175610-3 2.372830-3 1.886870-3 1.806130-3572833102
 1.868630-3 1.545720-3 2.359900-3 1.537840-3 1.596370-3 0.000000+0572833102
 3.401930-3 2.444360-3 2.279290-3 2.370740-3 1.958590-3 2.989980-3572833102
 1.947030-3 2.026610-3 0.000000+0 2.267330-3 1.989240-3 1.949760-3572833102
 1.647100-3 2.494680-3 1.617180-3 1.718630-3 0.000000+0 2.152470-3572833102
 1.957590-3 1.558050-3 2.394310-3 1.564730-3 1.603350-3 0.000000+0572833102
 2.316060-3 1.674220-3 2.457340-3 1.587250-3 1.713230-3 0.000000+0572833102
 1.593640-3 2.044400-3 1.329110-3 1.370750-3 0.000000+0 4.144240-3572833102
 2.022380-3 2.119610-3 0.000000+0 1.549840-3 1.413310-3 0.000000+0572833102
 2.109240-3 0.000000+0 0.000000+0                                 572833102
 0.000000+0 0.000000+0          0          1         20         10572833102
 1.000000-5 0.000000+0 2.000000+2 2.345650-1 3.000000+2 3.374160-1572833102
 5.000000+2 2.215830-1 1.000000+3 1.670580-1 2.200000+3 9.646030-2572833102
 4.750000+3 1.530000-1 1.000000+4 2.445570-1 2.000000+4 0.000000+0572833102
 2.000000+7 0.000000+0                                            572833102
 0.000000+0 0.000000+0          1          5        325         25572833102
 1.000000-5 2.000000+4 4.000000+4 6.000000+4 8.000000+4 1.000000+5572833102
 2.000000+5 4.000000+5 6.000000+5 8.000000+5 1.000000+6 2.000000+6572833102
 3.000000+6 4.000000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6572833102
 9.000000+6 1.000000+7 1.200000+7 1.400000+7 1.600000+7 1.800000+7572833102
 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102
 0.000000+0 4.373210-3 4.257650-3 4.190390-3 4.171110-3 4.148810-3572833102
 4.161440-3 3.619020-3 2.457010-3 1.445370-3 1.531360-3 8.031520-4572833102
 1.043540-3 1.200720-3 1.215670-3 1.172440-3 1.144140-3 1.175140-3572833102
 1.194820-3 1.182890-3 1.164400-3 1.307070-3 1.656590-3 1.955650-3572833102
 4.480110-3 4.497000-3 4.428800-3 4.177620-3 4.221330-3 3.656530-3572833102
 2.681970-3 1.606600-3 1.534450-3 8.608040-4 1.089460-3 1.307050-3572833102
 1.346700-3 1.277400-3 1.206560-3 1.240720-3 1.298720-3 1.259650-3572833102
 1.263580-3 1.392600-3 1.800410-3 2.077550-3 4.599630-3 4.605670-3572833102
 4.281150-3 4.258790-3 3.657810-3 2.547890-3 1.545390-3 1.541850-3572833102
 8.321630-4 1.060020-3 1.239290-3 1.268460-3 1.219390-3 1.174940-3572833102
 1.203660-3 1.237920-3 1.216770-3 1.206490-3 1.344290-3 1.716880-3572833102
 2.010090-3 4.742920-3 4.459620-3 4.291190-3 3.680760-3 2.327280-3572833102
 1.405120-3 1.562540-3 7.833310-4 1.022270-3 1.137270-3 1.142630-3572833102
 1.123670-3 1.124920-3 1.148600-3 1.141660-3 1.150330-3 1.115430-3572833102
 1.269810-3 1.584080-3 1.905780-3 4.605220-3 4.273810-3 3.718160-3572833102
 2.442330-3 1.420210-3 1.576280-3 8.058040-4 1.059450-3 1.197410-3572833102
 1.202570-3 1.168140-3 1.156000-3 1.186970-3 1.191920-3 1.190400-3572833102
 1.162340-3 1.315120-3 1.652630-3 1.970050-3 4.650030-3 3.785000-3572833102
 2.297160-3 1.362930-3 1.607150-3 8.019660-4 1.059440-3 1.177090-3572833102
 1.176780-3 1.153250-3 1.155170-3 1.183480-3 1.175920-3 1.183830-3572833102
 1.148000-3 1.307290-3 1.631080-3 1.960810-3 4.191990-3 3.716890-3572833102
 2.220880-3 1.586800-3 9.805650-4 1.200080-3 1.483940-3 1.552120-3572833102
 1.462450-3 1.354480-3 1.388990-3 1.478980-3 1.419310-3 1.438570-3572833102
 1.569140-3 2.050230-3 2.338900-3 6.013730-3 5.613730-3 1.648130-3572833102
 1.136680-3 1.104900-3 1.402660-3 1.605800-3 1.597240-3 1.462240-3572833102
 1.416670-3 1.530810-3 1.479160-3 1.507310-3 1.624740-3 2.135960-3572833102
 2.441640-3 9.596400-3 2.815990-3 9.693480-4 1.103800-3 1.023690-3572833102
 1.074000-3 1.221770-3 1.341960-3 1.284610-3 1.173230-3 1.279090-3572833102
 1.172100-3 1.398410-3 1.645660-3 2.135440-3 5.787750-3 1.657310-3572833102
 2.205260-3 2.617220-3 2.669340-3 2.556840-3 2.464170-3 2.529390-3572833102
 2.598890-3 2.555700-3 2.531740-3 2.824340-3 3.602280-3 4.223850-3572833102
 3.633110-3 4.267780-3 4.641970-3 4.800960-3 4.840430-3 4.849650-3572833102
 4.869290-3 4.846410-3 4.898520-3 4.754970-3 5.395380-3 6.744660-3572833102
 8.112480-3 5.966690-3 6.739140-3 6.708790-3 6.302550-3 5.970150-3572833102
 6.109290-3 6.347980-3 6.211780-3 6.184360-3 6.864580-3 8.790700-3572833102
 1.027570-2 8.063440-3 8.221160-3 7.668950-3 7.008490-3 6.952500-3572833102
 7.356010-3 7.186020-3 7.197940-3 7.924890-3 1.020480-2 1.189920-2572833102
 8.717380-3 8.454870-3 7.820360-3 7.406320-3 7.592980-3 7.632630-3572833102
 7.502130-3 8.380310-3 1.061400-2 1.263480-2 8.701350-3 8.438740-3572833102
 7.794530-3 7.432500-3 7.718140-3 7.386790-3 8.453560-3 1.047490-2572833102
 1.273760-2 8.742840-3 8.359810-3 7.410060-3 7.578710-3 7.242570-3572833102
 8.353630-3 1.032520-2 1.250930-2 8.829350-3 8.074660-3 7.385510-3572833102
 7.413820-3 8.264620-3 1.053090-2 1.236150-2 8.774640-3 7.717270-3572833102
 7.483530-3 8.327420-3 1.057710-2 1.255790-2 8.949390-3 7.595650-3572833102
 8.383470-3 1.059340-2 1.256530-2 8.513730-3 8.420100-3 1.022820-2572833102
 1.243340-2 1.077640-2 1.217860-2 1.364410-2 1.682810-2 1.804110-2572833102
 2.429660-2                                                       572833102
                                                                  572833  0
                                                                  5728 0  0
                                                                     0 0  0
 5.91410E+4 1.39697E+2          0          0         34         105925 1451
 0.0        0.0                 0          0          0          65925 1451
 1.00000E+0 2.00000E+7          0          0         10       20025925 1451
 3.00000E+2 0.0                 1          0        122          35925 1451
 59-Pr-141 FEI        EVAL-Jan97 K.I.Zolotarev et al.             5925 1451
                      DIST-Feb2004                                5925 1451
----IRDF-2002         MATERIAL 5925                               5925 1451
-----INCIDENT NEUTRON DATA                                        5925 1451
------ENDF-6 FORMAT                                               5925 1451
***************************************************************** 5925 1451
 59-PR-141 FEI        EVAL-Jan97 K.I.Zolotarev et al.             5925 1451
                      DIST-Jan02                       20020123   5925 1451
----BROND-2           MATERIAL 5925                               5925 1451
-----INCIDENT NEUTRON DATA                                        5925 1451
------ENDF-6 FORMAT                                               5925 1451
------Russian Reactor Dosimetry File RRDF-2002                    5925 1451
***************************************************************** 5925 1451
Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko      5925 1451
***************************************************************** 5925 1451
MF= 3                                                             5925 1451
MT= 16 - Pr-141(n,2n)Pr-140 reaction                              5925 1451
--------------------------------------                            5925 1451
    Excitation function for the  Pr-141(n,2n)Pr-140  reaction in  5925 1451
the energy region from threshold to 20 MeV was evaluated by means 5925 1451
of statistical analysis of experimental cross section data [1-19] 5925 1451
and data from systematics  of (n,2n)  excitation function [20] in 5925 1451
the energy regions 9.5 - 12.0 MeV and 18.0 - 20.0 MeV.            5925 1451
    All experimental data were renormalized to  the new standards 5925 1451
for monitor reactions cross sections and decay data.              5925 1451
    The final procedure of evaluation has been carried out within 5925 1451
the framework of generalized least squares method. Rational func- 5925 1451
tion was used as model function [21].  Calculations was performed 5925 1451
by means of Pade-2 code [22].                                     5925 1451
    U-235 thermal fission [23]   and   Cf-252 spontaneous fission 5925 1451
neutron spectra [24]  averaged cross-sections calculated from the 5925 1451
evaluated Pr141(n,2n)Pr140 excitation function are the following: 5925 1451
                                                                  5925 1451
         --------------------------------------------             5925 1451
              TYPE OF SPECTRUM     I  CS>, mb (calc.)             5925 1451
         --------------------------I-----------------             5925 1451
         U-235  neutron  fission   I     1.0922                   5925 1451
         CF-252 spontan. fission   I     1.9843                   5925 1451
                                                                  5925 1451
MF=33                                                             5925 1451
MT= 16 -(n,2n) cross section cov. matrix                          5925 1451
----------------------------------------                          5925 1451
    Uncertainties in the evaluated excitation function for the re-5925 1451
action Pr-141(n,2n)Pr-140 are given in the form of relative cova- 5925 1451
riance matrix for the 18-neutron energy groups (LB=5). Covariance 5925 1451
matrix  of  uncertainties  was  calculated  simultaneously  with  5925 1451
recommended cross section data by means of PADE-2 code.           5925 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  5925 1451
given in the 33-file are the following:                           5925 1451
                                                                  5925 1451
    6.47369E-06    6.50089E-06    6.56924E-06    6.70314E-06      5925 1451
    6.90792E-06    7.20594E-06    7.64163E-06    8.30022E-06      5925 1451
    9.35563E-06    1.12079E-05    1.49213E-05    2.36154E-05      5925 1451
    4.48092E-05    3.28699E-04    6.81059E-03    1.70941E-02      5925 1451
    2.55798E-02    1.53270E-01                                    5925 1451
                                                                  5925 1451
References :                                                      5925 1451
 1. E.B.Paul, R.L.Clarke  Canadian J. Phys., v.31,p.267, 1953     5925 1451
 2. J.M.Ferguson, W.E.Thompson  Phys. Rev., v.118, p.228, 1960    5925 1451
 3. R.G.Wille, R.W.Fink  Phys. Rev., v.118, p.242, 1960           5925 1451
 4. L.A.Rayburn  Phys. Rev., v.122, p.168, 1961                   5925 1451
 5. C.S.Khurana, H.S.Hans  Nucl. Phys., v.28, p.560, 1961         5925 1451
 6. D.R.Koehler, W.L.Alford  Report NP-11667, 1962                5925 1451
 7. M.Cevolani, S.Petralia  Nucl. Sci. Eng., v.26, p.1328, 1962   5925 1451
 8. L.A.Rayburn  Bull. American Phys. Soc., v.8, p.60, Jan. 1963  5925 1451
 9. M.P.Menon, M.Y.Cuypers  Phys. Rev., v.156, p.1340, 1967       5925 1451
10. P.Cuzzocrea e.a. Nuovo Cimento, Sec.B, v.52, n.2, p.476, 1967 5925 1451
11. M.Bormann e.a.  Nucl. Phys., Sec.A, v.115, p.309, July 1968   5925 1451
12. G.Peto e.a.  Acta Physica Hung., v.24, p.93, April 1968       5925 1451
13. A.Chatterjee e.a.  Proc. of 12th Nucl. Phys. and Solid State  5925 1451
    Phys. Sympos., Roorkee, India, 28-31 December 1969, v.2,p.117 5925 1451
14. A.Bari   Dissertation Abstracts, sec.B, v.32, p.5091, 1972    5925 1451
15. J.Araminowicz, J.Dresler  Prog. Rep. INR-1464, p.14, May 1973 5925 1451
16. R.A.Sigg, P.K.Kuroda  Inorg. Nucl. Chem., v.37, p.631, 1975   5925 1451
17. M.Valkonen  Report JU-RR-1/1976, Jyvaeskylae University, 1976 5925 1451
18. S.Murahira et al.  Report INDC(JPN)-175/U, p.171, Nov. 1995   5925 1451
19. Y.Kasugai, Y.Ikeda, Y.Uno Proc. of Int. Conf. on Nuclear Data 5925 1451
    for Science and Technology, Trieste, Italy, 19-24 May 1997,   5925 1451
    v.1, p.635                                                    5925 1451
20. V.N.Manokhin  Report INDC(CCP)-397, IAEA, Vienna, 1997        5925 1451
21. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 5925 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 5925 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  5925 1451
22. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          5925 1451
23. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 5925 1451
    Library, MAT=9228, MF=5, MT=18, eval. April 1989              5925 1451
24. W.Mannhart   IAEA-TECDOC-410, p.158, Vienna, 1987             5925 1451
******************************************************************5925 1451
 The Q values and threshold energies were updated prior to pro-   5925 1451
 cessing through the codes to comply with the values obtained     5925 1451
 using the NNDC calculation program which is based on the 1995    5925 1451
 Update to the Atomic mass Evaluation.                            5925 1451
                                                                  5925 1451
 File 2 added to the pointwise file containing only the effective 5925 1451
 scattering radius with no resonance parameters given.            5925 1451
 Taken from ENDF/B-VI                                             5925 1451
******************************************************************5925 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************5925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     5925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     5925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************5925 1451
 Data Doppler Broadened to 300.000000 Kelvin                      5925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     5925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     5925 1451
 ***************** Program FIXUP (Version 2002-1) ****************5925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 5925 1451
 Corrected Thresholds-----------------------------------------Yes 5925 1451
 Extended Cross Sections to 20 MeV----------------------------No  5925 1451
 Allow Cross Section Deletion---------------------------------No  5925 1451
 Allow Cross Section Reconstruction---------------------------No  5925 1451
 Make All Cross Sections Non-Negative-------------------------Yes 5925 1451
 Delete Energies Not in Ascending Order-----------------------Yes 5925 1451
 Deleted Duplicate Points-------------------------------------Yes 5925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 5925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 5925 1451
 Allow Creation of Missing Sections---------------------------No  5925 1451
 Allow Insertion of Energy Points-----------------------------No  5925 1451
 Create Uniform Energy Grid-----------------------------------No  5925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 5925 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************5925 1451
 Unshielded Group Averages Using  640 Groups                      5925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     5925 1451
                                1        451        129          15925 1451
                                3         16         39          15925 1451
                               33         16         38          15925 1451
                                                                  5925 1  0
                                                                  5925 0  0
 5.91410E+4 1.39697E+2          0          0          0          05925 3 16
-9.39700E+6-9.39700E+6          0          0          1        1075925 3 16
        107          1                                            5925 3 16
 9400000.00 .000102039 9500000.00 .004115304 9600000.00 .0164385955925 3 16
 9700000.00 .038856950 9800000.00 .070499450 9900000.00 .1102883505925 3 16
 10000000.0 .157013000 10100000.0 .209402000 10200000.0 .2661885005925 3 16
 10300000.0 .326164000 10400000.0 .388315750 10500000.0 .4513412505925 3 16
 10600000.0 .514366750 10700000.0 .577392250 10800000.0 .6395045005925 3 16
 10900000.0 .699850500 11000000.0 .758343500 11100000.0 .8147340005925 3 16
 11200000.0 .868854500 11300000.0 .920604500 11400000.0 .9699370005925 3 16
 11500000.0 1.01684800 11600000.0 1.06136500 11700000.0 1.102987505925 3 16
 11800000.0 1.14290250 11900000.0 1.18068500 12000000.0 1.216335005925 3 16
 12100000.0 1.25004500 12200000.0 1.28181500 12300000.0 1.311847505925 3 16
 12400000.0 1.34014250 12500000.0 1.36689000 12600000.0 1.392090005925 3 16
 12700000.0 1.41561500 12800000.0 1.43746500 12900000.0 1.459315005925 3 16
 13000000.0 1.47946167 13100000.0 1.49790500 13200000.0 1.516348335925 3 16
 13300000.0 1.53338000 13400000.0 1.54900000 13500000.0 1.564620005925 3 16
 13600000.0 1.57889750 13700000.0 1.59183250 13800000.0 1.604767505925 3 16
 13900000.0 1.61770250 14000000.0 1.62927600 14100000.0 1.639488005925 3 16
 14200000.0 1.64970000 14300000.0 1.65991200 14400000.0 1.670124005925 3 16
 14500000.0 1.67916600 14600000.0 1.68703800 14700000.0 1.694910005925 3 16
 14800000.0 1.70278200 14900000.0 1.71065400 15000000.0 1.717540835925 3 16
 15100000.0 1.72344250 15200000.0 1.72934417 15300000.0 1.735245835925 3 16
 15400000.0 1.74114750 15500000.0 1.74704917 15600000.0 1.752039295925 3 16
 15700000.0 1.75611786 15800000.0 1.76019643 15900000.0 1.764275005925 3 16
 16000000.0 1.76835357 16100000.0 1.77243214 16200000.0 1.776510715925 3 16
 16300000.0 1.77971063 16400000.0 1.78203188 16500000.0 1.784353135925 3 16
 16600000.0 1.78667438 16700000.0 1.78899563 16800000.0 1.791316885925 3 16
 16900000.0 1.79363813 17000000.0 1.79595938 17100000.0 1.797471675925 3 16
 17200000.0 1.79817500 17300000.0 1.79887833 17400000.0 1.799581675925 3 16
 17500000.0 1.80028500 17600000.0 1.80098833 17700000.0 1.800834175925 3 16
 17800000.0 1.79982250 17900000.0 1.79881083 18000000.0 1.797799175925 3 16
 18100000.0 1.79678750 18200000.0 1.79577583 18300000.0 1.793587005925 3 16
 18400000.0 1.79022100 18500000.0 1.78685500 18600000.0 1.783489005925 3 16
 18700000.0 1.78012300 18800000.0 1.77536333 18900000.0 1.769210005925 3 16
 19000000.0 1.76305667 19100000.0 1.75515667 19200000.0 1.745510005925 3 16
 19300000.0 1.73586333 19400000.0 1.72389250 19500000.0 1.709597505925 3 16
 19600000.0 1.69321500 19700000.0 1.67284000 19800000.0 1.648080005925 3 16
 19900000.0 1.61751500 20000000.0 0.0                             5925 3 16
                                                                  5925 3  0
                                                                  5925 0  0
 5.91410E+4 1.39697E+2          0          0          0          1592533 16
 0.000000+0 0.000000+0          0         16          0          1592533 16
 0.000000+0 0.000000+0          1          5        210         20592533 16
 1.000000-5 9.400000+6 1.050000+7 1.100000+7 1.150000+7 1.200000+7592533 16
 1.250000+7 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7592533 16
 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7592533 16
 1.850000+7 1.900000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16
 0.000000+0 0.000000+0 0.000000+0 7.629710-2 5.270640-2 3.376320-2592533 16
 2.074520-2 1.252350-2 7.563810-3 4.706180-3 3.188710-3 2.533760-3592533 16
 2.446920-3 2.749080-3 3.334790-3 4.148410-3 5.171510-3 6.419360-3592533 16
 7.945970-3 9.861990-3 1.237550-2 4.016490-2 2.809550-2 1.900490-2592533 16
 1.265780-2 8.366950-3 5.519320-3 3.674800-3 2.539500-3 1.923500-3592533 16
 1.707930-3 1.823370-3 2.237300-3 2.948350-3 3.986210-3 5.417190-3592533 16
 7.354640-3 9.962330-3 2.145550-2 1.571000-2 1.127530-2 7.970050-3592533 16
 5.552600-3 3.813640-3 2.597590-3 1.796650-3 1.340690-3 1.189230-3592533 16
 1.326580-3 1.760550-3 2.524460-3 3.682770-3 5.338130-3 7.621620-3592533 16
 1.236030-2 9.413210-3 7.029290-3 5.152190-3 3.707030-3 2.623120-3592533 16
 1.845220-3 1.335170-3 1.071090-3 1.046710-3 1.272260-3 1.777090-3592533 16
 2.614640-3 3.867430-3 5.635080-3 7.582390-3 5.943050-3 4.583860-3592533 16
 3.484360-3 2.620050-3 1.966010-3 1.503080-3 1.220170-3 1.115420-3592533 16
 1.197700-3 1.489140-3 2.029030-3 2.877480-3 4.106360-3 4.916750-3592533 16
 3.993780-3 3.224900-3 2.597250-3 2.101980-3 1.730500-3 1.477540-3592533 16
 1.342690-3 1.331920-3 1.459470-3 1.750620-3 2.244570-3 2.989940-3592533 16
 3.456170-3 2.967610-3 2.561130-3 2.226210-3 1.959650-3 1.758980-3592533 16
 1.623750-3 1.556590-3 1.564460-3 1.660280-3 1.865240-3 2.208980-3592533 16
 2.737410-3 2.515350-3 2.327170-3 2.163680-3 2.023420-3 1.905210-3592533 16
 1.808580-3 1.734270-3 1.685040-3 1.667360-3 1.695630-3 2.469300-3592533 16
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 2.705480-3 2.499580-3 3.001600-3 3.118840-3 3.217400-3 3.275390-3592533 16
 3.472370-3 3.866030-3 4.312850-3 4.687830-3 5.673930-3 7.407790-3592533 16
                                                                  592533  0
                                                                  5925 0  0
                                                                     0 0  0
 6.40000E+4 1.55901E+2         -1          0         34         106400 1451
 0.0        0.0                 0          0          0          66400 1451
 1.00000E+0 2.00000E+7          0          0         10       20026400 1451
 3.00000E+2 0.0                 1          0         59          26400 1451
 64-Gd-  0  NDS       IAEA-JAN04                                  6400 1451
                      DIST-Feb2004                                6400 1451
----IRDF-2002                                                     6400 1451
-----INCIDENT NEUTRON DATA                                        6400 1451
------ENDF-6 FORMAT                                               6400 1451
 *****************************************************************6400 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************6400 1451
DOSIMETRY  Assembled at NDS from the ENDF/B-VI evaluations for    6400 1451
           64-Gd-152,154,155,156,157,158 and 160. The files       6400 1451
           were processed through the 2002 pre-processing codes   6400 1451
           LINEAR,RECENT,SIGMA1 and FIXUP to produce pointwise    6400 1451
           files prior to input to MIXER.                         6400 1451
                                                                  6400 1451
MIXER INPUT FILE                                                  6400 1451
 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451
GADOLINIUM,MET.   DENSITY  7.901 G/CC. CONSTITUENTS:  160-21.86   6400 1451
 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65      6400 1451
 ----------------------------------------                         6400 1451
 Composition                                                      6400 1451
 ----------------------------------------                         6400 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                         6400 1451
 ----------------------------------------                         6400 1451
   64152  3    1  .001999808   .015264700                         6400 1451
   64154  3    1  .021799790   .168591500                         6400 1451
   64155  3    1  .147999470   1.15202110                         6400 1451
   64156  3    1  .204700250   1.60365810                         6400 1451
   64157  3    1  .156499785   1.23392290                         6400 1451
   64158  3    1  .248400703   1.97099140                         6400 1451
   64160  3    1  .218600194   1.75653450                         6400 1451
 ----------------------------------------                         6400 1451
   64000  3    1  1.00000000   7.90098420                         6400 1451
 ----------------------------------------                         6400 1451
 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451
GADOLINIUM,MET.   DENSITY  7.901 G/CC. CONSTITUENTS:  160-21.86   6400 1451
 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65      6400 1451
 ----------------------------------------                         6400 1451
 Composition                                                      6400 1451
 ----------------------------------------                         6400 1451
 Isotope MF   MT  Atom-Fract     Grams/cc                         6400 1451
 ----------------------------------------                         6400 1451
   64152  3  102  .001999808   .015264700                         6400 1451
   64154  3  102  .021799790   .168591500                         6400 1451
   64155  3  102  .147999470   1.15202110                         6400 1451
   64156  3  102  .204700250   1.60365810                         6400 1451
   64157  3  102  .156499785   1.23392290                         6400 1451
   64158  3  102  .248400703   1.97099140                         6400 1451
   64160  3  102  .218600194   1.75653450                         6400 1451
 ----------------------------------------                         6400 1451
   64000  3  102  1.00000000   7.90098420                         6400 1451
 ----------------------------------------                         6400 1451
 *****************************************************************6400 1451
                                                                  6400 1451
                                                                  6400 1451
                                                                  6400 1451
                                                                  6400 1451
 *****************************************************************6400 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************6400 1451
 Unshielded Group Averages Using  640 Groups                      6400 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     6400 1451
                                1        451         65          06400 1451
                                3          1        217          06400 1451
                                                                  6400 1  0
                                                                  6400 0  0
 6.40000E+4 1.55901E+2          0          0          0          06400 3  1
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        641          1                                            6400 3  1
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 2500000.00 6.37078706 2600000.00 6.32704875 2700000.00 6.283322386400 3  1
 2800000.00 6.23959601 2900000.00 6.19314033 3000000.00 6.142457286400 3  1
 3100000.00 6.09066487 3200000.00 6.03878275 3300000.00 5.986103296400 3  1
 3400000.00 5.93170828 3500000.00 5.87555926 3600000.00 5.819069596400 3  1
 3700000.00 5.76268152 3800000.00 5.70629345 3900000.00 5.649905386400 3  1
 4000000.00 5.59371016 4100000.00 5.53770780 4200000.00 5.481792206400 3  1
 4300000.00 5.43159956 4400000.00 5.38486083 4500000.00 5.341090516400 3  1
 4600000.00 5.30023290 4700000.00 5.25937631 4800000.00 5.218520406400 3  1
 4900000.00 5.18113675 5000000.00 5.14730080 5100000.00 5.113730106400 3  1
 5200000.00 5.08444163 5300000.00 5.06068140 5400000.00 5.036943056400 3  1
 5500000.00 5.01320469 5600000.00 4.99003581 5700000.00 4.971389656400 3  1
 5800000.00 4.95371758 5900000.00 4.93604551 6000000.00 4.918532966400 3  1
 6100000.00 4.90117992 6200000.00 4.88925335 6300000.00 4.878335386400 3  1
 6400000.00 4.86670072 6500000.00 4.85506606 6600000.00 4.843433476400 3  1
 6700000.00 4.83429814 6800000.00 4.82738682 6900000.00 4.820475506400 3  1
 7000000.00 4.81365421 7100000.00 4.80698616 7200000.00 4.803311816400 3  1
 7300000.00 4.80127939 7400000.00 4.79924696 7500000.00 4.800718896400 3  1
 7600000.00 4.80616541 7700000.00 4.81161581 7800000.00 4.817066216400 3  1
 7900000.00 4.82251661 8000000.00 4.83000689 8100000.00 4.840432106400 3  1
 8200000.00 4.85095573 8300000.00 4.86147935 8400000.00 4.872002986400 3  1
 8500000.00 4.88346054 8600000.00 4.89585205 8700000.00 4.908243556400 3  1
 8800000.00 4.92063505 8900000.00 4.93302656 9000000.00 4.945436686400 3  1
 9100000.00 4.95800229 9200000.00 4.97062120 9300000.00 4.983244306400 3  1
 9400000.00 4.99586785 9500000.00 5.00849140 9600000.00 5.021114956400 3  1
 9700000.00 5.03373850 9800000.00 5.04636205 9900000.00 5.058985606400 3  1
 10000000.0 5.07160915 10100000.0 5.08423271 10200000.0 5.096717116400 3  1
 10300000.0 5.10846077 10400000.0 5.12009543 10500000.0 5.131730106400 3  1
 10600000.0 5.14336476 10700000.0 5.15499943 10800000.0 5.166634096400 3  1
 10900000.0 5.17825492 11000000.0 5.18969603 11100000.0 5.201081876400 3  1
 11200000.0 5.21246771 11300000.0 5.22385355 11400000.0 5.231251816400 3  1
 11500000.0 5.23109654 11600000.0 5.23005744 11700000.0 5.217738096400 3  1
 11800000.0 5.20419925 11900000.0 5.19066042 12000000.0 5.177083466400 3  1
 12100000.0 5.16346838 12200000.0 5.15342648 12300000.0 5.149966776400 3  1
 12400000.0 5.14665396 12500000.0 5.14653915 12600000.0 5.149235366400 3  1
 12700000.0 5.15193157 12800000.0 5.15462778 12900000.0 5.157324006400 3  1
 13000000.0 5.16002021 13100000.0 5.16593640 13200000.0 5.179399976400 3  1
 13300000.0 5.19317636 13400000.0 5.20695274 13500000.0 5.220360286400 3  1
 13600000.0 5.22656845 13700000.0 5.22998405 13800000.0 5.230243256400 3  1
 13900000.0 5.22806018 14000000.0 5.22587386 14100000.0 5.223647106400 3  1
 14200000.0 5.22138800 14300000.0 5.21912889 14400000.0 5.216869796400 3  1
 14500000.0 5.21461069 14600000.0 5.21235098 14700000.0 5.207864546400 3  1
 14800000.0 5.20129401 14900000.0 5.19472347 15000000.0 5.188152936400 3  1
 15100000.0 5.18158239 15200000.0 5.17501185 15300000.0 5.168441316400 3  1
 15400000.0 5.16187077 15500000.0 5.15530023 15600000.0 5.148720686400 3  1
 15700000.0 5.14212068 15800000.0 5.13551972 15900000.0 5.128918776400 3  1
 16000000.0 5.12306018 16100000.0 5.11794396 16200000.0 5.112827746400 3  1
 16300000.0 5.10771151 16400000.0 5.10259529 16500000.0 5.097479076400 3  1
 16600000.0 5.09236285 16700000.0 5.08724663 16800000.0 5.082130406400 3  1
 16900000.0 5.07701418 17000000.0 5.07189796 17100000.0 5.066781746400 3  1
 17200000.0 5.06166552 17300000.0 5.05654929 17400000.0 5.051433076400 3  1
 17500000.0 5.04631685 17600000.0 5.04120063 17700000.0 5.036084416400 3  1
 17800000.0 5.03096818 17900000.0 5.02585196 18000000.0 5.020675146400 3  1
 18100000.0 5.01543771 18200000.0 5.01020029 18300000.0 5.004962876400 3  1
 18400000.0 4.99972544 18500000.0 4.99481980 18600000.0 4.990245936400 3  1
 18700000.0 4.98580967 18800000.0 4.98405926 18900000.0 4.983345976400 3  1
 19000000.0 4.98287553 19100000.0 4.98264794 19200000.0 4.982420356400 3  1
 19300000.0 4.98219276 19400000.0 4.98196517 19500000.0 4.981737576400 3  1
 19600000.0 4.98150998 19700000.0 4.98128239 19800000.0 4.981054806400 3  1
 19900000.0 4.98082721 20000000.0 0.0                             6400 3  1
                                                                  6400 3  0
                                                                  6400 0  0
                                                                     0 0  0
 6.91690E+4 1.67483E+2          0          0         34         106925 1451
 0.0        0.0                 0          0          0          66925 1451
 1.00000E+0 2.00000E+7          0          0         10       20026925 1451
 3.00000E+2 0.0                 1          0         88          36925 1451
 69-Tm-169 SRI        EVAL-JAN96 N.ODANO (SHIP RES. INST.)        6925 1451
                      DIST-Feb2004                                6925 1451
----IRDF-2002         MATERIAL 6925                               6925 1451
-----INCIDENT NEUTRON DATA                                        6925 1451
------ENDF-6 FORMAT                                               6925 1451
****************************************************************  6925 1451
 69-TM-169 SRI        EVAL-JAN96 N.ODANO (SHIP RES. INST.)        6925 1451
                      DIST-JUL98                                  6925 1451
----JENDL/D-99        MATERIAL 6925                               6925 1451
****************************************************************  6925 1451
HISTORY                                                           6925 1451
96-01   EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 6925 1451
        N.ODANO (SHIP RES. INST.).                                6925 1451
97-09   COMPILED TO JENDL DOSIMETRY FILE 99.                      6925 1451
                                                                  6925 1451
             ==== POINT-WISE DATA FILE ====                       6925 1451
                                                                  6925 1451
 <DOSIMETRY CROSS SECTION>                                        6925 1451
      TM-169 (N,2N) TM-168 (HALF-LIFE = 93.1 D)                   6925 1451
                                                                  6925 1451
MF=1  GENERAL INFORMATION                                         6925 1451
 MT=451  DESCRIPTIVE DATA AND DICTIONARY                          6925 1451
                                                                  6925 1451
MF=2  RESONANCE PARAMETERS                                        6925 1451
 MT=151  PARAMETERS                                               6925 1451
     ONLY SPIN AND SCATTERING RADIUS ARE GIVEN.                   6925 1451
                                                                  6925 1451
MF=3 NEUTRON CROSS SECTIONS                                       6925 1451
 MT=16 (N,2N) CROSS SECTION                                       6925 1451
       EXPERIMENTAL DATA/1-13/ IN NESTOR-2/14/ WERE TAKEN FOR     6925 1451
       THE EVALUATION USING GMA CODE/15/.                         6925 1451
                                                                  6925 1451
MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS                       6925 1451
 MT=16  GENERATED USING THE GMA CODE.                             6925 1451
                                                                  6925 1451
REFERENCES                                                        6925 1451
 1) W.DILG+ : NUCL. PHYS. A, 118, 9 (1968).                       6925 1451
 2) D.S.MATHER+ : AWRE-O-72/,72 (1972).                           6925 1451
 3) D.R.NETHERWAY : NUCL. PHYS. A, 190, 635 (1972).               6925 1451
 4) R.VOS+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 18,       6925 1451
    775(EB6) (1973).        6931 1451                             6925 1451
 5) S.M.QAIM : NUCL. PHYS. A, 224, 319 (1974).                    6925 1451
 6) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975).                 6925 1451
 7) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977).                  6925 1451
 8) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 6925 1451
    MEV, UPTON, USA, 12-14 MAY, 1980, P.399 (1980).               6925 1451
 9) J.LAUREC+ : CEA-R-5109 (1981).                                6925 1451
10) L.R.GREENWOOD : ASTM-STP-956, 743 (1987).                     6925 1451
11) LU HANLIN+ : INDC(CRP)-16 (1989).                             6925 1451
12) WANG XIUYUAN+ : A COLLECTION OF ACADEMIC EXCHANGE REPORT IN   6925 1451
    THE FIELDS OD RADIOCHEMISTRY, CHEMICAL INDUSTRY AND ISOTOPE,  6925 1451
    NUCLEAR INSTITUTE OF SICHUAN (1989).                          6925 1451
13) C.KONNO+ : JAERI 1329 (1993).                                 6925 1451
14) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED.      6925 1451
15) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS     6925 1451
    AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS-       6925 1451
    51363, P.249 (1981).                                          6925 1451
****************************************************************  6925 1451
  The Q values and threshold energies were updated prior to pro-  6925 1451
  cessing through the codes to comply with the values obtained    6925 1451
  using the NNDC calculation program which is based on the 1995   6925 1451
  Update to the Atomic mass Evaluation.                           6925 1451
****************************************************************  6925 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************6925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     6925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     6925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************6925 1451
 Data Doppler Broadened to 300.000000 Kelvin                      6925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     6925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     6925 1451
 ***************** Program FIXUP (Version 2002-1) ****************6925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 6925 1451
 Corrected Thresholds-----------------------------------------Yes 6925 1451
 Extended Cross Sections to 20 MeV----------------------------No  6925 1451
 Allow Cross Section Deletion---------------------------------No  6925 1451
 Allow Cross Section Reconstruction---------------------------No  6925 1451
 Make All Cross Sections Non-Negative-------------------------Yes 6925 1451
 Delete Energies Not in Ascending Order-----------------------Yes 6925 1451
 Deleted Duplicate Points-------------------------------------Yes 6925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 6925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 6925 1451
 Allow Creation of Missing Sections---------------------------No  6925 1451
 Allow Insertion of Energy Points-----------------------------No  6925 1451
 Create Uniform Energy Grid-----------------------------------No  6925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 6925 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************6925 1451
 Unshielded Group Averages Using  640 Groups                      6925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     6925 1451
                                1        451         95          16925 1451
                                3         16         44          16925 1451
                               33         16         23          16925 1451
                                                                  6925 1  0
                                                                  6925 0  0
 6.91690E+4 1.67483E+2          0          0          0          06925 3 16
-8.03340E+6-8.03340E+6          0          0          1        1216925 3 16
        121          1                                            6925 3 16
 8000000.00 .000162775 8100000.00 .006767183 8200000.00 .0167043106925 3 16
 8300000.00 .026641437 8400000.00 .070615833 8500000.00 .1486275006925 3 16
 8600000.00 .226639167 8700000.00 .304650833 8800000.00 .3826625006925 3 16
 8900000.00 .460674167 9000000.00 .535816000 9100000.00 .6080880006925 3 16
 9200000.00 .680360000 9300000.00 .752632000 9400000.00 .8249040006925 3 16
 9500000.00 .897176000 9600000.00 .969448000 9700000.00 1.041720006925 3 16
 9800000.00 1.11399200 9900000.00 1.18626400 10000000.0 1.235935006925 3 16
 10100000.0 1.26300500 10200000.0 1.29007500 10300000.0 1.317145006925 3 16
 10400000.0 1.34421500 10500000.0 1.37128500 10600000.0 1.398355006925 3 16
 10700000.0 1.42542500 10800000.0 1.45249500 10900000.0 1.479565006925 3 16
 11000000.0 1.50758000 11100000.0 1.53654000 11200000.0 1.565500006925 3 16
 11300000.0 1.59446000 11400000.0 1.62342000 11500000.0 1.652380006925 3 16
 11600000.0 1.68134000 11700000.0 1.71030000 11800000.0 1.739260006925 3 16
 11900000.0 1.76822000 12000000.0 1.79406500 12100000.0 1.816795006925 3 16
 12200000.0 1.83952500 12300000.0 1.86225500 12400000.0 1.884985006925 3 16
 12500000.0 1.90771500 12600000.0 1.93044500 12700000.0 1.953175006925 3 16
 12800000.0 1.97590500 12900000.0 1.99863500 13000000.0 2.009486676925 3 16
 13100000.0 2.00846000 13200000.0 2.00743333 13300000.0 2.006406676925 3 16
 13400000.0 2.00538000 13500000.0 2.00435333 13600000.0 2.003326676925 3 16
 13700000.0 2.00230000 13800000.0 2.00127333 13900000.0 2.000246676925 3 16
 14000000.0 1.99922000 14100000.0 1.99819333 14200000.0 1.997166676925 3 16
 14300000.0 1.99614000 14400000.0 1.99511333 14500000.0 1.994686676925 3 16
 14600000.0 1.99486000 14700000.0 1.99503333 14800000.0 1.995206676925 3 16
 14900000.0 1.99538000 15000000.0 1.99555333 15100000.0 1.995726676925 3 16
 15200000.0 1.99590000 15300000.0 1.99607333 15400000.0 1.996246676925 3 16
 15500000.0 1.99642000 15600000.0 1.99659333 15700000.0 1.996766676925 3 16
 15800000.0 1.99694000 15900000.0 1.99711333 16000000.0 1.994380006925 3 16
 16100000.0 1.98874000 16200000.0 1.98310000 16300000.0 1.977460006925 3 16
 16400000.0 1.97182000 16500000.0 1.96618000 16600000.0 1.960540006925 3 16
 16700000.0 1.95490000 16800000.0 1.94926000 16900000.0 1.943620006925 3 16
 17000000.0 1.91730500 17100000.0 1.87031500 17200000.0 1.823325006925 3 16
 17300000.0 1.77633500 17400000.0 1.72934500 17500000.0 1.682355006925 3 16
 17600000.0 1.63536500 17700000.0 1.58837500 17800000.0 1.541385006925 3 16
 17900000.0 1.49439500 18000000.0 1.45198500 18100000.0 1.414155006925 3 16
 18200000.0 1.37632500 18300000.0 1.33849500 18400000.0 1.300665006925 3 16
 18500000.0 1.26283500 18600000.0 1.22500500 18700000.0 1.187175006925 3 16
 18800000.0 1.14934500 18900000.0 1.11151500 19000000.0 1.081435006925 3 16
 19100000.0 1.05910500 19200000.0 1.03677500 19300000.0 1.014445006925 3 16
 19400000.0 .992115000 19500000.0 .969785000 19600000.0 .9474550006925 3 16
 19700000.0 .925125000 19800000.0 .902795000 19900000.0 .8804650006925 3 16
 20000000.0 0.0                                                   6925 3 16
                                                                  6925 3  0
                                                                  6925 0  0
 6.91690E+4 1.67483E+2          0          0          0          1692533 16
 0.000000+0 0.000000+0          0         16          0          1692533 16
 0.000000+0 0.000000+0          1          5        120         15692533 16
 1.000000-5 8.000000+6 8.240180+6 8.700000+6 9.500000+6 1.050000+7692533 16
 1.150000+7 1.250000+7 1.375000+7 1.525000+7 1.650000+7 1.750000+7692533 16
 1.850000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0692533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.544490-3692533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16
 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16
 1.544490-3 2.161260-4 8.021960-5 8.018380-5 8.965270-5 1.308120-4692533 16
 8.611730-5 7.978640-5 1.636990-4 1.667380-4 9.654600-5 2.752760-4692533 16
 1.762320-3 1.773580-4 1.789170-4 1.261310-4 1.126210-4 7.632130-5692533 16
 6.414960-5 1.179690-4 1.202010-4 7.090170-5 1.789260-4 2.253400-3692533 16
 3.400620-4 1.941300-4 1.014070-4 7.258540-5 4.974090-5 6.811760-5692533 16
 7.143290-5 4.130170-5 6.744160-5 2.055720-3 1.967720-4 1.028640-4692533 16
 7.314980-5 5.047840-5 6.867560-5 7.201810-5 4.176900-5 6.741150-5692533 16
 7.739520-4 1.059610-4 7.502940-5 7.712840-5 9.702920-5 1.017520-4692533 16
 5.552930-5 7.629130-5 2.742330-4 8.094930-5 8.206160-5 1.132040-4692533 16
 1.180210-4 6.526070-5 1.110700-4 1.449620-4 5.756050-5 7.774620-5692533 16
 8.052350-5 5.484240-5 7.438890-5 4.765490-4 1.022420-4 1.031120-4692533 16
 1.117260-4 7.429010-5 3.972050-4 1.424560-4 1.295420-4 1.435500-4692533 16
 4.368100-4 1.091050-4 1.436310-4 2.619390-3 1.132960-4 1.091640-3692533 16
                                                                  692533  0
                                                                  6925 0  0
                                                                     0 0  0
 7.31810E+4 1.79393E+2          0          0         34         107328 1451
 0.0        0.0                 0          0          0          67328 1451
 1.00000E+0 2.00000E+7          0          0         10       20027328 1451
 3.00000E+2 0.0                 1          0        117          37328 1451
 73-Ta-181 NAIG       EVAL-MAR87 N.YAMAMURO                       7328 1451
                      DIST-Feb2004                                7328 1451
----IRDF-2002         MATERIAL 7328         REVISION 2            7328 1451
-----INCIDENT NEUTRON DATA                                        7328 1451
------ENDF-6 FORMAT                                               7328 1451
****************************************************************  7328 1451
 73-TA-181 NAIG       EVAL-MAR87 N.YAMAMURO                       7328 1451
                      DIST-SEP89 REV2-FEB94                       7328 1451
----JENDL-3.2         MATERIAL 7328         REVISION 2            7328 1451
 <DOSIMETRY CROSS SECTION>                                        7328 1451
      TA-181 CAPTURE TA-181(HALF-LIFE = 114.43D)                  7328 1451
****************************************************************  7328 1451
HISTORY                                                           7328 1451
76-03 THE EVALUATION FOR JENDL-1 /1/ WAS MADE BY H.YAMAKOSHI      7328 1451
      (SHIP RESEARCH INSTITUTE) AND JENDL-1 COMPILATION GROUP.    7328 1451
83-03 JENDL-1 DATA WERE ADOPTED FOR JENDL-2 AND EXTENDED TO 20    7328 1451
      MEV. MF=5 WAS REVISED, AND UNRESOLVED RESONANCE PARAMETERS  7328 1451
      WERE ADDED BY Y.KIKUCHI (JAERI) /2/.                        7328 1451
83-11 COMMENT DATA WERE ADDED.                                    7328 1451
87-03 THE EVALUATION FOR JENDL-3 WAS MADE BY N.YAMAMURO (NAIG).   7328 1451
      RESONANCE PARAMETERS WERE ADDED BY NEW EXPERIMENTAL DATA.   7328 1451
      NEUTRON CROSS SECTIONS, EXCEPT TOTAL AND ELASTIC SCATTERING 7328 1451
      CROSS SECTIONS, AND ENERGY DISTRIBUTIONS OF SECONDARY       7328 1451
      NEUTRONS AND PHOTONS WERE CALCULATED WITH GNASH /3/ AND     7328 1451
      CASTHY /4/ CODES.                                           7328 1451
94-02 JENDL-3.2.                                                  7328 1451
      COMPILED BY T.NAKAGAWA (NDC/JAERI)                          7328 1451
                                                                  7328 1451
     *****   MODIFIED PARTS FOR JENDL-3.2   ********************  7328 1451
      (12,102)                                                    7328 1451
             DATA WERE DETERMINED FROM ENERGY BALANCE.            7328 1451
     ***********************************************************  7328 1451
                                                                  7328 1451
                                                                  7328 1451
MF=1   GENERAL INFORMATION                                        7328 1451
  MT=451     DESCRIPTIVE DATA AND DICTIONARY                      7328 1451
                                                                  7328 1451
MF=2   RESONANCE PARAMETERS                                       7328 1451
  MT=151     RESOLVED AND UNRESOLVED RESONANCE PARAMETERS         7328 1451
  RESOLVED PARAMETERS FOR MLBW FORMULA                            7328 1451
                                                                  7328 1451
      THE ENERGY REGION FROM 1.0E-5EV TO 1.0 KEV. PARAMETERS WERE 7328 1451
      TAKEN FROM REFS./5,6,7/ FOR POSITIVE RESONANCES, AND FROM   7328 1451
      ENDF/B-IV FOR A NEGATIVE RESONANCE. THE RADIATIVE WIDTH OF  7328 1451
      0.059EV WAS ASSUMED FOR THE RESONANCE WHOSE RADIATIVE WIDTH 7328 1451
      WAS UNKNOWN.                                                7328 1451
                                                                  7328 1451
  UNRESOLVED PARAMETERS                                           7328 1451
                                                                  7328 1451
      IN THE ENERGY RANGE FROM 1 TO 100KEV, PARAMETERS WERE       7328 1451
      DETERMINED TO REPRODUCE THE MEASURED CAPTURE CROSS SECTIONS 7328 1451
      /6,8/. THE PARAMETERS ARE AS FOLLOWS,                       7328 1451
                                                                  7328 1451
         R= 7.8 FM , DOBS= 4.2 EV , RADIATIVE WIDTH= 0.065 EV,    7328 1451
         SO= 1.7E-04  S1= 2.0E-05  S2= 2.3E-04  NL= 3             7328 1451
                                                                  7328 1451
      CALCULATED 2200-M/SEC CROSS SECTIONS AND RESONANCE INTEGRALS7328 1451
                        2200-M/SEC      RES. INTEG                7328 1451
          ELASTIC         5.65 B            -                     7328 1451
          CAPTURE        20.67 B         660.43 B                 7328 1451
          TOTAL          26.32 B            -                     7328 1451
                                                                  7328 1451
MF=3   NEUTRON CROSS SECTIONS                                     7328 1451
  MT=1   TOTAL                                                    7328 1451
    EVALUATED FROM EXPERIMENTAL DATA.                             7328 1451
  MT=2   ELASTIC SCATTERING                                       7328 1451
    (TOTAL CROSS SECTION) - (REACTION CROSS SECTION)              7328 1451
  MT=102  RADIATIVE CAPTURE CROSS SECTION                         7328 1451
    CALCULATED WITH CASTHY/4/.                                    7328 1451
                                                                  7328 1451
REFERENCES                                                        7328 1451
 1) IGARASI S. ET AL.: JAERI 1261 (1979)                          7328 1451
 2) (ED.) NAKAGAWA T.: JAERI-M 84-103 (1984)                      7328 1451
 3) YOUNG,P.G. AND ARTHUR,E.D.: "GNASH, A PREEQUILIBRIUM          7328 1451
    STATISTICAL NUCLEAR-MODEL CODE FOR CALCULATION OF CROSS       7328 1451
    SECTIONS AND EMISSION SPECTRA", LA-6974 (1977).               7328 1451
 4) IGARASHI,S. AND FUKAHORI,T.: JAERI 1321 (1991).               7328 1451
 5) MUGHABGHAB,S.F, AND GARDER,D.I.: BNL325, 3RD ED. (1973).      7328 1451
 6) MACKLIN,R,L,: NUCL,SCI,ENG., 86,362 (1984).                   7328 1451
 7) TSUBONE,I., NAKAJIMA,Y. AND KANDA,Y.: PRIVATE COMMUNICATION   7328 1451
 8) YAMAMURO,N., SAITO,K., EMOTO,T., WADA,T., FUJITA,Y. AND       7328 1451
    KOBAYASHI,K.: J.NUCL.SCI.TECHNOL., 17, 582 (1980).            7328 1451
 9) WILMORE,D. AND HODGSON,P.E.: NUCL, PHYS., 55, 673 (1964).     7328 1451
10) FIRESTONE,R.B.* NUCL, DATA SHEETS 43, 289 (1984).             7328 1451
 ***************************************************************  7328 1451
  The Q values and threshold energies were updated prior to pro-  7328 1451
  cessing through the codes to comply with the values obtained    7328 1451
  using the NNDC calculation program which is based on the 1995   7328 1451
  Update to the Atomic mass Evaluation.                           7328 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************7328 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     7328 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     7328 1451
 ***************** Program RECENT (VERSION 2002-1) ***************7328 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7328 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     7328 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************7328 1451
 Data Doppler Broadened to 300.000000 Kelvin                      7328 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7328 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     7328 1451
 ***************** Program FIXUP (Version 2002-1) ****************7328 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 7328 1451
 Corrected Thresholds-----------------------------------------Yes 7328 1451
 Extended Cross Sections to 20 MeV----------------------------No  7328 1451
 Allow Cross Section Deletion---------------------------------No  7328 1451
 Allow Cross Section Reconstruction---------------------------No  7328 1451
 Make All Cross Sections Non-Negative-------------------------Yes 7328 1451
 Delete Energies Not in Ascending Order-----------------------Yes 7328 1451
 Deleted Duplicate Points-------------------------------------Yes 7328 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 7328 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 7328 1451
 Allow Creation of Missing Sections---------------------------No  7328 1451
 Allow Insertion of Energy Points-----------------------------No  7328 1451
 Create Uniform Energy Grid-----------------------------------No  7328 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 7328 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************7328 1451
 Unshielded Group Averages Using  640 Groups                      7328 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     7328 1451
                                1        451        124          17328 1451
                                3        102        217          17328 1451
                               33        102         16          17328 1451
                                                                  7328 1  0
                                                                  7328 0  0
 7.31810E+4 1.79393E+2          0          0          0          07328 3102
 6.06296E+6 6.06296E+6          0          0          1        6417328 3102
        641          1                                            7328 3102
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 .018999999 23.5405789 .020000000 22.9587048 .021000000 22.42625467328 3102
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 18400000.0 3.65908E-8 18500000.0 3.57061E-8 18600000.0 3.48673E-87328 3102
 18700000.0 3.40452E-8 18800000.0 3.32291E-8 18900000.0 3.24595E-87328 3102
 19000000.0 3.16995E-8 19100000.0 3.09575E-8 19200000.0 3.02550E-87328 3102
 19300000.0 2.95540E-8 19400000.0 2.88663E-8 19500000.0 2.82189E-87328 3102
 19600000.0 2.75742E-8 19700000.0 2.69479E-8 19800000.0 2.63519E-87328 3102
 19900000.0 2.57564E-8 20000000.0 0.0                             7328 3102
                                                                  7328 3  0
                                                                  7328 0  0
 7.31810E+4 1.79393E+2          0          0          0          1732833102
 0.000000+0 0.000000+0          0        102          0          1732833102
 0.000000+0 0.000000+0          0          1         74         37732833102
 1.000000-5 9.000000-4 1.000000+0 2.500000-3 5.000000+0 2.500000-3732833102
 1.000000+1 1.000000-2 2.000000+1 1.000000-2 3.000000+1 2.250000-2732833102
 5.000000+1 2.250000-2 7.000000+1 2.250000-2 1.000000+2 4.000000-2732833102
 2.000000+2 4.000000-2 3.000000+2 4.000000-2 5.000000+2 4.000000-2732833102
 7.000000+2 4.000000-2 1.000000+3 2.250000-2 2.000000+3 2.250000-2732833102
 3.000000+3 1.000000-2 5.000000+3 1.000000-2 7.000000+3 1.000000-2732833102
 1.000000+4 1.000000-2 2.000000+4 1.000000-2 3.000000+4 1.000000-2732833102
 5.000000+4 1.000000-2 7.000000+4 1.000000-2 1.000000+5 4.000000-2732833102
 2.000000+5 4.000000-2 3.000000+5 4.000000-2 5.000000+5 4.000000-2732833102
 7.000000+5 4.000000-2 1.000000+6 1.000000-2 2.000000+6 1.000000-2732833102
 3.000000+6 1.000000-2 4.000000+6 4.000000-2 6.000000+6 4.000000-2732833102
 7.000000+6 4.000000-2 1.000000+7 9.000000-2 1.500000+7 9.000000-2732833102
 2.000000+7 0.000000+0                                            732833102
                                                                  732833  0
                                                                  7328 0  0
                                                                     0 0  0
 7.41860E+4 1.84357E+2          0          0         34         107443 1451
 0.0        0.0                 0          0          0          67443 1451
 1.00000E+0 2.00000E+7          0          0         10       20027443 1451
 3.00000E+2 0.0                 1          0        199          37443 1451
 74-W -186 FEI,LANL+  EVAL-SEP01 K.I.Zolotarev,M.B.CHADWICK etal. 7443 1451
                      DIST-Feb2004                                7443 1451
----IRDF-2002         MATERIAL 7443                               7443 1451
-----INCIDENT NEUTRON DATA                                        7443 1451
------ENDF-6 FORMAT                                               7443 1451
***************************************************************** 7443 1451
 74-W -186 FEI        EVAL-Sep01 K.I.Zolotarev                    7443 1451
                      DIST-Jan04                                  7443 1451
----BROND-3           MATERIAL 7443         Revision 1, Jan. 2004 7443 1451
-----INCIDENT NEUTRON DATA                                        7443 1451
------ENDF-6 FORMAT                                               7443 1451
------Russian Reactor Dosimetry File RRDF-2002                    7443 1451
***************************************************************** 7443 1451
 **********   Start of (N,G) bibliographical component ********** 7443 1451
***************************************************************** 7443 1451
***************************************************************** 7443 1451
Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev             7443 1451
***************************************************************** 7443 1451
                                                                  7443 1451
                    ----- MF=2 MT=151 -----                       7443 1451
                                                                  7443 1451
RESOLVED RESONANCE PARAMETERS                                     7443 1451
    Resolved MLBW resonance parameters up to  8.5 keV  were based 7443 1451
on data from ref.[1-2]. Spins of resonances, which were not iden- 7443 1451
tified, have been assigned by taking into account 2*J+1 law  and  7443 1451
approximate linear growth of the number of levels with given spin 7443 1451
and parity from energy. Average gamma-widths 0.05 eV (L=0 and L=1)7443 1451
were assigned to those resonances, widths for which were absent.  7443 1451
Fictitious resonances were added in the energy region 4 - 8.5 keV 7443 1451
to obtain  the average capture cross section values  observed  in 7443 1451
the experiments [3],[4],[5],[6].                                  7443 1451
                                                                  7443 1451
UNRESOLVED RESONANCE PARAMETERS                                   7443 1451
    Unresolved resonance parameters in  the region  8.5 - 100 keV 7443 1451
were evaluated  with using 'EVPAR' code (Hauser-Feshbach-Moldauer 7443 1451
statistical model)[7]. They were calculated from optimized fit to 7443 1451
the measured capture [4,5,6,8,9] and total [10,11] cross sections 7443 1451
data.  Average resolved resonance  parameters  have been  used as 7443 1451
zero approximation.  Contribution of  F-wave in the energy region 7443 1451
30 - 100 keV given as smoothed background cross section in File 3.7443 1451
                                                                  7443 1451
                    ----- MF=3 MT=102 -----                       7443 1451
    Capture cross sections from 1.000E-5 eV to 100 keV are recon- 7443 1451
structed from evaluated  MLBW resolved  and  unresolved resonance 7443 1451
parameters. Contribution of F-wave in the neutron energy interval 7443 1451
30 - 100 keV is added as described above.                         7443 1451
    Data base for the evaluation W-186(n,g)W-187 excitation func- 7443 1451
tion in the neutron energy range 100 keV - 20 MeV was formed from 7443 1451
microscopic experimental data [4-6],[8-9],[12-25] and data recei- 7443 1451
ved from theoretical model calculations. Experimental data [8-9], 7443 1451
[16-20], [22], [24] were corrected to the new standards.  Uncer-  7443 1451
tainty for cross section values measured by Bartolome et al. [4]  7443 1451
was evaluated as +-16%.  Theoretical model calculations  was done 7443 1451
by means of GNASH code [26].  New modified  version of GNASH  was 7443 1451
utilized for calculations [27].  Data from GNASH calculation were 7443 1451
dominant in the neutron energy range 4 - 20 MeV.                  7443 1451
    The evaluation  W-186(n,g)W-187  excitation function  in  the 7443 1451
energy range above 100 keV has been carried out within the frame- 7443 1451
work  of generalized least squares method , rational function was 7443 1451
used as model function [28]. Procedure of calculation recommended 7443 1451
cross section data was performed by means of PADE-2 code [29].    7443 1451
   Capture cross section at En=0.0253 eV and Resonance Integral   7443 1451
(0.5 to 2.0E+7 eV) from present evaluation are given below in the 7443 1451
Table 1. in a comparison with data from compilations [1] and [30].7443 1451
                                                      Table 1.    7443 1451
        --------------------------------------------------        7443 1451
          Parameters            Values, barn    References        7443 1451
        --------------------------------------------------        7443 1451
        Capture cross section   38.47 +- 0.85   This eval.        7443 1451
        at En=0.0253 eV         37.9  +- 0.6     [ 1]             7443 1451
                                37    +- 2       [30]             7443 1451
                                38.5  +- 0.5     [31]             7443 1451
        --------------------------------------------------        7443 1451
        Resonance Integral      482.2           This eval.        7443 1451
                                485   +- 15      [ 1]             7443 1451
                                485   +- 15      [30]             7443 1451
                                485   +- 15      [31]             7443 1451
        --------------------------------------------------        7443 1451
                                                                  7443 1451
                    ----- MF=33 MT=102-----                       7443 1451
    Uncertainties in  the evaluated excitation function  for the  7443 1451
reaction  W-186(n,g)W-187  are  given  in  the three  independent 7443 1451
matrixes.                                                         7443 1451
    In the energy range 1.000E-05 - 100 eV uncertainties are pre- 7443 1451
sented in the form of relative covariance matrix for the 14-neut- 7443 1451
ron energy groups (LB=5).  Eigenvalues  of this 6-th digits rela- 7443 1451
tive covariance matrix are the following:                         7443 1451
                                                                  7443 1451
    1.14625E-10    1.09510E-07    7.28111E-07    8.61003E-06      7443 1451
    4.80743E-05    5.34800E-05    5.91642E-05    6.60535E-05      7443 1451
    8.44288E-05    1.48138E-04    1.95936E-04    2.39607E-04      7443 1451
    8.15874E-04    8.38515E-03                                    7443 1451
                                                                  7443 1451
    In the energy range 0.1 - 8.5 keV  uncertainties are given in 7443 1451
the form  of diagonal matrix  of uncertainties  for 19-th neutron 7443 1451
energy intervals (LB=1)                                           7443 1451
    In the energy range  8.5 keV - 20 MeV  uncertainties are pre- 7443 1451
sented in the form of relative covariance matrix for the 24-neut- 7443 1451
ron energy groups (LB=5).  Eigenvalues  of this 6-th digits rela- 7443 1451
tive covariance matrix are the following:                         7443 1451
                                                                  7443 1451
    6.81049E-10    1.45633E-05    5.83401E-05    1.88615E-04      7443 1451
    6.00477E-04    7.08017E-04    9.54276E-04    1.08032E-03      7443 1451
    1.26144E-03    1.35976E-03    1.50022E-03    1.66258E-03      7443 1451
    1.66943E-03    2.15756E-03    2.87259E-03    3.83276E-03      7443 1451
    6.81420E-03    9.70769E-03    9.99331E-03    1.17664E-02      7443 1451
    1.48009E-02    1.93593E-02    2.04003E-02    2.26206E-01      7443 1451
                                                                  7443 1451
References :                                                      7443 1451
 1. S.F.Mughabghab et al.  Neutron Cross Sections, vol.1, part B, 7443 1451
    New York, Academic Press, 1984                                7443 1451
 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,7443 1451
    ed. H.Schopper, Springer, 1998, pp.w19-w23                    7443 1451
 3. S.V.Kapchigashev, Ju.P.Popov, F.L.Shapiro  EXFOR 40755.005    7443 1451
 4. Z.M.Bartolome et al.  Nucl. Sci. Eng., v.37, pp.137-156, 1969 7443 1451
 5. R.L.Macklin et al.  Nucl. Sci. Eng., v.84, p.98, June 1983    7443 1451
 6. M.V.Bokhovko et al.  Voprosy Atomnoj Nauki i Tekhniki, Ser.:  7443 1451
    Yadernye Konstanty, v.1, p.39, February 1986                  7443 1451
 7. G.N.Manturov et al.  Voprosy Atomnoy Nauki i Tekhniki, Ser.:  7443 1451
    Jadernye Konstanty, v.1, p.50, 1983                           7443 1451
 8. Yu.Ya.Stavisskii, V.A.Tolstikov   Atomnaja Energija, v.9,     7443 1451
    p.401, 1960                                                   7443 1451
 9. V.N.Kononov et al.  Yadernaja Fizika, v.4, no.2, p.282, 1966  7443 1451
10. P.T.Guenther. et al.  Phys. Rev. C., v.26, p.2433, 1982       7443 1451
11. V.N.Kononov  et al.  Sov J. of Nuclear Physics, v.46, part 1, 7443 1451
    p.51, July 1987                                               7443 1451
12. R.Allen et al.  Nature, v.161, p.727, May 1948                7443 1451
13. L.E.Beghian, H.H.Halban  Nature, v.163, p.366, March 1949     7443 1451
14. J.L.Perkin et al.  Proc. Phys. Soc., v.72, p.505, 1958        7443 1451
15. A.I.Leipunskij et al. Second UN Conf. on the Peaceful Uses of 7443 1451
    Atomic Energy, Geneva, 1-13 September 1958, v.15, p.50(2219)  7443 1451
16. M.V.Pasechnik et al.  Second UN Conf. on the Peaceful Uses of 7443 1451
    Atomic Energy, Geneva, 1-13 September 1958, v.15, p.18(2030)  7443 1451
17. A.E.Johnsrud et al.  Phys. Rev., v.116, p.927, 1959           7443 1451
18. G.G.Zaikin et al.  Jaderno-Fizicheskie Issledovanija (USSR    7443 1451
    progress report), v.6, p.103, 1968                            7443 1451
19. M.Lindner et al.  Nucl. Sci. Eng., v.59, p.381, April 1976    7443 1451
20. O.Schwerer et al.  Nucl. Phys., v.A264, p.105, June 1976      7443 1451
21. M.Valkonen  Report JU-RR-1/1976, Jyvaeskylae University, 1976 7443 1451
    M.Valkonen et al.  J. Inorg. Nucl. Chem., v.36, p.715, 1974   7443 1451
22. G.Magnusson et al.  Physica Scripta, v.21, p.21, January 1980 7443 1451
23. J.Voignier et al.  Report CEA-R-5089, August 1981 ;           7443 1451
    J.Voignier et al.  Nucl. Sci. Eng., v.93, p.43, 1986          7443 1451
24. Yu.N.Trofimov   1st International Conf. on Neutron Physics,   7443 1451
    Kiev, 14-16 September 1987, v.3, p.331                        7443 1451
25. Zhang Guohui, Lu Hanlin et al. Communication of Nuclear Data  7443 1451
    Progress, No. 23, Atomic Energy Press, Beijing, June, 2000    7443 1451
26. P.G.Young, E.D.Arthur  A Preequilibrium Statistical Nuclear   7443 1451
    Model Code for Calculation of Cross Section and Emission      7443 1451
    Spectra. Report LA-6947, Los Alamos, 1977                     7443 1451
27. E.L.Trykov, G.Ya.Tertychnyi  Private communication, IPPE,     7443 1451
    Obninsk, May 1999                                             7443 1451
28. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 7443 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 7443 1451
    Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105  7443 1451
29. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          7443 1451
30. A.V.Ignatyuk et al.  Landolt Bornstein New Series, v.I/16A,   7443 1451
    Part 1, ed. H.Schopper, Springer, 1998, p.8-24                7443 1451
31. S.F.Mughabghab et al.  Termal Neutron Capture Cross Sections  7443 1451
    Resonance Integrals and G-factors, Report INDC(NDS)-440,      7443 1451
    IAEA, Vienna, February 2003                                   7443 1451
***************************************************************** 7443 1451
 **********    End of (N,G) bibliographical component  ********** 7443 1451
***************************************************************** 7443 1451
  74-W -186 LANL,ANL  EVAL-OCT96 M.B.CHADWICK,P.G.YOUNG,ARTHUR    7443 1451
 Ch96a,Ch96b,Ar80,Ch99DIST-SEP 1 REV1-                 20010926   7443 1451
----ENDF/B-VI         MATERIAL 7443                               7443 1451
  MF=3 MT=  1  Total Cross Section                                7443 1451
       MT=  2  Elastic Scattering Cross Section                   7443 1451
***************************************************************** 7443 1451
                                                                  7443 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************7443 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     7443 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     7443 1451
 ***************** Program RECENT (VERSION 2002-1) ***************7443 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7443 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     7443 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************7443 1451
 Data Doppler Broadened to 300.000000 Kelvin                      7443 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7443 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     7443 1451
 ***************** Program FIXUP (Version 2002-1) ****************7443 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 7443 1451
 Corrected Thresholds-----------------------------------------Yes 7443 1451
 Extended Cross Sections to 20 MeV----------------------------No  7443 1451
 Allow Cross Section Deletion---------------------------------No  7443 1451
 Allow Cross Section Reconstruction---------------------------No  7443 1451
 Make All Cross Sections Non-Negative-------------------------Yes 7443 1451
 Delete Energies Not in Ascending Order-----------------------Yes 7443 1451
 Deleted Duplicate Points-------------------------------------Yes 7443 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 7443 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 7443 1451
 Allow Creation of Missing Sections---------------------------No  7443 1451
 Allow Insertion of Energy Points-----------------------------No  7443 1451
 Create Uniform Energy Grid-----------------------------------No  7443 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 7443 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************7443 1451
 Unshielded Group Averages Using  640 Groups                      7443 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     7443 1451
                                1        451        206          17443 1451
                                3        102        217          17443 1451
                               33        102         94          17443 1451
                                                                  7443 1  0
                                                                  7443 0  0
 7.41860E+4 1.84357E+2          0          0          0          07443 3102
 5.46700E+6 5.46700E+6          0          0          1        6417443 3102
        641          1                                            7443 3102
 .000100000 603.335846 .000105000 589.267631 .000110000 575.9026837443 3102
 .000115000 563.587364 .000120000 549.190298 .000127500 533.2169627443 3102
 .000135000 518.623372 .000142500 505.212229 .000150000 490.6976817443 3102
 .000160000 475.577316 .000170000 461.819668 .000180000 449.1912287443 3102
 .000190000 437.441542 .000200000 426.713841 .000210000 416.6208847443 3102
 .000220000 407.246173 .000230000 398.535864 .000240000 388.3117177443 3102
 .000255000 377.076879 .000270000 368.407585 .000280000 358.7570467443 3102
 .000300000 346.987888 .000320000 336.331282 .000340000 326.5715237443 3102
 .000360000 317.582189 .000380000 309.376115 .000400000 300.8208007443 3102
 .000425000 292.066023 .000450000 284.066608 .000475000 276.7330497443 3102
 .000500000 269.836551 .000525000 263.533426 .000550000 257.5790457443 3102
 .000575000 252.041159 .000600000 246.365181 .000630000 240.5260707443 3102
 .000660000 235.183230 .000690000 230.057942 .000720000 224.6043127443 3102
 .000760000 218.756748 .000800000 213.327421 .000840000 208.3517907443 3102
 .000880000 203.622726 .000920000 199.287405 .000960000 195.1354907443 3102
 .001000000 190.831817 .001050000 186.339264 .001100000 182.1277957443 3102
 .001150000 178.247247 .001200000 173.669665 .001275000 168.6387387443 3102
 .001350000 164.039287 .001425000 159.746960 .001500000 155.1870537443 3102
 .001600000 150.419482 .001700000 146.067368 .001800000 142.0423177443 3102
 .001900000 138.365382 .002000000 134.956017 .002100000 131.7566167443 3102
 .002200000 128.826606 .002300000 126.024764 .002400000 122.8276247443 3102
 .002550000 119.274676 .002700000 116.495377 .002800000 113.4751727443 3102
 .003000000 109.757167 .003200000 106.374065 .003400000 103.2810967443 3102
 .003600000 100.461353 .003800000 97.8587164 .004000000 95.14697857443 3102
 .004250000 92.3949345 .004500000 89.8831489 .004750000 87.53526427443 3102
 .005000000 85.3797497 .005250000 83.3803562 .005500000 81.48937897443 3102
 .005750000 79.7549491 .006000000 77.9328522 .006300000 76.11093107443 3102
 .006600000 74.3872531 .006900000 72.7894516 .007200000 71.05385907443 3102
 .007600000 69.1997423 .008000000 67.5065760 .008400000 65.92103847443 3102
 .008800000 64.4499949 .009200000 63.0772891 .009600000 61.76737067443 3102
 .010000000 60.4112555 .010500000 58.9776762 .011000000 57.66673047443 3102
 .011500000 56.4163054 .012000000 54.9836997 .012750000 53.39855427443 3102
 .013500000 51.9309308 .014250000 50.5851860 .015000000 49.15023247443 3102
 .016000001 47.6438249 .017000001 46.2621264 .017999999 45.00019547443 3102
 .018999999 43.8399023 .020000000 42.7512128 .021000000 41.75505657443 3102
 .022000000 40.8167414 .023000000 39.9428153 .024000000 38.91634697443 3102
 .025500000 37.8095093 .027000001 36.9325200 .028000001 35.97969527443 3102
 .029999999 34.8066273 .032000002 33.7378775 .034000002 32.76244257443 3102
 .035999998 31.8726360 .037999999 31.0529462 .039999999 30.19880417443 3102
 .042500000 29.3331880 .045000002 28.5416236 .047499999 27.80334387443 3102
 .050000001 27.1230983 .052499998 26.4935146 .055000000 25.89745037443 3102
 .057500001 25.3521109 .059999999 24.7788712 .063000001 24.20585197443 3102
 .066000000 23.6660150 .068999998 23.1633318 .071999997 22.61846327443 3102
 .075999998 22.0351607 .079999998 21.5031559 .083999999 21.00417257443 3102
 .088000000 20.5414942 .092000000 20.1120657 .096000001 19.70136977443 3102
 .100000001 19.2765995 .104999997 18.8267371 .109999999 18.41600767443 3102
 .115000002 18.0238516 .119999997 17.5751000 .127499998 17.07955187443 3102
 .135000005 16.6203733 .142499998 16.1993084 .150000006 15.75106177443 3102
 .159999996 15.2818325 .170000002 14.8558108 .180000007 14.46999417443 3102
 .189999998 14.0996627 .200000003 13.7648418 .209999993 13.45402097443 3102
 .219999999 13.1616894 .230000004 12.8935802 .239999995 12.57700827443 3102
 .254999995 12.2297639 .270000011 11.9645082 .280000001 11.66692567443 3102
 .300000012 11.3057050 .319999993 10.9795359 .340000004 10.68698387443 3102
 .360000014 10.4145184 .379999995 10.1634427 .400000006 9.902794257443 3102
 .425000012 9.63866553 .449999988 9.39853042 .474999994 9.175161707443 3102
 .500000000 8.97023658 .524999976 8.78189600 .550000012 8.603760717443 3102
 .574999988 8.44231333 .600000024 8.27353196 .629999995 8.107680887443 3102
 .660000026 7.94505970 .689999998 7.80132218 .720000029 7.639667707443 3102
 .759999990 7.47024298 .800000012 7.31302988 .839999974 7.169505637443 3102
 .879999995 7.04137512 .920000017 6.91613741 .959999979 6.800998867443 3102
 1.00000000 6.68266043 1.04999995 6.55552625 1.10000002 6.444120707443 3102
 1.14999998 6.33912340 1.20000005 6.21961754 1.27499998 6.083371227443 3102
 1.35000002 5.96232397 1.42499995 5.85656846 1.50000000 5.739618607443 3102
 1.60000002 5.62159377 1.70000005 5.51970966 1.79999995 5.424408957443 3102
 1.89999998 5.34155762 2.00000000 5.26976356 2.09999990 5.202706237443 3102
 2.20000005 5.14198715 2.29999995 5.08795501 2.40000010 5.030501777443 3102
 2.54999995 4.96748721 2.70000005 4.92518185 2.79999995 4.880380877443 3102
 3.00000000 4.83453936 3.20000005 4.79788239 3.40000010 4.775360897443 3102
 3.59999990 4.75819048 3.79999995 4.75230086 4.00000000 4.758569807443 3102
 4.25000000 4.77106693 4.50000000 4.79690169 4.75000000 4.833598287443 3102
 5.00000000 4.87885727 5.25000000 4.93616534 5.50000000 5.000960757443 3102
 5.75000000 5.07713265 6.00000000 5.17233026 6.30000019 5.288142407443 3102
 6.59999990 5.41949101 6.90000010 5.56814022 7.19999981 5.762945927443 3102
 7.59999990 6.01650068 8.00000000 6.30471484 8.39999962 6.635955477443 3102
 8.80000019 7.01430912 9.19999981 7.44727175 9.60000038 7.942231267443 3102
 10.0000000 8.58836367 10.5000000 9.44261389 11.0000000 10.47294877443 3102
 11.5000000 11.7309584 12.0000000 13.7561740 12.7500000 17.08032167443 3102
 13.5000000 22.0166931 14.2500000 29.8026254 15.0000000 46.73659427443 3102
 16.0000000 94.5260670 17.0000000 309.787436 18.0000000 5638.537747443 3102
 19.0000000 1815.84329 20.0000000 172.939121 21.0000000 62.58393787443 3102
 22.0000000 31.8812393 23.0000000 19.1580069 24.0000000 11.70253047443 3102
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 79-Au-197 LANL/IRK   EVAL-JAN84 P.G.YOUNG,VONACH ET AL           7925 1451
                      DIST-Feb2004                                7925 1451
----IRDF-2002         MATERIAL 7925                               7925 1451
-----INCIDENT NEUTRON DATA                                        7925 1451
------ENDF-6 FORMAT                                               7925 1451
 *****************************************************************7925 1451
 79-AU-197 LANL       EVAL-JAN84 P.G.YOUNG                        7925 1451
 LA-10069-PR          DIST-SEP91 REV1-JUL91            19930129   7925 1451
----ENDF/B-VI         MATERIAL 7925         REVISION 1            7925 1451
 *****************************************************************7925 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************7925 1451
DOSIMETRY                                                         7925 1451
******************************************************************7925 1451
                                                                  7925 1451
                  MOD1 OF ENDF/B-VI                               7925 1451
                                                                  7925 1451
     The following revisions were made for MOD1 of ENDF/B-VI:     7925 1451
                                                                  7925 1451
 1. MF=1,MT=451 - Comments were added regarding estimated         7925 1451
    (expanded) covariance for the Standards Cross Sections.       7925 1451
 2. MF=3,MT=102 - Q-value corrected.                              7925 1451
                                                                  7925 1451
 *****************************************************************7925 1451
++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451
         MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING    7925 1451
         EVALUATION                                               7925 1451
                79-AU-197 IRK-VIENNA EVAL-APR90                   7925 1451
                                     DIST-JUN90                   7925 1451
                IRK-EVAL.NLIB 25  7925                            7925 1451
++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451
*************** SUMMARY OF ENDF/B-VI EVALUATION ******************7925 1451
                                                                  7925 1451
   A new evaluation of all neutron and gamma-ray data above the   7925 1451
 resonance region is joined with the endf/b-v resolved resonance  7925 1451
 region evaluation and with the Version VI standard cross section 7925 1451
 for the (n,gamma) reaction below a neutron energy of 2.5 MeV.    7925 1451
                                                                  7925 1451
*************** GENERAL DESCRIPTION ******************************7925 1451
                                                                  7925 1451
              P.G.Young and E.D.Arthur                            7925 1451
                                                                  7925 1451
   the new evaluation for files 3,4,5,12,13,14,15 is based on     7925 1451
 statistical theory, hauser-feshbach, preequilibrium calculations 7925 1451
 with the comnuc and gnash codes (ref1,2). deformed optical poten-7925 1451
 tial of delaroche and ecis coupled-channel code were used to cal-7925 1451
 culate neutron transmission coefficients and total and elastic   7925 1451
 elastic cross sections (ref3,4). gamma-ray strength functions    7925 1451
 were obtained by fitting morgan n,xg data (ref5) at 0.4 and 6.5  7925 1451
 mev. calculated results were used for all major reactions except 7925 1451
 total cross section. for total, the theoretical cross section    7925 1451
 was used as prior in covariance analysis of experimental data    7925 1451
 using glucs code (ref6). more details on experimental data used  7925 1451
 are given below and in main reference for evaluation (ref 7).    7925 1451
                                                                  7925 1451
 **************************************************************** 7925 1451
                                                                  7925 1451
                        STANDARDS COVARIANCES                     7925 1451
                                                                  7925 1451
 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 7925 1451
 expressed the concern that the uncertainties resulting from the  7925 1451
 combination of R-matrix and simultaneous evaluations might have  7925 1451
 led to uncertainties that are too small.  As a result, the       7925 1451
 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 7925 1451
 a set of expanded covariance estimates for the standard cross    7925 1451
 section reactions.  These uncertainties are estimates such that  7925 1451
 if a modern day experiment were performed on a given standard    7925 1451
 cross section using the best techniques, approximately 2/3 of    7925 1451
 the results should fall within these expanded uncertainties. The 7925 1451
 expanded uncertainties for the Au-197(n,gamma) cross section are 7925 1451
 given in the following table and are compared to values from the 7925 1451
 combined output of the standards covariance analysis:            7925 1451
                                                                  7925 1451
    Energy Range    Estimated Uncertainty    Combined Analysis    7925 1451
      (keV)              (percent)               (percent)        7925 1451
                                                                  7925 1451
    2.53E-05                 0.14                   0.14          7925 1451
     200 - 500               3.0                    1.31          7925 1451
     500 - 1000              3.5                    2.1           7925 1451
    1000 - 2500              4.5                    2.0           7925 1451
                                                                  7925 1451
                                                                  7925 1451
*************** mf=2 resonance parameters ************************7925 1451
                                                                  7925 1451
 mt=151    resolved resonance parameters    given from 1.0e-05ev  7925 1451
           to 2 kev based on ref8 and references therein          7925 1451
           and a bound level. some of the reson. spin assignments 7925 1451
           from ref9. from 2 to 4.827 kev the parameters are based7925 1451
           on macklin et al and hoffman et al normalized data.    7925 1451
           see refs 10 and 11.                                    7925 1451
           thermal cross sections are as follows:                 7925 1451
                capture = 98.71 b                                 7925 1451
                scattering = 6.84 b                               7925 1451
                total = 105.55 b                                  7925 1451
           the absorption resonance integral is 1559 b            7925 1451
                                                                  7925 1451
*************** mf=3 smooth neutron cross sections ***************7925 1451
                                                                  7925 1451
 mt=  1  total cross section.  based on glucs covariance analysis 7925 1451
    using deformed optical model calculation as the prior and     7925 1451
    experimental data from refs 12-22, 29 for fitting.            7925 1451
 mt=  2  elastic cross section.  difference of mt=1 and sum of    7925 1451
    all nonelastic cross sections. closely approximates theore-   7925 1451
    tical results.                                                7925 1451
 mt=102  (n,gamma) cross section.  below 2.5 MeV, adopted the     7925 1451
    ENDF/B-VI standard cross section (Ref.30,31) down to the      7925 1451
    resonance region.  At higher energies, the theoretical cal-   7925 1451
    culations were adjusted to agree with experimental data. A    7925 1451
    semi-direct component normalized to an average of experimental7925 1451
    data at 14 MeV was included above En = 6 MeV.                 7925 1451
    at higher energies, use theoretical calculations, which agree 7925 1451
    reasonably with available exp. data.  above 5 mev, calculation7925 1451
    includes semi-direct component normalized to average of       7925 1451
    14 mev data.                                                  7925 1451
                                                                  7925 1451
*************** mf=33 neutron cross section covariances **********7925 1451
                                                                  7925 1451
 mt=  1  total cross section covariance from glucs analysis.      7925 1451
                                                                  7925 1451
*************** references ***************************************7925 1451
                                                                  7925 1451
  1. c.l.dunford. ai-aec-12931(1970)                              7925 1451
  2. p.g.young, e.d.arthur, la-6947 (1977).                       7925 1451
  3. j.p.delaroche, harwell conference (1978)p.366.               7925 1451
  4. j.raynal, iaea smr-9/8 (1970).                               7925 1451
  5. g.l.morgan, e.newman, ornl-tm-4973 (1975).                   7925 1451
  6. d.m.hetrick, c.y.fu, ornl/tm-7341 (1980).                    7925 1451
  7. p.g.young, e.d.arthur, in la-10069-pr (1984)p.12.            7925 1451
  8. s.f.mughabghab and d.i.garber bnl-325,3rd edn,vol i(1973).   7925 1451
  9. a.lottin and a.jain conf on nuclear structure study with     7925 1451
          neutrons,budapest,1972 p34 and private communication.   7925 1451
 10. r.macklin et al. phys. rev/c 11,1270(1975) and private       7925 1451
          communication.                                          7925 1451
 11. m.m. hoffman et al. 71knoxville conf., 2, 868(1971)          7925 1451
 12. w.poenitz et al., nuc.sci.eng. 78, 333(1981).                7925 1451
 13. d.g.foster jr., d.glasgow, phys.rev. c3, 576(1971).          7925 1451
 14. k.k.seth,phys.letters,16,306(1965).                          7925 1451
 15. s.c.snowdon, phys.rev. 90, 615(1953).                        7925 1451
 16. j.f.whalen,anl-7210,16(1966).                                7925 1451
 17. n.nereson, phys.rev. 94, 1678(1954).                         7925 1451
 18. a.bratenahl et al., phys.rev. 110, 927(1958).                7925 1451
 19. j.p.conner,phys.rev.109,1268(1958).                          7925 1451
 20. j.h.coon,phys.rev.88,562(1952).                              7925 1451
 21. j.m.peterson,phys.rev.120,521(1960).                         7925 1451
 22. e.g.bilpuch,private communication(1959).                     7925 1451
 23. j.frehaut et al, proc. 10-50 mev conf, bnl-ncs-51245 (1980)  7925 1451
          page 399.                                               7925 1451
 24. l.r.veeser et al, phys.rev. c16, 1792(1977).                 7925 1451
 25. b.p.bayhurst et al, phys.rev. c12, 451(1975).                7925 1451
 26. r.j.prestwood and b.p.bayhurst,phys.rev.121,1438(1961).      7925 1451
 27. c.kalbach and f.mann, bnl-ncs-5/245,p.689 (1980).            7925 1451
 28. v.j.orphan et al, ga-10248 (1970).                           7925 1451
 29. d.c.larson, proc. 10-50 mev conf, bnl-ncs-51245 (1980) p.277.7925 1451
 30. A.Carlson et al., Nuc.Data for Basic & Applied Science,      7925 1451
     Santa Fe, NM (1985) p.1429.                                  7925 1451
 31. W.Poenitz, ANL-West, personnal communication (1989).         7925 1451
                                                                  7925 1451
******************************************************************7925 1451
                                                                  7925 1451
                                                                  7925 1451
                                                                  7925 1451
                                                                  7925 1451
 ***************** PROGRAM FIXUP (VERSION 86-2) ******************7925 1451
 *RECONSTRUCTED MT NUMBERS                                        7925 1451
   4 =+( 51, 91)                                                  7925 1451
 103 =+(700,718)                                                  7925 1451
 104 =+(720,738)                                                  7925 1451
 105 =+(740,758)                                                  7925 1451
 106 =+(760,778)                                                  7925 1451
 107 =+(780,798)                                                  7925 1451
 101 =+(102,114)                                                  7925 1451
  27 =+( 18, 18)+(101,101)                                        7925 1451
   3 =+(  4,  4)+(  6,  9)+( 16, 17)+( 22, 37)                    7925 1451
  19 =+( 18, 18)-( 20, 21)-( 38, 38)                              7925 1451
   1 =+(  2,  3)                                                  7925 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************7925 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     7925 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     7925 1451
 ***************** Program RECENT (VERSION 2002-1) ***************7925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7925 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     7925 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************7925 1451
 Data Doppler Broadened to 300.000000 Kelvin                      7925 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     7925 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     7925 1451
 ***************** Program FIXUP (Version 2002-1) ****************7925 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 7925 1451
 Corrected Thresholds-----------------------------------------Yes 7925 1451
 Extended Cross Sections to 20 MeV----------------------------No  7925 1451
 Allow Cross Section Deletion---------------------------------No  7925 1451
 Allow Cross Section Reconstruction---------------------------No  7925 1451
 Make All Cross Sections Non-Negative-------------------------Yes 7925 1451
 Delete Energies Not in Ascending Order-----------------------Yes 7925 1451
 Deleted Duplicate Points-------------------------------------Yes 7925 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 7925 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 7925 1451
 Allow Creation of Missing Sections---------------------------No  7925 1451
 Allow Insertion of Energy Points-----------------------------No  7925 1451
 Create Uniform Energy Grid-----------------------------------No  7925 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 7925 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************7925 1451
 Unshielded Group Averages Using  640 Groups                      7925 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     7925 1451
                                1        451        211          27925 1451
                                3         16         43          17925 1451
                                3        102        217          27925 1451
                               33         16         71          17925 1451
                               33        102         26          17925 1451
                                                                  7925 1  0
                                                                  7925 0  0
 7.91970E+4 1.95274E+2          0          0          0          07925 3 16
-8.08000E+6-8.08000E+6          0          0          1        1207925 3 16
        120          1                                            7925 3 16
 8100000.00 .005579896 8200000.00 .023230150 8300000.00 .0417860347925 3 16
 8400000.00 .071166000 8500000.00 .104974000 8600000.00 .1404245007925 3 16
 8700000.00 .179890000 8800000.00 .220456917 8900000.00 .2812400007925 3 16
 9000000.00 .350293333 9100000.00 .419346667 9200000.00 .4888333337925 3 16
 9300000.00 .560920000 9400000.00 .633440000 9500000.00 .7059600007925 3 16
 9600000.00 .778480000 9700000.00 .851136750 9800000.00 .9246140007925 3 16
 9900000.00 .998228000 10000000.0 1.07184200 10100000.0 1.145456007925 3 16
 10200000.0 1.21760825 10300000.0 1.28099000 10400000.0 1.342910007925 3 16
 10500000.0 1.40483000 10600000.0 1.46675000 10700000.0 1.524579007925 3 16
 10800000.0 1.55786200 10900000.0 1.58705400 11000000.0 1.616246007925 3 16
 11100000.0 1.64543800 11200000.0 1.67711725 11300000.0 1.723720007925 3 16
 11400000.0 1.77281000 11500000.0 1.82190000 11600000.0 1.870990007925 3 16
 11700000.0 1.91616500 11800000.0 1.93785000 11900000.0 1.955620007925 3 16
 12000000.0 1.97339000 12100000.0 1.99116000 12200000.0 2.008416687925 3 16
 12300000.0 2.02259345 12400000.0 2.03625690 12500000.0 2.049920347925 3 16
 12600000.0 2.06358379 12700000.0 2.07724724 12800000.0 2.090910697925 3 16
 12900000.0 2.10411356 13000000.0 2.10718360 13100000.0 2.106108407925 3 16
 13200000.0 2.10503320 13300000.0 2.10395800 13400000.0 2.102882807925 3 16
 13500000.0 2.10180760 13600000.0 2.10313545 13700000.0 2.106866367925 3 16
 13800000.0 2.11048693 13900000.0 2.11168000 14000000.0 2.111880007925 3 16
 14100000.0 2.11336363 14200000.0 2.11798500 14300000.0 2.122594637925 3 16
 14400000.0 2.12380800 14500000.0 2.12363200 14600000.0 2.124854387925 3 16
 14700000.0 2.12949500 14800000.0 2.13413834 14900000.0 2.135423147925 3 16
 15000000.0 2.13533400 15100000.0 2.13524486 15200000.0 2.135155717925 3 16
 15300000.0 2.13506657 15400000.0 2.13497743 15500000.0 2.134888297925 3 16
 15600000.0 2.13479914 15700000.0 2.13272231 15800000.0 2.118719337925 3 16
 15900000.0 2.10272867 16000000.0 2.08673800 16100000.0 2.070747337925 3 16
 16200000.0 2.05475667 16300000.0 2.03876600 16400000.0 2.022775337925 3 16
 16500000.0 2.00678467 16600000.0 1.99079400 16700000.0 1.974803337925 3 16
 16800000.0 1.95881267 16900000.0 1.94282200 17000000.0 1.926831337925 3 16
 17100000.0 1.91084067 17200000.0 1.89294833 17300000.0 1.863646007925 3 16
 17400000.0 1.83244200 17500000.0 1.80123800 17600000.0 1.770034007925 3 16
 17700000.0 1.73883000 17800000.0 1.70762600 17900000.0 1.676422007925 3 16
 18000000.0 1.64521800 18100000.0 1.61401400 18200000.0 1.582810007925 3 16
 18300000.0 1.55160600 18400000.0 1.52040200 18500000.0 1.483501767925 3 16
 18600000.0 1.44090529 18700000.0 1.39830882 18800000.0 1.355712357925 3 16
 18900000.0 1.31311588 19000000.0 1.27051941 19100000.0 1.227922947925 3 16
 19200000.0 1.18532647 19300000.0 1.14579943 19400000.0 1.124689007925 3 16
 19500000.0 1.10664800 19600000.0 1.08860700 19700000.0 1.070566007925 3 16
 19800000.0 1.05252500 19900000.0 1.03448400 20000000.0 0.0       7925 3 16
                                                                  7925 3  0
 7.91970E+4 1.95274E+2          0          0          0          07925 3102
 6.51238E+6 6.51238E+6          0          0          1        6417925 3102
        641          1                                            7925 3102
 .000100000 1537.29252 .000105000 1501.44899 .000110000 1467.397307925 3102
 .000115000 1436.01995 .000120000 1399.33870 .000127500 1358.641447925 3102
 .000135000 1321.45967 .000142500 1287.29063 .000150000 1250.310417925 3102
 .000160000 1211.78680 .000170000 1176.73521 .000180000 1144.560717925 3102
 .000190000 1114.62516 .000200000 1087.29347 .000210000 1061.579027925 3102
 .000220000 1037.69457 .000230000 1015.50287 .000240000 989.4543447925 3102
 .000255000 960.831039 .000270000 938.744134 .000280000 914.1573887925 3102
 .000300000 884.173150 .000320000 857.023494 .000340000 832.1588477925 3102
 .000360000 809.257102 .000380000 788.350966 .000400000 766.5551987925 3102
 .000425000 744.251380 .000450000 723.872005 .000475000 705.1891857925 3102
 .000500000 687.619941 .000525000 671.562452 .000550000 656.3934707925 3102
 .000575000 642.285605 .000600000 627.826066 .000630000 612.9510547925 3102
 .000660000 599.340433 .000690000 586.284077 .000720000 572.3914287925 3102
 .000760000 557.495391 .000800000 543.664916 .000840000 530.9903397925 3102
 .000880000 518.943943 .000920000 507.900667 .000960000 497.3246297925 3102
 .001000000 486.362115 .001050000 474.918645 .001100000 464.1913267925 3102
 .001150000 454.307055 .001200000 442.647469 .001275000 429.8334017925 3102
 .001350000 418.118588 .001425000 407.186240 .001500000 395.5727517925 3102
 .001600000 383.430809 .001700000 372.347369 .001800000 362.0972177925 3102
 .001900000 352.734017 .002000000 344.052418 .002100000 335.9056627925 3102
 .002200000 328.445141 .002300000 321.311073 .002400000 313.1706647925 3102
 .002550000 304.124389 .002700000 297.047972 .002800000 289.3582737925 3102
 .003000000 279.892094 .003200000 271.278970 .003400000 263.4050697925 3102
 .003600000 256.227562 .003800000 249.603436 .004000000 242.7024357925 3102
 .004250000 235.699851 .004500000 229.309233 .004750000 223.3359257925 3102
 .005000000 217.852235 .005250000 212.765682 .005500000 207.9548987925 3102
 .005750000 203.542253 .006000000 198.906549 .006300000 194.2713607925 3102
 .006600000 189.886414 .006900000 185.822166 .007200000 181.4082817925 3102
 .007600000 176.693912 .008000000 172.390086 .008400000 168.3883327925 3102
 .008800000 164.712392 .009200000 161.105934 .009600000 157.8468347925 3102
 .010000000 154.330462 .010500000 150.747366 .011000000 147.3700377925 3102
 .011500000 144.221144 .012000000 140.584965 .012750000 136.5312257925 3102
 .013500000 132.824467 .014250000 129.435415 .015000000 125.7629757925 3102
 .016000001 121.935514 .017000001 118.450410 .017999999 115.2633947925 3102
 .018999999 112.286014 .020000000 109.561077 .021000000 107.0224547925 3102
 .022000000 104.638524 .023000000 102.465150 .024000000 99.85883977925 3102
 .025500000 97.0380020 .027000001 94.8246650 .028000001 92.42430877925 3102
 .029999999 89.4684790 .032000002 86.7734053 .034000002 84.31140787925 3102
 .035999998 82.0650200 .037999999 79.9947802 .039999999 77.83657157925 3102
 .042500000 75.6484100 .045000002 73.6514483 .047499999 71.79363757925 3102
 .050000001 70.0886291 .052499998 68.5162423 .055000000 67.03301467925 3102
 .057500001 65.6958772 .059999999 64.2644203 .063000001 62.81128337925 3102
 .066000000 61.4578421 .068999998 60.2146434 .071999997 58.83941687925 3102
 .075999998 57.4021479 .079999998 56.0670142 .083999999 54.82132787925 3102
 .088000000 53.6810646 .092000000 52.5913686 .096000001 51.59775847925 3102
 .100000001 50.5297939 .104999997 49.4450140 .109999999 48.42801377925 3102
 .115000002 47.4819972 .119999997 46.3875128 .127499998 45.16472307925 3102
 .135000005 44.0503141 .142499998 43.0362057 .150000006 41.93990547925 3102
 .159999996 40.8057609 .170000002 39.7944610 .180000007 38.83650607925 3102
 .189999998 37.9911792 .200000003 37.1728634 .209999993 36.44175077925 3102
 .219999999 35.7867340 .230000004 35.1426385 .239999995 34.42348767925 3102
 .254999995 33.6154139 .270000011 33.0136267 .280000001 32.33737407925 3102
 .300000012 31.5248164 .319999993 30.8047948 .340000004 30.16142567925 3102
 .360000014 29.5868954 .379999995 29.0404626 .400000006 28.51089197925 3102
 .425000012 27.9618870 .449999988 27.4882584 .474999994 27.04839567925 3102
 .500000000 26.6687770 .524999976 26.3039381 .550000012 25.99200867925 3102
 .574999988 25.7259568 .600000024 25.4414972 .629999995 25.17634067925 3102
 .660000026 24.9381122 .689999998 24.7209218 .720000029 24.52089797925 3102
 .759999990 24.3100987 .800000012 24.1596553 .839999974 24.02185247925 3102
 .879999995 23.9374044 .920000017 23.8642712 .959999979 23.81519297925 3102
 1.00000000 23.8008509 1.04999995 23.8273729 1.10000002 23.88641777925 3102
 1.14999998 23.9797452 1.20000005 24.1184559 1.27499998 24.36020907925 3102
 1.35000002 24.6743302 1.42499995 25.0446067 1.50000000 25.56475497925 3102
 1.60000002 26.2693686 1.70000005 27.0828354 1.79999995 28.03658207925 3102
 1.89999998 29.1320605 2.00000000 30.3825863 2.09999990 31.79511597925 3102
 2.20000005 33.4102728 2.29999995 35.2433462 2.40000010 37.91256777925 3102
 2.54999995 41.7098948 2.70000005 45.5197409 2.79999995 51.18692347925 3102
 3.00000000 60.9691962 3.20000005 74.5978554 3.40000010 94.41603557925 3102
 3.59999990 124.824071 3.79999995 174.919844 4.00000000 286.0010437925 3102
 4.25000000 617.301275 4.50000000 2459.39283 4.75000000 18409.26697925 3102
 5.00000000 4282.77397 5.25000000 728.833098 5.50000000 291.2255427925 3102
 5.75000000 154.998188 6.00000000 91.9087531 6.30000019 58.06983507925 3102
 6.59999990 39.7909171 6.90000010 28.8500384 7.19999981 20.93280267925 3102
 7.59999990 15.1949028 8.00000000 11.4939378 8.39999962 8.979720707925 3102
 8.80000019 7.19682944 9.19999981 5.88842827 9.60000038 4.902696997925 3102
 10.0000000 4.06426786 10.5000000 3.35862380 11.0000000 2.822208247925 3102
 11.5000000 2.40506334 12.0000000 2.00713668 12.7500000 1.646472067925 3102
 13.5000000 1.37815674 14.2500000 1.17294347 15.0000000 .9909460337925 3102
 16.0000000 .833272462 17.0000000 .713942300 18.0000000 .6219160237925 3102
 19.0000000 .549403977 20.0000000 .491655073 21.0000000 .4450035537925 3102
 22.0000000 .407044594 23.0000000 .375831363 24.0000000 .3445336307925 3102
 25.5000000 .315411599 27.0000000 .296723395 28.0000000 .2802886707925 3102
 30.0000000 .265673723 32.0000000 .258550107 34.0000000 .2583786107925 3102
 36.0000000 .265360233 38.0000000 .280505870 40.0000000 .3109225867925 3102
 42.5000000 .375780029 45.0000000 2.12704336 47.5000000 .6352437287925 3102
 50.0000000 .940577311 52.5000000 1.74316513 55.0000000 5.482857417925 3102
 57.5000000 129.955698 60.0000000 550.969463 63.0000000 3.949048027925 3102
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                                                                  792533  0
                                                                  7925 0  0
                                                                     0 0  0
 8.01990E+4 1.97259E+2          0          0         34         108034 1451
 0.0        0.0                 0          0          0          68034 1451
 1.00000E+0 2.00000E+7          0          0         10       20028034 1451
 3.00000E+2 0.0                 1          0         78          48034 1451
 80-Hg-199 JAERI      EVAL-MAR90 K.SAKURAI                        8034 1451
                      DIST-Feb2004                                8034 1451
----IRDF-2002         MATERIAL 8034                               8034 1451
-----INCIDENT NEUTRON DATA                                        8034 1451
------ENDF-6 FORMAT                                               8034 1451
***************************************************************   8034 1451
 80-HG-199 JAERI      EVAL-MAR90 K.SAKURAI                        8034 1451
                      DIST-JUL91                                  8034 1451
----JENDL/D-99        MATERIAL 8034                               8034 1451
***************************************************************   8034 1451
HISTORY                                                           8034 1451
90-03  EVALUATION WAS MADE BY K. SAKURA(JAERI).                   8034 1451
90-10  COMPILED TO JENDL DOSIMETRY FILE VERSION 1.                8034 1451
91-07  DESCRIPTIVE DATA WERE ADDED.                               8034 1451
                                                                  8034 1451
            ===== POINT-WISE DATA FILE =====                      8034 1451
                                                                  8034 1451
 <DOSIMETRY CROSS SECTION>                                        8034 1451
      HG-199 (N,N')HG-199M (HALF-LIFE = 42.6 M)                   8034 1451
                                                                  8034 1451
MF=1  GENERAL INFORMATION                                         8034 1451
 MT=451  DESCRIPTIVE DATA AND DICTIONARY                          8034 1451
                                                                  8034 1451
MF=2  RESONANCE PARAMETERS                                        8034 1451
 MT=151  PARAMETERS                                               8034 1451
     ONLY SPIN AND SCATTERING RADIUS ARE GIVEN.                   8034 1451
                                                                  8034 1451
MF=3   NEUTRON CROSS SECTIONS                                     8034 1451
 MT=51  HG-199(N,N')HG-199M CROSS SECTION                         8034 1451
     ISOMERIC STATE = 0.531 MEV 13/2+, HALF-LIFE = 42.6 M         8034 1451
     EVALUATION WAS BASED ON THE EXPERIMENTAL DATA REPORTED IN    8034 1451
     REF./1/.                                                     8034 1451
                                                                  8034 1451
MF=33  COVARIANCES OF NEUTRON CROSS SECTIONS                      8034 1451
 MT=51                                                            8034 1451
    ESTIMATED FROM THE EXPERIMENTAL DATA.                         8034 1451
                                                                  8034 1451
REFERENCE                                                         8034 1451
 1) SAKURAI K., ET AL.: J. NUCL. SCI. TECHNOL., 19, 775 (1982).   8034 1451
                                                                  8034 1451
***************************************************************   8034 1451
                    ----- MF=3 MT=51  -----                       8034 1451
  For the IRDF-2002 file this reaction was converted at           8034 1451
  IAEA/NDS.                                                       8034 1451
  The reaction MF/MT=3/51 was converted to MF/MT=10/ 4            8034 1451
  The corresponding co-variance files were also converted         8034 1451
  The reaction MF/MT=33/51 was converted to MF/MT=40/ 4           8034 1451
***************************************************************   8034 1451
 The threshold energy was recalculated from 5.25000+5 to          8034 1451
 5.33691+5 to satisfy code FIZCON criteria                        8034 1451
                                                                  8034 1451
                                                                  8034 1451
***************************************************************   8034 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************8034 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     8034 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     8034 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************8034 1451
 Data Doppler Broadened to 300.000000 Kelvin                      8034 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     8034 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     8034 1451
 ***************** Program FIXUP (Version 2002-1) ****************8034 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 8034 1451
 Corrected Thresholds-----------------------------------------Yes 8034 1451
 Extended Cross Sections to 20 MeV----------------------------No  8034 1451
 Allow Cross Section Deletion---------------------------------No  8034 1451
 Allow Cross Section Reconstruction---------------------------No  8034 1451
 Make All Cross Sections Non-Negative-------------------------Yes 8034 1451
 Delete Energies Not in Ascending Order-----------------------Yes 8034 1451
 Deleted Duplicate Points-------------------------------------Yes 8034 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 8034 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 8034 1451
 Allow Creation of Missing Sections---------------------------No  8034 1451
 Allow Insertion of Energy Points-----------------------------No  8034 1451
 Create Uniform Energy Grid-----------------------------------No  8034 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 8034 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************8034 1451
 Unshielded Group Averages Using  640 Groups                      8034 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     8034 1451
                                1        451         86          08034 1451
                                8          4          2          08034 1451
                               10          4         72          08034 1451
                               40          4         17          08034 1451
                                                                  8034 1  0
                                                                  8034 0  0
 8.01990E+4 1.97259E+2          0          0          1          08034 3291
 0.0       -5.30999E+5      80199          0          1        2058034 3291
        205          1                                            8034 3291
 525000.000 .000774677 550000.000 .004195311 575000.000 .0078359368034 3291
 600000.000 .011840624 630000.000 .016209374 660000.000 .0205781248034 3291
 690000.000 .024946875 720000.000 .030024219 760000.000 .0354000008034 3291
 800000.000 .040600000 840000.000 .045800000 880000.000 .0510000008034 3291
 920000.000 .056200000 960000.000 .061400000 1000000.00 .0714000008034 3291
 1100000.00 .086200000 1200000.00 .102100000 1300000.00 .1246000008034 3291
 1400000.00 .148200000 1500000.00 .174000000 1600000.00 .2020000008034 3291
 1700000.00 .229000000 1800000.00 .250000000 1900000.00 .2700000008034 3291
 2000000.00 .290000000 2100000.00 .310000000 2200000.00 .3292500008034 3291
 2300000.00 .344000000 2400000.00 .358000000 2500000.00 .3750000008034 3291
 2600000.00 .395000000 2700000.00 .414250000 2800000.00 .4290000008034 3291
 2900000.00 .443000000 3000000.00 .457000000 3100000.00 .4710000008034 3291
 3200000.00 .484000000 3300000.00 .491000000 3400000.00 .4970000008034 3291
 3500000.00 .506000000 3600000.00 .518000000 3700000.00 .5295000008034 3291
 3800000.00 .538000000 3900000.00 .546000000 4000000.00 .5540000008034 3291
 4100000.00 .562000000 4200000.00 .569500000 4300000.00 .5740000008034 3291
 4400000.00 .578000000 4500000.00 .582000000 4600000.00 .5860000008034 3291
 4700000.00 .590000000 4800000.00 .594000000 4900000.00 .5980000008034 3291
 5000000.00 .600400000 5100000.00 .601200000 5200000.00 .6023000008034 3291
 5300000.00 .605200000 5400000.00 .608400000 5500000.00 .6120000008034 3291
 5600000.00 .616000000 5700000.00 .619550000 5800000.00 .6204000008034 3291
 5900000.00 .620800000 6000000.00 .622900000 6100000.00 .6267000008034 3291
 6200000.00 .630500000 6300000.00 .634300000 6400000.00 .6381000008034 3291
 6500000.00 .640000000 6600000.00 .640000000 6700000.00 .6400000008034 3291
 6800000.00 .640000000 6900000.00 .640000000 7000000.00 .6400000008034 3291
 7100000.00 .640000000 7200000.00 .640000000 7300000.00 .6400000008034 3291
 7400000.00 .640000000 7500000.00 .640000000 7600000.00 .6400000008034 3291
 7700000.00 .640000000 7800000.00 .640000000 7900000.00 .6400000008034 3291
 8000000.00 .637000000 8100000.00 .631000000 8200000.00 .6256250008034 3291
 8300000.00 .624000000 8400000.00 .623000000 8500000.00 .6220000008034 3291
 8600000.00 .621000000 8700000.00 .619625000 8800000.00 .6160000008034 3291
 8900000.00 .612000000 9000000.00 .606000000 9100000.00 .5980000008034 3291
 9200000.00 .590500000 9300000.00 .586000000 9400000.00 .5820000008034 3291
 9500000.00 .576000000 9600000.00 .568000000 9700000.00 .5595000008034 3291
 9800000.00 .548000000 9900000.00 .536000000 10000000.0 .5260000008034 3291
 10100000.0 .518000000 10200000.0 .510500000 10300000.0 .5060000008034 3291
 10400000.0 .502000000 10500000.0 .494000000 10600000.0 .4820000008034 3291
 10700000.0 .470500000 10800000.0 .462000000 10900000.0 .4540000008034 3291
 11000000.0 .442000000 11100000.0 .426000000 11200000.0 .4105000008034 3291
 11300000.0 .398000000 11400000.0 .386000000 11500000.0 .3740000008034 3291
 11600000.0 .362000000 11700000.0 .349000000 11800000.0 .3300000008034 3291
 11900000.0 .310000000 12000000.0 .296000000 12100000.0 .2880000008034 3291
 12200000.0 .279500000 12300000.0 .268000000 12400000.0 .2560000008034 3291
 12500000.0 .244000000 12600000.0 .232000000 12700000.0 .2205000008034 3291
 12800000.0 .212000000 12900000.0 .204000000 13000000.0 .1960000008034 3291
 13100000.0 .188000000 13200000.0 .180500000 13300000.0 .1760000008034 3291
 13400000.0 .172000000 13500000.0 .168000000 13600000.0 .1640000008034 3291
 13700000.0 .160000000 13800000.0 .156000000 13900000.0 .1520000008034 3291
 14000000.0 .148000000 14100000.0 .144000000 14200000.0 .1400000008034 3291
 14300000.0 .136000000 14400000.0 .132000000 14500000.0 .1280000008034 3291
 14600000.0 .124000000 14700000.0 .120250000 14800000.0 .1180000008034 3291
 14900000.0 .116000000 15000000.0 .114000000 15100000.0 .1120000008034 3291
 15200000.0 .110000000 15300000.0 .108000000 15400000.0 .1060000008034 3291
 15500000.0 .104600000 15600000.0 .103800000 15700000.0 .1029500008034 3291
 15800000.0 .101800000 15900000.0 .100600000 16000000.0 .0996000008034 3291
 16100000.0 .098800000 16200000.0 .098050000 16300000.0 .0976000008034 3291
 16400000.0 .097200000 16500000.0 .096400000 16600000.0 .0952000008034 3291
 16700000.0 .093950000 16800000.0 .092400000 16900000.0 .0908000008034 3291
 17000000.0 .089800000 17100000.0 .089400000 17200000.0 .0890000008034 3291
 17300000.0 .088600000 17400000.0 .088200000 17500000.0 .0878000008034 3291
 17600000.0 .087400000 17700000.0 .087000000 17800000.0 .0866000008034 3291
 17900000.0 .086200000 18000000.0 .085800000 18100000.0 .0854000008034 3291
 18200000.0 .085000000 18300000.0 .084600000 18400000.0 .0842000008034 3291
 18500000.0 .083800000 18600000.0 .083400000 18700000.0 .0830000008034 3291
 18800000.0 .082600000 18900000.0 .082200000 19000000.0 .0818000008034 3291
 19100000.0 .081400000 19200000.0 .081050000 19300000.0 .0810000008034 3291
 19400000.0 .081000000 19500000.0 .080800000 19600000.0 .0804000008034 3291
 19700000.0 .080050000 19800000.0 .080000000 19900000.0 .0800000008034 3291
 20000000.0 0.0                                                   8034 3291
                                                                  8034 3  0
                                                                  8034 0  0
 8.01990E+4 1.97259E+2          0          0          0          1803433291
 0.0000E+00 0.0000E+00          0        291          0          2803433291
 0.0000E+00 0.0000E+00          0          1         62         31803433291
 1.000000-5 0.000000+0 5.250000+5 2.464000-1 7.500000+5 8.640000-2803433291
 1.000000+6 1.600000-2 2.000000+6 1.600000-2 2.250000+6 2.200000-2803433291
 3.000000+6 2.200000-2 3.500000+6 1.600000-2 4.000000+6 1.600000-2803433291
 5.000000+6 1.600000-2 6.000000+6 1.600000-2 7.000000+6 1.600000-2803433291
 7.250000+6 8.640000-2 8.000000+6 8.640000-2 9.000000+6 8.640000-2803433291
 1.000000+7 8.640000-2 1.100000+7 8.640000-2 1.200000+7 8.640000-2803433291
 1.300000+7 8.640000-2 1.325000+7 3.640000-2 1.375000+7 6.400000-3803433291
 1.400000+7 6.400000-3 1.450000+7 3.640000-2 1.500000+7 3.640000-2803433291
 1.525000+7 8.640000-2 1.600000+7 8.640000-2 1.650000+7 2.464000-1803433291
 1.700000+7 2.464000-1 1.800000+7 2.464000-1 1.900000+7 2.464000-1803433291
 2.000000+7 0.000000+0                                            803433291
 0.000000+0 0.000000+0          0          1          6          3803433291
 1.000000-5 0.000000+0 5.250000+5 3.600000-3 2.000000+7 0.000000+0803433291
                                                                  803433  0
                                                                  8034 0  0
                                                                     0 0  0
 8.22040E+4 2.02220E+2          0          0         34         108225 1451
 0.0        0.0                 0          0          0          68225 1451
 1.00000E+0 2.00000E+7          0          0         10       20028225 1451
 3.00000E+2 0.0                 1          0        147          48225 1451
 82-Pb-204 FEI        EVAL-Oct01 K.I.Zolotarev                    8225 1451
                      DIST-Feb2004                                8225 1451
----IRDF-2002         MATERIAL 8225                               8225 1451
-----INCIDENT NEUTRON DATA                                        8225 1451
------ENDF-6 FORMAT                                               8225 1451
***************************************************************** 8225 1451
 82-Pb-204 FEI        EVAL-Oct01 K.I.Zolotarev                    8225 1451
                      DIST-Oct01                       20011014   8225 1451
----BROND-2           MATERIAL 8225                               8225 1451
------Russian Reactor Dosimetry File RRDF-2002                    8225 1451
***************************************************************** 8225 1451
                    ----- MF=3 MT=71  -----                       8225 1451
  For the IRDF-2002 file this reaction was converted at IAEA/NDS  8225 1451
  The reaction MF/MT=3/71 was converted to MF/MT=10/ 4            8225 1451
  The corresponding co-variance files were also converted         8225 1451
  The reaction MF/MT=33/71 was converted to MF/MT=40/ 4           8225 1451
***************************************************************** 8225 1451
Author  of evaluation: K.I.Zolotarev                              8225 1451
***************************************************************** 8225 1451
                                                                  8225 1451
                    ----- MF=3 MT=71  -----                       8225 1451
                                                                  8225 1451
    In the section MT= 71 are given evaluated  cross section data 8225 1451
of neutron inelastic excitation of the 67.2-min isomeric state in 8225 1451
in the  lead-204 in the energy range  from threshold  to  20 MeV. 8225 1451
The isomeric state is the 21-excited level in Pb-204.  This level 8225 1451
has energy 2185.79 keV  with spin and parity 9- [1].              8225 1451
    Data base for the evaluation Pb-204(n,n')Pb-204m  excitation  8225 1451
function was formed from microscopic experimental data [2-9] and  8225 1451
data received from theoretical model calculations. Experimental   8225 1451
data [3-9] were corrected to the new standards. Data of Decowski  8225 1451
et al. [6] were renormalized also to the result of precise measu- 8225 1451
rements of Ryves et al. at 14.3 MeV [8]. Uncertainty for cross    8225 1451
section value measured by Bornemisza-Pauspertl et al. at 2.80 MeV 8225 1451
[2] was evaluated as +-35 %.  Theoretical model calculations was  8225 1451
done by means of GNASH code [10]. New modified  version of GNASH  8225 1451
was used for calculations [11].                                   8225 1451
    The evaluation Pb-204(n,n')Pb-204m excitation function  from  8225 1451
threshold to 20 MeV has been carried out within the framework of  8225 1451
generalized least squares method , rational function was used as  8225 1451
model function [12]. Procedure of calculation recommended cross   8225 1451
section data was performed by means of PADE-2 code [13].          8225 1451
    Integral experimental data for U-235 neutron fission spectrum 8225 1451
[14,15] and Cf-252 spontaneous fission neutron spectrum [16,17]   8225 1451
was used for testing evaluated Pb-204(n,n')Pb-204m excitation fun-8225 1451
ction. Data for U-235 thermal fission neutron spectrum and Cf-252 8225 1451
spontaneous fission neutron spectrum were taken from ref.[18] and 8225 1451
[19],  respectively.  The  results  of  testing are given in  the 8225 1451
table 1.                                                          8225 1451
                                                     Table 1.     8225 1451
 ---------------------------------------------------------------  8225 1451
    TYPE OF SPECTRUM   I <CS>,MB (calc.) I   <CS>, MB (measured)  8225 1451
 ----------------------I-----------------I----------------------  8225 1451
                       I                 I                        8225 1451
 U-235 neutron fission I   17.770        I 18.900 +- 2.000 [14]   8225 1451
                       I                 I 19.010 +- 1.536 [15]   8225 1451
                       I                 I                        8225 1451
 ----------------------I-----------------I----------------------  8225 1451
                       I                 I                        8225 1451
 CF-252 spont. fission I   20.373        I 20.900 +- 1.202 [16]   8225 1451
                       I                 I 20.850 +- 0.920 [17]   8225 1451
 ---------------------------------------------------------------  8225 1451
                                                                  8225 1451
                    ----- MF=33 MT=71 -----                       8225 1451
    Uncertainties in  the evaluated excitation function  for the  8225 1451
reaction  Pb-204(n,n')Pb-204m  are given  in the form of relative 8225 1451
covariance matrix for the 32-neutron energy groups (LB=5).  Cova- 8225 1451
riance matrix of uncertainties was calculated simultaneously with 8225 1451
recommended cross section data by means of PADE-2 code.           8225 1451
    Eigenvalues  of the  6-th digits  relative covariance matrix  8225 1451
given in the 33-file are the following:                           8225 1451
    2.56972E-09    3.22598E-09    3.49414E-09    4.00985E-09      8225 1451
    4.91484E-09    5.78671E-09    6.70005E-09    8.71427E-09      8225 1451
    1.03160E-08    1.33154E-08    1.55381E-08    2.07583E-08      8225 1451
    2.40281E-08    3.05913E-08    4.05801E-08    4.76532E-08      8225 1451
    5.93691E-08    9.33863E-08    1.58183E-07    2.72581E-06      8225 1451
    1.76176E-04    7.87730E-04    1.22783E-03    2.11803E-03      8225 1451
    3.22408E-03    4.16999E-03    4.89344E-03    1.08231E-02      8225 1451
    2.01870E-02    3.34051E-02    7.41942E-02                     8225 1451
                                                                  8225 1451
References :                                                      8225 1451
 1. R.B.Firestone   Table of Isotopes, Eighth edition, Vol. 1,    8225 1451
    John Wiley & Sons, Inc., New York, 1995                       8225 1451
 2. P.Bornemisza-Pauspertl, J.Karolyi, G.Peto Atomki Kozl., v.10, 8225 1451
    no.2, p.112, July 1968                                        8225 1451
 3. J.Csikai, G.Peto  Acta Phys. Hung., v.23, p.87, May 1967      8225 1451
 4. A.K.Hankla, R.W.Fink, J.H.Hamilton  Nucl. Phys., v.A180,      8225 1451
    p.157, January 1972                                           8225 1451
 5. G.N.Maslov, F.Nasyrov, N.F.Pashkin Yadernye Konstanty, v.9,   8225 1451
    p.50, Obninsk, 1972                                           8225 1451
 6. P.Decowski et al.  Nucl. Phys., v.A204, p.121, April 1973     8225 1451
 7. D.L.Smith, J.W.Meadows  Report ANL-NDM-37, December 1977      8225 1451
 8. T.B.Ryves, P.Kolkowski, A.C.Hooley  Annals Nucl. Energy,      8225 1451
    v.17, p.107, 1990                                             8225 1451
 9. I.Kimura, K.Kobayashi  Nucl. Sci. Eng., v.106, p.332, 1990    8225 1451
10. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear    8225 1451
    Model Code for Calculation of Cross Section and Emission      8225 1451
    Spectra. Report LA-6947, Los Alamos, 1977                     8225 1451
11. E.L.Trykov, G.Ya.Tertychnyi  Private communication, IPPE,     8225 1451
    Obninsk, May 1999                                             8225 1451
12. S.Badikov,N.Rabotnov,K.Zolotarev Proc. of NEANSC Specialist's 8225 1451
    Meeting on Evaluation and Processing of Covariance Data, Oak  8225 1451
    Ridge , USA, 7-9 September 1992, OECD, Paris, 1993, p.105     8225 1451
13. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          8225 1451
14. I.Kimura, K.Kobayashi, T.Shibata  Nucl. Sci. Technol., v.8,   8225 1451
    p.59, February 1971                                           8225 1451
15. A.K.Brodskaja et al.  Jadernyje Konstanty, no.23, v.4,        8225 1451
    October 1976                                                  8225 1451
16. J.Csikai, Z.Dezso  Proc. Educational Seminar on the use of    8225 1451
    Cf-252, Karlsruhe, 14-18 April 1975, p.29                     8225 1451
17. K.Kobayashi et al.  Report KURRI-AR-17, p.15, 1984  ;         8225 1451
    K.Kobayashi et al.  Progress Report NEANDC(J)-106/U, p.41,    8225 1451
    September 1984                                                8225 1451
18. L.W.Weston et al. Evaluated Neutron Data File for U-235,      8225 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    8225 1451
19. W.Mannhart  Report IAEA-TECDOC-410, IAEA, Vienna, 1987, p158  8225 1451
***************************************************************** 8225 1451
 File 2 added to the pointwise file containing only the effective 8225 1451
 scattering radius with no resonance parameters given.            8225 1451
 Taken from JENDL-3.2                                             8225 1451
                                                                  8225 1451
***************************************************************** 8225 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************8225 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     8225 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     8225 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************8225 1451
 Data Doppler Broadened to 300.000000 Kelvin                      8225 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     8225 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     8225 1451
 ***************** Program FIXUP (Version 2002-1) ****************8225 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 8225 1451
 Corrected Thresholds-----------------------------------------Yes 8225 1451
 Extended Cross Sections to 20 MeV----------------------------No  8225 1451
 Allow Cross Section Deletion---------------------------------No  8225 1451
 Allow Cross Section Reconstruction---------------------------No  8225 1451
 Make All Cross Sections Non-Negative-------------------------Yes 8225 1451
 Delete Energies Not in Ascending Order-----------------------Yes 8225 1451
 Deleted Duplicate Points-------------------------------------Yes 8225 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 8225 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 8225 1451
 Allow Creation of Missing Sections---------------------------No  8225 1451
 Allow Insertion of Energy Points-----------------------------No  8225 1451
 Create Uniform Energy Grid-----------------------------------No  8225 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 8225 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************8225 1451
 Unshielded Group Averages Using  640 Groups                      8225 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     8225 1451
                                1        451        155          18225 1451
                                8          4          2          18225 1451
                               10          4         63          18225 1451
                               40          4         98          18225 1451
                                                                  8225 1  0
                                                                  8225 0  0
 8.22040E+4 2.02220E+2          0          0          1          08225 3291
 0.0       -2.18579E+6      82204          0          1        1808225 3291
        180          1                                            8225 3291
 2100000.00 1.34936E-8 2200000.00 3.51923E-5 2300000.00 .0001891968225 3291
 2400000.00 .001283871 2500000.00 .004581835 2600000.00 .0080065758225 3291
 2700000.00 .011515626 2800000.00 .015279183 2900000.00 .0192366028225 3291
 3000000.00 .023323413 3100000.00 .027458638 3200000.00 .0315938638225 3291
 3300000.00 .035729088 3400000.00 .039819900 3500000.00 .0438663008225 3291
 3600000.00 .047857600 3700000.00 .051793800 3800000.00 .0556669508225 3291
 3900000.00 .059477050 4000000.00 .063287150 4100000.00 .0670321088225 3291
 4200000.00 .070711925 4300000.00 .074391742 4400000.00 .0780715588225 3291
 4500000.00 .081751375 4600000.00 .085431192 4700000.00 .0891182908225 3291
 4800000.00 .092812670 4900000.00 .096507050 5000000.00 .1002014308225 3291
 5100000.00 .103895810 5200000.00 .107668750 5300000.00 .1115202508225 3291
 5400000.00 .115371750 5500000.00 .119223250 5600000.00 .1231751678225 3291
 5700000.00 .127227500 5800000.00 .131279833 5900000.00 .1354421678225 3291
 6000000.00 .139714500 6100000.00 .143986833 6200000.00 .1483836678225 3291
 6300000.00 .152905000 6400000.00 .157426333 6500000.00 .1620780008225 3291
 6600000.00 .166860000 6700000.00 .171642000 6800000.00 .1765623758225 3291
 6900000.00 .181621125 7000000.00 .186679875 7100000.00 .1917386258225 3291
 7200000.00 .196883429 7300000.00 .202114286 7400000.00 .2073451438225 3291
 7500000.00 .212576000 7600000.00 .217806857 7700000.00 .2230377148225 3291
 7800000.00 .228268571 7900000.00 .233326500 8000000.00 .2382115008225 3291
 8100000.00 .243096500 8200000.00 .247725500 8300000.00 .2520985008225 3291
 8400000.00 .256166250 8500000.00 .259928750 8600000.00 .2632957508225 3291
 8700000.00 .266267250 8800000.00 .268757750 8900000.00 .2707672508225 3291
 9000000.00 .272370000 9100000.00 .273276000 9200000.00 .2735920008225 3291
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 9900000.00 .258890750 10000000.0 .254652250 10100000.0 .2499567508225 3291
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 11100000.0 .190766000 11200000.0 .184529333 11300000.0 .1784821678225 3291
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 13200000.0 .095205225 13300000.0 .092443475 13400000.0 .0898044508225 3291
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 13800000.0 .080394125 13900000.0 .078299575 14000000.0 .0762947258225 3291
 14100000.0 .074379575 14200000.0 .072544700 14300000.0 .0707901008225 3291
 14400000.0 .069123583 14500000.0 .067545150 14600000.0 .0659667178225 3291
 14700000.0 .064479067 14800000.0 .063082200 14900000.0 .0616853338225 3291
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 15600000.0 .053400017 15700000.0 .052390450 15800000.0 .0513808838225 3291
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 16800000.0 .042869338 16900000.0 .042109113 17000000.0 .0413804408225 3291
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 17700000.0 .036630108 17800000.0 .035984792 17900000.0 .0353394758225 3291
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 19800000.0 .023673350 19900000.0 .023057650 20000000.0 0.0       8225 3291
                                                                  8225 3  0
                                                                  8225 0  0
 8.22040E+4 2.02220E+2          0          0          0          1822533291
 0.0000E+00 0.0000E+00          0        291          0          1822533291
 0.0000E+00 0.0000E+00          1          5        561         33822533291
 1.000000-5 2.100000+6 2.500000+6 3.000000+6 3.500000+6 4.000000+6822533291
 4.500000+6 5.000000+6 5.500000+6 6.000000+6 6.500000+6 7.000000+6822533291
 7.500000+6 8.000000+6 8.500000+6 9.000000+6 9.500000+6 1.000000+7822533291
 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7 1.300000+7822533291
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 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 6.053160-3822533291
 4.115730-3 3.116670-3 2.961290-3 3.015440-3 3.068430-3 3.122180-3822533291
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 3.469890-4 4.076110-4 3.881820-4 2.944740-4 1.518610-4-6.200680-6822533291
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 2.104450-3 1.567490-3 1.212250-3 1.125220-3 1.782580-3 3.259190-3822533291
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 2.545590-3 4.012270-3 5.735180-3 2.716800-3 2.616640-3 2.520100-3822533291
 2.426860-3 2.449610-3 2.727590-3 3.350830-3 4.472360-3 2.879730-3822533291
 3.109940-3 3.340090-3 3.428280-3 3.287320-3 3.123280-3 3.473200-3822533291
 3.658080-3 4.259390-3 4.573780-3 4.243340-3 3.475180-3 2.932480-3822533291
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 1.989260-2 1.732600-2 2.174600-2                                 822533291
                                                                  822533  0
                                                                  8225 0  0
                                                                     0 0  0
 9.02320E+4 2.30044E+2          0          1         34         109040 1451
 0.0        1.00000E+0          0          0          0          69040 1451
 1.00000E+0 2.00000E+7          0          0         10       20029040 1451
 3.00000E+2 0.0                 1          0        208          59040 1451
 90-Th-232 BNL,ANL+   EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+   9040 1451
                      DIST-Feb2004                                9040 1451
----IRDF-2002         MATERIAL 9040                               9040 1451
-----INCIDENT NEUTRON DATA                                        9040 1451
------ENDF-6 FORMAT                                               9040 1451
******************************************************************9040 1451
 90-TH-232 BNL,ANL+   EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+   9040 1451
                      DIST-FEB90                       19900205   9040 1451
----ENDF/B-VI         MATERIAL 9040                               9040 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9040 1451
DOSIMETRY                                                         9040 1451
******************************************************************9040 1451
 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC             9040 1451
 *         *          *          *          *          *         *9040 1451
                                                                  9040 1451
 PRINCIPAL EVALUATORS- J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH9040 1451
                       AND J.F.WHALEN(ANL)                        9040 1451
                       R.J.HOWERTON(LLL)                          9040 1451
                       B.R.LEONARD ET.AL.(BNW)                    9040 1451
                       G.DE SAUSSURE,R.L.MACKLIN AND R.GWIN(ORNL) 9040 1451
                       D.K. OLSEN (ORNL)                          9040 1451
 THIS EVALUATION ASSEMBLED BY M.R.BHAT(BNL) USING THE FOLLOWING   9040 1451
 EVALUATIONS                                                      9040 1451
                                                                  9040 1451
      NU-BAR(PROMPT) - R.GWIN(ORNL)(REF 4)                        9040 1451
      DELAYED NEUTRONS -M.BRADY AND T.ENGLAND (REF 5)             9040 1451
      ENERGY RELEASE IN FISSION - R.SHER ET.AL.(STANFORD)(REF 6)  9040 1451
      THERMAL RANGE - B.R.LEONARD JR.ET.AL.(REF 2) WITH CHANGES BY9040 1451
                      M.R.BHAT(REF 31)                            9040 1451
      RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37)             9040 1451
                                                                  9040 1451
      UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN 9040 1451
                       (REF 3)WITH CHANGES BY M.R.BHAT(REF 31)    9040 1451
      SMOOTH CROSS-SECTIONS 50KEV TO 20MEV -                      9040 1451
      TOTAL - A.B.SMITH(ANL)(REF1)                                9040 1451
      FISSION - W.P.POENITZ(ANL)(REF1)                            9040 1451
      CAPTURE - W.P.POENITZ(ANL)(REF1)                            9040 1451
      INELASTIC AND OTHER CROSS-SECTIONS - J.W.MEADOWS ET.AL.(ANL)9040 1451
                                                            (REF1)9040 1451
      FISSION NEUTRON SPECTRUM - M.R.BHAT(BNL) (REF 31)           9040 1451
      GAMMA-RAY PRODUCTION - R.J.HOWERTON(LLL)                    9040 1451
                                                                  9040 1451
                             MF = 1                               9040 1451
                                                                  9040 1451
 MT=452(NU-BAR TOTAL) CONSISTENT WITH MT=455 AND MT=456           9040 1451
 MT=455(DELAYED NEUTRON YIELDS) (REF 5)                           9040 1451
 MT=456(PROMPT NU-BAR) BASED ON CF-252 NU-BAR(PROMPT)=3.757       9040 1451
      EVALUATION BY GWIN(ORNL) BASED ON A STRAIGHT LINE LEAST-    9040 1451
      SQUARES FIT TO THE AVAILABLE DATA LISTED IN REF.4.          9040 1451
 MT=458(ENERGY RELEASE IN FISSION)EVALUATION BY R.SHER ET.AL(STAN-9040 1451
      FORD)(REF 6)                                                9040 1451
                                                                  9040 1451
                             MF = 2                               9040 1451
                                                                  9040 1451
 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) RESONANCE        9040 1451
      PARAMETERS ARE UNCERTAINTY WEIGHTED AVERAGE OF AVAILABLE    9040 1451
      DATA AND THE TWO BOUND RESONANCES OF REF 2 WITH MF=3 FILES  9040 1451
      FOR TRUNCATION, MISSED P-WAVES, 1/V CAPTURE, AND SMOOTH     9040 1451
      CONNECTION TO THE END OF THE THERMAL REGION CROSS SECTIONS  9040 1451
      AT 5 EV. AVERAGE CAPTURE WIDTH = 2.44 X 10-2 EV, S-WAVE     9040 1451
      STRENGTH FUNCTION = 0.826 X 10-4, P-WAVE STRENGTH FUNCTION  9040 1451
      ASSUMED = 1.6 X 10-4, 5 TO 4000 EV INFINITLY DILUTE CAPTURE 9040 1451
      RESONANCE INTEGRAL = 81.58 B.                               9040 1451
                                                                  9040 1451
 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN(REF3)9040 1451
      EVALUATION OF CAPTURE CROSS-SECTION USED FROM 4 TO 25KEV.   9040 1451
      ABOVE 25KEV THIS WAS JOINED SMOOTHLY TO THE POENITZ EVALUAT-9040 1451
      ION FROM 40KEV TO 50KEV. THESE WERE THEN FITTED USING THE   9040 1451
      CODE UR(REF30) TO GIVE THE UNRESOLVED RESONANCE             9040 1451
      PARAMETERS. IN THIS FIT THE S- AND D-WAVE PARAMETERS WERE   9040 1451
      KEPT CONSTANT AT THE VALUES GIVEN BY DE SAUSSURE AND MACKLIN9040 1451
      THE P-WAVE GAMMA WIDTH WAS INCREASED TO 2.52E-02EV AND THE  9040 1451
      P-WAVE REDUCED NEUTRON WIDTH VARIED TO FIT THE CAPTURE      9040 1451
      CROSS-SECTION.                                              9040 1451
      THE CAPTURE CROSS-SECTION INPUT USED TO EXTRACT THE         9040 1451
      UNRESOLVED RESONANCE PARAMETERS IN FILE 2/151 ARE GIVEN HERE9040 1451
      FROM 4KEV TO 50KEV AT 1KEV INTERVAL. THESE ARE -            9040 1451
      1.130,0.982,0.908,0.850,0.803,0.764,0.730,0.702,0.677,0.655,9040 1451
      0.636,0.618,0.601,0.585,0.571,0.557,0.545,0.533,0.521,0.511,9040 1451
      0.500,0.491,0.482,0.473,0.465,0.459,0.452,0.448,0.443,0.439,9040 1451
      0.435,0.431,0.429,0.426,0.424,0.422,0.420,0.420,0.420,0.420,9040 1451
      0.417,0.414,0.411,0.407,0.401,0.395,0.390.                  9040 1451
                                                                  9040 1451
                             MF = 3                               9040 1451
                                                                  9040 1451
 THERMAL REGION(1.0E-05 TO 5.0EV) - MAINLY BASED ON LEONARD(REF 2)9040 1451
      EVALUATION WITH THE FOLLOWING CHANGES. THE CAPTURE CROSS-   9040 1451
      SECTION IN REF 2 NORMALISED TO 7.40B AT 2.53E-02 EV AND USED9040 1451
      FROM 1.0E-05EV TO 2.53E-02EV. FROM 2.6E-02EV TO 14EV THE FIT9040 1451
      TO THEIR CAPTURE DATA BY CHRIEN AND LIOU(REF 32) USED. THE  9040 1451
      SCATTERING CROSS-SECTION WAS CHANGED TO ACCOMMODATE CHANGES 9040 1451
      IN CAPTURE USING TOTAL SLIGHTLY MODIFIED FROM THAT GIVEN IN 9040 1451
      REF 2. THESE SMALL CHANGES ARE DESCRIBED IN REF 31.         9040 1451
      2.53E-02EV CROSS-SECTION   CAPTURE    7.4 B                 9040 1451
      DILUTE CAPTURE RESONANCE INTEGRAL    86.1 B                 9040 1451
      (0.5 TO 2.0E+07 EV)                                         9040 1451
 TOTAL CROSS-SECTION FROM 50KEV TO 20MEV - A.B.SMITH(ANL)(REF1)   9040 1451
      MAINLY BASED ON REF9-13 FROM 0.025 TO 1 MEV,FROM 1 TO 5MEV  9040 1451
      BASED ON REF 9-11,14-16.FROM 5 TO 15MEV EVALUATION BASED ON 9040 1451
      REF11,15,17.FROM 15 TO 20MEV EVALUATION BASED ON NUCLEAR    9040 1451
      MODEL CALCULATIONS DESCRIBED IN REF18 AND 19.TOTAL CROSS-   9040 1451
      SECTION MODIFIED BY ABOUT 3PER-CENT BETWEEN 50 TO 150KEV    9040 1451
      JOIN SMOOTHLY WITH THAT CALCULATED FROM THE UNRESOLVED      9040 1451
      RESONANCE PARAMETERS BELOW 50KEV.                           9040 1451
 ELASTIC SCATTERING CROSS-SECTION FROM 50KEV TO 20MEV OBTAINED AS 9040 1451
      THE DIFFERENCE BETWEEN TOTAL AND OTHER PARTIAL CROSSSECTIONS9040 1451
 FISSION CROSS-SECTION UP TO 20MEV - W.P.POENITZ(REF1) DETAILS OF 9040 1451
      THE EVALUATION AND THE DATA SETS USED ARE IN REF1.          9040 1451
      FISSION CROSS-SECTION CHANGED TO BE CONSISTENT WITH         9040 1451
      U-235(N,F) EVALUATION GIVEN IN REF 33                       9040 1451
 CAPTURE CROSS-SECTION 50KEV -20MEV - W.P.POENITZ. DETAILS OF THE 9040 1451
      EVALUATION AND THE DATA SETS USED ARE GIVEN IN REF1.        9040 1451
      CAPTURE CROSS-SECTION CHANGED TO BE CONSISTENT WITH         9040 1451
      U-235(N,F) EVALUATION GIVEN IN REF 33                       9040 1451
 INELASTIC SCATTERING CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF 1)  9040 1451
      EVALUATION BASED MAINLY ON THE DATA IN REF 20-25. INTER-    9040 1451
      POLATED BETWEEN EXPERIMENTAL DATA AND EXTRAPOLATED TO 20MEV 9040 1451
      USING COUPLED CHANNEL NUCLEAR MODEL CALCULATIONS OF REF26.  9040 1451
 (N,2N) AND (N,3N) CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF1)      9040 1451
      EXPERIMENTAL DATA USED(REF1) RENORMALISED TO A COMMON SET OF9040 1451
      STANDARDS BEFORE EVALUATION.                                9040 1451
                                                                  9040 1451
                              MF = 33                             9040 1451
  MT = 18,102 BASED ON ERROR ESTIMATES IN THESE CROSS SECTIONS IN 9040 1451
  REF. 32,1                                                       9040 1451
                                                                  9040 1451
                       PRINCIPAL REFERENCES                       9040 1451
                                                                  9040 1451
 1 J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH,J.F.WHALEN AND     9040 1451
   R.J.HOWERTON ANL/NDM-35(FEB.1978)                              9040 1451
 2 D.F.NEWMAN,B.R.LEONARD JR.ET.AL. EPRI NP-222,MAY1977           9040 1451
 3 G.DE SAUSSURE AND R.L.MACKLIN ORNL/TM-6161(ENDF-255)DEC.1977   9040 1451
 4 R.GWIN ORNL/TM-6245(ENDF-262)MAY1978.                          9040 1451
 5 M.C.BRADY AND T.R.ENGLAND NSE 103,129(1989)                    9040 1451
   **DATA INSERTED INTO FILE AT BNL**                             9040 1451
 6. R.SHER + C.BECK EPRI 1771/81 + REV. 1/83 + PC TO MAGURNO 2/83 9040 1451
 7 S.F.MUGHABGHAB AND D.I.GARBER,BNL-325(3RD EDN)1973             9040 1451
 8 H.DERRIEN,NEANDC(E)163U(1975)                                  9040 1451
 9 J.F.WHALEN AND A.B.SMITH,NUC.SCI.AND ENG.67,129(1978)          9040 1451
 10J.MEADOWS,A.SMITH AND J.WHALEN,PRIV.COMM.1977                  9040 1451
 11U.FASOLI ET.AL.NUCL.PHYS.A151,369(1970)                        9040 1451
 12M.DIVADEENAM ET.AL. DISS.ABS.28,3834(1968)                     9040 1451
 13C.UTTLEY ET.AL.66PARIS,1,165,1966                              9040 1451
 14L.GREEN ET.AL.PROC.CONF.ON NUCL.CROSS-SECTIONS AND TECHNOLOGY, 9040 1451
                                 KNOXVILLE,1,325(1971)            9040 1451
 15D.FOSTER ET.AL.PRIV.COMM(1967)                                 9040 1451
 16R.BATCHELOR ET.AL.NUCL.PHYS.65,236(1956)                       9040 1451
 17J.COON ET.AL.PHYS.REV.88,562(1952)                             9040 1451
 18P.GUENTHER ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(19779040 1451
 19C.PHILIS ET.AL. ANL/NDM-28(1977)                               9040 1451
 20A.SMITH PHYS.REV.126,718(1962)                                 9040 1451
 21W.MCMURRAY ET.AL.SOUTHERN UNIV.NUCLEAR INST.REPORT,SUNI-41(19759040 1451
 22J.HAOUAT ETAL.PROC.INER.CONF.ON INTERACTIONS OF NEUTRONS WITH  9040 1451
                                 NUCLEI,CONF-760715(1976)         9040 1451
 23R.BATCHELOR AND J.TOWLE,NUCL.PHYS.65,236(1965)                 9040 1451
 24R.BATCHELOR AND J.TOWLE,PROC.PHYS.SOC.73,193(1959)             9040 1451
 25A.SMITH,PRIV.COMM,1970 AND 1977                                9040 1451
 26A.SMITH ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(1977)  9040 1451
 27M.SEGEV.ET.AL.TRANS.AM.NUCL.SOC,22,679(1975)AND PRIV.COMM(1976)9040 1451
 28R.J.HOWERTON AND R.J.DOYAS,NUCL.SCI AND ENG,46,414(1971)       9040 1451
 29YU.A.VASIL'EV ET.AL.PHYSICS OF NUCLEAR FISSION,N.A.PERFILOV AND9040 1451
   V.P.EISMONT(EDS),ISRAEL PROGRAM OF SCI.TRANS.JERUSALEM(1964)   9040 1451
 30E.M.PENNINGTON,PRIVATE COMMUNICATION(1973)                     9040 1451
 31M.R.BHAT,ENDF-268(TO BE PUBLISHED)                             9040 1451
 32R.E.CHRIEN AND H.LIOU(1978) TO BE PUBLISHED                    9040 1451
 33W.P.POENITZ,ANL/NDM-45(1978) TO BE PUBLISHED                   9040 1451
 34R.W.PEELE AND F.C.MAIENSCHEIN, NUCL.SCI.ENG. 40, 485 (1970)    9040 1451
 35R.J. HOWERTON, D.E. CULLEN, R.C. HAIGHT, M.H. MACGREGOR, S.T.  9040 1451
   PERKINS, AND E.F. PLECHARTY, "THE LLL EVALUATED NUCLEAR DATA   9040 1451
   LIBRARY (ENDL): EVALUATION TECHNIQUES, REACTION INDEX, AND     9040 1451
   DESCRIPTIONS OF INDIVIDUAL EVALUATIONS," UCRL-50400, VOL. 15,  9040 1451
   PART A, LAWERENCE LIVERMORE LABORATORY (1975).                 9040 1451
 36S.T.PERKINS,R.C.HAIGHT AND R.J.HOWERTON,NUCL.SCI.ENG.57,1(1975)9040 1451
 37D.K.OLSEN, ORNL/TM-8056(1982), ENDF-319.                       9040 1451
******************************************************************9040 1451
                                                                  9040 1451
                                                                  9040 1451
                                                                  9040 1451
                                                                  9040 1451
******************************************************************9040 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9040 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9040 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9040 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9040 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9040 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9040 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9040 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9040 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9040 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9040 1451
 ***************** Program FIXUP (Version 2002-1) ****************9040 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9040 1451
 Corrected Thresholds-----------------------------------------Yes 9040 1451
 Extended Cross Sections to 20 MeV----------------------------No  9040 1451
 Allow Cross Section Deletion---------------------------------No  9040 1451
 Allow Cross Section Reconstruction---------------------------No  9040 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9040 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9040 1451
 Deleted Duplicate Points-------------------------------------Yes 9040 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9040 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9040 1451
 Allow Creation of Missing Sections---------------------------No  9040 1451
 Allow Insertion of Energy Points-----------------------------No  9040 1451
 Create Uniform Energy Grid-----------------------------------No  9040 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9040 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9040 1451
 Unshielded Group Averages Using  640 Groups                      9040 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9040 1451
                                1        451        217          19040 1451
                                3         18         72          09040 1451
                                3        102        217          09040 1451
                               33         18          9          09040 1451
                               33        102          7          09040 1451
                                                                  9040 1  0
                                                                  9040 0  0
 9.02320E+4 2.30044E+2          0          0          0          09040 3 18
 1.88470E+8 1.88470E+8          0          0          1        2069040 3 18
        206          1                                            9040 3 18
 500000.000 1.37500E-6 525000.000 4.12499E-6 550000.000 6.87498E-69040 3 18
 575000.000 9.62500E-6 600000.000 1.44501E-5 630000.000 2.13500E-59040 3 18
 660000.000 2.82500E-5 690000.000 4.12173E-5 720000.000 8.03773E-59040 3 18
 760000.000 .000130050 800000.000 .000154600 840000.000 .0002595459040 3 18
 880000.000 .000567766 920000.000 .001115025 960000.000 .0011350869040 3 18
 1000000.00 .001805998 1100000.00 .003257634 1200000.00 .0100931179040 3 18
 1300000.00 .038089744 1400000.00 .064670533 1500000.00 .0994199279040 3 18
 1600000.00 .097810452 1700000.00 .084170147 1800000.00 .0952105919040 3 18
 1900000.00 .116499192 2000000.00 .118500645 2100000.00 .1289995989040 3 18
 2200000.00 .130499193 2300000.00 .117750737 2400000.00 .1157501309040 3 18
 2500000.00 .118800036 2600000.00 .122399807 2700000.00 .1254663819040 3 18
 2800000.00 .125333781 2900000.00 .129000649 3000000.00 .1374367209040 3 18
 3100000.00 .138499460 3200000.00 .138262606 3300000.00 .1396000289040 3 18
 3400000.00 .141199830 3500000.00 .141599831 3600000.00 .1408000329040 3 18
 3700000.00 .140250187 3800000.00 .141200072 3900000.00 .1423999139040 3 18
 4000000.00 .142899857 4100000.00 .142699906 4200000.00 .1424999559040 3 18
 4300000.00 .142300003 4400000.00 .142100052 4500000.00 .1422143549040 3 18
 4600000.00 .142642911 4700000.00 .143071468 4800000.00 .1435000259040 3 18
 4900000.00 .143928582 5000000.00 .144517803 5100000.00 .1454997569040 3 18
 5200000.00 .145249710 5300000.00 .143749811 5400000.00 .1417500209040 3 18
 5500000.00 .139250337 5600000.00 .138500176 5700000.00 .1390002899040 3 18
 5800000.00 .141500578 5900000.00 .148001097 6000000.00 .1620014499040 3 18
 6100000.00 .182000830 6200000.00 .203750523 6300000.00 .2272505289040 3 18
 6400000.00 .252499823 6500000.00 .276599003 6600000.00 .2977987789040 3 18
 6700000.00 .318998553 6800000.00 .340198328 6900000.00 .3613981039040 3 18
 7000000.00 .376248272 7100000.00 .384748835 7200000.00 .3904989219040 3 18
 7300000.00 .393498528 7400000.00 .391373544 7500000.00 .3841239679040 3 18
 7600000.00 .376874391 7700000.00 .369631916 7800000.00 .3644204229040 3 18
 7900000.00 .360806729 8000000.00 .356899918 8100000.00 .3526999919040 3 18
 8200000.00 .348500063 8300000.00 .344300135 8400000.00 .3401002089040 3 18
 8500000.00 .336600256 8600000.00 .333800279 8700000.00 .3310003029040 3 18
 8800000.00 .328200325 8900000.00 .325400348 9000000.00 .3236003459040 3 18
 9100000.00 .322800318 9200000.00 .322000290 9300000.00 .3212002629040 3 18
 9400000.00 .320400235 9500000.00 .320200157 9600000.00 .3206000309040 3 18
 9700000.00 .320999903 9800000.00 .321399775 9900000.00 .3217996489040 3 18
 10000000.0 .321099623 10100000.0 .319299700 10200000.0 .3174997779040 3 18
 10300000.0 .315699855 10400000.0 .313899932 10500000.0 .3121000099040 3 18
 10600000.0 .310300086 10700000.0 .308500164 10800000.0 .3067002419040 3 18
 10900000.0 .304900318 11000000.0 .304000321 11100000.0 .3040002509040 3 18
 11200000.0 .304000179 11300000.0 .304000107 11400000.0 .3040000369040 3 18
 11500000.0 .304000033 11600000.0 .304000099 11700000.0 .3040001669040 3 18
 11800000.0 .304000232 11900000.0 .304000298 12000000.0 .3048003069040 3 18
 12100000.0 .306400257 12200000.0 .308000208 12300000.0 .3096001599040 3 18
 12400000.0 .311200110 12500000.0 .313025081 12600000.0 .3150750729040 3 18
 12700000.0 .317125064 12800000.0 .319175055 12900000.0 .3212250479040 3 18
 13000000.0 .323275038 13100000.0 .325325030 13200000.0 .3273750219040 3 18
 13300000.0 .329425013 13400000.0 .331475004 13500000.0 .3335501219040 3 18
 13600000.0 .335650362 13700000.0 .337750604 13800000.0 .3398508469040 3 18
 13900000.0 .341951087 14000000.0 .346751036 14100000.0 .3542506929040 3 18
 14200000.0 .361750348 14300000.0 .369250003 14400000.0 .3767496599040 3 18
 14500000.0 .384249315 14600000.0 .389874281 14700000.0 .3936245569040 3 18
 14800000.0 .397374831 14900000.0 .401125106 15000000.0 .4053668939040 3 18
 15100000.0 .410100194 15200000.0 .414833495 15300000.0 .4195667969040 3 18
 15400000.0 .424300097 15500000.0 .429033398 15600000.0 .4337666999040 3 18
 15700000.0 .438483313 15800000.0 .443099809 15900000.0 .4476996189040 3 18
 16000000.0 .451299505 16100000.0 .453899472 16200000.0 .4564994389040 3 18
 16300000.0 .459099404 16400000.0 .461699371 16500000.0 .4642993379040 3 18
 16600000.0 .466899304 16700000.0 .469499270 16800000.0 .4720992369040 3 18
 16900000.0 .474699203 17000000.0 .475649295 17100000.0 .4749495149040 3 18
 17200000.0 .474249732 17300000.0 .473549950 17400000.0 .4728501699040 3 18
 17500000.0 .472150387 17600000.0 .471450606 17700000.0 .4707508249040 3 18
 17800000.0 .470051042 17900000.0 .469351261 18000000.0 .4713512799040 3 18
 18100000.0 .476051098 18200000.0 .480750917 18300000.0 .4854507359040 3 18
 18400000.0 .490150554 18500000.0 .494850373 18600000.0 .4995501929040 3 18
 18700000.0 .504250010 18800000.0 .508949829 18900000.0 .5136496489040 3 18
 19000000.0 .517499576 19100000.0 .520499613 19200000.0 .5234996509040 3 18
 19300000.0 .526499687 19400000.0 .529499724 19500000.0 .5324997619040 3 18
 19600000.0 .535499798 19700000.0 .538499835 19800000.0 .5414998729040 3 18
 19900000.0 .544499909 20000000.0 0.0                             9040 3 18
                                                                  9040 3  0
 9.02320E+4 2.30044E+2          0          0          0          09040 3102
 4.78640E+6 4.78640E+6          0          0          1        6419040 3102
        641          1                                            9040 3102
 .000100000 117.717596 .000105000 114.972465 .000110000 112.3645769040 3102
 .000115000 109.961479 .000120000 107.152220 .000127500 104.0353799040 3102
 .000135000 101.187548 .000142500 98.5704503 .000150000 95.73802059040 3102
 .000160000 92.7873368 .000170000 90.1026352 .000180000 87.63829839040 3102
 .000190000 85.3454282 .000200000 83.2519344 .000210000 81.28235679040 3102
 .000220000 79.4530263 .000230000 77.7534639 .000240000 75.75853689040 3102
 .000255000 73.5659823 .000270000 71.8742409 .000280000 69.99118079040 3102
 .000300000 67.6943744 .000320000 65.6147172 .000340000 63.71014439040 3102
 .000360000 61.9559112 .000380000 60.3543582 .000400000 58.68475839040 3102
 .000425000 56.9764152 .000450000 55.4155706 .000475000 53.98450369040 3102
 .000500000 52.6386001 .000525000 51.4083118 .000550000 50.24608029040 3102
 .000575000 49.1651529 .000600000 48.0573418 .000630000 46.91773229040 3102
 .000660000 45.8748715 .000690000 44.8744625 .000720000 43.81004259040 3102
 .000760000 42.6686716 .000800000 41.6088204 .000840000 40.63725899040 3102
 .000880000 39.7138277 .000920000 38.8674685 .000960000 38.05690769040 3102
 .001000000 37.2166694 .001050000 36.3394262 .001100000 35.51694219040 3102
 .001150000 34.7590912 .001200000 33.8650807 .001275000 32.88238449040 3102
 .001350000 31.9838228 .001425000 31.1452380 .001500000 30.25449569040 3102
 .001600000 29.3230418 .001700000 28.4726712 .001800000 27.68617629040 3102
 .001900000 26.9677133 .002000000 26.3015200 .002100000 25.67635889040 3102
 .002200000 25.1039807 .002300000 24.5566911 .002400000 23.93225549040 3102
 .002550000 23.2382004 .002700000 22.6952333 .002800000 22.10554159040 3102
 .003000000 21.3795011 .003200000 20.7185473 .003400000 20.11387849040 3102
 .003600000 19.5622492 .003800000 19.0528074 .004000000 18.52179629040 3102
 .004250000 17.9828720 .004500000 17.4910479 .004750000 17.03148329040 3102
 .005000000 16.6099462 .005250000 16.2191227 .005500000 15.84959409040 3102
 .005750000 15.5105884 .006000000 15.1545074 .006300000 14.79878309040 3102
 .006600000 14.4627117 .006900000 14.1518952 .007200000 13.81281779040 3102
 .007600000 13.4492601 .008000000 13.1166977 .008400000 12.80447929040 3102
 .008800000 12.5138121 .009200000 12.2436716 .009600000 11.98660089040 3102
 .010000000 11.7204259 .010500000 11.4390181 .011000000 11.18162339040 3102
 .011500000 10.9360800 .012000000 10.6546914 .012750000 10.34322669040 3102
 .013500000 10.0546958 .014250000 9.78988762 .015000000 9.507197929040 3102
 .016000001 9.21021411 .017000001 8.93772944 .017999999 8.688957049040 3102
 .018999999 8.46027176 .020000000 8.24568794 .021000000 8.049197419040 3102
 .022000000 7.86409138 .023000000 7.69167626 .024000000 7.489465889040 3102
 .025500000 7.27213432 .027000001 7.10017272 .028000001 6.913551159040 3102
 .029999999 6.68370492 .032000002 6.47421280 .034000002 6.282990349040 3102
 .035999998 6.10855617 .037999999 5.94753456 .039999999 5.779328459040 3102
 .042500000 5.60831761 .045000002 5.45204742 .047499999 5.306473969040 3102
 .050000001 5.17257368 .052499998 5.04852697 .055000000 4.930998419040 3102
 .057500001 4.82340078 .059999999 4.71022229 .063000001 4.596858059040 3102
 .066000000 4.48999040 .068999998 4.39041832 .071999997 4.282415399040 3102
 .075999998 4.16667382 .079999998 4.06076698 .083999999 3.961125819040 3102
 .088000000 3.86836194 .092000000 3.78228946 .096000001 3.699986769040 3102
 .100000001 3.61485380 .104999997 3.52466255 .109999999 3.442182359040 3102
 .115000002 3.36336348 .119999997 3.27302265 .127499998 3.173399779040 3102
 .135000005 3.08098779 .142499998 2.99586987 .150000006 2.904812439040 3102
 .159999996 2.80887282 .170000002 2.72093103 .180000007 2.640377819040 3102
 .189999998 2.56634734 .200000003 2.49683305 .209999993 2.433374679040 3102
 .219999999 2.37339008 .230000004 2.31735697 .239999995 2.252175989040 3102
 .254999995 2.17907301 .270000011 2.12333892 .280000001 2.061599899040 3102
 .300000012 1.98578742 .319999993 1.91625736 .340000004 1.852888159040 3102
 .360000014 1.79473111 .379999995 1.74025445 .400000006 1.684892419040 3102
 .425000012 1.62861092 .449999988 1.57516393 .474999994 1.526449959040 3102
 .500000000 1.48180440 .524999976 1.43952136 .550000012 1.400689819040 3102
 .574999988 1.36377395 .600000024 1.32608723 .629999995 1.288288039040 3102
 .660000026 1.25244201 .689999998 1.21802661 .720000029 1.181535689040 3102
 .759999990 1.14277962 .800000012 1.10644514 .839999974 1.073211149040 3102
 .879999995 1.04172353 .920000017 1.01264674 .959999979 .9851716089040 3102
 1.00000000 .956038107 1.04999995 .926183156 1.10000002 .8979742459040 3102
 1.14999998 .872395836 1.20000005 .842109589 1.27499998 .8086236599040 3102
 1.35000002 .778257586 1.42499995 .749757624 1.50000000 .7199226839040 3102
 1.60000002 .688643210 1.70000005 .660468927 1.79999995 .6345953909040 3102
 1.89999998 .611022622 2.00000000 .588899851 2.09999990 .5687421669040 3102
 2.20000005 .550394153 2.29999995 .532947154 2.40000010 .5131081369040 3102
 2.54999995 .491271807 2.70000005 .474584932 2.79999995 .4565969309040 3102
 3.00000000 .434425625 3.20000005 .415050727 3.40000010 .3974006379040 3102
 3.59999990 .381584534 3.79999995 .367323626 4.00000000 .3527724399040 3102
 4.25000000 .338295876 4.50000000 .325110757 4.75000000 .3133694609040 3102
 5.00000000 .303267228 5.25000000 .293953576 5.50000000 .2855665319040 3102
 5.75000000 .277770758 6.00000000 .270102388 6.30000019 .2625471839040 3102
 6.59999990 .255929492 6.90000010 .249813290 7.19999981 .2436649449040 3102
 7.59999990 .238586990 8.00000000 .503553581 8.39999962 .2863362769040 3102
 8.80000019 .223111173 9.19999981 .218540561 9.60000038 .2151443099040 3102
 10.0000000 .212235602 10.5000000 .209717999 11.0000000 .2080073389040 3102
 11.5000000 .207247833 12.0000000 .207791935 12.7500000 .2951547429040 3102
 13.5000000 .214394588 14.2500000 .221544185 15.0000000 .2361351519040 3102
 16.0000000 .264004204 17.0000000 .314148900 18.0000000 .4122856149040 3102
 19.0000000 .647190716 20.0000000 1.58908644 21.0000000 348.6497699040 3102
 22.0000000 13.7818597 23.0000000 580.272945 24.0000000 2.930193819040 3102
 25.5000000 .555913332 27.0000000 .309736817 28.0000000 .2130329379040 3102
 30.0000000 .157303513 32.0000000 .132189787 34.0000000 .1188181339040 3102
 36.0000000 .164088285 38.0000000 .135759158 40.0000000 .1265483069040 3102
 42.5000000 .102371570 45.0000000 .157196272 47.5000000 .1233692639040 3102
 50.0000000 .122941177 52.5000000 .179861901 55.0000000 .2502567379040 3102
 57.5000000 90.2939115 60.0000000 .912830407 63.0000000 .6321927599040 3102
 66.0000000 24.3366888 69.0000000 271.925256 72.0000000 .5815312739040 3102
 76.0000000 .210880950 80.0000000 .142075695 84.0000000 .1178710979040 3102
 88.0000000 .177175958 92.0000000 .103142449 96.0000000 .1451483209040 3102
 100.000000 .156948948 105.000000 .177049679 110.000000 61.35374129040 3102
 115.000000 .833155910 120.000000 52.4531072 127.500000 12.61246379040 3102
 135.000000 .096591513 142.500000 .464631869 150.000000 .6322294319040 3102
 160.000000 2.92064219 170.000000 40.5825351 180.000000 .1441355559040 3102
 190.000000 34.1464386 200.000000 .313053176 210.000000 .3491740809040 3102
 220.000000 25.1147614 230.000000 .154139326 240.000000 15.73711019040 3102
 255.000000 11.5444632 270.000000 .185698563 280.000000 9.542452139040 3102
 300.000000 8.64438446 320.000000 12.5427253 340.000000 8.447136309040 3102
 360.000000 14.2073100 380.000000 .242211404 400.000000 3.519085549040 3102
 425.000000 .087542524 450.000000 6.86556759 475.000000 5.928064139040 3102
 500.000000 1.28130689 525.000000 3.35649766 550.000000 4.146925839040 3102
 575.000000 2.75726884 600.000000 1.01066055 630.000000 3.533309959040 3102
 660.000000 10.4884715 690.000000 4.60324876 720.000000 3.198388619040 3102
 760.000000 1.35002046 800.000000 3.20385549 840.000000 3.029109269040 3102
 880.000000 2.14459626 920.000000 1.92581674 960.000000 4.248621159040 3102
 1000.00000 2.48660615 1050.00000 1.22423539 1100.00000 2.257732309040 3102
 1150.00000 1.35929281 1200.00000 2.92156323 1275.00000 1.828611139040 3102
 1350.00000 2.82915468 1425.00000 1.66774418 1500.00000 2.565043429040 3102
 1600.00000 2.50133524 1700.00000 1.68617223 1800.00000 2.282139959040 3102
 1900.00000 2.12353389 2000.00000 1.37174720 2100.00000 2.056500689040 3102
 2200.00000 1.83149119 2300.00000 1.65067293 2400.00000 1.467774939040 3102
 2550.00000 1.84524578 2700.00000 1.83833978 2800.00000 1.384372559040 3102
 3000.00000 1.53838922 3200.00000 1.35952475 3400.00000 .8649848279040 3102
 3600.00000 1.15591066 3800.00000 1.40748008 4000.00000 1.111035839040 3102
 4250.00000 1.07128158 4500.00000 1.03449171 4750.00000 1.000201769040 3102
 5000.00000 .973593704 5250.00000 .953520883 5500.00000 .9350279759040 3102
 5750.00000 .918114982 6000.00000 .900391280 6300.00000 .8819826279040 3102
 6600.00000 .865208821 6900.00000 .849255855 7200.00000 .8319234349040 3102
 7600.00000 .813541191 8000.00000 .796481488 8400.00000 .7805729709040 3102
 8800.00000 .765755262 9200.00000 .751972679 9600.00000 .7383556069040 3102
 10000.0000 .724537391 10500.0000 .710518034 11000.0000 .6972472909040 3102
 11500.0000 .684725159 12000.0000 .670198070 12750.0000 .6541666739040 3102
 13500.0000 .639803940 14250.0000 .626106419 15000.0000 .6108198579040 3102
 16000.0000 .594277804 17000.0000 .579237312 18000.0000 .5651981659040 3102
 19000.0000 .552160936 20000.0000 .540125394 21000.0000 .5281135479040 3102
 22000.0000 .517106375 23000.0000 .506579531 24000.0000 .4942983439040 3102
 25500.0000 .480763374 27000.0000 .469986373 28000.0000 .4597073769040 3102
 30000.0000 .448673655 32000.0000 .439883360 34000.0000 .4323360589040 3102
 36000.0000 .427078591 38000.0000 .422812572 40000.0000 .4208046729040 3102
 42500.0000 .418363187 45000.0000 .410433795 47500.0000 .3974149129040 3102
 50000.0000 .382527128 52500.0000 .367607866 55000.0000 .3526886039040 3102
 57500.0000 .337769341 60000.0000 .324008124 63000.0000 .3119999999040 3102
 66000.0000 .299999999 69000.0000 .288736879 72000.0000 .2780678019040 3102
 76000.0000 .266022623 80000.0000 .256066783 84000.0000 .2480740989040 3102
 88000.0000 .240299624 92000.0000 .234074339 96000.0000 .2281518469040 3102
 100000.000 .222706480 105000.000 .218100000 110000.000 .2135000009040 3102
 115000.000 .208900000 120000.000 .203365159 127500.000 .1984077539040 3102
 135000.000 .193825772 142500.000 .189258045 150000.000 .1852126519040 3102
 160000.000 .181541436 170000.000 .177744490 180000.000 .1728635829040 3102
 190000.000 .167733061 200000.000 .163506608 210000.000 .1605000009040 3102
 220000.000 .157500000 230000.000 .154500000 240000.000 .1510044389040 3102
 255000.000 .148884890 270000.000 .147521763 280000.000 .1460064759040 3102
 300000.000 .144700087 320000.000 .144100124 340000.000 .1435002079040 3102
 360000.000 .142900291 380000.000 .142300374 400000.000 .1430003639040 3102
 425000.000 .145000258 450000.000 .147000153 475000.000 .1490017709040 3102
 500000.000 .151942351 525000.000 .155672651 550000.000 .1594029529040 3102
 575000.000 .163122586 600000.000 .164940353 630000.000 .1650000009040 3102
 660000.000 .165000000 690000.000 .164661912 720000.000 .1629724359040 3102
 760000.000 .160945120 800000.000 .157571511 840000.000 .1527142539040 3102
 880000.000 .147855060 920000.000 .142800070 960000.000 .1376001919040 3102
 1000000.00 .130600244 1100000.00 .121800229 1200000.00 .1134002049040 3102
 1300000.00 .107400117 1400000.00 .101800020 1500000.00 .0959999459040 3102
 1600000.00 .089999894 1700000.00 .083749872 1800000.00 .0760000279040 3102
 1900000.00 .068000212 2000000.00 .061800293 2100000.00 .0574002709040 3102
 2200000.00 .053000247 2300000.00 .048600224 2400000.00 .0442002019040 3102
 2500000.00 .040800183 2600000.00 .038400170 2700000.00 .0360001579040 3102
 2800000.00 .033600144 2900000.00 .031200131 3000000.00 .0294001229040 3102
 3100000.00 .028200118 3200000.00 .027000114 3300000.00 .0258001109040 3102
 3400000.00 .024600106 3500000.00 .023400102 3600000.00 .0222000989040 3102
 3700000.00 .021000094 3800000.00 .019800090 3900000.00 .0186000869040 3102
 4000000.00 .017750083 4100000.00 .017250081 4200000.00 .0167500799040 3102
 4300000.00 .016250077 4400000.00 .015750075 4500000.00 .0152500729040 3102
 4600000.00 .014750070 4700000.00 .014250068 4800000.00 .0137500669040 3102
 4900000.00 .013250064 5000000.00 .012750062 5100000.00 .0122500609040 3102
 5200000.00 .011750058 5300000.00 .011250056 5400000.00 .0107500549040 3102
 5500000.00 .010250051 5600000.00 .009750049 5700000.00 .0092500479040 3102
 5800000.00 .008750045 5900000.00 .008250043 6000000.00 .0078950419040 3102
 6100000.00 .007685040 6200000.00 .007475039 6300000.00 .0072650389040 3102
 6400000.00 .007055037 6500000.00 .006845036 6600000.00 .0066350359040 3102
 6700000.00 .006425034 6800000.00 .006215033 6900000.00 .0060050329040 3102
 7000000.00 .005795030 7100000.00 .005585029 7200000.00 .0053750289040 3102
 7300000.00 .005165027 7400000.00 .004955026 7500000.00 .0047450259040 3102
 7600000.00 .004535024 7700000.00 .004325023 7800000.00 .0041150229040 3102
 7900000.00 .003905021 8000000.00 .003755020 8100000.00 .0036650209040 3102
 8200000.00 .003575019 8300000.00 .003485019 8400000.00 .0033950199040 3102
 8500000.00 .003305019 8600000.00 .003215018 8700000.00 .0031250189040 3102
 8800000.00 .003035018 8900000.00 .002945018 9000000.00 .0028550179040 3102
 9100000.00 .002765017 9200000.00 .002675017 9300000.00 .0025850179040 3102
 9400000.00 .002495016 9500000.00 .002405016 9600000.00 .0023150169040 3102
 9700000.00 .002225016 9800000.00 .002135015 9900000.00 .0020450159040 3102
 10000000.0 .001993765 10100000.0 .001981265 10200000.0 .0019687649040 3102
 10300000.0 .001956264 10400000.0 .001943764 10500000.0 .0019312639040 3102
 10600000.0 .001918763 10700000.0 .001906263 10800000.0 .0018937629040 3102
 10900000.0 .001881262 11000000.0 .001868762 11100000.0 .0018562619040 3102
 11200000.0 .001843761 11300000.0 .001831261 11400000.0 .0018187609040 3102
 11500000.0 .001806260 11600000.0 .001793760 11700000.0 .0017812599040 3102
 11800000.0 .001768759 11900000.0 .001756259 12000000.0 .0017437589040 3102
 12100000.0 .001731258 12200000.0 .001718758 12300000.0 .0017062579040 3102
 12400000.0 .001693757 12500000.0 .001681257 12600000.0 .0016687569040 3102
 12700000.0 .001656256 12800000.0 .001643756 12900000.0 .0016312559040 3102
 13000000.0 .001618755 13100000.0 .001606255 13200000.0 .0015937549040 3102
 13300000.0 .001581254 13400000.0 .001568754 13500000.0 .0015562539040 3102
 13600000.0 .001543753 13700000.0 .001531253 13800000.0 .0015187529040 3102
 13900000.0 .001506252 14000000.0 .001499169 14100000.0 .0014975029040 3102
 14200000.0 .001495835 14300000.0 .001494169 14400000.0 .0014925029040 3102
 14500000.0 .001490835 14600000.0 .001489169 14700000.0 .0014875029040 3102
 14800000.0 .001485835 14900000.0 .001484169 15000000.0 .0014825029040 3102
 15100000.0 .001480835 15200000.0 .001479169 15300000.0 .0014775029040 3102
 15400000.0 .001475835 15500000.0 .001474169 15600000.0 .0014725029040 3102
 15700000.0 .001470835 15800000.0 .001469169 15900000.0 .0014675029040 3102
 16000000.0 .001465835 16100000.0 .001464169 16200000.0 .0014625029040 3102
 16300000.0 .001460835 16400000.0 .001459169 16500000.0 .0014575029040 3102
 16600000.0 .001455835 16700000.0 .001454169 16800000.0 .0014525029040 3102
 16900000.0 .001450835 17000000.0 .001449169 17100000.0 .0014475029040 3102
 17200000.0 .001445835 17300000.0 .001444169 17400000.0 .0014425029040 3102
 17500000.0 .001440835 17600000.0 .001439169 17700000.0 .0014375029040 3102
 17800000.0 .001435835 17900000.0 .001434169 18000000.0 .0014325029040 3102
 18100000.0 .001430835 18200000.0 .001429169 18300000.0 .0014275029040 3102
 18400000.0 .001425835 18500000.0 .001424169 18600000.0 .0014225029040 3102
 18700000.0 .001420835 18800000.0 .001419169 18900000.0 .0014175029040 3102
 19000000.0 .001415835 19100000.0 .001414169 19200000.0 .0014125029040 3102
 19300000.0 .001410835 19400000.0 .001409169 19500000.0 .0014075029040 3102
 19600000.0 .001405835 19700000.0 .001404169 19800000.0 .0014025029040 3102
 19900000.0 .001400835 20000000.0 0.0                             9040 3102
                                                                  9040 3  0
                                                                  9040 0  0
 9.02320E+4 2.30044E+2          0          0          0          1904033 18
 0.000000+0 0.000000+0          0         18          0          2904033 18
 0.000000+0 0.000000+0          0          1          6          3904033 18
 1.000000-5 0.000000+0 5.000000+5 1.600000-3 2.000000+7 0.000000+0904033 18
 0.000000+0 0.000000+0          0          1         24         12904033 18
 1.000000-5 0.000000+0 5.000000+5 1.024000-1 8.000000+5 1.440000-2904033 18
 9.000000+5 1.000000-2 1.000000+6 4.900000-3 1.200000+6 3.600000-3904033 18
 1.400000+6 1.600000-3 1.600000+6 3.600000-3 3.000000+6 1.600000-3904033 18
 4.000000+6 3.600000-3 6.000000+6 1.000000-2 2.000000+7 0.000000+0904033 18
                                                                  904033  0
 9.02320E+4 2.30044E+2          0          0          0          1904033102
 0.000000+0 0.000000+0          0        102          0          2904033102
 0.000000+0 0.000000+0          0          1          6          3904033102
 1.000000-5 0.000000+0 1.500000+1 2.500000-3 2.000000+7 0.000000+0904033102
 0.000000+0 0.000000+0          0          1         10          5904033102
 1.000000-5 2.000000-4 3.000000-2 6.400000-3 1.500000+1 1.000000-2904033102
 1.000000+6 4.000000-2 2.000000+7 0.000000+0                      904033102
                                                                  904033  0
                                                                  9040 0  0
                                                                     0 0  0
 9.22350E+4 2.33024E+2          0          1         34         109228 1451
 0.0        1.00000E+0          0          0          0          69228 1451
 1.00000E+0 2.00000E+7          0          0         10       20029228 1451
 3.00000E+2 0.0                 1          0        414          39228 1451
 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ   9228 1451
                      DIST-Feb2004                                9228 1451
----IRDF-2002         MATERIAL 9228                               9228 1451
-----INCIDENT NEUTRON DATA                                        9228 1451
------ENDF-6 FORMAT                                               9228 1451
 **************************************************************** 9228 1451
 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ   9228 1451
                      DIST-NOV98 REV5-OCT97            19981109   9228 1451
----ENDF/B-VI         MATERIAL 9228         REVISION 5            9228 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9228 1451
DOSIMETRY                                                         9228 1451
******************************************************************9228 1451
                                                                  9228 1451
 ENDF/B-VI MOD 6 Revision, October 1997, L.C. Leal, H. Derrien,   9228 1451
                 N.M. Larson, and R.Q. Wright (ORNL)              9228 1451
                                                                  9228 1451
 1. New resonance parameter analysis; File 2, MT=151.             9228 1451
 2. Total and prompt-nubar revised below 2 eV; File 1, MT=451,    9228 1451
    452,456.                                                      9228 1451
                                                                  9228 1451
 ---------------------------------------------------------------- 9228 1451
 File 2                                                           9228 1451
   MT=151  Resonance parameters, from a new analysis by Leal      9228 1451
     et al. [LE97], using the multilevel R-matrix analysis code   9228 1451
     SAMMY [LA96].  Energy range for U235 is 0 to 2.25 keV.       9228 1451
     For the first time, integral data were fitted during the     9228 1451
       analysis process:  Thermal cross sections (fission,        9228 1451
       capture, and elastic), Westcott g-factors (fission and     9228 1451
       absorption) are from the ENDF/B-6 standards [CA93], and    9228 1451
       the K1 value is from Hardy [HA79].                         9228 1451
     Thermal parameters obtained in the present evaluation,       9228 1451
       first using the microscopic experimental data only, and    9228 1451
       second including the integral data as well, are compared   9228 1451
       to the SAMMY input in the following Table:                 9228 1451
                                                                  9228 1451
       Parameter        SAMMY input       Fit to    Fit to diff.  9228 1451
                          value          diff data    & integ.    9228 1451
                                          alone         data      9228 1451
       -----------    -----------------  --------   ------------  9228 1451
       Fission        584.25 +/- 1.11    582.28      584.88       9228 1451
       Capture         98.96 +/- 0.74     99.18       98.66       9228 1451
       Scattering      15.46 +/- 1.06     15.44       15.12       9228 1451
       Westcott gf    0.9771 +/- 0.0008    0.9743      0.9764     9228 1451
       Westcott ga    0.9790 +/- 0.0008    0.9774      0.9785     9228 1451
       Westcott gg                         0.9956      0.9910     9228 1451
       K1(barn)       722.70 +/- 3.90    717.48      722.43       9228 1451
                                                                  9228 1451
     The final adjustment of nu by SAMMY to the recommended K1    9228 1451
       value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission   9228 1451
       and absorption cross sections calculated from the final    9228 1451
       resonance parameters.                                      9228 1451
     In the following Tables, the fission and capture cross       9228 1451
       sections calculated in this evaluation with the code       9228 1451
       SAMMY are compared with experimental data.                 9228 1451
                                                                  9228 1451
       Experimental and calculated total cross sections.          9228 1451
                                                                  9228 1451
       Energy Range    Calculated  Schrack   Weston   Weston      9228 1451
           (eV)          (b.eV)     (b.eV)   (b.eV)   (b.eV)      9228 1451
      ---------------  ----------  -------   ------   ------      9228 1451
         0.5 -   20.0    910.4      929.9                         9228 1451
        20.0 -   60.0   1867.8     1882.8    1869.9               9228 1451
        60.0 -  100.0    954.0      968.0     954.2               9228 1451
       100.0 -  200.0   2032.7     2092.7    2089.5   2073.9      9228 1451
       200.0 -  300.0   2062.2     2007.0    2060.0   2054.6      9228 1451
       300.0 -  400.0   1280.8     1321.6    1297.1   1292.9      9228 1451
       400.0 -  500.0   1333.2     1391.5    1351.8   1347.9      9228 1451
       500.0 -  600.0   1489.2     1467.9    1499.2   1494.3      9228 1451
       600.0 -  700.0   1126.6     1156.4    1134.1   1132.6      9228 1451
       700.0 -  800.0   1088.7     1085.8    1093.3   1075.7      9228 1451
       800.0 -  900.0    797.6      784.0     813.0    804.9      9228 1451
       900.0 - 1000.0    724.4      723.9     738.2    721.4      9228 1451
      1000.0 - 2000.0   7036.1                        7054.2      9228 1451
                                                                  9228 1451
       Experimental and calculated capture cross sections.        9228 1451
                                                                  9228 1451
       Energy Range     Calculated   De Saussure     Perez        9228 1451
           (eV)           (b.eV)       (b.eV)       (b.eV)        9228 1451
      ---------------   ----------   -----------    ------        9228 1451
        0.5  -   20.0      653.5         647                      9228 1451
       20.0  -   60.0     1066.1        1084         1057         9228 1451
       60.0  -  100.0      490.2         477          504         9228 1451
       100.0 -  200.0     1158.8        1148         1138         9228 1451
       200.0 -  300.0      907.8         904          940         9228 1451
       300.0 -  400.0      660.2         658          642         9228 1451
       400.0 -  500.0      495.9         506          478         9228 1451
       500.0 -  600.0      533.3         506          562         9228 1451
       600.0 -  700.0      494.8         481          449         9228 1451
       700.0 -  800.0      490.1         513          475         9228 1451
       800.0 -  900.0      439.8         444          397         9228 1451
       900.0 - 1000.0      504.2         542          482         9228 1451
      1000.0 - 1100.0      509.6         522          463         9228 1451
      1100.0 - 1200.0      413.7         395          332         9228 1451
      1200.0 - 1300.0      340.4         372          267         9228 1451
      1300.0 - 1400.0      304.1         304          225         9228 1451
      1400.0 - 1500.0      355.7         301          254         9228 1451
      ---------------   ----------   -----------    ------        9228 1451
        20.0 - 1500.0     9164.7        9046         8665         9228 1451
                                                                  9228 1451
    The fission and capture resonance integral calculated from    9228 1451
 the present evaluation are 276.04 b and 140.49 b, respectively,  9228 1451
 giving a capture-to-fission ratio (alpha value) of 0.509 in      9228 1451
 excellent agreement with the value obtained from integral        9228 1451
 measurements.                                                    9228 1451
    The following energy-differential data were included in the   9228 1451
 analysis:                                                        9228 1451
  (1) Transmission data of Harvey et al. [HA86] on the ORELA      9228 1451
      18-meter flight path, with sample thickness of 0.03269      9228 1451
      atoms/barn, cooled to 77 K (0.4 to 68 eV).                  9228 1451
  (2) Transmission data of Harvey et al. [HA86] on the ORELA      9228 1451
      80-meter flight path, with sample thickness of 0.00233      9228 1451
      atoms/barn, cooled to 77 K (4 to 2250 eV).                  9228 1451
  (3) Transmission data of Harvey et al. [HA86] on the ORELA      9228 1451
      80-meter flight path, with sample thickness of 0.03269      9228 1451
      atoms/barn, cooled to 77 K (4 to 2250 eV).                  9228 1451
  (4) Fission data of Schrack [SC88] on the RPI Linac at 8.4      9228 1451
      meter flight path (0.02 to 20 eV).                          9228 1451
 (5,6) Fission and capture data of de Saussure et al. [DE67]      9228 1451
      on the ORELA 25.2-meter flight path (0.01 to 2250 eV).      9228 1451
 (7,8) Fission and capture data of Perez et al. [PE73] on the     9228 1451
      ORELA 39-meter flight path (0.01 to 100 eV).                9228 1451
  (9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter  9228 1451
      flight path (0.01 to 20 eV).                                9228 1451
 (10) Transmission data of Spencer et al. [SP84] on the ORELA     9228 1451
     ORELA 18-meter flight path, sample thickness of 0.001468     9228 1451
     atom/barn (0.01 to 1.0 eV).                                  9228 1451
 (11) Fission data of Wagemans et al. [WA88] on the Geel 18-      9228 1451
      meter flight path (0.001 to 1.0 eV)                         9228 1451
 (12,13) Absorption and fission data of Gwin [GW96] at ORELA      9228 1451
      (0.01 to 4.0 eV).                                           9228 1451
 (14) Fission data of Weston and Todd [WE84] on the ORELA         9228 1451
      18.9-meter flight path (14 to 2250 eV).                     9228 1451
 (15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to    9228 1451
      1.0 eV).                                                    9228 1451
 (16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to      9228 1451
      0.15 eV).                                                   9228 1451
 (17) Fission data of Weston and Todd [WE92] on the ORELA         9228 1451
      86.5-meter flight path (100 to 2000 eV).                    9228 1451
 (18) Fission yield data of Moxon et al. [MO92] at ORELA          9228 1451
      (0.01 to 50.0 eV).                                          9228 1451
                                                                  9228 1451
 ---------------------------------------------------------------- 9228 1451
 REFERENCES FOR RESOLVED RESONANCE REGION                         9228 1451
                                                                  9228 1451
 [CA93] A. Carlson, W.P. Poenitz, G.M. Hale et al., "The ENDF/B-6 9228 1451
    Neutron Cross Section Measurements Standards," National       9228 1451
    Institute of Standards and Technology report NISTIR-5177      9228 1451
    (1993)                                                        9228 1451
 [DE67] G. de Saussure, R. Gwin, L.W. Weston, and R.W. Ingle,     9228 1451
    "Simultaneous Measurements of the Neutron Fission and Capture 9228 1451
    Cross Section for 235U for Incident Neutron Energy from       9228 1451
    0.04 eV to 3 keV," Oak Ridge National Laboratory report       9228 1451
    ORNL/TM-1804 (1967)                                           9228 1451
 [GW84] R. Gwin, R.R. Spencer, R.W. Ingle, J.H. Todd, and S.W.    9228 1451
    Scoles, Nuc.Sci.Eng. 88, 37 (1984)                            9228 1451
 [GW96] R. Gwin, To be published in Nuclear Science Engineering   9228 1451
 [HA79] J. Hardy, Brookhaven National Laboratory, report          9228 1451
    BNL-NCS-51123 [ENDF-300] (1979) Sec. B.1                      9228 1451
 [HA86] J.A. Harvey, N.W. Hill, F.G. Perey et al., Nuclear Data   9228 1451
    for Science and Technology, Proc. Int. Conf. May 30-June 3,   9228 1451
    1988, Mito, Japan. (Saikon Publishing, 1988) p. 115           9228 1451
 [LA96] N.M. Larson, "Updated Users' Guide to SAMMY" report       9228 1451
    ORNL/TM-9179/R3 (1996)                                        9228 1451
 [LE97] L.C. Leal, H. Derrien, N.M. Larson, R.Q. Wright,          9228 1451
    "R-Matrix Analysis of 235U Neutron Transmission and Cross     9228 1451
    Sections in the Energy Range 0 eV to 2.25 keV," Oak Ridge     9228 1451
    National Laboratory report ORNL/TM-13516 (1997).              9228 1451
 [MO92] M.C. Moxon, J.A. Harvey, and N.W. Hill, private           9228 1451
    communication, Oak Ridge National Laboratory (1992).          9228 1451
 [PE73] R.B. Perez, G. de Saussure, and E.G. Silver, Nucl.Sci.    9228 1451
    Eng. 52, 46 (1973)                                            9228 1451
 [SC88] R.A. Schrack, "Measurement of the 235U(n,f) Reaction from 9228 1451
    Thermal to 1 keV," Nuclear Data for Science and Technology,   9228 1451
    Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon           9228 1451
    Publishing, 1988) p. 101                                      9228 1451
 [SP84] R.R. Spencer, J.A. Harvey, N.W. Hill, and L. Weston,      9228 1451
    Nucl.Sci.Eng. 96, 318 (1987)                                  9228 1451
 [WA87] J.A. Wartena, H. Weigmann, and C. Burkholz, report  IAEA  9228 1451
    Tecdoc 491 (1987) p.123                                       9228 1451
 [WA88] C. Wagemans, P. Schillebeeckx, A.J. Deruyter, and R.      9228 1451
    Barthelemy, "Subthermal Fission Cross Section Measurements    9228 1451
    for 233U and 239Pu," Nuclear Data for Science and Technology, 9228 1451
    Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon           9228 1451
    Publishing, 1988) p. 91                                       9228 1451
 [WE84] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 88, 567 (1984)   9228 1451
 [WE90] H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach,    9228 1451
    and J.A. Wartena, The Physics of Reactors, Proc. Int.  Conf., 9228 1451
    Marseille, 1990, Vol.1 (1990) p. 133                          9228 1451
 [WE92] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992)  9228 1451
                                                                  9228 1451
 **************************************************************** 9228 1451
                                                                  9228 1451
 ENDF/B-VI MOD 5 Revision 4 (R.Q.Wright, ORNL, V. McLane, NNDC,   9228 1451
                 October, 1996)                                   9228 1451
                                                                  9228 1451
 File 1, MT=451: Update of comments.                              9228 1451
 File 1, MT=456: Update of prompt nubar.                          9228 1451
                                                                  9228 1451
 *****************************************************************9228 1451
                                                                  9228 1451
 ENDF/B-VI MOD 4 Revision, February 1995, C. Lubitz (KAPL)        9228 1451
                                                                  9228 1451
 The resonance parameters below 900 eV were adjusted as described 9228 1451
 in C.R. Lubitz, "A modification to ENDF6 U235 to increase        9228 1451
    epithermal alpha and K1", Proceedings of the Int. Conf. on    9228 1451
    Nuclear Data for Science and Technology, Gatlinburg, TN, May  9228 1451
    9-13, Vol. 2 (American Nuclear Soc., 1994) page 646.          9228 1451
                                                                  9228 1451
 *****************************************************************9228 1451
                                                                  9228 1451
 ENDF/B-VI MOD 3 Revision 2                                       9228 1451
                                                                  9228 1451
 Missing File 1 section 458 added.                                9228 1451
                                                                  9228 1451
 *****************************************************************9228 1451
                                                                  9228 1451
 ENDF/B-VI MOD 2 Revision, L. Weston (ORNL)                       9228 1451
                                                                  9228 1451
 FILE 1: The uncertainties on the fission cross section           9228 1451
 recommended by the Standards Committee of CSEWG are listed. The  9228 1451
 updated description of File 2, the resolved resonance parameters,9228 1451
 is given.                                                        9228 1451
                                                                  9228 1451
 FILE 2: The resolved resonance parameters have been changed      9228 1451
 extensively. The neutron energy region below 4 eV has been made  9228 1451
 a separate group. Eta decreases with decreasing neutron energy   9228 1451
 below 0.1 eV. At the higher neutron energies, the resonance      9228 1451
 parameters are more refined.                                     9228 1451
                                                                  9228 1451
 FILE 3: Smooth cross sections above 100 keV have minor           9228 1451
 corrections.                                                     9228 1451
                                                                  9228 1451
 FILE 31: Nubar covariance files were updated.                    9228 1451
                                                                  9228 1451
 FILE 33: All covariance files were removed as correct new files  9228 1451
 are not yet available.                                           9228 1451
 *****************************************************************9228 1451
                                                                  9228 1451
 ENDF/B-VI, MOD 1 Evaluation, April 1989                          9228 1451
                                                                  9228 1451
 Principal evaluators:                                            9228 1451
  Thermal parameters: Standards Committee of CSEWG.               9228 1451
  Resolved resonance region: (0-2250 eV) L.C. Leal (U.Tenn.),     9228 1451
    G. deSaussure (ORNL), R.B. Perez (U.Tenn.), N.M. Larson       9228 1451
    (ORNL), M.S.Moore (LANL), and R.Q. Wright (ORNL).             9228 1451
    Replaced by revision 6.                                       9228 1451
  Unresolved resonance region and File 3 below 100 keV and the    9228 1451
    capture cross section above 100 keV: W.P. Poenitz (ANL) and   9228 1451
    L.W. Weston (ORNL)                                            9228 1451
  Fission cross section above 100 keV: W.P. Poenitz (ANL) and the 9228 1451
    Standards Committee of CSEWG.                                 9228 1451
  Model calculations and fits above 100 keV: P.G. Young (LANL),   9228 1451
    R.E. MacFarlane (LANL), and E.D. Arthur (LANL).               9228 1451
  Covariance files: R.W. Peelle (ORNL).  All files reoved in      9228 1451
    revision 1, except MT=456.                                    9228 1451
                                                                  9228 1451
 File 1 Descriptive and Nubar Information ----------------------- 9228 1451
   MT=452 Total nubar.  Replaced in revision 6.                   9228 1451
   MT=455 Delayed neutron yields from England [EN89].             9228 1451
   MT=456 Prompt neutron yields. Replaced in revision 6.          9228 1451
                                                                  9228 1451
 FISSION CROSS SECTION UNCERTAINTIES:  The CSEWG Standards        9228 1451
 Committee supplied the following uncertainties with the these    9228 1451
 statements.  "These uncertainties are estimates such that if a   9228 1451
 modern day experiment were performed today on a given standard   9228 1451
 using the best techniques, those results should fall within these9228 1451
 expanded uncertainties (2/3 of the time).  They take into account9228 1451
 inconsistencies and concerns about R-matrix parameters.  Note    9228 1451
 that it is not assumed that the uncertainties are totally        9228 1451
 correlated within the energy ranges given."                      9228 1451
                                                                  9228 1451
          Energy (keV)          Estimated         Comb Result (%) 9228 1451
                             Uncertainty (%)                      9228 1451
           2.53E-05                0.2                (0.19)      9228 1451
           150-600                 1.5                            9228 1451
           600-1000                1.6                (0.60)      9228 1451
           1000-3000               1.8                            9228 1451
           3000-6000               2.3                (0.69)      9228 1451
           6000-10000              2.2                            9228 1451
           10000-12000             1.8                (1.14)      9228 1451
           12000-14000             1.2                            9228 1451
           14000-14500             0.8                (0.55)      9228 1451
           14500-15000             1.5                            9228 1451
           15000-16000             2.0                (0.97)      9228 1451
           16000-17000             2.5                            9228 1451
           17000-19000             3.0                (1.26)      9228 1451
           19000-20000             4.0                            9228 1451
                                                                  9228 1451
 File 2.  RESOLVED AND UNRESOLVED RESONANCE PARAMETERS ---------- 9228 1451
   MT=151                                                         9228 1451
     Resolved resonance region: replaced in revision 6.           9228 1451
     Unresolved resonance region:                                 9228 1451
        The unresolved resonance region was derived by a FITACS   9228 1451
     (Fritz Froehner code) fit by L.W. Weston to the Standards    9228 1451
     Committee recommendation for the fission cross section and   9228 1451
     new capture evaluation based on newer alpha measurements     9228 1451
     (see ANL-83-4 supplement).  These results were then fit      9228 1451
     with URES (Ed Pennington code) so ENDF would reproduce.      9228 1451
     unresolved resonance region extends from 2.25 to 25 keV      9228 1451
     and is used only for self shielding calculations.  Dilute    9228 1451
     cross sections are taken from File 3 which shows             9228 1451
     experimentally observed structure carried over from          9228 1451
     version 5 up to 100 keV.                                     9228 1451
                                                                  9228 1451
 File 3. SMOOTH CROSS SECTIONS ---------------------------------- 9228 1451
    Model calculations from LANL: P.G. Young, R.E.MacFarlane,     9228 1451
 E.D.Arthur.  The evaluation above 100 keV is based on a detailed 9228 1451
 theoretical analysis utilizing the available experimental data.  9228 1451
 Coupled channel optical model calculations with the ECIS code    9228 1451
 [Ra70] were used to provide the total, elastic, and inelastic    9228 1451
 cross sections to the first 3 members of the ground state        9228 1451
 rotational band, as well as neutron elastic and inelastic        9228 1451
 angular distributions to the rotational levels.  The ECIS code   9228 1451
 was also used to calculate neutron transmission coefficients.    9228 1451
 Hauser-Feshbach statistical theory calculations were carried out 9228 1451
 with the GNASH [Ar88,Yo77] and COMNUC [Du70] code systems,       9228 1451
 including preequilibrium and fission.  DWBA calculations were    9228 1451
 performed with the DWUCK code [Ku70] for several vibrational     9228 1451
 levels, using B(El) values inferred from (d,d') data on U234,    9228 1451
 U235, U238, as well as Coulomb excitation measurements.  A weak  9228 1451
 coupling model [Pe69] was used to apply the U234 and U238        9228 1451
 results to states in U235.                                       9228 1451
    A preliminary description of the analysis was given at the    9228 1451
 Mito conference [Yo88].                                          9228 1451
                                                                  9228 1451
 MT=1   SUM OF PARTIAL CROSS SECTIONS FROM 2.25 TO 100 KEV.       9228 1451
   0.10 to 20 Mev, obtained from a covariance analysis of         9228 1451
   available experimental data, using an initial or prior cross   9228 1451
   section from the coupled-channel optical model analysis.       9228 1451
   Experimental data used include [Fo71], [Ve80], [Bo72], [Po81], 9228 1451
   [Gr73], [Sc74], [Po83], [Pe60], [Wh65], [Ca73], and [Br58].    9228 1451
   The GLUCS code was used for analysis [He80].                   9228 1451
 MT=2   UNCHANGED FROM VERSION 5 FROM 2.25 TO 100 KEV.            9228 1451
   0.12 to 20 MeV, based on subtraction of (MT=4,16,17,18,37,102) 9228 1451
   from MT=1.                                                     9228 1451
 MT=18  2.2 to 120 keV: average values from simultaneous          9228 1451
   evaluation of W.P. Poenitz with structure carried over from    9228 1451
   version 5.  Renormalizaion was over 3 ranges per decade.       9228 1451
   0.10 to 20 MeV: standards evaluation by CSEWG Standards        9228 1451
   Committee.                                                     9228 1451
                                                                  9228 1451
 ---------------------------------------------------------------- 9228 1451
 REFERENCES                                                       9228 1451
                                                                  9228 1451
 [Ar88] E.D. Arthur, Los Alamos National Laboratory report        9228 1451
    LA-UR-88-382 (1988)                                           9228 1451
 [Bo72] K. Boeckhoff et al., J.Nuc.En. 26, 91 (1972)              9228 1451
 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958)            9228 1451
 [Ca73] J. Cabe et al., report CEA-R-4524 (1973)                  9228 1451
 [Du70] C.L. Dunford, report AI-AEC-12931 (1970)                  9228 1451
 [En89] T.R. England et al, Los Alamos National Laboratory        9228 1451
    reports LA-11151-MS(88), LA-11534T(89); LAUR-88-4118 to be    9228 1451
    published in Nucl.Sci.Eng. (1989)                             9228 1451
 [Fo71] D. Foster and D. Glasgow, Phys.Rev. C 3, 576 (1971)       9228 1451
 [Fr86] J. Frehaut, report NEANDC(E) 238/L (1986)                 9228 1451
 [Gr73] L. Green et al., report USNDC-9 (1973) p.170              9228 1451
 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Laboratory     9228 1451
    report ORNL/TM-7341 (1980)                                    9228 1451
 [Ku70] P.D. Kunz, "DWUCK: A Distorted-Wave Born Approximation    9228 1451
    Program," unpublished report                                  9228 1451
 [Ma88] D.G.Madland, Nuclear Data for Science and Technology,     9228 1451
    Proc. Int. Conf., Mito, Japan, May 30-June 3, 1988, Mito,     9228 1451
    Japan. (Saikon Publishing, 1988) p.759                        9228 1451
 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960)             9228 1451
 [Pe69] R.J. Peterson, Ann.Phys. 53, 40 (1069)                    9228 1451
 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981)            9228 1451
 [Po83] W. Poenitz et al., Argonne National Laboratory report     9228 1451
    ANL-NDM-80 (1983)                                             9228 1451
 [Ra70] J. Raynal, report IAEA SMR-9/8 (1972) p.281               9228 1451
 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974)           9228 1451
 [Ut66] C. Uttley et al., Paris Conf. (1966) v1, p165             9228 1451
 [Ve80] V. Vertebnyj et al., report YFI-16,8(1973)                9228 1451
 [Wh65] W. Whalen et al., Argonne National Laboratory report      9228 1451
    ANL-7110 (1965) p.15                                          9228 1451
 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National           9228 1451
    Laboratory report LA-6947 (1977).                             9228 1451
 [Yo88] P.G. Young and E.D. Arthur, Nuclear Data for Science and  9228 1451
    Technology, Proc. Int. Conf., Mito, Japan, May 30-June 3,     9228 1451
    1988, Mito, Japan. (Saikon Publishing, 1988) p.603            9228 1451
                                                                  9228 1451
 *****************************************************************9228 1451
                                                                  9228 1451
                                                                  9228 1451
                                                                  9228 1451
                                                                  9228 1451
 ************************ C O N T E N T S *********************** 9228 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9228 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9228 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9228 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9228 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9228 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9228 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9228 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9228 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9228 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9228 1451
 ***************** Program FIXUP (Version 2002-1) ****************9228 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9228 1451
 Corrected Thresholds-----------------------------------------Yes 9228 1451
 Extended Cross Sections to 20 MeV----------------------------No  9228 1451
 Allow Cross Section Deletion---------------------------------No  9228 1451
 Allow Cross Section Reconstruction---------------------------No  9228 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9228 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9228 1451
 Deleted Duplicate Points-------------------------------------Yes 9228 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9228 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9228 1451
 Allow Creation of Missing Sections---------------------------No  9228 1451
 Allow Insertion of Energy Points-----------------------------No  9228 1451
 Create Uniform Energy Grid-----------------------------------No  9228 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9228 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9228 1451
 Unshielded Group Averages Using  640 Groups                      9228 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9228 1451
                                1        451        421          69228 1451
                                3         18        217          29228 1451
                               33         18         53          19228 1451
                                                                  9228 1  0
                                                                  9228 0  0
 9.22350E+4 2.33024E+2          0          0          0          09228 3 18
 1.93720E+8 1.93720E+8          0          0          1        6419228 3 18
        641          1                                            9228 3 18
 .000100000 9805.44565 .000105000 9576.58959 .000110000 9359.172989228 3 18
 .000115000 9158.81971 .000120000 8924.60065 .000127500 8664.732699228 3 18
 .000135000 8427.28509 .000142500 8209.06581 .000150000 7972.884059228 3 18
 .000160000 7726.83530 .000170000 7502.95387 .000180000 7297.437119228 3 18
 .000190000 7106.20951 .000200000 6931.59545 .000210000 6767.315239228 3 18
 .000220000 6614.73347 .000230000 6472.96795 .000240000 6306.559359228 3 18
 .000255000 6123.66483 .000270000 5982.52302 .000280000 5825.396019228 3 18
 .000300000 5633.76253 .000320000 5460.23314 .000340000 5301.297839228 3 18
 .000360000 5154.89776 .000380000 5021.23541 .000400000 4881.864359228 3 18
 .000425000 4739.23564 .000450000 4608.94667 .000475000 4489.456319228 3 18
 .000500000 4377.06284 .000525000 4274.34196 .000550000 4177.297809228 3 18
 .000575000 4087.03116 .000600000 3994.50545 .000630000 3899.312979228 3 18
 .000660000 3812.19928 .000690000 3728.62825 .000720000 3639.695339228 3 18
 .000760000 3544.32250 .000800000 3455.75135 .000840000 3374.564279228 3 18
 .000880000 3297.39250 .000920000 3226.63610 .000960000 3158.862739228 3 18
 .001000000 3088.58871 .001050000 3015.21513 .001100000 2946.416769228 3 18
 .001150000 2883.00805 .001200000 2808.19450 .001275000 2725.942999228 3 18
 .001350000 2650.71743 .001425000 2580.49435 .001500000 2505.864519228 3 18
 .001600000 2427.81125 .001700000 2356.53647 .001800000 2290.600769228 3 18
 .001900000 2230.35443 .002000000 2174.48226 .002100000 2122.043409228 3 18
 .002200000 2074.01496 .002300000 2028.08069 .002400000 1975.639259228 3 18
 .002550000 1917.33528 .002700000 1871.70821 .002800000 1822.088059228 3 18
 .003000000 1760.97131 .003200000 1705.32414 .003400000 1654.419039228 3 18
 .003600000 1607.99265 .003800000 1565.14127 .004000000 1520.490199228 3 18
 .004250000 1475.15152 .004500000 1433.80806 .004750000 1395.182699228 3 18
 .005000000 1359.72924 .005250000 1326.84435 .005500000 1295.742349228 3 18
 .005750000 1267.21002 .006000000 1237.23029 .006300000 1207.238559228 3 18
 .006600000 1178.79652 .006900000 1152.32374 .007200000 1123.623289228 3 18
 .007600000 1093.00492 .008000000 1065.01491 .008400000 1038.751729228 3 18
 .008800000 1014.31842 .009200000 991.628105 .009600000 970.0474319228 3 18
 .010000000 947.719780 .010500000 924.123851 .011000000 902.5498069228 3 18
 .011500000 881.968820 .012000000 858.377920 .012750000 832.2540279228 3 18
 .013500000 808.046326 .014250000 785.825850 .015000000 762.1113209228 3 18
 .016000001 737.201126 .017000001 714.343317 .017999999 693.4602519228 3 18
 .018999999 674.247444 .020000000 656.207027 .021000000 639.6710829228 3 18
 .022000000 624.062750 .023000000 609.484077 .024000000 592.3042479228 3 18
 .025500000 573.671892 .027000001 558.869598 .028000001 542.7086499228 3 18
 .029999999 522.732344 .032000002 504.440711 .034000002 487.6563799228 3 18
 .035999998 472.255930 .037999999 457.986634 .039999999 443.0352159228 3 18
 .042500000 427.789733 .045000002 413.809066 .047499999 400.7416609228 3 18
 .050000001 388.666130 .052499998 377.437964 .055000000 366.7721629228 3 18
 .057500001 356.975025 .059999999 346.658820 .063000001 336.3192579228 3 18
 .066000000 326.567189 .068999998 317.474766 .071999997 307.6173299228 3 18
 .075999998 297.063784 .079999998 287.441943 .083999999 278.4211999228 3 18
 .088000000 270.063384 .092000000 262.378406 .096000001 255.0731149228 3 18
 .100000001 247.618727 .104999997 239.782344 .109999999 232.7116519228 3 18
 .115000002 226.023391 .119999997 218.531499 .127499998 210.3819449228 3 18
 .135000005 203.246366 .142499998 196.997225 .150000006 190.7646869228 3 18
 .159999996 184.898339 .170000002 180.357774 .180000007 177.1051079228 3 18
 .189999998 175.126287 .200000003 174.462125 .209999993 175.1938609228 3 18
 .219999999 177.178431 .230000004 180.252611 .239999995 185.0329859228 3 18
 .254999995 190.769754 .270000011 193.939661 .280000001 193.1709989228 3 18
 .300000012 184.231407 .319999993 168.852362 .340000004 151.7589049228 3 18
 .360000014 135.928082 .379999995 122.589373 .400000006 110.5176789228 3 18
 .425000012 99.9215048 .449999988 91.5947961 .474999994 84.94332229228 3 18
 .500000000 79.5238600 .524999976 75.0078680 .550000012 71.20982789228 3 18
 .574999988 67.9934101 .600000024 65.0089965 .629999995 62.28403759228 3 18
 .660000026 60.0415616 .689999998 58.2212189 .720000029 56.58891869228 3 18
 .759999990 55.3543587 .800000012 54.8208531 .839999974 55.12350979228 3 18
 .879999995 56.4393925 .920000017 59.2267685 .959999979 64.23533029228 3 18
 1.00000000 74.2339430 1.04999995 92.4090720 1.10000002 106.6311909228 3 18
 1.14999998 84.1961980 1.20000005 40.5506560 1.27499998 23.32765609228 3 18
 1.35000002 18.9406099 1.42499995 16.6043776 1.50000000 15.48397319228 3 18
 1.60000002 14.7261739 1.70000005 14.1849222 1.79999995 13.86143009228 3 18
 1.89999998 15.5494953 2.00000000 22.6032031 2.09999990 13.26896039228 3 18
 2.20000005 11.8173435 2.29999995 11.1169809 2.40000010 10.19335919228 3 18
 2.54999995 8.62381024 2.70000005 10.1648690 2.79999995 19.79615379228 3 18
 3.00000000 41.6625435 3.20000005 23.3431674 3.40000010 43.65022679228 3 18
 3.59999990 36.2943279 3.79999995 2.85022773 4.00000000 3.174980859228 3 18
 4.25000000 3.85193210 4.50000000 4.72371501 4.75000000 12.31006079228 3 18
 5.00000000 6.82445083 5.25000000 13.2267853 5.50000000 16.76028149228 3 18
 5.75000000 17.0446690 6.00000000 42.0229076 6.30000019 61.81895219228 3 18
 6.59999990 6.06136848 6.90000010 47.1281131 7.19999981 6.296274059228 3 18
 7.59999990 3.13158452 8.00000000 10.2897118 8.39999962 245.1606749228 3 18
 8.80000019 206.950433 9.19999981 68.4065317 9.60000038 25.68573689228 3 18
 10.0000000 23.9732579 10.5000000 13.3149403 11.0000000 13.29925449228 3 18
 11.5000000 33.1829564 12.0000000 113.363678 12.7500000 23.18694299228 3 18
 13.5000000 65.5441042 14.2500000 24.7410127 15.0000000 21.58862779228 3 18
 16.0000000 37.0450220 17.0000000 14.7065015 18.0000000 38.46404379228 3 18
 19.0000000 187.981636 20.0000000 30.3272760 21.0000000 43.55740589228 3 18
 22.0000000 19.9103987 23.0000000 67.9364956 24.0000000 47.81544969228 3 18
 25.5000000 44.6604029 27.0000000 37.5121468 28.0000000 11.47197099228 3 18
 30.0000000 21.2250310 32.0000000 48.1749933 34.0000000 153.0713249228 3 18
 36.0000000 11.8018034 38.0000000 49.6850700 40.0000000 32.98915849228 3 18
 42.5000000 32.1073233 45.0000000 29.1617658 47.5000000 36.16726859228 3 18
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                                                                  922833  0
                                                                  9228 0  0
                                                                     0 0  0
 9.22380E+4 2.36005E+2          0          1         34          49237 1451
 0.0        1.00000E+0          0          0          0          69237 1451
 1.00000E+0 2.00000E+7          0          0         10       20029237 1451
 3.00000E+2 0.0                 1          0        423          59237 1451
 92-U -238 ORNL+,JAER EVAL-NOV89 L.W.WESTON ET AL.,Y.KANDA ET AL. 9237 1451
                      DIST-OCT98 REV3-FEB97            19981007   9237 1451
----IRDF-2002         MATERIAL 9237         REVISION 3            9237 1451
-----INCIDENT NEUTRON DATA                                        9237 1451
------ENDF-6 FORMAT                                               9237 1451
Relative covariances MF33 MT102 in the energy range up to 100 eV  9237 1451
were increased to account for the systematic difference between   9237 1451
the present ENDF/B-VI value of 2.718 barns and the newly          9237 1451
recommended value of 2.683 barns for the thermal capture cross    9237 1451
section. The correction amounts to ((2.718-2.683)/2.683)**2.      9237 1451
Ref.: A.Trkov et al: Re-visiting the U-238 Thermal Capture Cross  9237 1451
Section and Gamma-ray Emission Probabilities of Np-239 Decay,     9237 1451
to be published in Nuclear Science and Engineering, 2005.         9237 1451
******************************************************************9237 1451
 92-U -238 ORNL,LANL+ EVAL-NOV89 L.W.WESTON,P.G.YOUNG,W.POENITZ   9237 1451
                      DIST-OCT98 REV3-FEB97            19981007   9237 1451
----ENDF/B-VI         MATERIAL 9237         REVISION 3            9237 1451
*****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9237 1451
DOSIMETRY                                                         9237 1451
******************************************************************9237 1451
 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL.                   9237 1451
                      DIST-JUL98                                  9237 1451
----JENDL/D-99        MATERIAL 9237                               9237 1451
                                                                  9237 1451
 <DOSIMETRY CROSS SECTION>                                        9237 1451
      U -238 FISSION                                              9237 1451
 **************************************************************** 9237 1451
                                                                  9237 1451
 ENDF/B-VI MOD 4 Revision, February 1997, V.McLane (NNDC)         9237 1451
                                                                  9237 1451
 MT=1 changes                                                     9237 1451
    MF=451: Updated comments, references.                         9237 1451
    MF=455: Corrected TAB1 bookkeeping (minor revision).          9237 1451
 MT=2 changes                                                     9237 1451
    MF=151                                                        9237 1451
      Added l-dependent scattering radius equal to scattering     9237 1451
        radius for resolved region.                               9237 1451
      Corrected energy upper limit in unresolved region.          9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
                                                                  9237 1451
 ENDF/B-VI MOD 3 Revision, January 1993, L. Weston (ORNL)         9237 1451
                                                                  9237 1451
 MF=2 changes                                                     9237 1451
    Resolved region (to 10 keV): Replaces the preliminary         9237 1451
      evaluation of Moxon and Sowerby with the final version      9237 1451
      [Mo??].                                                     9237 1451
    Unresolved region: Uses Froehner's evaluation.                9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
                                                                  9237 1451
 ENDF/B-VI MOD 2 Revision 1, January 1991 (NNDC)                  9237 1451
                                                                  9237 1451
   Covariance files for total, elastic scattering, fission and    9237 1451
 capture cross sections removed, since the correct files were and 9237 1451
 still are not yet available.                                     9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
                                                                  9237 1451
 ENDF/B-VI MOD 1 Evaluation, November 1989 (ORNL, LANL, ANL)      9237 1451
                                                                  9237 1451
 Principal evaluators                                             9237 1451
 --------------------                                             9237 1451
 RESOLVED RESONANCE REGION:    (10-5 to 10,000 eV)                9237 1451
    M.G.Sowerby and M. C. Moxon (Harwell)                         9237 1451
 UNRESOLVED RESONANCE REGION:  (10 TO 149 keV)                    9237 1451
    F.H.Froehner(KFK) and W.P.Poenitz(ANL)                        9237 1451
 REACTIONS ABOVE 149 keV:                                         9237 1451
    MODEL CODE CALCULATIONS: (LANL)-P.G.Young and R.E.MacFarlane  9237 1451
    FISSION: W.P.Poenitz(ANL)                                     9237 1451
 NU-BAR DELAYED AND DELAYED  NEUTRON SPECTRA:                     9237 1451
    KAISER AND CARPENTER (Ka78)                                   9237 1451
 GAMMA PRODUCTION FILES: R. HOWERTON (LLL)                        9237 1451
 UNCERTAINTY FILES: NOT YET AVAILABLE                             9237 1451
                                                                  9237 1451
 ---------------------------------------------------------------- 9237 1451
                                                                  9237 1451
 ENDF/B-VI EVALUATION ABOVE THE UNRESOLVED RESONANCE REGION       9237 1451
             P.G.Young and R.E.MacFarlane (LANL)                  9237 1451
              Updated by R.Q. Wright (ORNL)                       9237 1451
                                                                  9237 1451
   Above the unresolved resonance region, new evaluations were    9237 1451
 performed of the neutron total, (n,2n), (,n,3n), (n,4n), (n,f),  9237 1451
 (n,nf), (n,2nf), (n,3nf), and (n,gamma) cross sections as well as9237 1451
 prompt nubar.  The elastic and inelastic data from ENDF/B-V were 9237 1451
 carried over for Version VI.                                     9237 1451
   To provide the new data, coupled channel optical model calcu-  9237 1451
 lations were performed with the ECIS code (Ra70) for the lowest  9237 1451
 3 members of the U238 ground state rotational band. These calcu- 9237 1451
 lations were used to provide initial (prior) values for a covar- 9237 1451
 iance analysis of the total cross section and to provide neutron 9237 1451
 transmission coefficients for nuclear reaction theory            9237 1451
 calculations with the GNASH (Ar88,Yo77) and COMNUC (Du70) Hauser-9237 1451
 Feshbach statistical/fission/preequilibrium codes.  These theory 9237 1451
 calculations were used to provide the MF=6 neutron distributions 9237 1451
 from the (n,2n), (n,3n), and (n,4n) reactions as well as prior   9237 1451
 values for covariance analyses of the cross sections for those   9237 1451
 reactions.  Additionally, the above analyses plus DWBA calcu-    9237 1451
 lations were used to check the ENDF/B-V evaluation of elastic    9237 1451
 and inelastic scattering.  While some differences were found,    9237 1451
 the earlier work was generally found to be reliable, and we      9237 1451
 decided to carry over the ENDF/B-V data because of the effort    9237 1451
 taken to match experimental data, both at lower energies and at  9237 1451
 14 MeV.                                                          9237 1451
                                                                  9237 1451
 MF=1 DESCRIPTIVE INFORMATION                                     9237 1451
                                                                  9237 1451
 MF=2 RESONANCE PARAMETERS (Updated for MOD 3)                    9237 1451
    MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS           9237 1451
       RESOLVED RESONANCE PARAMETERS (1.0E-5 eV to 10.0 keV)      9237 1451
         Reich-Moore analysis from new evaluation by Moxon and    9237 1451
           Sowerby [1].  Present parameters are a preliminary set 9237 1451
           due to use of preliminary capture yield data of        9237 1451
           Macklin et al. [2].  Parameters above 4.0 kev corrected9237 1451
            for background and normalisation errors in [2].       9237 1451
         Parameters 0f the bound resonances evaluated so as to    9237 1451
           reproduce the following values at 2200m/s              9237 1451
                                                                  9237 1451
          Reaction    Cross section (barns)    Reference          9237 1451
           Elastic            9.38               [3]              9237 1451
           Fission           11.0E-6             [4]              9237 1451
           Capture            2.708              [5],[6]          9237 1451
       RESOLVED RESONANCE RANGE (10.0 to 149 keV)                 9237 1451
         Coherent calculation of point cross sections and of      9237 1451
          energy-dependent average resonance parameters from      9237 1451
          evaluated average resonance parameters for E = 0 with   9237 1451
          Hauser-Feshbach code FITACS [7]                         9237 1451
         In the unresolved region the point cross sections given  9237 1451
          in MF=3 are considered as correct.  The unresolved      9237 1451
          parameters in MF=2 are only to be used for self-        9237 1451
          shielding calculations.  The use of the parameters with 9237 1451
          ENDF processing codes does not lead to cross sections   9237 1451
          consistent with MF=3 as the codes use a more primative  9237 1451
          version of the formalism.                               9237 1451
         Cross-sections used to determine the parameters          9237 1451
           TOTAL: determined mainly by the following data.        9237 1451
             below 45 keV: Uttley et al. [19], Byoun et al. [20], 9237 1451
                  Kononov and Poletaev [9], Tsubone et al. [8],   9237 1451
             above 45 keV: Smith et al. [10]                      9237 1451
           ELASTIC SCATTERING:                                    9237 1451
             Calculated as (total)-(partial cross-sections)       9237 1451
           TOTAL INELASTIC AND INELASTIC LEVELS MT=51 AND 52      9237 1451
             Determined mainly by the following data:             9237 1451
             Winters et al. [11], Murzin et al. [12], Tsang and   9237 1451
             Brugger [13], Smith [14], with level scheme from     9237 1451
             Table of Isotopes [15]                               9237 1451
               Level No.  Energy(MeV)    Spin-parity              9237 1451
                  G.S.     0.0              0+                    9237 1451
                   1       0.0448           2+                    9237 1451
                   2       0.1484           4+                    9237 1451
                   3       0.3072           6+                    9237 1451
           FISSION                                                9237 1451
             Extrapolated downward from JEF-1.                    9237 1451
           CAPTURE CROSS-SECTION                                  9237 1451
             Determined mainly by the post-1966 data compiled by  9237 1451
             Poenitz [16], by the measurement of Kazakov [17] and 9237 1451
             by the mean parameters (radiation width, level       9237 1451
             density) extracted from resolved resonance           9237 1451
             parameters [7].                                      9237 1451
                                                                  9237 1451
   2200 M/S VALUES AND RESONANCE INTEGRALS GIVEN BY EVALUATION    9237 1451
                                                                  9237 1451
                   2200m/s values   Resonance Integral            9237 1451
                       (barns)         (barns)                    9237 1451
        Total          12.32            597.4                     9237 1451
        Elastic         9.60            319.4                     9237 1451
        Fission         6.46E-05          1.7E-03                 9237 1451
        Capture         2.72            278.0                     9237 1451
                                                                  9237 1451
 ---------------------------------------------------------------- 9237 1451
 REFERENCES FOR MF=2                                              9237 1451
                                                                  9237 1451
  1. M.C. Moxon and M.G. Sowerby, Pr0c 1988 Int.Reac.Phys.Conf.   9237 1451
     Jackson Hole (1988) Vol.1, p,281                             9237 1451
  2. R.L. Macklin et al, Conf. 1988 Mito, Japan, (1988) p.71      9237 1451
  3. S.F. Mughabghab, Neutron Cross Sections, Vol.1, Part B,      9237 1451
     Z=61-100 (Academic Press, 1984)                              9237 1451
  4. P. d'Hondt et al, Ann.Nucl.EN., 11, 485 (1984)               9237 1451
  5. W.P. Poenitz, report IAEA-TECDOC-335 (1985) p.426 and        9237 1451
     private communication (1987)                                 9237 1451
  6. A.D. Carlson et al, Proc.Conf. 1985 Santa Fe, Vol.2, p.1429  9237 1451
     (1985)                                                       9237 1451
  7. F.H. Froehner, PROC 1988 INT.REAC.PHYS. CONF., Jackson Hole  9237 1451
     (1988) Vol.III, p 171                                        9237 1451
  8. I. Tsubone et al, Nucl.Sci.Eng. 88, 579 (1984)               9237 1451
  9. V.N. Kononov and E.D. Poletaev, 1973  KIEV CONF 2, 199 (1984)9237 1451
 10. A.B. Smith et al, private communication (1988)               9237 1451
 11. R.R. Winters et al, Nucl.Sci.Eng. 78, 147 (1981)             9237 1451
 12. A.W. Murzin et al, Yad.Konst. 4, 30 (1986)                   9237 1451
 13. F.Y. Tsang and R.M. Brugger, Nucl.Sci.Eng. 65, 70 (1978)     9237 1451
 14. A.B. Smith, report INDC/P(84)-28, unpublished (1984)         9237 1451
 15. C.M. Lederer, Ed., Table of Isotopes, 7th edition, (1978)    9237 1451
 16. W.P. Poenitz et al, Argonne National Lab. report ANL-83-4    9237 1451
     suppl. (1983)                                                9237 1451
 17. L.E. Kazakov et al, Yad.Konst. 3, 37 (1986)                  9237 1451
 18. F.C. Difilippo et al, Phys.Rev.C 21, 1400 (1980)             9237 1451
 19. C.A. Uttley et al., Nucl.Data for Reactors, Proc.Int.Conf.   9237 1451
     Paris 1966 (IAEA, 1966) Vol.I, p.165                         9237 1451
 20. T.Y. Byoun et al., PROC.MEET. REACTOR PHYS. AND SHIELD CALC.,9237 1451
     Kiamesha Lake, NY, 1972 (U.S.Nucl.Reg.Comm., 1972) Vol.1,    9237 1451
     p.1115                                                       9237 1451
                                                                  9237 1451
 MF=3 SMOOTH CROSS SECTIONS                                       9237 1451
    MT=102 Radiative Capture Cross Section.  10 to 149 keV,       9237 1451
       evaluation by F.H. Froehner (KFK) which is based on FITACS 9237 1451
       fit to the experimental data.  0.15 to 20 MeV, taken       9237 1451
       directly from simultaneous standards analysis [Ca85],      9237 1451
       [Po89].                                                    9237 1451
                                                                  9237 1451
 ---------------------------------------------------------------- 9237 1451
 REFERENCES                                                       9237 1451
                                                                  9237 1451
 [Al61] K. Allen et al., J.Nuc.En. 14, 100 (1961)                 9237 1451
 [An58] G. Antropov et al., At.En. 5, 456 (1958)                  9237 1451
 [Ar88] E.D. Arthur, Los Alamos National Lab. report LA-UR-88-382 9237 1451
    (1988)                                                        9237 1451
 [Ba66] D. Barr, Los Aalamos National Lab., private communication 9237 1451
    to R.Howerton (1966)                                          9237 1451
 [Ba65] R. Batchelor et al., Nuc.Phys. 65, 236 (1965)             9237 1451
 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958)            9237 1451
 [Ca73] J. Cabe et al., CEN Saclay report CEA-R-4524 (1973)       9237 1451
 [Ca85] A. Carlson et al., Nucl. Data for Basic & Applied Science,9237 1451
       Proc.Conf., Santa Fe, NM 1985 [] p.1429.                   9237 1451
 [Di80] F.C. Difillippo et al., Phys.Rev.C 21, 1400 (1980)        9237 1451
 [Du70] C.L. Dunford, Atomics Int. report AI-AEC-12931 (1970)     9237 1451
 [Fo71] D. Foster and D. Glasgow, Phys.Rev.C 3, 576 (1971)        9237 1451
 [Fr80a] J. Frehaut et al., Nucl.Sci.Eng. 74, 29 (1980)           9237 1451
 [Fr80b] J. Frehaut et al., Brookhaven National Lab. report       9237 1451
    BNL-NCS-512457 (1980) p.399                                   9237 1451
 [Fr86] J. Frehaut, NEANDC(E) 238/L (1986)                        9237 1451
 [Ha73] S. Hayes et al., Nuc.Sci.Eng. 50, 243 (1973)              9237 1451
 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Lab. report    9237 1451
    ORNL/TM-7341 (1980)                                           9237 1451
 [Ka78] R. Kaiser and S. Carpenter (ANL-West) private             9237 1451
    communication (1978).                                         9237 1451
 [Ko80] N. Kornilov et al., report ZFK-410 (1980) p.68            9237 1451
 [La73] J. Landrum et al., Phys.Rev.C 8, 1938 (1973)              9237 1451
 [Li79] P. Lisowski et al., (LANL WNR measurement) private        9237 1451
    communication (1979)                                          9237 1451
 [Ma69] D. Mather, report AWRE-O-47 (1969)                        9237 1451
 [Ma72] D. Mather et al., report AWRE-O-72 (1972).                9237 1451
 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960)             9237 1451
 [Pe61] J. Perkin, J.Nuc.En. 14, 69 (1961)                        9237 1451
 [Ph56] J. Phillips, report AERE-NP/R-2033 (1956)                 9237 1451
 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981)            9237 1451
 [Po83] W.Poenitz et al., Argonne National Lab. report            9237 1451
    ANL-NDM-80 (1983)                                             9237 1451
 [Po89] W. Poenitz, ANL-West, private communication (1989)        9237 1451
 [Ra70] J. Raynal, report IAEA SMR-9/8 (1970)                     9237 1451
 [Ro57] L. Rosen et al., Los Alamos National Lab. report LA-2111  9237 1451
    (1958)                                                        9237 1451
 [Ry80] T. Ryves, J.Phys.G 6, 771 (1980)                          9237 1451
 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974)           9237 1451
 [Sh78] R. Shamu et al., private communication (1978)             9237 1451
 [Sl77] R. Slovacek et al., Nuc.Sci.Eng. 62, 455 (1977)           9237 1451
 [Sm82] A. Smith et al., Argonne National Lab. report ANL/NDM-74  9237 1451
    (1982)                                                        9237 1451
 [Ut66] C. Uttley et al., Paris Conf. (1966) vol.1, p.165         9237 1451
 [Ve78] L. Veeser and E. Arthur, Harwell Nuc.Data Conf. (1978)    9237 1451
    p 1054                                                        9237 1451
 [Wh62] P. White et al., J.Nuc.En.A/B 16, 261 (1962)              9237 1451
 [Wh71] J. Whalen et al., Argonne National Lab. report ANL-7710   9237 1451
    (1971) p.9                                                    9237 1451
 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National Lab.      9237 1451
    report LA-6947 (1977)                                         9237 1451
 [Yo78] Chou You-Pu, report HSJ-77091 (1978)                      9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
          Modifications applied by the IRDF evaluators            9237 1451
          H.Vonach and S.Tagesen, IRK Vienna                      9237 1451
             MT=102 Capture Cross Section                         9237 1451
          Uncertainties from simultaneous standards analysis      9237 1451
          used up to En=2.3+6 ev                                  9237 1451
          Uncertainties for En=2.3+6 to 2.0+7 ev estimated from   9237 1451
          existing experimental data (1)                          9237 1451
          REFERENCES:                                             9237 1451
          1.Neutron Cross Sections, vol2, V.McLane, C.L.Dunford,  9237 1451
          P.F.Rose, Academic Press(1988)                          9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
                                                                  9237 1451
 ENDF/B-V MAT 1393 Evaluation, November 1981 (ANL, LLNL)          9237 1451
                                                                  9237 1451
 Description of data carried over from ENDF/B-V                   9237 1451
                                                                  9237 1451
 Principal Evaluators:                                            9237 1451
  E.M.Pennington (ANL), A.B.Smith (ANL), W.P.Poenitz (ANL)        9237 1451
  R.Howerton (LLL) - gamma production files                       9237 1451
 Contributing evaluators:                                         9237 1451
  R.Kaiser and S.Carpenter (ANL-west) - nu-bar delayed and        9237 1451
   delayed neutron spectra                                        9237 1451
                                                                  9237 1451
MF = 3 smooth cross sections                                      9237 1451
   Cross sections above 45 kev were evaluated by A.Smith.         9237 1451
   more details are provided under each reaction type below. Most 9237 1451
   of the evaluation above 45 kev is described in ANL/NDM-32.     9237 1451
 ---------------------------------------------------------------- 9237 1451
 REFERENCES                                                       9237 1451
  1. R.E. Kaiser and S.G. Carpenter (ANL-west) private comm.      9237 1451
     (March 1978)                                                 9237 1451
  2. S.A. Cox, Argonne National Lab. report ANL/NDM-5 (1974)      9237 1451
  3. C. Besant et al, Sem. Fast Pulsed Reactors CONF-760111 (1976)9237 1451
  4. F. Manero and V. Konshin, At.En.Review 10, #4 (1972)         9237 1451
  5. M. Soleilhac, revised data received from NNCSC               9237 1451
  6. B. Nurpeisov et al., Sov.At.En. translation 807 (Mar 1976)   9237 1451
  7. G. deSaussure et al., Prog.Nucl.En. 3, 87 (1979)             9237 1451
  8. D. Olsen et al., Nuc.Sci.Eng. 62, 479 (1977)                 9237 1451
  9. P. Guenther, D. Havel, A. Smith Argonne National Lab. report 9237 1451
     ANL-NDM-22 (1976)                                            9237 1451
 10. E. Pennington, W. Poenitz, A. Smith, Trans.Am.Nucl.Soc. 26,  9237 1451
     591 (1977)                                                   9237 1451
 11. W.P. Poenitz Argonne National Lab. report ANL/NDM-45 (1979)  9237 1451
 12. W.P. Poenitz and A.B. Smith, ed. Argonne National Lab. report9237 1451
     ANL-76-90 (1976)                                             9237 1451
 13. R. Slovacek et al., Nucl.Sci.Eng. 62, 455 (1977)             9237 1451
 14. M. Caner, M. Segev, W. Yiftah, Nucl.Sci.Eng. 59, 395 (1976)  9237 1451
 15. E. Kujawski and L. Stewart, Trans.Am.Nucl.Soc. 24, 453 (1976)9237 1451
 16. L.W. Weston private comm. to B.A. Magurno Nov.12, 1982       9237 1451
 17. R. Sher and ?. Beck report EPRI NP-1771/81 and rev. 1/83,    9237 1451
     and private communication to Magurno, February 1983          9237 1451
                                                                  9237 1451
 **************************************************************** 9237 1451
 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL.                   9237 1451
                      DIST-JUL98                                  9237 1451
----JENDL/D-99        MATERIAL 9237                               9237 1451
-----INCIDENT NEUTRON DATA                                        9237 1451
------ENDF-6 FORMAT                                               9237 1451
HISTORY                                                           9237 1451
87-01 SIMULTANEOUS EVALUATION FOR FISSION AND CAPTURE CROSS       9237 1451
      SECTIONS WAS COMPLETED IN THE ENERGY RANGE ABOVE 50 KEV.    9237 1451
93-03 FISSION SPECTRA CALCULATED BY T.OHSAWA(KINKI UNIV.)         9237 1451
      COMPILED BY T.NAKAGAWA (NDC/JAERI)                          9237 1451
98-04 COMPILED TO JENDL DOSIMETRY FILE 99.                        9237 1451
                                                                  9237 1451
            ===== POINT-WISE DATA FILE =====                      9237 1451
                                                                  9237 1451
 <DOSIMETRY CROSS SECTION>                                        9237 1451
      U -238 FISSION                                              9237 1451
                                                                  9237 1451
MF=1  GENERAL INFORMATION                                         9237 1451
  MT=451  DESCRIPTIVE DATA AND DIRECTORY RECORDS                  9237 1451
                                                                  9237 1451
                                                                  9237 1451
   2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS.    9237 1451
                       2200 M/S(B)     RES. INTEG.(B)             9237 1451
           FISSION      11.8E-6           1.72                    9237 1451
                                                                  9237 1451
 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.800 PER-CENT      9237 1451
 ***************** PROGRAM RECENT (VERSION 89-1) *****************9237 1451
 RESONANCE CONTRIBUTION RECONSTRUCTED TO WITHIN    1.000 PER-CENT 9237 1451
 COMBINED DATA NOT THINNED (ALL RESONANCE + BACKGROUND DATA KEPT) 9237 1451
 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451
 DATA LINEARIZED TO WITHIN AN ACCURACY OF     0.800 PER-CENT      9237 1451
                                                                  9237 1451
MF=3  NEUTRON CROSS SECTIONS                                      9237 1451
     BELOW 10 KEV, NO BACKGROUND CROSS SECTIONS WERE GIVEN.       9237 1451
     ABOVE 10 KEV, CROSS SECTIONS WERE EVALUATED AS FOLLOWS, AND  9237 1451
     THEY WERE REPRESENTED WITH THE UNRESOLVED RESONANCE          9237 1451
     PARAMETERS EXCEPT THE FISSION CROSS SECTION.                 9237 1451
                                                                  9237 1451
  MT=18  FISSION                                                  9237 1451
     BELOW 100 KEV : TAKEN FROM EXPERIMENTAL DATA /12/.           9237 1451
     100 - 600 KEV : EVALUATED ON THE BASIS OF THE DATA OF        9237 1451
          DIFFILIPPO ET AL. /13/, BEHRENS AND CARLSON /14/,       9237 1451
          NORDBORG ET AL. /15/ AND MEADOWS /16,17/.               9237 1451
     ABOVE 600 KEV : RESULTS OF SIMULTANEOUS EVALUATION /18/ MADE 9237 1451
          BY CONSIDERING THE EXPERIMENTAL DATA OF REFS./14,15,    9237 1451
          16,17,19,20,21,22,23,24,25,26,27,28/.                   9237 1451
                                                                  9237 1451
MF=33                                                             9237 1451
 MT=18   Taken from JENDL/D-99                                    9237 1451
                                                                  9237 1451
                                                                  9237 1451
REFERENCES                                                        9237 1451
12) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 63, 153 (1977).       9237 1451
13) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 68, 43 (1978).        9237 1451
14) BEHRENS J.W. AND CARLSON G.W.: NUCL. SCI. ENG., 63, 250       9237 1451
    (1977).                                                       9237 1451
15) NORDBORG C. ET AL: ANL-76-90, 128 (1976).                     9237 1451
16) MEADOWS J.W.: NUCL. SCI. ENG., 58, 255 (1975).                9237 1451
17) MEADOWS J.W.: NUCL. SCI. ENG., 49, 310 (1972).                9237 1451
18) KANDA Y. ET AL.: 1985 SANTA FE, 2, 1567 (1986).               9237 1451
19) CANCE M. AND GRENIER G.:NUCL. SCI. ENG., 68, 197 (1978).      9237 1451
20) BILAUD P. ET AL.:1958 GENEVA,16, 106, 5809 (1958).            9237 1451
21) ADAMOV V.M. ET AL.: 1977 NBS, 313 (1977).                     9237 1451
22) ARLT R. ET AL.: KE, 24, 48, 8102 (1981).                      9237 1451
23) CIERJACKS S. ET AL.:1976 ANL, 94 (1976).                      9237 1451
24) GOVERDOVSKII A.A. ET AL.: 1983 KIEV, 2, 159  (1983).          9237 1451
25) ANDROSENKO S.D. ET AL.: 1983 KIEV, 2, 153  (1983).            9237 1451
26) FURSOV B.I. ET AL.: SOV. ATOM. ENERG., 43, 808  (1978).       9237 1451
27) POENITZ W.P. AND ARMANI R.J.: J. NUCL. ENEG., 26, 483 (1972). 9237 1451
28) POENITZ W.P.: NUCL. SCI. ENG., 57, 300 (1975).                9237 1451
                                                                  9237 1451
 ************************ C O N T E N T S *********************** 9237 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9237 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9237 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9237 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9237 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9237 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9237 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9237 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9237 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9237 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9237 1451
 ***************** Program FIXUP (Version 2002-1) ****************9237 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9237 1451
 Corrected Thresholds-----------------------------------------Yes 9237 1451
 Extended Cross Sections to 20 MeV----------------------------No  9237 1451
 Allow Cross Section Deletion---------------------------------No  9237 1451
 Allow Cross Section Reconstruction---------------------------No  9237 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9237 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9237 1451
 Deleted Duplicate Points-------------------------------------Yes 9237 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9237 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9237 1451
 Allow Creation of Missing Sections---------------------------No  9237 1451
 Allow Insertion of Energy Points-----------------------------No  9237 1451
 Create Uniform Energy Grid-----------------------------------No  9237 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9237 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9237 1451
 Unshielded Group Averages Using  640 Groups                      9237 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9237 1451
                                1        451        432          49237 1451
                                3         18        217          19237 1451
                                3        102        217          39237 1451
                               33         18         11          09237 1451
                               33        102        335          09237 1451
                                                                  9237 1  0
                                                                  9237 0  0
 9.22380E+4 2.36005E+2          0          0          0          09237 3 18
 1.89629E+8 1.89629E+8          0          0          1        6419237 3 18
        641          1                                            9237 3 18
 .000100000 .000184668 .000105000 .000180362 .000110000 .0001762729237 3 18
 .000115000 .000172502 .000120000 .000168096 .000127500 .0001632069237 3 18
 .000135000 .000158740 .000142500 .000154635 .000150000 .0001501939237 3 18
 .000160000 .000145564 .000170000 .000141353 .000180000 .0001374889237 3 18
 .000190000 .000133892 .000200000 .000130609 .000210000 .0001275209237 3 18
 .000220000 .000124650 .000230000 .000121984 .000240000 .0001188559237 3 18
 .000255000 .000115416 .000270000 .000112762 .000280000 .0001098089237 3 18
 .000300000 .000106207 .000320000 .000102945 .000340000 9.99571E-59237 3 18
 .000360000 9.72059E-5 .000380000 9.46943E-5 .000400000 9.20757E-59237 3 18
 .000425000 8.93960E-5 .000450000 8.69475E-5 .000475000 8.47028E-59237 3 18
 .000500000 8.25920E-5 .000525000 8.06627E-5 .000550000 7.88401E-59237 3 18
 .000575000 7.71450E-5 .000600000 7.54077E-5 .000630000 7.36205E-59237 3 18
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 8.112564-9 3.906122-7 3.906122-7 1.311645-6 1.311645-6 2.413183-6923733102
 2.460964-6 2.983461-6 2.527120-6 2.714317-6 2.678308-6 2.660081-6923733102
 2.675881-6 2.151841-6 1.976851-6 1.818055-6 1.777526-6 1.655458-6923733102
 1.544691-6 1.487994-6 1.275018-6 1.067313-6 5.311681-7 9.632330-7923733102
 4.184709-7 4.184709-7 1.447843-6 1.447843-6 2.694459-6 2.892719-6923733102
 3.638571-6 3.135139-6 3.366454-6 3.245738-6 3.114898-6 3.130073-6923733102
 2.531963-6 2.346474-6 2.151492-6 2.097494-6 1.949340-6 1.812555-6923733102
 1.745052-6 1.498504-6 1.254898-6 6.248403-7 1.132928-6 4.626174-7923733102
 4.626174-7 1.622848-6 1.622848-6 2.966232-6 3.432597-6 5.688083-6923733102
 4.582886-6 4.979980-6 4.583791-6 4.133409-6 4.046871-6 3.336078-6923733102
 3.184438-6 2.909915-6 2.819586-6 2.613597-6 2.422280-6 2.330329-6923733102
 2.006674-6 1.680714-6 8.371374-7 1.517885-6 3.932794-7 3.932794-7923733102
 1.447838-6 1.447838-6 2.703966-6 3.436486-6 5.122691-6 7.174253-6923733102
 7.805172-6 6.856467-6 5.367642-6 5.021746-6 4.167291-6 4.055483-6923733102
 3.677896-6 3.512186-6 3.229974-6 2.963649-6 2.846886-6 2.472893-6923733102
 2.071395-6 1.033842-6 1.872878-6 3.474893-7 3.474893-7 1.359861-6923733102
 1.359861-6 2.624447-6 3.461609-6 5.090528-6 7.493914-6 1.303279-5923733102
 9.074728-6 6.403962-6 5.835266-6 4.948797-6 4.855329-6 4.434647-6923733102
 4.239867-6 3.869714-6 3.523156-6 3.383203-6 2.954499-6 2.475535-6923733102
 1.238298-6 2.240272-6 3.961709-7 3.961709-7 1.433665-6 1.433665-6923733102
 2.696034-6 3.187881-6 4.215297-6 6.164439-6 8.092111-6 1.595267-5923733102
 7.928151-6 6.248201-6 5.376552-6 5.152228-6 4.748594-6 4.567367-6923733102
 4.148828-6 3.750376-6 3.598700-6 3.180284-6 2.654147-6 1.330140-6923733102
 2.401390-6 4.044586-7 4.044586-7 1.438398-6 1.438398-6 2.637152-6923733102
 2.986730-6 3.951442-6 5.080911-6 6.277739-6 8.271959-6 1.115239-5923733102
 6.831780-6 5.665925-6 5.170668-6 4.727049-6 4.588867-6 4.184250-6923733102
 3.792206-6 3.631848-6 3.199669-6 2.663915-6 1.332716-6 2.410612-6923733102
 3.224936-7 3.224936-7 1.303452-6 1.303452-6 2.510809-6 3.013809-6923733102
 3.729339-6 4.588372-6 5.496114-6 6.247491-6 6.780534-6 1.172814-5923733102
 6.754721-6 5.548898-6 5.143174-6 4.988147-6 4.475989-6 3.949614-6923733102
 3.759107-6 3.293664-6 2.787971-6 1.399763-6 2.531463-6 3.234167-7923733102
 3.234167-7 1.226796-6 1.226796-6 2.304837-6 2.690726-6 3.269777-6923733102
 4.184483-6 4.900121-6 5.925724-6 5.771941-6 6.867382-6 1.171719-5923733102
 6.496188-6 5.483618-6 5.208147-6 4.680770-6 4.045318-6 3.850792-6923733102
 3.412132-6 2.893050-6 1.458853-6 2.631127-6 2.501066-7 2.501066-7923733102
 1.000070-6 1.000070-6 1.905809-6 2.305040-6 2.833609-6 4.010246-6923733102
 4.773239-6 5.619265-6 5.534104-6 5.597130-6 5.939952-6 1.146934-5923733102
 7.166277-6 6.145922-6 5.447677-6 4.727558-6 4.508064-6 4.065641-6923733102
 3.446592-6 1.727079-6 3.135060-6 1.051057-7 1.051057-7 5.873906-7923733102
 5.873906-7 1.177779-6 1.468241-6 1.856267-6 2.638892-6 2.864619-6923733102
 3.679720-6 3.697025-6 4.211474-6 4.426577-6 6.835702-6 1.262842-5923733102
 8.885556-6 6.385032-6 5.671047-6 5.461539-6 5.000133-6 4.284651-6923733102
 2.129101-6 3.932657-6 9.722399-8 9.722399-8 5.084818-7 5.084818-7923733102
 1.012793-6 1.258993-6 1.601016-6 2.265194-6 2.355179-6 3.161445-6923733102
 3.147990-6 3.677312-6 3.676768-6 5.524721-6 6.855775-6 1.243189-5923733102
 9.611692-6 6.597824-6 6.445379-6 5.871086-6 5.042240-6 2.532897-6923733102
 4.637333-6                                                       923733102
 3.000000+0 0.000000+0       9437         18          0          1923733102
 0.000000+0 0.000000+0          0          6        331         15923733102
 1.000000-5 3.000000-1 1.000000+2 1.000000+3 1.000000+4 3.750000+4923733102
 9.000000+4 1.350000+5 2.050000+5 2.750000+5 4.125000+5 6.250000+5923733102
 9.500000+5 1.500000+6 2.300000+6 1.000000-5 3.000000-1 1.000000+2923733102
 1.000000+3 1.000000+4 3.750000+4 9.000000+4 1.350000+5 2.050000+5923733102
 2.750000+5 4.125000+5 6.250000+5 9.500000+5 1.325000+6 1.900000+6923733102
 2.900000+6 4.600000+6 6.750000+6 9.500000+6 1.350000+7 1.550000+7923733102
 2.000000+7 1.756601-8 1.756601-8 2.307246-8 3.049692-8 6.589630-8923733102
 4.306795-8 3.891639-8 9.316174-8 6.233225-8 5.014352-8 1.132461-7923733102
 7.828227-8 7.217931-8 7.466687-8 8.420934-8 8.280880-8 7.986202-8923733102
 7.099759-8 6.653936-8 3.729051-8 6.229219-8 1.756601-8 1.756601-8923733102
 2.307246-8 3.049692-8 6.589630-8 4.306795-8 3.891639-8 9.316174-8923733102
 6.233225-8 5.014352-8 1.132461-7 7.828227-8 7.217931-8 7.466687-8923733102
 8.420934-8 8.280880-8 7.986202-8 7.099759-8 6.653936-8 3.729051-8923733102
 6.229219-8 4.462786-7 4.462786-7 2.837186-6 2.806743-6 3.223859-6923733102
 2.752009-6 3.010122-6 2.961952-6 2.971455-6 3.034000-6 2.467666-6923733102
 2.328588-6 2.168345-6 2.148547-6 2.101845-6 1.956477-6 1.868093-6923733102
 1.606852-6 1.416914-6 7.693215-7 1.329355-6 4.845855-7 4.845855-7923733102
 3.132689-6 3.256939-6 3.913624-6 3.392811-6 3.695923-6 3.566611-6923733102
 3.470179-6 3.539297-6 2.892230-6 2.746992-6 2.550140-6 2.519340-6923733102
 2.457250-6 2.281159-6 2.177644-6 1.876005-6 1.653521-6 8.985493-7923733102
 1.551922-6 5.522040-7 5.522040-7 3.438471-6 3.867718-6 6.089964-6923733102
 4.935918-6 5.382186-6 4.995669-6 4.606240-6 4.582258-6 3.812452-6923733102
 3.710496-6 3.431793-6 3.369575-6 3.273834-6 3.032319-6 2.894236-6923733102
 2.499881-6 2.202559-6 1.199479-6 2.068463-6 5.048252-7 5.048252-7923733102
 3.168782-6 3.887495-6 5.527578-6 7.574905-6 8.200554-6 7.254595-6923733102
 5.857552-6 5.588665-6 4.658222-6 4.595782-6 4.214876-6 4.082617-6923733102
 3.923727-6 3.600930-6 3.433465-6 2.983538-6 2.618851-6 1.420304-6923733102
 2.460591-6 4.740287-7 4.740287-7 3.176372-6 3.970602-6 5.519170-6923733102
 7.847616-6 1.357481-5 9.459232-6 6.909114-6 6.411306-6 5.445994-6923733102
 5.421315-6 4.991993-6 4.829055-6 4.591193-6 4.185679-6 3.990721-6923733102
 3.479556-6 3.045590-6 1.645466-6 2.863388-6 4.904450-7 4.904450-7923733102
 3.224962-6 3.648894-6 4.494154-6 6.372683-6 8.250372-6 1.645021-5923733102
 8.292258-6 6.662842-6 5.744732-6 5.594043-6 5.180347-6 5.028297-6923733102
 4.732302-6 4.281294-6 4.081069-6 3.583898-6 3.108881-6 1.671810-6923733102
 2.921791-6 5.179343-7 5.179343-7 3.228833-6 3.502100-6 4.376715-6923733102
 5.429168-6 6.657624-6 8.720918-6 1.189946-5 7.362235-6 6.140660-6923733102
 5.733672-6 5.281981-6 5.181792-6 4.906823-6 4.452004-6 4.239362-6923733102
 3.730595-6 3.251817-6 1.751799-6 3.052687-6 4.618064-7 4.618064-7923733102
 3.101498-6 3.554217-6 4.180992-6 4.990470-6 5.929547-6 6.708322-6923733102
 7.279768-6 1.250699-5 7.214284-6 6.057901-6 5.665926-6 5.554190-6923733102
 5.173895-6 4.593276-6 4.355488-6 3.817554-6 3.360511-6 1.809253-6923733102
 3.170271-6 4.124357-7 4.124357-7 2.762236-6 3.088725-6 3.571202-6923733102
 4.445007-6 5.195247-6 6.309300-6 6.113009-6 7.239937-6 1.216836-5923733102
 6.836683-6 5.833379-6 5.606253-6 5.194231-6 4.520603-6 4.285438-6923733102
 3.770336-6 3.302310-6 1.771589-6 3.122687-6 2.810233-7 2.810233-7923733102
 2.155689-6 2.505228-6 2.936489-6 4.045053-6 4.836048-6 5.855992-6923733102
 5.702776-6 5.664378-6 6.012916-6 1.167280-5 7.220186-6 6.264512-6923733102
 5.695446-6 4.919647-6 4.676586-6 4.184389-6 3.636112-6 1.925731-6923733102
 3.444378-6 1.203301-7 1.203301-7 1.262618-6 1.521675-6 1.824114-6923733102
 2.555311-6 2.803257-6 3.856687-6 3.718516-6 4.195658-6 4.391636-6923733102
 6.740464-6 1.247719-5 8.932404-6 6.456683-6 5.685773-6 5.465747-6923733102
 4.982233-6 4.350095-6 2.231388-6 4.170753-6 6.444943-8 6.444943-8923733102
 9.519869-7 1.166006-6 1.423474-6 2.034209-6 2.151504-6 3.242044-6923733102
 3.021133-6 3.535045-6 3.502486-6 5.328298-6 6.687729-6 1.212462-5923733102
 9.427675-6 6.431210-6 6.282109-6 5.700808-6 4.973386-6 2.534591-6923733102
 4.757867-6                                                       923733102
                                                                  923733  0
                                                                  9237 0  0
                                                                     0 0  0
 9.32370E+4 2.35011E+2          0          1         34         109346 1451
 0.0        1.00000E+0          0          0          0          69346 1451
 1.00000E+0 2.00000E+7          0          0         10       20029346 1451
 3.00000E+2 0.0                 1          0        345          39346 1451
 93-Np-237 FEI/LANL   EVAL-May96 K.Zolotarev et al.,P.Young et al.9346 1451
                      DIST-Feb2004                                9346 1451
----IRDF-2002         MATERIAL 9346                               9346 1451
-----INCIDENT NEUTRON DATA                                        9346 1451
------ENDF-6 FORMAT                                               9346 1451
***************************************************************** 9346 1451
 93-Np-237 FEI        EVAL-May96  K.Zolotarev et al.              9346 1451
                      DIST-Sep03                                  9346 1451
----BROND-3           MATERIAL 9346         Revision 2, Oct. 2003 9346 1451
***************************************************************** 9346 1451
------Russian Reactor Dosimetry File RRDF-2002                    9346 1451
***************************************************************** 9346 1451
Authors of evaluation: K.I.Zolotarev, S.A.Badikov, A.I.Blokhin,   9346 1451
                       A.V.Ignatyuk                               9346 1451
***************************************************************** 9346 1451
                                                                  9346 1451
                    ----- MF=2 MT=151 -----                       9346 1451
                                                                  9346 1451
RESONANCE PARAMETERS                                              9346 1451
--------------------                                              9346 1451
Resolved Resonance Region : 1.0E-5 - 600 eV                       9346 1451
    Resolved MLBW resonance parameters up to 600 eV were based on 9346 1451
on data from ref. [1].                                            9346 1451
    A statistical method  of  the resonance analysis showed  that 9346 1451
the missing of the resonances becomes essential above  the energy 9346 1451
of 150 eV. However only weak resonances are missed with the neut- 9346 1451
ron widths less than the average width by the factor 5-10.        9346 1451
    The resonances with widths close to average or above it  were 9346 1451
identified without  the noticeable  missing up  to the  energy of 9346 1451
600 eV.  So as these resonances  give dominant contribution  into 9346 1451
the neutron cross sections the resolved resonances region was ex- 9346 1451
panded up to 600 eV  and  average contribution of the missed weak 9346 1451
resonances was taken into account by addition  of  relevant cross 9346 1451
section background in the file MF=3.  (For fission cross sections 9346 1451
this background is equal to zero).                                9346 1451
                                                                  9346 1451
Unresolved Resonance Region : 600 eV - 6 keV                      9346 1451
    For the optical-statistical calculation of the cross sections 9346 1451
in the unresolved resonance region were used the following para-  9346 1451
meters:                                                           9346 1451
       Average resonance spacing - D0 = 0.57+-0.03 eV             9346 1451
       Neutron strengh function  - S0 = (0.97+-0.07)*104          9346 1451
       Average radiation width   - Gg = 40.0+-1.2 meV             9346 1451
This values were obtained  as  a result of  the analysis resolwed 9346 1451
MLBW resonance parameters taking into account the missed resonan- 9346 1451
ces correction.                                                   9346 1451
    Considerable attention was paid to  the  analysis of fission  9346 1451
widths and  elimination  of contradictions  in the description of 9346 1451
intermediate structure of averaged fission cross sections for the 9346 1451
neutron energies above 100 eV. The new experimental data [2-4] on 9346 1451
fission cross sections were included in the analysis.  They  were 9346 1451
obtained by LANL and JAERI physicists on their neutron spectrome- 9346 1451
ter [2,3]  and by Dubna-Obninsk collaboration on the pulsed reac- 9346 1451
tor of JINR [4].                                                  9346 1451
    The resulting set of resonance parameters reproduces well the 9346 1451
observed intermediate structure of fission cross sections (100 eV-9346 1451
6 keV).  Corresponding group  fission cross sections are 30-50 %  9346 1451
lower than  JENDL-3.2  evaluation [5]  and  approximately  twice  9346 1451
higher than ENDF/B-VI ones [6].  Corrected fission widths of  the 9346 1451
neutron resonances below 100 eV also result in fission cross sec- 9346 1451
tions which are lower than JENDL-3.2 values [5].                  9346 1451
    Calculated  from  the adopted resonance parameters  2200 M/S  9346 1451
cross sections and resonance integrals are agree rather well with 9346 1451
the recommended evaluated experimental data [1] :                 9346 1451
                                                                  9346 1451
  Value             This evaluation    Experimental data [1]      9346 1451
  Total               190.23 barn                                 9346 1451
  Elastic             14.17  barn                                 9346 1451
  Fission             0.0215 barn      0.0215+-0.0024 barn        9346 1451
  Capture             176.04 barn       175.9+-2.9    barn        9346 1451
  Fission res. int.   6.930  barn         6.9+-1.0    barn        9346 1451
  Capture res. int.   642.30 barn         640+-50     barn        9346 1451
                                                                  9346 1451
                    ----- MF=3 MT= 18 -----                       9346 1451
                                                                  9346 1451
FISSION CROSS SECTIONS                                            9346 1451
----------------------                                            9346 1451
    Np-237 fission cross sections from 1.0E-5 eV to 6 keV are re- 9346 1451
constructed from evaluated MLBW resolved and unresolved resonance 9346 1451
parameters.                                                       9346 1451
    For the evaluation of the  Np-237(n,f)  cross sections in the 9346 1451
neutron energy region of  6 keV - 20 MeV  available  experimental 9346 1451
data [7- 46] were analyzed.  During this  procedure  experimental 9346 1451
data  if it  was possible  were  corrected to the new recommended 9346 1451
cross section data for monitor reactions used in the measurements 9346 1451
and to the new recommended decay data.                            9346 1451
    The top priority was given to absolute measurements  where no 9346 1451
reference cross sections were used to determine the neutron  flux 9346 1451
and to time-of-flight experiments  with simultaneous registration 9346 1451
of the fission and monitoring reaction events.                    9346 1451
    In ten experiments the Np-237(n,f) cross sections were measu- 9346 1451
red relatively to the fission of U-235.  The use  as the standard 9346 1451
of the U-235 fission cross section from  ENDF/B-VI  [47]  instead 9346 1451
of the old one (ENDF/B-V) results in decrease of the  Np-237(n,f) 9346 1451
cross sections in average  by 2 %  in 0.1 - 2.0 MeV  range and by 9346 1451
1.5 % in 2.0 - 3.0 MeV range.                                     9346 1451
    In the work of Kupriyanov et al. [20] the  Np-237(n,f)  cross 9346 1451
sections  were  measured against  the fission  cross sections  of 9346 1451
Pu-239. At the present time there are no recommended Pu-239 cross 9346 1451
sections as a standard data.  The data  from  two libraries were  9346 1451
used to get  absolute values: JENDL-3.2 [48] and ENDF/B-VI [49].  9346 1451
Below 1.6 MeV the results of Ref. [20] disagree with the integral 9346 1451
experiments no matter which data are used for  Pu-239(n,f) .  The 9346 1451
analysis of Ref. [20] data demonstrated that  the  good agreement 9346 1451
with  the rest of  data may be obtained for the energy  region of 9346 1451
1.6-7.0 MeV if  Pu-239(n,f) reaction cross section data are taken 9346 1451
from JENDL-3.2 library.                                           9346 1451
    The ratio of Np-237 to U-235  fission cross sections measured 9346 1451
by Behrens et al. [23] in 0.11 - 18.89 MeV energy range was multi-9346 1451
plied by factor of 1.051. This normalization factor was  obtained 9346 1451
from the values of the functional <sNp237(n,f)> / <sU235(n,f)> in 9346 1451
the 1 - 5 MeV  energy range evaluated before.  There is a  lot of 9346 1451
experimental data which are in  good agreement in  this interval. 9346 1451
The ratio of the Np-237(n,f) evaluated averaged cross section for 9346 1451
the neutron spectrum of  Cf-252 spontaneous  fission, known  from 9346 1451
many works, to the averaged data  of Behrens et al. [23] is equal 9346 1451
to 1.055 that confirms our renormalization of these data.         9346 1451
    The analysis of experimental data on the ratio of Np-237  and 9346 1451
U-235 fission cross sections  indicates that the relative  energy 9346 1451
trends of <sNp237(n,f)> / <sU235(n,f)> measured by Terayama et al.9346 1451
[36] in the energy range of 4.19 - 6.99 MeV  and  by Goverdovskiy 9346 1451
et al. [32]  in the 5.66 - 10.06 MeV energy range coincides  with 9346 1451
the  results  of other  authors.  To make them agree  in absolute 9346 1451
values Terayama et al. and Goverdovskiy et al. results were multi-9346 1451
plied by 0.96 and 1.079, respectively.  The data  on  the  Np-237 9346 1451
fission cross sections obtained by Terayama et al. on T(p,n)He-3  9346 1451
neutron source for the  0.70 - 2.99 MeV  energy range are in good 9346 1451
agreement with the results of Refs. [38-40]. So they were correc- 9346 1451
ted only according to  the new data  on the  monitoring  reaction 9346 1451
U-235(n,f)  [46].                                                 9346 1451
    Experimental data of Meadows et al. [26] as well as  that  of 9346 1451
Kupriyanov et al. [20] are  systematically  too low  below  1 MeV 9346 1451
that   contradicts   to  the   evaluated   integral   experiments 9346 1451
available.  Above 1 MeV  Meadows et al. [26] data well agree with 9346 1451
the results of other authors  so they were included in the final  9346 1451
evaluation only above this energy. Data of Kupriyanov et al. [20] 9346 1451
were taken into account in the evaluation above 1.6 MeV.          9346 1451
    Statistical analysis  of the input experimental cross section 9346 1451
data for  Np-237(n,f) reaction in 0.1-20 MeV neutron energy range 9346 1451
was carried out  by means of PADE-2 code [50].  Rational function 9346 1451
was used as the model function [51].                              9346 1451
    Averaged cross sections for  U-235  thermal fission [52]  and 9346 1451
Cf-252 spontaneous fission neutron spectra [53] calculated  from  9346 1451
the evaluated Np-237(n,f) excitation function are the following:  9346 1451
                                                                  9346 1451
  9346 1451
    TYPE OF SPECTRUM    <CS>,mb (calc.)    <CS>, mb (measured)  9346 1451
  9346 1451
 U-235 neutron fission     1356.2        1353.0 +- 24.0   [54]  9346 1451
                                         1350.0 +- 24.0   [55]  9346 1451
  9346 1451
 CF-252 spont. fission     1359.9        1361.0 +- 21.6   [55]  9346 1451
  9346 1451
                                                                  9346 1451
    Evaluated Np-237 fission cross section is in a good agreement 9346 1451
with the integral experiment data both  for U-235 thermal fission 9346 1451
neutron spectrum and for Cf-252 spontaneous fission neutron spec- 9346 1451
trum.                                                             9346 1451
    The evaluated Np-237(n,f) cross section agrees within the ac- 9346 1451
curacy with integral experimental data for SIGNA-SIGMA field.     9346 1451
    The Np-237 fission cross sections measured in  CFRMF facility 9346 1451
are lower than calculated ones for any fission cross section eva- 9346 1451
luation. This apparently indicates the necessity of more detailed 9346 1451
and careful analysis of the CFRMF experimental  data  accuracy.   9346 1451
                                                                  9346 1451
                    ----- MF=33 MT= 18-----                       9346 1451
    Uncertainties in  the evaluated excitation function  for the  9346 1451
Np-237(n,f) reaction are given in the two independent matrixes.   9346 1451
    In the  energy range 1.000E-05 eV - 0.1 MeV uncertainties are 9346 1451
given in the form of diagonal matrix  of uncertainties  for  6-th 9346 1451
neutron energy intervals (LB=1)                                   9346 1451
    In the energy range 0.1 - 20 MeV  uncertainties are presented 9346 1451
in  the form  of relative covariance matrix  for the  48 -neutron 9346 1451
energy groups (LB=5).  Eigenvalues  of this  6-th digits relative 9346 1451
covariance matrix are the following:                              9346 1451
                                                                  9346 1451
    1.49939E-07    1.52531E-07    1.53109E-07    1.53318E-07      9346 1451
    1.54049E-07    1.55336E-07    1.56641E-07    1.58051E-07      9346 1451
    1.58347E-07    1.60763E-07    1.63743E-07    1.64776E-07      9346 1451
    1.67513E-07    1.69820E-07    1.73330E-07    1.75751E-07      9346 1451
    1.77557E-07    2.05370E-07    2.66364E-07    2.70615E-07      9346 1451
    2.75409E-07    2.80956E-07    2.93138E-07    3.15582E-07      9346 1451
    3.35128E-07    4.86134E-07    7.89097E-07    3.07453E-06      9346 1451
    1.65989E-05    9.67332E-05    1.24983E-04    1.51845E-04      9346 1451
    4.05040E-04    4.14092E-04    5.11273E-04    5.63190E-04      9346 1451
    7.24776E-04    8.29295E-04    9.68770E-04    9.77438E-04      9346 1451
    1.07709E-03    1.13969E-03    1.57856E-03    1.84372E-03      9346 1451
    2.04068E-03    2.65362E-03    3.89983E-03    1.60049E-02      9346 1451
                                                                  9346 1451
References:                                                       9346 1451
 1. S.F.Mughabghab  Neutron Cross Sections. N.Y.-London, Academic 9346 1451
    Press, 1984, v.1, part B.                                     9346 1451
 2. J.Kimura  Nucl. Sci. Techn., v. 30, p.863, 1993               9346 1451
 3. A.D.Carlson et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451
    Science and Technology, Gatlinburg, Tennessee, USA, 9-13 May  9346 1451
    1994, Ed. J.Dickens, ANS, Oak Ridge, 1994, v.1, p.40          9346 1451
 4. A.A.Goverdovskiy et al.  Jadernaja Fizika (Rus.), v.58, p.27, 9346 1451
    January 1995                                                  9346 1451
 5. Y.Kikuchi  Proc. of an Int. Conf. on Nuclear Data for Science 9346 1451
    and Technology, Gatlinburg, Tennessee, USA, 9-13 May 1994,    9346 1451
    Ed. J.Dickens, ANS, Oak Ridge, 1994, v.2, p.692               9346 1451
 6. P.Young, E.Arthur, F.Mann  ENDF/B-VI library, MAT 9346, 1990  9346 1451
 7. E.D.Klema  Phys. Rev., v.72, p.88, July 1947                  9346 1451
 8. R.L.Henkel  Report LA-2114, 1957                              9346 1451
    R.L.Henkel  Report LA-2122, June 1957                         9346 1451
 9. A.N.Protopopov et al. Atomnaja Energija (Sov.),v.4,p.190,1958 9346 1451
10. H.W.Schmitt, R.B.Murrey  Phys. Rev., v.116, p.1575, 1959      9346 1451
10. R.B.Murrey, H.W.Schmitt  Phys. Rev., v.115, p.1707, 1959      9346 1451
11. B.M.Gokhberg et al.  Doklady Akademii Nauk SSSR, v.128,(6),   9346 1451
    p.1157, 1959                                                  9346 1451
12. V.M.Pankratov et al. Atomnaya Energiya, v.9, p.399, 1960 ???  9346 1451
13. V.M.Pankratov  Atomnaja Energiya (Sov.), v.14, p.177, 1963    9346 1451
14. J.A.Grundl  Nucl. Sci. Eng., v.30, p.39, October 1967         9346 1451
15. P.H.White, G.P.Warner  J. of Nucl. Energy, v.21, p.671, 1967  9346 1451
16. W.E.Stein, R.K.Smith, H.L.Smith   Proceedings Conf. Neutron   9346 1451
    Cross Sections and Technology, Washington, D.C., March 4-7,   9346 1451
    1968, NBS Special Publication 299 , p.627 ,  U.S. National    9346 1451
    Bureau of Standards, 1968                                     9346 1451
17. R.H.Iyer, R.Sampathkumar Proc. of 12th Symp. on Nuclear Phys. 9346 1451
    and Solid State Phys., Roorkee, India, 28-31 December 1969,   9346 1451
    v.2, p.289 ;                                                  9346 1451
    R.H.Iyer, R.Sampathkumar  Prog. Report BARC/I-79, p.55, 1970  9346 1451
18. W.K.Brown et al. Nucl. Phys., vA156, p.609, November 1970     9346 1451
19. R.J.Jiacoletti, W.K.Brown, H.G.Olson.  Nucl. Sci. Eng., v.48, 9346 1451
    p.412, August 1972                                            9346 1451
20. V.M.Kupriyanov et al.  Atomnaya Energiya, v.45, p.440, 1978   9346 1451
21. D.J.Grady et al.  Proc. of an Int. Conf. on Nuclear Cross     9346 1451
    Sections for Technology, Knoxville, Tennessee, 22-26 October  9346 1451
    1979, NBS Special Publ.594, p.976, September 1980             9346 1451
22. R.Arlt et al.   Report ZFK-410, p.122, January 1980 ;         9346 1451
    R.Arlt et al.   Kernenergie, v.24, p.48, February 1981        9346 1451
23. J.W.Behrens, J.C.Browne, J.C.Malden   Nucl. Sci. Eng., v.80,  9346 1451
    p.393, 1982                                                   9346 1451
24. M.Cance, G.Grenier   Proc of an Int. Conf. on Nuclear Data    9346 1451
    for Science and Technology, Antwerp, 6-10 September 1982,     9346 1451
    Dr. Reidel Publishing Company, 1983, p.51                     9346 1451
25. M.Varnagy, S.Juhasz, J.Csikai  Nucl. Instr. Meth., v.196,     9346 1451
    p.465, May 1982                                               9346 1451
26. J.W.Meadows  Nucl. Sci. Eng., v.85, p.271, November 1983      9346 1451
27. I.D.Alkhazov, E.A.Ganza, L.V.Drapchinskij et al. Proceedings  9346 1451
    of the 3-d All-union Conference  on the  Neutron Radiation    9346 1451
    Metrology at Reactors and Accelerators, v.2, p.201, Moscow,   9346 1451
    CNIIATOMINFORM, 1983                                          9346 1451
28. K.R.Zasadny et al.  J. Trans. Amer. Nucl. Soc., v.47, p.425,  9346 1451
    November 1984                                                 9346 1451
29. Wu Jingxia et al. Chinese J. of Nuclear Physics, v.6, p.369,  9346 1451
    November 1984                                                 9346 1451
30. I.Garlea et al. Revue Roumaine de Physique, v.29, p.421, 1984 9346 1451
31. A.A.Goverdovski et al.  Preprint FEI-1552, Obninsk, 1984      9346 1451
32. A.A.Goverdovskij et al. VANT, Serija: Yadernye Konstanty,     9346 1451
    issue 3(57), p.13, September 1984                             9346 1451
33. K.Kanda et al.  Report JAERI-M-85-035, p.220, 1985            9346 1451
34. R.Arlt et al.  Isotopenpraxis, v.21, p.344, 1985              9346 1451
35. A.A.Goverdovskij et al.  Atomnaja Energija (Sov.), v.58, no.2,9346 1451
    p.137, February 1985                                          9346 1451
36. H.Terayama et al.   Progress Report NETU-47, 1986 ;           9346 1451
    H.Terayama et al.   Progress Report NEANDC(J)-122, Sep. 1986  9346 1451
37. I.D.Alkhazov et al.   VANT, Serija: Yadernye Konstanty, v.4,  9346 1451
    p.19, December 1986                                           9346 1451
38. V.A.Kalinin et al.  VANT, Serija: Yadernye Konstanty, v.4,    9346 1451
    p.3, December 1987                                            9346 1451
39. P.W.Lisowski et al.   Proc. of an Int. Conf. on Nuclear Data  9346 1451
    for Science and Technology, Mito, Japan, 30 May-3 June 1988,  9346 1451
    pp. 97-99                                                     9346 1451
40. J.W.Meadows  Annals of Nucl. Energy, v.15, p.421, August 1988 9346 1451
41. J.W.Meadows, D.L.Smith, L.P.Geraldo   Annals of Nucl. Energy, 9346 1451
    v.16, p.471, September 1989                                   9346 1451
36. L.Desdin, S.Szegedi, J.Csikai  Acta Physica Hungaria, v.65,   9346 1451
    p.271, 1989                                                   9346 1451
43. P.W.Lisowski et al.  Proc. of the Conference on Fifty Years   9346 1451
    with Nuclear Fission, NIST, Gaithersburg, MD, 1989, p.443     9346 1451
44. K.Merla et al.   Proc. of an Int. Conf. on Nuclear Data for   9346 1451
    Science and Technology, Julich, FRG, 13-17 May 1991. Springer 9346 1451
    Verlag, Berlin - Heidelberg, 1992, p.510-513                  9346 1451
45. A.D.Carlson et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451
    Science and Technology, Gatlinburg, Tennessee, USA, May 9-13, 9346 1451
    1994, Vol.1, pp. 40-42                                        9346 1451
46. T.Iwasaki et al. Nucl. Sci. Technology, v.36, p.127, 1999     9346 1451
47. L.Weston et al. ORNL,LANL eval. April 1989, ENDF/B-VI Library 9346 1451
    MAT=9228;                                                     9346 1451
    H.Conde ed. Nuclear Data Standards for Nuclear Measurements,  9346 1451
    Report NEANDC-311 U, p.51-58, OECD, Paris, 1992.              9346 1451
48. M.Kawai et al. eval. NAIG, JENDL-3.2 library, MAT=9437,       9346 1451
    Rev. 2, February 1993                                         9346 1451
49. P.Young et al. eval. ORNL, LANL , ENDF/B-VI library, MAT=9437,9346 1451
    Rev.1, January 1993                                           9346 1451
50. S.A.Badikov et al.  Preprint FEI-1686, Obninsk, 1985          9346 1451
51. S.Badikov, N.Rabotnov, K.Zolotarev  Proc. of NEANSC Speciali- 9346 1451
    st's Meeting on Evaluation and Processing of Covariance Data, 9346 1451
    Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105                9346 1451
52. L.W.Weston et al.  Evaluated Neutron Data  for  Uranium-235,  9346 1451
    ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989    9346 1451
53. W.Mannhart  IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987        9346 1451
54. W.Mannhart  Prog. Report INDC(Ger)-045, pp.40-43, June 1999   9346 1451
55. W.Mannhart  Validation of Differential Cross Sections with    9346 1451
    Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 9346 1451
    September 2002                                                9346 1451
******************************************************************9346 1451
                    ----- MF=3 MT= 1 -----                        9346 1451
                    ----- MF=3 MT= 2 -----                        9346 1451
 93-NP-237 LANL       EVAL-APR90 P.YOUNG, E.ARTHUR, F.MANN        9346 1451
                      DIST-SEP91 REV1-JUL91            19910806   9346 1451
----ENDF/B-VI         MATERIAL 9346         REVISION 1            9346 1451
 MT=1   Neutron Total Cross Section.  Derrien evaluation (De80,   9346 1451
          De89) used from 1.0E-5 eV to 0.008 MeV. Above 0.008 MeV,9346 1451
          coupled channel optical model calculations (ECIS code)  9346 1451
          with modified Lagrange (Ha80) potential used as prior   9346 1451
          in a covariance analysis of measurement by Lisowski et  9346 1451
          al. from 3 to 20 MeV (Li90).                            9346 1451
 MT=2   All Energies, based on subtraction of MT=4,16,17,18,37,   9346 1451
          and 102 from MT=1.                                      9346 1451
******************************************************************9346 1451
 The cross sections for total and elastic were modified between   9346 1451
 6 and 10 Kev for continuity with the data of Zolatarev file 18.  9346 1451
 The number of interpolation schemes is set to 1.                 9346 1451
                                                                  9346 1451
******************************************************************9346 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9346 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9346 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9346 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9346 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9346 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9346 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9346 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9346 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9346 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9346 1451
 ***************** Program FIXUP (Version 2002-1) ****************9346 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9346 1451
 Corrected Thresholds-----------------------------------------Yes 9346 1451
 Extended Cross Sections to 20 MeV----------------------------No  9346 1451
 Allow Cross Section Deletion---------------------------------No  9346 1451
 Allow Cross Section Reconstruction---------------------------No  9346 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9346 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9346 1451
 Deleted Duplicate Points-------------------------------------Yes 9346 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9346 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9346 1451
 Allow Creation of Missing Sections---------------------------No  9346 1451
 Allow Insertion of Energy Points-----------------------------No  9346 1451
 Create Uniform Energy Grid-----------------------------------No  9346 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9346 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9346 1451
 Unshielded Group Averages Using  640 Groups                      9346 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9346 1451
                                1        451        352          19346 1451
                                3         18        217          19346 1451
                               33         18        220          19346 1451
                                                                  9346 1  0
                                                                  9346 0  0
 9.32370E+4 2.35011E+2          0          0          0          09346 3 18
 1.96370E+8 1.96370E+8          0          0          1        6419346 3 18
        641          1                                            9346 3 18
 .000100000 .350176171 .000105000 .342008374 .000110000 .3342488439346 3 18
 .000115000 .327098660 .000120000 .318739733 .000127500 .3094655159346 3 18
 .000135000 .300992335 .000142500 .293205616 .000150000 .2847781559346 3 18
 .000160000 .275998821 .000170000 .268010600 .000180000 .2606779419346 3 18
 .000190000 .253855440 .000200000 .247626226 .000210000 .2417655159346 3 18
 .000220000 .236321789 .000230000 .231263736 .000240000 .2253264939346 3 18
 .000255000 .218802285 .000270000 .213767829 .000280000 .2081634209346 3 18
 .000300000 .201328458 .000320000 .195139402 .000340000 .1894710319346 3 18
 .000360000 .184249952 .000380000 .179483628 .000400000 .1745142929346 3 18
 .000425000 .169428883 .000450000 .164781972 .000475000 .1605217589346 3 18
 .000500000 .156515349 .000525000 .152853437 .000550000 .1493940069346 3 18
 .000575000 .146176393 .000600000 .142878415 .000630000 .1394855349346 3 18
 .000660000 .136380844 .000690000 .133402485 .000720000 .1302330909346 3 18
 .000760000 .126834596 .000800000 .123678992 .000840000 .1207868839346 3 18
 .000880000 .118037967 .000920000 .115517707 .000960000 .1131039389346 3 18
 .001000000 .110601667 .001050000 .107989435 .001100000 .1055404969346 3 18
 .001150000 .103283729 .001200000 .100621374 .001275000 .0976949739346 3 18
 .001350000 .095019215 .001425000 .092521895 .001500000 .0898685859346 3 18
 .001600000 .087094186 .001700000 .084561306 .001800000 .0822185809346 3 18
 .001900000 .080078306 .002000000 .078093524 .002100000 .0762307479346 3 18
 .002200000 .074524537 .002300000 .072892776 .002400000 .0710302379346 3 18
 .002550000 .068959741 .002700000 .067339620 .002800000 .0655782189346 3 18
 .003000000 .063409053 .003200000 .061434563 .003400000 .0596289969346 3 18
 .003600000 .057982841 .003800000 .056463429 .004000000 .0548803339346 3 18
 .004250000 .053273676 .004500000 .051806876 .004750000 .0504353269346 3 18
 .005000000 .049175399 .005250000 .048005871 .005500000 .0468991099346 3 18
 .005750000 .045882941 .006000000 .044814857 .006300000 .0437458799346 3 18
 .006600000 .042737357 .006900000 .041807271 .007200000 .0407925889346 3 18
 .007600000 .039704533 .008000000 .038708909 .008400000 .0377738219346 3 18
 .008800000 .036902821 .009200000 .036092823 .009600000 .0353216709346 3 18
 .010000000 .034522688 .010500000 .033677652 .011000000 .0329042329346 3 18
 .011500000 .032166171 .012000000 .031319810 .012750000 .0303825409346 3 18
 .013500000 .029514065 .014250000 .028717004 .015000000 .0278662799346 3 18
 .016000001 .026972189 .017000001 .026151028 .017999999 .0253998709346 3 18
 .018999999 .024708250 .020000000 .024058602 .021000000 .0234632449346 3 18
 .022000000 .022901813 .023000000 .022378150 .024000000 .0217623259346 3 18
 .025500000 .021096998 .027000001 .020569412 .028000001 .0199954169346 3 18
 .029999999 .019288082 .032000002 .018643086 .034000002 .0180538369346 3 18
 .035999998 .017515017 .037999999 .017016249 .039999999 .0164941339346 3 18
 .042500000 .015963022 .045000002 .015477840 .047499999 .0150259329346 3 18
 .050000001 .014610201 .052499998 .014224066 .055000000 .0138574899346 3 18
 .057500001 .013520775 .059999999 .013166258 .063000001 .0128110779346 3 18
 .066000000 .012476011 .068999998 .012163493 .071999997 .0118242259346 3 18
 .075999998 .011460453 .079999998 .011127722 .083999999 .0108147869346 3 18
 .088000000 .010523590 .092000000 .010253805 .096000001 .0099960989346 3 18
 .100000001 .009730400 .104999997 .009449537 .109999999 .0091940429346 3 18
 .115000002 .008949963 .119999997 .008669777 .127499998 .0083608469346 3 18
 .135000005 .008075380 .142499998 .007814857 .150000006 .0075394509346 3 18
 .159999996 .007252813 .170000002 .006994118 .180000007 .0067619469346 3 18
 .189999998 .006553643 .200000003 .006362157 .209999993 .0061938979346 3 18
 .219999999 .006038306 .230000004 .005902139 .239999995 .0057570899346 3 18
 .254999995 .005614804 .270000011 .005523163 .280000001 .0054538269346 3 18
 .300000012 .005427699 .319999993 .005504867 .340000004 .0057314309346 3 18
 .360000014 .006195397 .379999995 .007082282 .400000006 .0092244399346 3 18
 .425000012 .015646793 .449999988 .036793181 .474999994 .0629237879346 3 18
 .500000000 .038738402 .524999976 .015415600 .550000012 .0081228339346 3 18
 .574999988 .005516367 .600000024 .004228004 .629999995 .0034929969346 3 18
 .660000026 .003067240 .689999998 .002794695 .720000029 .0025850439346 3 18
 .759999990 .002428195 .800000012 .002330981 .839999974 .0022797239346 3 18
 .879999995 .002270076 .920000017 .002304836 .959999979 .0023934069346 3 18
 1.00000000 .002585116 1.04999995 .002981698 1.10000002 .0037717339346 3 18
 1.14999998 .005572217 1.20000005 .016859753 1.27499998 .0991445359346 3 18
 1.35000002 .033359471 1.42499995 .061337230 1.50000000 .0187992959346 3 18
 1.60000002 .003197779 1.70000005 .001991181 1.79999995 .0023677749346 3 18
 1.89999998 .012818236 2.00000000 .004020261 2.09999990 .0012228239346 3 18
 2.20000005 .000943953 2.29999995 .000841134 2.40000010 .0007797769346 3 18
 2.54999995 .000745924 2.70000005 .000738829 2.79999995 .0007565669346 3 18
 3.00000000 .000834398 3.20000005 .001050203 3.40000010 .0017939659346 3 18
 3.59999990 .011680717 3.79999995 .091743390 4.00000000 .0036430029346 3 18
 4.25000000 .001531565 4.50000000 .001529026 4.75000000 .0111573539346 3 18
 5.00000000 .001980189 5.25000000 .003737615 5.50000000 .0759425739346 3 18
 5.75000000 .170694857 6.00000000 .005174401 6.30000019 .0095837429346 3 18
 6.59999990 .005934798 6.90000010 .003879601 7.19999981 .0247172669346 3 18
 7.59999990 .004021327 8.00000000 .010239189 8.39999962 .0025321519346 3 18
 8.80000019 .040897396 9.19999981 .009833115 9.60000038 .0008205629346 3 18
 10.0000000 .001721497 10.5000000 .037224375 11.0000000 .0217181639346 3 18
 11.5000000 .000693415 12.0000000 .001188167 12.7500000 .0016957539346 3 18
 13.5000000 .000370329 14.2500000 .000461172 15.0000000 .0021672699346 3 18
 16.0000000 .005761308 17.0000000 .006465396 18.0000000 .0011551539346 3 18
 19.0000000 .005610998 20.0000000 .002301894 21.0000000 .0092803679346 3 18
 22.0000000 .010744339 23.0000000 .004631749 24.0000000 .0492887639346 3 18
 25.5000000 .220758492 27.0000000 .005746784 28.0000000 .0130291269346 3 18
 30.0000000 .703819197 32.0000000 .007119277 34.0000000 .0060040889346 3 18
 36.0000000 .334555382 38.0000000 2.71431094 40.0000000 1.241150879346 3 18
 42.5000000 .056762665 45.0000000 .246180569 47.5000000 .0284064919346 3 18
 50.0000000 .217279104 52.5000000 .007701922 55.0000000 .0039022969346 3 18
 57.5000000 .013388890 60.0000000 .014392911 63.0000000 .0075685689346 3 18
 66.0000000 .012411994 69.0000000 .002766569 72.0000000 .0017795579346 3 18
 76.0000000 .000883391 80.0000000 .007293836 84.0000000 .0061816369346 3 18
 88.0000000 .007212762 92.0000000 .001742301 96.0000000 .0011406509346 3 18
 100.000000 .007206525 105.000000 .014011554 110.000000 .0091785959346 3 18
 115.000000 .716468072 120.000000 .031411800 127.500000 .0054040429346 3 18
 135.000000 .007133476 142.500000 .003782246 150.000000 .0046127629346 3 18
 160.000000 .004272124 170.000000 .011623731 180.000000 .0474976779346 3 18
 190.000000 .165258393 200.000000 .540697276 210.000000 .0427710969346 3 18
 220.000000 .046115274 230.000000 .194836017 240.000000 .1362023409346 3 18
 255.000000 .034890906 270.000000 .039706012 280.000000 .0949865719346 3 18
 300.000000 .009460424 320.000000 .072428790 340.000000 .0097993869346 3 18
 360.000000 .153479242 380.000000 .010383970 400.000000 .0213377549346 3 18
 425.000000 .061291069 450.000000 .033054528 475.000000 .0289526589346 3 18
 500.000000 .010694900 525.000000 .010747355 550.000000 .0204234889346 3 18
 575.000000 .039419651 600.000000 .047147866 630.000000 .0474235209346 3 18
 660.000000 .047710784 690.000000 .048004194 720.000000 .0483567159346 3 18
 760.000000 .048767863 800.000000 .049279992 840.000000 .0498880019346 3 18
 880.000000 .050490191 920.000000 .051086299 960.000000 .0516761769346 3 18
 1000.00000 .051492531 1050.00000 .050563851 1100.00000 .0497096019346 3 18
 1150.00000 .048921004 1200.00000 .048020463 1275.00000 .0470578449346 3 18
 1350.00000 .046156813 1425.00000 .045360119 1500.00000 .0438091259346 3 18
 1600.00000 .041562433 1700.00000 .039503223 1800.00000 .0376042849346 3 18
 1900.00000 .035843944 2000.00000 .034705986 2100.00000 .0341605869346 3 18
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 2.58606- 4 2.47719- 4 2.49552- 4 2.60818- 4 2.72381- 4 2.74633- 4934633 18
 2.70133- 4 2.64278- 4 2.52348- 4 2.35029- 4 2.19388- 4 2.10192- 4934633 18
 2.08928- 4 2.15759- 4 2.30403- 4 2.51458- 4 2.74351- 4 2.89076- 4934633 18
 2.81580- 4 2.43112- 4 1.82232- 4 1.23164- 4 8.82266- 5 8.44909- 5934633 18
 1.05958- 4 1.42371- 4 1.85081- 4 2.28612- 4 2.70114- 4 3.08401- 4934633 18
 3.58860- 4 4.15907- 4 3.28465- 4 3.39717- 4 3.26799- 4 2.97683- 4934633 18
 2.64420- 4 2.45363- 4 2.59299- 4 2.94738- 4 2.98878- 4 2.64694- 4934633 18
 2.39961- 4 2.36401- 4 2.37467- 4 2.33676- 4 2.26134- 4 2.20733- 4934633 18
 2.24362- 4 2.41945- 4 2.71982- 4 3.01322- 4 3.06029- 4 2.67303- 4934633 18
 1.93877- 4 1.21155- 4 8.20412- 5 8.50317- 5 1.18749- 4 1.66778- 4934633 18
 2.16653- 4 2.61545- 4 2.98786- 4 3.28046- 4 3.57990- 4 3.80289- 4934633 18
 4.03814- 4 4.22809- 4 3.94479- 4 3.31138- 4 2.65890- 4 2.51579- 4934633 18
 3.05080- 4 3.36539- 4 2.89424- 4 2.41579- 4 2.36856- 4 2.50646- 4934633 18
 2.58092- 4 2.53699- 4 2.43778- 4 2.39098- 4 2.49783- 4 2.78957- 4934633 18
 3.14754- 4 3.28802- 4 2.94168- 4 2.15340- 4 1.32144- 4 8.55738- 5934633 18
 8.82762- 5 1.27665- 4 1.83830- 4 2.41225- 4 2.91240- 4 3.30538- 4934633 18
 3.58801- 4 3.82042- 4 3.88901- 4 4.84277- 4 4.91261- 4 4.40374- 4934633 18
 3.49179- 4 2.75765- 4 2.90154- 4 3.48240- 4 3.34665- 4 2.75003- 4934633 18
 2.43248- 4 2.45756- 4 2.59855- 4 2.70717- 4 2.75308- 4 2.77843- 4934633 18
 2.85341- 4 3.02338- 4 3.23710- 4 3.30579- 4 3.00692- 4 2.32563- 4934633 18
 1.54873- 4 1.03539- 4 9.49706- 5 1.23380- 4 1.73546- 4 2.31453- 4934633 18
 2.87969- 4 3.38421- 4 3.81074- 4 4.29495- 4 4.72217- 4 5.57329- 4934633 18
 5.61852- 4 4.83304- 4 3.46535- 4 2.68354- 4 3.23738- 4 3.75273- 4934633 18
 3.30999- 4 2.64206- 4 2.34578- 4 2.42221- 4 2.68234- 4 2.96892- 4934633 18
 3.19372- 4 3.32551- 4 3.37057- 4 3.34048- 4 3.20737- 4 2.89638- 4934633 18
 2.37446- 4 1.75519- 4 1.25445- 4 1.03707- 4 1.13744- 4 1.49423- 4934633 18
 2.01619- 4 2.62310- 4 3.25822- 4 3.88640- 4 4.77013- 4 5.82359- 4934633 18
 6.47331- 4 6.28928- 4 4.68876- 4 2.80692- 4 2.74535- 4 3.76575- 4934633 18
 3.79260- 4 3.00839- 4 2.37506- 4 2.23378- 4 2.48346- 4 2.91552- 4934633 18
 3.34110- 4 3.62044- 4 3.67635- 4 3.50775- 4 3.17798- 4 2.75835- 4934633 18
 2.28270- 4 1.78541- 4 1.35636- 4 1.10939- 4 1.11186- 4 1.36234- 4934633 18
 1.81486- 4 2.41036- 4 3.09547- 4 3.82894- 4 4.95670- 4 6.43341- 4934633 18
 7.07560- 4 6.18587- 4 3.85226- 4 2.53712- 4 3.15080- 4 3.68519- 4934633 18
 3.32449- 4 2.69756- 4 2.32252- 4 2.31316- 4 2.58765- 4 3.00781- 4934633 18
 3.42482- 4 3.69355- 4 3.69506- 4 3.38026- 4 2.81091- 4 2.14757- 4934633 18
 1.57354- 4 1.21495- 4 1.11377- 4 1.24852- 4 1.56848- 4 2.01899- 4934633 18
 2.55368- 4 3.13766- 4 3.74637- 4 4.66943- 4 5.87737- 4 7.05021- 4934633 18
 6.00325- 4 3.65527- 4 2.45354- 4 2.69836- 4 3.08775- 4 3.08003- 4934633 18
 2.81394- 4 2.52772- 4 2.38472- 4 2.47291- 4 2.80683- 4 3.29499- 4934633 18
 3.70382- 4 3.71371- 4 3.14563- 4 2.19576- 4 1.33640- 4 9.32197- 5934633 18
 1.02809- 4 1.45576- 4 2.01834- 4 2.58151- 4 3.07952- 4 3.49183- 4934633 18
 3.82085- 4 4.17720- 4 4.48871- 4 7.68546- 4 6.41955- 4 3.49689- 4934633 18
 2.04361- 4 2.22276- 4 2.79420- 4 3.09615- 4 3.05936- 4 2.84298- 4934633 18
 2.65430- 4 2.67335- 4 2.97649- 4 3.43520- 4 3.68953- 4 3.36988- 4934633 18
 2.48764- 4 1.51052- 4 9.46217- 5 9.63766- 5 1.41923- 4 2.07376- 4934633 18
 2.73942- 4 3.31237- 4 3.75312- 4 4.05894- 4 4.28578- 4 4.29956- 4934633 18
 8.20572- 4 6.63832- 4 3.85859- 4 2.29267- 4 2.05857- 4 2.44471- 4934633 18
 2.93239- 4 3.28924- 4 3.45667- 4 3.47058- 4 3.41336- 4 3.35468- 4934633 18
 3.27203- 4 3.03216- 4 2.53079- 4 1.86818- 4 1.30567- 4 1.05473- 4934633 18
 1.16279- 4 1.55568- 4 2.12285- 4 2.76912- 4 3.42950- 4 4.06580- 4934633 18
 4.92853- 4 5.90847- 4 8.28554- 4 6.74958- 4 4.22994- 4 2.53188- 4934633 18
 1.92841- 4 2.07941- 4 2.60164- 4 3.19879- 4 3.65641- 4 3.83410- 4934633 18
 3.68231- 4 3.25922- 4 2.69812- 4 2.12901- 4 1.63729- 4 1.28584- 4934633 18
 1.12792- 4 1.18833- 4 1.45362- 4 1.88447- 4 2.43400- 4 3.05984- 4934633 18
 3.72865- 4 4.76014- 4 6.11963- 4 7.49516- 4 6.08925- 4 4.15996- 4934633 18
 2.67354- 4 1.87169- 4 1.69933- 4 2.01766- 4 2.64388- 4 3.32589- 4934633 18
 3.73902- 4 3.59443- 4 2.86151- 4 1.87900- 4 1.12294- 4 8.49159- 5934633 18
 1.00836- 4 1.41343- 4 1.89882- 4 2.37043- 4 2.79133- 4 3.15565- 4934633 18
 3.47023- 4 3.86721- 4 4.31509- 4 6.36549- 4 5.39459- 4 3.99082- 4934633 18
 2.69959- 4 1.81145- 4 1.47503- 4 1.72772- 4 2.43875- 4 3.23360- 4934633 18
 3.56628- 4 3.08074- 4 2.00343- 4 1.00419- 4 5.98365- 5 8.23513- 5934633 18
 1.41452- 4 2.08688- 4 2.66476- 4 3.07850- 4 3.32256- 4 3.41989- 4934633 18
 3.34886- 4 3.02380- 4 5.57432- 4 4.93285- 4 3.90762- 4 2.86104- 4934633 18
 2.07425- 4 1.75261- 4 1.96933- 4 2.55178- 4 3.04268- 4 2.95024- 4934633 18
 2.20720- 4 1.28196- 4 7.22802- 5 7.26645- 5 1.15428- 4 1.76292- 4934633 18
 2.36288- 4 2.85266- 4 3.19732- 4 3.39905- 4 3.46390- 4 3.26467- 4934633 18
 5.20814- 4 4.90773- 4 4.23620- 4 3.40991- 4 2.65742- 4 2.19526- 4934633 18
 2.13609- 4 2.35429- 4 2.48687- 4 2.22352- 4 1.62363- 4 1.03524- 4934633 18
 7.51805- 5 8.40554- 5 1.20915- 4 1.72526- 4 2.28244- 4 2.81467- 4934633 18
 3.28873- 4 3.85829- 4 4.39684- 4 5.44209- 4 5.44137- 4 4.94518- 4934633 18
 4.07637- 4 3.07249- 4 2.25930- 4 1.89121- 4 1.91727- 4 1.97704- 4934633 18
 1.76050- 4 1.30261- 4 8.65519- 5 6.68813- 5 7.80470- 5 1.15832- 4934633 18
 1.72057- 4 2.38893- 4 3.10384- 4 4.17586- 4 5.51601- 4 6.17393- 4934633 18
 6.25532- 4 5.63821- 4 4.44969- 4 3.05856- 4 1.97860- 4 1.52979- 4934633 18
 1.54588- 4 1.57496- 4 1.36506- 4 1.01383- 4 7.50123- 5 7.28036- 5934633 18
 9.84158- 5 1.48103- 4 2.15526- 4 2.94497- 4 4.24091- 4 6.00908- 4934633 18
 7.01848- 4 6.99402- 4 6.10456- 4 4.54759- 4 2.85268- 4 1.63429- 4934633 18
 1.13789- 4 1.10696- 4 1.15827- 4 1.13531- 4 1.09834- 4 1.15820- 4934633 18
 1.37931- 4 1.77038- 4 2.30735- 4 2.95477- 4 4.06168- 4 5.64219- 4934633 18
 7.79050- 4 7.69067- 4 6.54107- 4 4.58342- 4 2.51297- 4 1.08549- 4934633 18
 5.80237- 5 7.41570- 5 1.16655- 4 1.60384- 4 1.97857- 4 2.30396- 4934633 18
 2.61252- 4 2.92868- 4 3.26450- 4 3.81206- 4 4.60124- 4 8.69495- 4934633 18
 8.52683- 4 6.90019- 4 4.27097- 4 1.77648- 4 4.11395- 5 3.38196- 5934633 18
 1.06327- 4 2.00867- 4 2.82548- 4 3.39481- 4 3.72530- 4 3.87081- 4934633 18
 3.88907- 4 3.77309- 4 3.51727- 4 9.64193- 4 9.01880- 4 6.57344- 4934633 18
 3.41072- 4 1.07066- 4 3.11372- 5 8.47740- 5 1.96925- 4 3.10075- 4934633 18
 3.95536- 4 4.45941- 4 4.64516- 4 4.58009- 4 4.14746- 4 3.25813- 4934633 18
 9.79517- 4 8.48395- 4 5.60783- 4 2.68204- 4 9.54045- 5 6.53598- 5934633 18
 1.30517- 4 2.32221- 4 3.30598- 4 4.06862- 4 4.56004- 4 4.79820- 4934633 18
 4.75834- 4 4.25947- 4 8.85772- 4 7.28861- 4 4.60726- 4 2.13428- 4934633 18
 6.81451- 5 3.16448- 5 7.13005- 5 1.50030- 4 2.40988- 4 3.28775- 4934633 18
 4.06173- 4 4.96655- 4 5.78294- 4 7.35670- 4 5.68859- 4 3.21582- 4934633 18
 1.03067- 4-2.84224- 5-6.78220- 5-3.62829- 5 4.05124- 5 1.41735- 4934633 18
 2.53158- 4 4.20532- 4 6.26698- 4 5.22812- 4 3.55656- 4 1.57839- 4934633 18
 2.11845- 6-8.22425- 5-9.78357- 5-6.04168- 5 1.27671- 5 1.07588- 4934633 18
 2.69172- 4 4.90480- 4 3.05310- 4 2.08198- 4 1.10496- 4 3.95566- 5934633 18
 3.42473- 6-1.04504- 6 1.90253- 5 5.63655- 5 1.32607- 4 2.49668- 4934633 18
 2.37424- 4 2.42348- 4 2.29411- 4 2.06211- 4 1.78849- 4 1.51149- 4934633 18
 1.25140- 4 9.14376- 5 5.55298- 5 3.49455- 4 4.09754- 4 4.23490- 4934633 18
 4.00001- 4 3.50544- 4 2.84715- 4 1.70068- 4 9.99255- 6 5.36145- 4934633 18
 5.97068- 4 6.00550- 4 5.60150- 4 4.89289- 4 3.47930- 4 1.33595- 4934633 18
 7.05793- 4 7.51272- 4 7.45658- 4 7.02138- 4 5.88980- 4 3.96187- 4934633 18
 8.46993- 4 8.93652- 4 9.01817- 4 8.60734- 4 7.52757- 4 1.00335- 3934633 18
 1.07891- 3 1.14035- 3 1.16290- 3 1.23137- 3 1.41398- 3 1.59674- 3934633 18
 1.79587- 3 2.24847- 3 3.07560- 3                                 934633 18
                                                                  934633  0
                                                                  9346 0  0
                                                                     0 0  0
 9.42390E+4 2.36999E+2          0          1         34         109437 1451
 0.0        1.00000E+0          0          0          0          69437 1451
 1.00000E+0 2.00000E+7          0          0         10       20029437 1451
 3.00000E+2 0.0                 1          0        564          39437 1451
 94-Pu-239 NAIG       EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA       9437 1451
                      DIST-Feb2004                                9437 1451
----IRDF-2002         MATERIAL 9437                               9437 1451
-----INCIDENT NEUTRON DATA                                        9437 1451
------ENDF-6 FORMAT                                               9437 1451
================================================================= 9437 1451
 94-PU-239 NAIG       EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA       9437 1451
                      DIST-SEP89 REV2-FEB93                       9437 1451
----JENDL-3.2         MATERIAL 9437                               9437 1451
================================================================= 9437 1451
HISTORY                                                           9437 1451
87-03 EVALUATION WAS MADE BY                                      9437 1451
        M.KAWAI AND K.HIDA(NAIG) :  CROSS SECTIONS ABOVE          9437 1451
                      RESONANCE REGION AND OTHER QUANTITIES,      9437 1451
        T.YOSHIDA(NAIG) : RESONANCE PARAMETERS AND BACKGROUND     9437 1451
                      CROSS SECTIONS,                             9437 1451
      DATA WERE COMPILED BY T.NAKAGAWA (JAERI).                   9437 1451
88-08 PARTLY MODIFIED.                                            9437 1451
       NU-BAR, RESOLVED RESONS., (N,2N).                          9437 1451
89-02 FP YIELDS WERE TAKEN FROM JNDC FP DECAY DATA FILE VERSION-2.9437 1451
89-03 UNRESOLVED RESONANCE PARAMETERS WERE SLIGHTLY MODIFIED.     9437 1451
93-02 JENDL-3.2.                                                  9437 1451
      RESONANCE PARAMETERS EVALUATED BY H.DERRIEN (JAERI)/1/.     9437 1451
      FISSION SPECTRA CALCULATED BY T.OHSAWA (KINKI UNIV.)        9437 1451
      COMPILED BY T.NAKAGAWA (NDC/JAERI)                          9437 1451
                                                                  9437 1451
     *****   MODIFIED PARTS FOR JENDL-3.2   ********************  9437 1451
      (2,151)       RESOLVED RESONANCE PARAMETERS UP TO 2.5 KEV   9437 1451
     ***********************************************************  9437 1451
                                                                  9437 1451
                                                                  9437 1451
MF=1  GENERAL INFORMATION                                         9437 1451
  MT=451  DESCRIPTIVE DATA AND DICTIONARY                         9437 1451
                                                                  9437 1451
MF=2  RESONANCE PARAMETERS                                        9437 1451
  MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS             9437 1451
   RESOLVED RES. PARAMETERS FOR REICH-MOORE FORMULA: UP TO 2.5 KEV9437 1451
      PARAMETERS WERE TAKEN FROM REF./1/ DETAILS OF EVALUATION    9437 1451
      ARE GIVEN IN APPENDIX.                                      9437 1451
   UNRESOLVED RESONANCES : FROM 2.5 TO 30 KEV.                    9437 1451
      THE ENERGY DEPENDENT S0, S1 AND FISSION WIDTH WERE DETER-   9437 1451
      MINED SO AS TO REPRODUCE THE EVALUATED TOTAL, CAPTURE AND   9437 1451
      FISSION CROSS SECTIONS.                                     9437 1451
                                                                  9437 1451
   2200-M/SEC CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS.  9437 1451
                       2200 M/S          RES. INTEG.              9437 1451
         TOTAL       1025.7   B             -                     9437 1451
         ELASTIC        7.970 B             -                     9437 1451
         FISSION      747.4   B            302.6 B                9437 1451
         CAPTURE      270.3   B            181.6 B                9437 1451
                                                                  9437 1451
MF=3  NEUTRON CROSS SECTIONS                                      9437 1451
     BELOW 2.5 KEV, CROSS SECTIONS WERE REPRESENTED WITH THE      9437 1451
     RESOLVED RESONANCE PARAMETERS.  BETWEEN 2.5 AND 30 KEV, CROSS9437 1451
     SECTIONS WERE REPLACED WITH UNRESOLVED RESONANCE PARAMETERS. 9437 1451
                                                                  9437 1451
  MT=1  TOTAL                                                     9437 1451
     BELOW 7 MEV, JENDL-2 EVALUATION WHICH WERE BASED ON THE      9437 1451
     EXPERIMENTS OF REFS./10,11,12,13,14/ WAS ADOPTED.  ABOVE     9437 1451
     7 MEV, EXPERIMENTAL DATA BY POENITZ /15/ WERE ADOPTED.       9437 1451
                                                                  9437 1451
  MT=2  ELASTIC SCATTERING                                        9437 1451
     CALCULATED AS (TOTAL) - (PARTIAL CROSS SECTIONS).            9437 1451
                                                                  9437 1451
  MT=18   FISSION                                                 9437 1451
     BELOW  50 KEV                                                9437 1451
        BASED ON MEASUREMENTS OF REF./31/ AND REF./32/.           9437 1451
     ABOVE  50 KEV                                                9437 1451
        SIMULTANEOUS EVALUATION WAS PERFORMED BY KANDA ET AL./33/ 9437 1451
                                                                  9437 1451
MF=33  COVARIANCES                                                9437 1451
  MT=18                                                           9437 1451
     TAKEN FROM JENDL/D-99.                                       9437 1451
                                                                  9437 1451
REFERENCES                                                        9437 1451
 1) DERRIEN H.: J. NUCL. SCI. TECHNOL., 30, 845 (1993).           9437 1451
10) UTTELY C.A.: EANDC(UK)-40 (1964).                             9437 1451
11) SCHWARTZ R.B. ET AL.: NUCL. SCI. ENG., 54, 322 (1974).        9437 1451
12) FOSTER D.G.JR. AND GLAGOW D.W.: PHYS. REV., C3, 576 (1971).   9437 1451
13) SMITH A.B. ET AL.: J. NUCL. ENERGY, 27, 317 (1973).           9437 1451
14) NADOLNY ET AL.: C00-3058-39, 33 (1973).                       9437 1451
15) POENITZ W.P. ET AL.: NUCL. SCI. ENG., 78, 333 (1981).         9437 1451
31) GAYTHER D.B.: 1975 WASHINGTON, 2, 560 (1975).                 9437 1451
32) WAGEMANS C. ET AL.: ANN. NUCL. ENERGY, 7, 495 (1980).         9437 1451
                                                                  9437 1451
                                                                  9437 1451
================================================================= 9437 1451
APPENDIX          RESONANCE DATA                                  9437 1451
================================================================= 9437 1451
                                                                  9437 1451
     THE PRESENT FILE CONTAINS THE RESONANCE PARAMETERS OBTAINED  9437 1451
FROM A SAMMY FIT ANALYSIS OF HIGH RESOLUTION EXPERIMENTAL DATA,   9437 1451
PERFORMED AT ORNL(OAK RIDGE NATIONNAL LABORATORY,USA) BY H.DERRIEN9437 1451
AND G.DE SAUSSURE AND AT JAERI(TOKAI-MURA RESEARCH ESTABLISHMENT, 9437 1451
JAPAN) BY H.DERRIEN.                                              9437 1451
   THE FILE CONTAINS THREE INDEPENDANT SECTIONS:                  9437 1451
   1/ THE FIRST CORRESPONDS TO THE ENERGY RANGE 0 KEV TO 1 KEV.   9437 1451
THE CORRESPONDING SET OF RESONANCE PARAMETRES CONTAINS 398 RESO-  9437 1451
NANCES IN THE ENERGY RANGE 0 KEV TO 1 KEV, 4 FICTICIOUS NEGATIVE  9437 1451
ENERGY RESONANCES AND 3 FICTICIOUS RESONANCES ABOVE 1 KEV;        9437 1451
   2/ THE SECOND CORRESPONDS TO THE ENERGY RANGE 1 KEV TO 2 KEV.  9437 1451
THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 435 RESON- 9437 1451
ANCES IN THE ENERGY RANGE 0.980 KEV TO 2.02 KEV, 3 FICTICIOUS     9437 1451
RESONANCES BELOW 0.9 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.02   9437 1451
KEV;                                                              9437 1451
   3/ THE THIRD CORRESPONDS TO THE ENERGY RANGE 2 KEV TO 2.5 KEV. 9437 1451
THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 218 RESO-  9437 1451
NANCES IN THE ENERGY RANGE 1.98 KEV TO 2.53 KEV, 3 FICTICIOUS     9437 1451
RESONANCES BELOW 1.98 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.53  9437 1451
KEV.                                                              9437 1451
   IN ALL SECTIONS THE FICTICIOUS RESONANCE PARAMETERS TAKE INTO  9437 1451
ACCOUNT THE CONTRIBUTION OF ALL THE EXTERNAL TRUNCATED RESONANCES 9437 1451
IN SUCH A WAY THAT NO TOTAL, SCATTERING, FISSION AND CAPTURE      9437 1451
SMOOTH FILES ARE NEEDED IN THE CORRESPONDING ENERGY RANGES FOR THE9437 1451
REPRODUCTION OF THE CROSS SECTIONS WITHIN THE EXPERIMENTAL ERRORS.9437 1451
   THE FOLLOWING EXPERIMENTAL DATA BASE HAS BEEN USED IN THE SAMMY9437 1451
FITS:                                                             9437 1451
  -ABSORPTION AND FISSION FROM R.GWIN ET AL./1,2/;                9437 1451
  -FISSION FROM R.GWIN ET AL./3,4/, J.BLONS/5/, L.W.WESTON ET     9437 1451
   AL./6,7/;                                                      9437 1451
  -TRANSMISSION FROM R.R.SPENCER ET AL./8/, J.A.HARVEY ET AL./9/  9437 1451
PRIOR TO THE FITS THE EXPERIMENTAL FISSION AND ABSORPTION CROSS   9437 1451
SECTIONS WERE NORMALISED, DIRECTLY OR INDIRECTLY TO THE 0.0253 EV 9437 1451
VALUES OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP/10/.   9437 1451
THE TRANSMISSION DATA WERE CONSIDERED AS ACCURATE ABSOLUTE MEASU- 9437 1451
REMENTS(R.R.SPENCER TOTAL CROSS SECTION AT 0.0253 EV IS 1025.0 B  9437 1451
IN EXCELLENT AGREEMENT WITH THE 1027.3 B STANDARD VALUE).         9437 1451
   DETAILS ON THE ANALYSIS ARE FOUND IN REFERENCES/11,12,13/      9437 1451
                                                                  9437 1451
***************************************************************** 9437 1451
         COMMENTS ON THE THERMAL AND LOW ENERGY RANGES            9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
     THE THERMAL CROSS SECTION VALUES CALCULATED AT 293 K BY THE  9437 1451
RESONANCE PARAMETERS OF THE FIRST SECTION ARE GIVEN IN THE FOLLO- 9437 1451
WING TABLE:                                                       9437 1451
                                                                  9437 1451
                                                                  9437 1451
              SAMMY      RESENDD              PROPOSED            9437 1451
                  293 K (BARN)        STANDARD VALUES(BARN)/10/   9437 1451
****************************************************************  9437 1451
FISSION       747.64     747.90           747.99+-1.87            9437 1451
CAPTURE       271.10     270.73           271.43+-2.14            9437 1451
SCATTERING      7.97       7.99             7.88+-0.97            9437 1451
TOTAL        1026.71    1026.62          1027.30+-5.00            9437 1451
****************************************************************  9437 1451
                                                                  9437 1451
     ONE SHOULD NOTE THAT THE 293 K CROSS SECTIONS CALCULATED AT  9437 1451
0.0253 EV DEPEND ON THE WAY THE DOPPLER BROADENING CALCULATION IS 9437 1451
PERFORMED. FOR INSTANCE USING A GAUSSIAN BROADENING FUNCTION WILL 9437 1451
GIVE A FISSION CROSS SECTION ABOUT 2.5 BARNS LARGER THAN THE ONE  9437 1451
OBTAINED FROM THE ACCURATE CALCULATION WHICH CONSERVES THE 1/V    9437 1451
SHAPE OF THE THERMAL CROSS SECTION. THE VALUES GIVEN IN THE TABLE 9437 1451
ABOVE WERE OBTAINED FROM SAMMY (LEAL-HWANG METHOD)/14,15/ AND FROM9437 1451
RESENDD WITH 0.1% FOR THE INTERPOLATION ACCURACY/16/.             9437 1451
    THE FOLLOWING TABLE SHOWS EXPERIMENTAL CROSS SECTIONS AVE-    9437 1451
RAGED OVER THE ENERGY RANGES 0.02 EV TO 0.06 EV AND 0.02 EV TO    9437 1451
0.65 EV, COMPARED TO THE CALCULATED VALUES:                       9437 1451
                                                                  9437 1451
                                                                  9437 1451
                  AVERAGE CROSS-SECTIONS(BARN)                    9437 1451
****************************************************************  9437 1451
 REFERENCES(1-10)   0.02 TO 0.06 EV        0.02 TO 0.65 EV        9437 1451
****************************************************************  9437 1451
                   EXP     CALC (293K)     EXP    CALC (293K)     9437 1451
GWIN71 FISS       631.41                  843.71                  9437 1451
GWIN76 FISS       631.41                  838.39                  9437 1451
GWIN84 FISS(*)    631.41  631.75(+0.05%)  837.18  838.69(+0.18%)  9437 1451
DERUYTER70 FISS   631.41                  859.43                  9437 1451
WAGEMANS80 FISS   631.41                  862.56                  9437 1451
WAGEMANS88 FISS   631.41                  841.80                  9437 1451
GWIN71 CAPTURE    243.84  243.22(-0.25%)  524.75  518.13(-1.26%)  9437 1451
GWIN76 ABSORPT(*) 875.90  874.29(-0.18%) 1359.96 1357.14(-0.21%)  9437 1451
SPENCER84 TOT(*)  883.20  882.86(-0.04%) 1361.69 1367.6 (+0.43%)  9437 1451
****************************************************************  9437 1451
(*)THESE DATA HAD THE LARGIEST WEIGHT IN THE THERMAL FIT. THE VA- 9437 1451
LUES BETWEEN THE PARENTHESES GIVE THE PERCENTAGE DEVIATION BETWEEN9437 1451
THE CALCULATED DATA AND THE EXPERIMENTAL DATA.                    9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
     THE VALUE OF 631.4 BARNS FOR ALL THE AVERAGED EXPERIMENTAL   9437 1451
FISSION CROSS SECTIONS IN THE ENERGY RANGE 0.02 EV TO 0.06 EV     9437 1451
CORRESPONDS TO THE RENORMALISATION OF THE FISSION EXPERIMENTS TO  9437 1451
748.0+-1. BARNS AT 0.0253 EV. ORNL DATA ARE CONSISTENT WITHIN 0.8%9437 1451
OVER THE ENERGY RANGE 0.02 EV TO 0.65 EV (I.E. OVER THE 0.3 EV    9437 1451
RESONANCE). DERUYTER 1970 AND WAGEMANS 1980 DATA ARE ABOUT 2.5%   9437 1451
LARGER AND WERE NOT INCLUDED IN THE SAMMY FIT. WHEN NORMALIZED ON 9437 1451
THE STANDARD VALUE AT 0.0253 EV, GWIN 76 ABSORPTION AGREES WITH   9437 1451
THE ABSORPTION OBTAINED FROM SPENCER TOTAL CROSS SECTION WITHIN   9437 1451
0.7% OVER THE 0.3 EV RESONANCE. THE PRESENT EVALUATION IS         9437 1451
ESSENTIALLY THE RESULT OF A CONSISTENT SAMMY ANALYSIS OF ALL THE  9437 1451
AVAILABLE ORNL DATA WITH A LARGER WEIGHT ON GWIN 1984 FISSION,    9437 1451
GWIN 1976 ABSORPTION AND SPENCER TRANSMISSION DATA.               9437 1451
     AFTER RENORMALIZATION OF THE CALCULATED FISSION CROSS SECTION9437 1451
ON THE PRELIMINARY 1991 WESTON AND TODD FISSION DATA(SEE NEXT     9437 1451
SECTION) A SLIGHT ADJUSTMENT OF THE NEGATIVE RESONANCE PARAMETERS 9437 1451
WAS PERFORMED TO KEEP THE VALUES CALCULATED AT 0.0253 EV IN CLOSE 9437 1451
AGREEMENT WITH THE STANDARD VALUES. THE 1988 DATA OF WAGEMANS ET  9437 1451
AL./17/ AGREE WITHIN 0.4% WITH THE CALCULATED VALUES OVER THE     9437 1451
ENERGY RANGE FROM 0.02 EV TO 0.65 EV AFTER ADJUSTMENT OF THE      9437 1451
ENERGY SCALE TO THE ORNL SCALE (THE DIFFERENCE WAS 0.27 EV AT 20  9437 1451
EV BETWEEN 1988 WAGEMANS AND ORNL SAMMY FIT ENERGY SCALES).       9437 1451
                                                                  9437 1451
***************************************************************** 9437 1451
         COMMENTS ON THE 0 KEV TO 1 KEV ENERGY RANGE.             9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
   AT THE END OF 1987, AN ANALYSIS WAS COMPLETED UP TO 1 KEV. IN A9437 1451
PRELIMINARY STEP, A CORRELATED FIT OF HARVEY TRANSMISSION DATA,   9437 1451
WESTON 84 FISSION DATA AND BLONS FISSION DATA WAS PERFORMED, WITH 9437 1451
POSSIBLE ADJUSTMENT OF THE NORMALIZATION COEFFICIENTS AND OF THE  9437 1451
BACKGROUND CORRECTIONS. THIS PRELIMINARY STEP HAS SHOWN THAT THIS 9437 1451
ADJUSTMENT WAS NOT NECESSARY TO HAVE CONSISTENCY BETWEEN HARVEY   9437 1451
DATA AND WESTON DATA. BLONS DATA NEEDED A LARGE READJUSTMENT OF   9437 1451
THE BACKGROUND AND OF THE NORMALIZATION. THEREFORE, THE FINAL FIT 9437 1451
WAS PERFORMED ONLY ON HARVEY TRANSMISSION DATA, GWIN 84 FISSION   9437 1451
DATA (BELOW 30 EV) AND WESTON 84 FISSION DATA, WITH NO BACKGROUND 9437 1451
AND NORMALISATION ADJUSTMENT. BLONS DATA, WHICH HAVE BETTER       9437 1451
RESOLUTION THAN WESTON 84 DATA, WERE USED ONLY TO OBTAIN MORE     9437 1451
ACCURATE FISSION WIDTHS OF SOME NARROW RESONANCES IN THE HIGH     9437 1451
ENERGY RANGE.                                                     9437 1451
   IN 1989, PRELIMINARY RESULTS OF THE 1988 WESTON FISSION        9437 1451
MEASUREMENT/7/ WERE INCLUDED IN THE SAMMY EXPERIMENTAL DATA       9437 1451
BASE.  ONE EXPECTED FROM THIS MEASUREMENT, WHICH WAS PERFORMED BY 9437 1451
USING A 86 M FLIGHT PATH WITH A RESOLUTION COMPARABLE TO THE      9437 1451
RESOLUTION OF HARVEY TRANSMISSION, A CONFIRMATION OF THE EXCELLENT9437 1451
QUALITY OF THE 1984 MEASUREMENT. A CONSISTENT SAMMY FIT OF HARVEY 9437 1451
TRANSMISSION, WESTON 84 FISSION AND PRELIMINARY WESTON 88 FISSION 9437 1451
WAS RESTARTED FROM THE PARAMETER AND COVARIANCE FILES OBTAINED IN 9437 1451
1987.  IT APPEARED THAT LARGE BACKGROUND AND NORMALISATION        9437 1451
CORRECTIONS WERE NEEDED ON THE NEW WESTON FISSION TO OBTAIN       9437 1451
CONSISTENCY WITH HARVEY TRANSMISSION DATA. THESE CORRECTIONS WERE 9437 1451
COMPARABLE TO THOSE FOUND ON BLONS DATA AND WERE NOT UNDERSTOOD BY9437 1451
THE AUTHORS OF THE EXPERIMENT. THE LAST SAMMY RUNS WERE PERFORMED 9437 1451
BY NOT ALLOWING BACKGROUND AND NORMALIZATION VARIATIONS ON HARVEY 9437 1451
TRANSMISSION AND WESTON 84 FISSION(VERY SMALL ERROR BARS WERE     9437 1451
ASSIGNED TO THE CORRESPONDING PARAMETERS IN THE COVARIANCE MATRIX)9437 1451
AND BY ALLOWING THESE VARIATIONS ON WESTON 88 DATA. A NEW SET OF  9437 1451
RESONANCE PARAMETERS WAS OBTAINED,WHICH WAS IMPROVED COMPARED TO  9437 1451
THE PREVIOUS SET DUE TO THE VERY HIGH RESOLUTION OF THE NEW WESTON9437 1451
FISSION MEASUREMENT.                                              9437 1451
   THE CALCULATED AVERAGE FISSION CROSS SECTION IN THE ENERGY     9437 1451
RANGE FROM 0.1 KEV TO 1.0 KEV WAS 3.7% SMALLER THAN THE VALUES    9437 1451
OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP, DUE TO THE   9437 1451
FACT THAT WESTON 84 DATA WERE 3.1% LOWER THAN THE AVERAGE STANDARD9437 1451
VALUE. A NEW MEASUREMENT WAS PERFORMED BY WESTON AND TODD IN 1991 9437 1451
/18/ IN ORDER TO CHECK THEIR 1984 DATA. A CAREFUL NORMALIZATION   9437 1451
OF THE DATA IN THE THERMAL ENERGY RANGE SHOWED THAT THE 1984 DATA 9437 1451
SHOULD BE RENORMALIZED BY ABOUT +3%. TO TAKE INTO ACCOUNT THIS    9437 1451
RENORMALIZATION, THE 1989 RESONANCE PARAMETERS WERE MODIFIED AT   9437 1451
JAERI/13/ IN THE FOLLOWING WAY:                                   9437 1451
   1/ INCREASE OF THE FISSION WIDTH BY 3% AND DECREASE OF THE     9437 1451
CAPTURE WIDTH BY A QUANTITY EQUAL TO THE VARIATION OF THE FISSION 9437 1451
WIDTH, IN THE NARROW RESONANCES(MAINLY 1+ RESONANCES); THAT DOES  9437 1451
NOT MODIFY THE TOTAL CROSS SECTION IN THE CORRESPONDING           9437 1451
RESONANCES;                                                       9437 1451
   2/ ADJUSTMENT OF THE NEUTRON WIDTH OF THE 0+ RESONANCES BY A   9437 1451
REFIT OF THE TRANSMISSION DATA AND OF THE RENORMALIZED WESTON AND 9437 1451
TODD 1984 DATA IN ENERGY RANGES WHERE THE CONTRIBUTION OF THE 0+  9437 1451
RESONANCES IS DOMINANT, AND INCREASE OF THE OTHER(SMALL) 0+       9437 1451
NEUTRON WIDTHS BY 3%. NO SEVERE INCONSISTENCY WAS OBSERVED BETWEEN9437 1451
THE TRANSMISSION DATA AND THE NEW FISSION DATA OVER THE DOMINANT  9437 1451
0+ RESONANCES;THE DIFFERENCES BETWEEN THE 1989 FITS OF THE TRANS- 9437 1451
MISSION AND THE NEW FITS WERE CONSISTENT WITHIN THE EXPERIMENTAL  9437 1451
ERROR BARS.                                                       9437 1451
   THE FOLLOWING TABLE SHOWS THE FISSION CROSS SECTIONS CALCULA-  9437 1451
TED FROM THE RESONANCE PARAMETERS, THE EXPERIMENTAL VALUES AND THE9437 1451
RESULTS OF THE ENDF/B-VI STANDARD EVALUATION GROUP AVERAGED IN THE9437 1451
SAME ENERGY INTERVALS. WESTON 1991 DATA ARE PRELIMINARY. WESTON   9437 1451
1984 DATA ARE NORMALIZED ON PRELIMINARY WESTON 1991:              9437 1451
                                                                  9437 1451
                           CROSS-SECTIONS(BARN)                   9437 1451
         *************************************************        9437 1451
            ENERGY    CALCUL   WESTON   WESTON   STANDARD         9437 1451
             (EV)               1991     1984                     9437 1451
         *************************************************        9437 1451
          0.010-10.    80.12    79.98                             9437 1451
              9-20     94.74    94.91                             9437 1451
             20-40     17.52    17.76    17.97                    9437 1451
             40-60     50.64    50.90    50.87                    9437 1451
             60-100    54.42    54.38    54.33                    9437 1451
            100-200    18.63    18.59    18.56    18.66           9437 1451
            200-300    17.85             17.89    17.88           9437 1451
            300-400     8.31              8.34     8.43           9437 1451
            400-500     9.59              9.58     9.57           9437 1451
         -------------------------------------------------        9437 1451
            200-500    11.92    11.93    11.93    11.96           9437 1451
         -------------------------------------------------        9437 1451
            500-600    15.39             15.57    15.86           9437 1451
            600-700     4.37              4.30     4.46           9437 1451
            700-800     5.51              5.53     5.63           9437 1451
            800-900     4.84              4.89     4.98           9437 1451
            900-1000    8.33              8.38     8.30           9437 1451
         -------------------------------------------------        9437 1451
            500-1000    7.69     7.73     7.73     7.79           9437 1451
         -------------------------------------------------        9437 1451
             20-1000   13.09    13.11    13.11                    9437 1451
         **************************************************       9437 1451
                                                                  9437 1451
                                                                  9437 1451
   GWIN 1971 AND 1976 ABSORPTION DATA WERE NOT INCLUDED IN THE    9437 1451
SAMMY FIT IN THE ENERGY RANGE ABOVE 1 EV. ACCURATE ABSORPTION     9437 1451
CROSS SECTIONS SHOULD BE CALCULATED FROM THE PARAMETERS OBTAINED  9437 1451
FROM THE ANALYSIS OF THE TRANSMISSION AND FISSION DATA. THE FOLLO-9437 1451
WING TABLE SHOWS THE CALCULATED AVERAGE VALUES OF THE CAPTURE, AB-9437 1451
SORPTION AND ALPHA COMPARED TO GWIN 1971 AND GWIN 1976 DATA. THE  9437 1451
CALCULATIONS WERE PERFORMED WITH RESENDD, 1.0 % ACCURACY:         9437 1451
                                                                  9437 1451
                     CROSS-SECTIONS(BARN)                         9437 1451
    *******************************************************       9437 1451
      ENERGY(EV)   CALC. VALUES (293K)     GWIN DATA              9437 1451
    *******************************************************       9437 1451
                  CAPT   ABSORP ALPHA     ABSORP ALPHA            9437 1451
       7.3- 16.0  76.61  196.04  0.64     208.00  0.74(*)         9437 1451
      16.0- 37.5  20.51   44.55  0.85      46.50  0.89(*)         9437 1451
      37.5- 50.0  48.72   70.00  2.29      83.15  2.96(*)         9437 1451
      50.0-100.0  33.60   92.13  0.57      92.84  0.63            9437 1451
     100.0-200.0  15.58   34.29  0.83      33.66  0.87            9437 1451
     200.0-300.0  15.85   33.68  0.89      34.69  0.94            9437 1451
     300.0-400.0   9.69   18.01  1.16      18.31  1.16            9437 1451
     400.0-500.0   3.96   13.56  0.41      13.56  0.44            9437 1451
     500.0-600.0  10.87   26.30  0.70      26.54  0.72            9437 1451
     600.0-700.0   6.53   10.90  1.49      11.57  1.54            9437 1451
     700.0-800.0   4.95   10.47  0.90      10.52  0.97            9437 1451
     800.0-900.0   3.65    8.50  0.75       9.30  0.82            9437 1451
     900.0-999.9   5.06   13.51  0.60      13.23  0.70            9437 1451
    ******************************************************        9437 1451
     (*) GWIN 1971 DATA                                           9437 1451
                                                                  9437 1451
   IF ONE EXCEPTS THE ENERGY RANGE 37.5-50 EV, THE CALCULATED AB- 9437 1451
SORPTION VALUES AGREE WELL WITH GWIN EXPERIMENTAL DATA; THEY ARE  9437 1451
ON AVERAGE 1.2% LOWER IN THE ENERGY RANGE FROM 50 EV TO 1000 EV.  9437 1451
                                                                  9437 1451
***************************************************************** 9437 1451
      COMMENTS ON THE 1 KEV TO 2 KEV ENERGY RANGE                 9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
   PRELIMINARY RESONANCE PARAMETERS WERE OBTAINED IN 1989 FROM THE9437 1451
ANALYSIS OF THE HARVEY THICK SAMPLE TRANSMISSION DATA AND OF THE  9437 1451
PRELIMINARY RESULTS OF WESTON 88 FISSION MEASUREMENT.  DUE TO LACK9437 1451
OF TIME, THE MEDIUM AND THIN SAMPLE TRANSMISSION DATA WERE NOT    9437 1451
INCLUDED IN THE SAMMY DATA BASE, AND THE CONTRIBUTION OF THE      9437 1451
TRUNCATED EXTERNAL RESONANCES WAS NOT CAREFULLY INVESTIGATED.     9437 1451
NEVERTHELESS, THE RESULTS WERE USED IN THE ENDF/B-VI FILE, ALONG  9437 1451
WITH A SMOOTH FILE IN ORDER TO AGREE WITH THE AVERAGE VALUES OF A 9437 1451
PREVIOUS ENDF/B-VI EVALUATION (THIS PRELIMINARY SET OF PARAMETERS 9437 1451
WAS CONSIDERED AS MORE USEFUL THAN THE STATISTICAL PARAMETERS IN  9437 1451
THE ENERGY RANGE 1 KEV TO 2 KEV FOR THE CALCULATION OF THE SELF-  9437 1451
SHIELDING FACTORS).                                               9437 1451
   THE ANALYSIS WAS RESTARTED IN APRIL 1991 AT JAERI(TOKAI        9437 1451
RESEARCH ESTABLISHMENT) WITH AN UPDATED VERSION OF SAMMY ADAPTED  9437 1451
BY T.NAKAGAWA TO THE FACOM 780. THE PRELIMINARY SET OF PARAMETERS 9437 1451
OBTAINED AT OAK RIDGE IN 1989 WAS USED AS PRIOR INFORMATIONS TO   9437 1451
START THE SAMMY CALCULATIONS. ALSO PRIOR TO THE ANALYSIS, THE CON-9437 1451
TRIBUTION OF THE EXTERNAL RESONANCES WAS CALCULATED BY USING THE  9437 1451
SET OF THE 0 KEV TO 1 KEV KNOWN RESONANCES, SHIFTED IN THE ENERGY 9437 1451
RANGES -1 KEV TO 0 KEV, 2 KEV TO 3KEV AND 3 KEV TO 4 KEV; EQUIVA- 9437 1451
LENT CONTRIBUTION WAS OBTAINED BY USING 3 FICTICIOUS RESONANCES   9437 1451
BELOW 1 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2 KEV(SEE DETAILS IN9437 1451
REF./13/). THE ANALYSIS WAS PERFORMED ON THE THICK AND MEDIUM     9437 1451
SAMPLE TRANSMISSIONS OF HARVEY DATA (THE THIN SAMPLE DATA WAS NOT 9437 1451
USEFUL IN THE HIGH ENERGY RANGE) AND ON THE 1988 FISSION DATA RE- 9437 1451
LEASED BY WESTON AT THE BEGINNING OF 1991/7/. THE DEFINITIVE      9437 1451
SAMMY FITS WERE PERFORMED IN APRIL 1992 AFTER RENORMALIZATION OF  9437 1451
THE 1988 DATA OF WESTON ON THE ENDF/B-VI STANDARD VALUES BETWEEN 19437 1451
KEV AND 2 KEV, IN AGREEMENT WITH THE 1991 NEW MEASUREMENTS OF     9437 1451
WESTON AND TODD.                                                  9437 1451
   THE AVERAGE CROSS SECTIONS CALCULATED FROM THE RESONANCE       9437 1451
PARAMETERS ARE COMPARED TO THE EXPERIMENTAL VALUES IN THE FOLLO-  9437 1451
WING TABLE:                                                       9437 1451
                                                                  9437 1451
                                                                  9437 1451
**************************************************************    9437 1451
                        CROSS-SECTIONS(BARN)                      9437 1451
           ---------------------------------------------------    9437 1451
                 TOTAL           FISSION         CAPTURE          9437 1451
  ENERGY    --------------    -------------    -------------      9437 1451
   KEV     CALC(A)   EXP(B)  CALC(A)   EXP(C) CALC(A)  EXP(D)     9437 1451
**************************************************************    9437 1451
  1.0-1.1   24.47    24.95    5.549    5.581   4.728   5.04       9437 1451
  1.1-1.2   22.82    23.10    5.985    6.017   3.757   2.95       9437 1451
  1.2-1.3   22.29    22.90    4.601    4.501   4.287   4.00       9437 1451
  1.3-1.4   22.63    22.85    6.997    6.997   3.012   2.52       9437 1451
  1.4-1.5   20.42    20.95    4.041    4.059   3.450   3.57       9437 1451
  1.5-1.6   18.30    18.95    2.564    2.613   3.521   3.89       9437 1451
  1.6-1.7   21.82    21.90    3.952    3.955   3.833   4.36       9437 1451
  1.7-1.8   21.26    21.35    3.400    3.425   4.091   4.37       9437 1451
  1.8-1.9   23.76    23.30    5.178    5.187   3.639   3.14       9437 1451
  1.9-2.0   18.48    18.90    2.152    2.180   3.205   4.06       9437 1451
**************************************************************    9437 1451
  1.0-2.0   21.63    21.92    4.442    4.446   3.752   3.79       9437 1451
**************************************************************    9437 1451
 (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY      9437 1451
     RESENDD FROM THE RESONANCE PARAMETERS.                       9437 1451
 (B) EXPERIMENTAL TOTAL CROSS SECTIONS FROM REFERENCE/19/.        9437 1451
 (C) WESTON AND TODD 1988 HIGH RESOLUTION FISSION CROSS SECTIONS  9437 1451
     FROM REFERENCE/7/ NORMALIZED TO ENDF/B-VI STANDARD IN THE    9437 1451
     ENERGY RANGE FROM 1.0 KEV TO 2.0 KEV.                        9437 1451
 (D) GWIN 1971 EXPERIMENTAL DATA NORMALIZED TO GWIN 1976 DATA.    9437 1451
**************************************************************    9437 1451
                                                                  9437 1451
   THE DIFFERENCE OF 1.3% BETWEEN THE AVERAGE CALCULATED TOTAL    9437 1451
CROSS SECTION AND THE AVERAGE EXPERIMENTAL CROSS SECTION IN THE   9437 1451
ENERGY RANGE FROM 1.0 KEV AND 2.0 KEV IS MAINLY DUE TO THE METHOD 9437 1451
OF EVALUATING THE TOTAL CROSS SECTION FROM THE EFFECTIVE CROSS    9437 1451
SECTION IN REFERENCE/19/. THE ACCURACY OF THE SAMMY FIT OF THE    9437 1451
EXPERIMENTAL TRANSMISSION DATA IS BETTER THAN 0.5% ON THE CROSS   9437 1451
SECTION. THE CALCULATED FISSION CROSS SECTIONS ARE IN VERY GOOD   9437 1451
AGREEMENT WITH THE EXPERIMENTAL DATA. THE CAPTURE DATA /1/ ARE    9437 1451
AVERAGE VALUES OBTAINED FROM THE DATA AVAILABLE IN THE EXFOR FILE 9437 1451
AND NORMALIZED TO GWIN 1976 AVERAGE VALUES; THERE ARE LARGE       9437 1451
DIFFERENCES BETWEEN THE CALCULATED DATA AND THE EXPERIMENTAL DATA 9437 1451
AVERAGED OVER 0.1 KEV INTERVALS; BUT ON THE INTERVAL FROM 1.0 KEV 9437 1451
TO 2.0 KEV THE AVERAGE VALUES ARE CONSISTENT WITHIN 1.0%.         9437 1451
                                                                  9437 1451
***************************************************************** 9437 1451
   COMMENTS ON THE 2.0 KEV TO 2.5 KEV REGION                      9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
   THIS ENERGY RANGE WAS ALSO ANALYSED AT JAERI /13/. NO          9437 1451
PRELIMINARY SET OF RESONANCE PARAMETERS WAS AVAILABLE PRIOR TO THE9437 1451
ANALYSIS. MORE THAN 90% OF THE RESONANCES, COMPARED TO THE LOW    9437 1451
ENERGY RANGE, COULD STILL BE IDENTIFIED IN THE TRANSMISSION DATA  9437 1451
BETWEEN 2 KEV AND 2.5 KEV. THEREFORE THE CORRELATED SAMMY ANALYSIS9437 1451
OF HARVEY TRANSMISSIONS AND WESTON FISSION WAS STILL FEASIBLE IN  9437 1451
THIS ENERGY RANGE. THE RESONANCE PARAMETERS OBTAINED ARE          9437 1451
CONSISTENT AND HAS NEARLY THE SAME STATISTICAL PROPERTIES AS THOSE9437 1451
OF THE RESONANCES IN THE 0 TO 2 KEV ENERGY RANGE. A QUITE GOOD FIT9437 1451
OF THE TRANSMISSION AND FISSION DATA WAS OBTAINED WITHOUT         9437 1451
BACKGROUND AND NORMALISATION ADJUSTMENT. HOWEVER, THE CALCULATED  9437 1451
FISSION CROSS SECTIONS ARE, ON AVERAGE, 1.4% LOWER THAN THE       9437 1451
EXPERIMENTAL VALUES.  THIS DIFFERENCE,WHICH HOWEVER IS NOT LARGER 9437 1451
THAN THE SYSTEMATIC ERRORS ON THE EXPERIMENTAL DATA, COULD BE DUE 9437 1451
TO THE DIFFICULTIES OF IDENTIFYING THE WIDE J=0+ RESONANCES IN THE9437 1451
EXPERIMENTAL DATA, BECAUSE THE EFFECTS OF THE INCREASING          9437 1451
RESOLUTION AND DOPPLER WIDTHS. PRIOR TO THE SAMMY FITS, THE       9437 1451
FISSION DATA OF WESTON AND TODD (1988 HIGH RESOLUTION DATA) WERE  9437 1451
NORMALIZED TO THE ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1   9437 1451
KEV TO 2 KEV.                                                     9437 1451
   THE CROSS SECTIONS,CALCULATED FROM THE RESONANCE PARAMETERS    9437 1451
AND AVERAGED OVER 0.1 KEV INTERVALS,ARE GIVEN IN THE FOLLOWING    9437 1451
TABLE:                                                            9437 1451
                                                                  9437 1451
**************************************************************    9437 1451
                         CR0SS-SECTIONS(BARN)                     9437 1451
                ------------------------------------------        9437 1451
                    TOTAL            FISSION       CAPTURE        9437 1451
      ENERGY    --------------    --------------   -------        9437 1451
       (KEV)   CALC(A)   EXP(B)  CALC(A)   EXP(C)  CALC(A)        9437 1451
**************************************************************    9437 1451
      2.0-2.1   17.34    17.30    2.034    2.062    3.223         9437 1451
      2.1-2.2   20.27    19.80    2.949    2.999    4.051         9437 1451
      2.2-2.3   19.34    19.10    2.357    2.393    3.324         9437 1451
      2.3-2.4   21.28    21.20    3.646    3.679    3.640         9437 1451
      2.4-2.5   20.03    20.60    3.956    4.024    3.128         9437 1451
**************************************************************    9437 1451
      2.0-2.5   19.65    19.60    2.989    3.031    3.473         9437 1451
**************************************************************    9437 1451
  (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY     9437 1451
      RESENDD, 1% ACCURACY, AT 300 K,FROM THE RESONANCE PARAME-   9437 1451
      TERS.                                                       9437 1451
  (B) AVERAGE TOTAL CROSS SECTIONS OBTAINED FROM THE AVERAGE      9437 1451
      EXPERIMENTAL EFFECTIVE CROSS SECTIONS IN REFERENCE/19/.     9437 1451
  (C) 1988 HIGH RESOLUTION DATA OF WESTON AND TODD NORMALIZED     9437 1451
      TO ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1 KEV TO     9437 1451
      2 KEV.                                                      9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
                                                                  9437 1451
***************************************************************** 9437 1451
      FISSION AND CAPTURE RESONANCE INTEGRALS                     9437 1451
***************************************************************** 9437 1451
                                                                  9437 1451
THE FISSION AND CAPTURE RESONANCE INTEGRALS ARE COMPARED TO       9437 1451
JENDL-3 DATA IN THE FOLLOWING TABLE:                              9437 1451
                                                                  9437 1451
   **********************************************************     9437 1451
     ENERGY RANGE(EV)    FISSION(BARN)      CAPTURE(BARN)         9437 1451
   **********************************************************     9437 1451
                        JENDL-3  PRESENT    JENDL-3  PRESENT      9437 1451
       0.5 -   5.0       85.725   84.879     28.651   28.723      9437 1451
       5.0 -  10.0       25.081   25.147     19.059   18.950      9437 1451
      10.0 -  50.0       96.856   99.715     77.181   74.686      9437 1451
      50.0 - 100.0       40.479   41.552     25.930   25.376      9437 1451
     100.0 - 301.0       19.677   20.252     17.952   17.729      9437 1451
     301.0 -1000.0       10.047   10.317      8.348    8.418      9437 1451
    1000.0 -2000.0        3.484    3.206      2.840    2.634      9437 1451
    2000.0 -2.E+07       17.783  (17.783)     5.224   (5.224)     9437 1451
   **********************************************************     9437 1451
         TOTAL          299.132  302.851    185.185  181.739      9437 1451
   **********************************************************     9437 1451
                                                                  9437 1451
   THE JENDL-3 RESONANCE PARAMETERS ARE THOSE OBTAINED IN 1987 IN 9437 1451
THE ENERGY RANGE 0 KEV TO 1 KEV. THEY ARE SLIGTHLY DIFFERENT FROM 9437 1451
THOSE PUBLISHED IN 1989. THAT EXPLAINS THE SMALL DIFFERENCES OB-  9437 1451
SERVED BETWEEN JENDL-3 AND THE PRESENT RESULTS IN THIS ENERGY RAN-9437 1451
GE. IN THE ENERGY RANGE 1 KEV TO 2 KEV JENDL-3 IS UNRESOLVED      9437 1451
RANGE.  THE FISSION AND CAPTURE RESONANCE INTEGRALS CALCULATED    9437 1451
FROM ENDF/B-V AND THOSE FOUND IN BNL-325 ARE THE FOLLOWING:       9437 1451
                                                                  9437 1451
      ENDF/B-V     FISSION: 302.13 B    CAPTURE: 194.10 B         9437 1451
       BNL-325     FISSION: 310+-10 B   CAPTURE: 200+-20 B        9437 1451
                                                                  9437 1451
   THE CONSEQUENCE OF CHANGING FROM THE OLD SETS OF RESONANCE     9437 1451
PARAMETERS(ENDF/B-V AND PREVIOUS SETS) TO THE NEW SET IS THAT     9437 1451
THE CAPTURE RESONANCE INTEGRAL WILL DECREASE BY 6.7% COMPARED     9437 1451
WITH ENDF/B-V VALUE.                                              9437 1451
                                                                  9437 1451
REFERENCES OF APPENDIX                                            9437 1451
 1) R.GWIN ET AL.,NUCL.SCI.ENG.,45,25(1971)                       9437 1451
 2) R.GWIN ET AL.,NUCL.SCI.ENG.,59,79(1976)                       9437 1451
 3) R.GWIN ET AL.,NUCL.SCI.ENG.,61,116(1976)                      9437 1451
 4) R.GWIN ET AL.,NUCL.SCI.ENG.,88,37(1984)                       9437 1451
 5) J.BLONS, NUCL.SCI.ENG.,51,130(1973)                           9437 1451
 6) L.W.WESTON ET AL.,NUCL.SCI.ENG.88,567(1984)                   9437 1451
 7) L.W.WESTON ET AL.,TO BE PUBLISHED(HIGH RESOLUTION 1988 DATA)  9437 1451
 8) R.R.SPENCER ET AL.,NUCL.SCI.ENG.,96,318(1987)                 9437 1451
 9) J.A.HARVEY ,MITO 1988,PAGE 115                                9437 1451
10) A.CARLSON ET AL.,PRELIMINARY RESULTS OF THE ENDF/B-6 STANDARD 9437 1451
   EVALUATION(SEPT 8 1987)                                        9437 1451
11) H.DERRIEN AND G. DE SAUSSURE,ORNL-TM-10986(1988)              9437 1451
12) H.DERRIEN ET AL.,NUCL.SCI.ENG.,106,434(1990)                  9437 1451
13) H.DERRIEN, J.NUCL>SCI.TECHNOL.,30,845(1993).                  9437 1451
14) N.M.LARSON ET AL.,ORNL/TM-7485,ORNL/TM-9179,ORNL/TM-9719/R1   9437 1451
15) L.LEAL AND R.N.HWANG,TRANS.AM.NUC.SOC.,55,340(1987)           9437 1451
16) T.NAKAGAWA,RESENDD A JAERI VERSION OF RESEND,JAER=-M 84-192   9437 1451
    (1984).                                                       9437 1451
17) C.WAGEMANS ET AL.,MITO 1988,PAGE 91                           9437 1451
18) L.W.WESTON,PRIVATE COMMUNICATION(1992)                        9437 1451
19) H.DERRIEN,J.NUCL.SCI.TECHNOL.,29,794(1992).                   9437 1451
   **********************************************************     9437 1451
                                                                  9437 1451
                                                                  9437 1451
                                                                  9437 1451
                                                                  9437 1451
   **********************************************************     9437 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9437 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9437 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9437 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9437 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9437 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9437 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9437 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9437 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9437 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9437 1451
 ***************** Program FIXUP (Version 2002-1) ****************9437 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9437 1451
 Corrected Thresholds-----------------------------------------Yes 9437 1451
 Extended Cross Sections to 20 MeV----------------------------No  9437 1451
 Allow Cross Section Deletion---------------------------------No  9437 1451
 Allow Cross Section Reconstruction---------------------------No  9437 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9437 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9437 1451
 Deleted Duplicate Points-------------------------------------Yes 9437 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9437 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9437 1451
 Allow Creation of Missing Sections---------------------------No  9437 1451
 Allow Insertion of Energy Points-----------------------------No  9437 1451
 Create Uniform Energy Grid-----------------------------------No  9437 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9437 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9437 1451
 Unshielded Group Averages Using  640 Groups                      9437 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9437 1451
                                1        451        571          19437 1451
                                3         18        217          19437 1451
                               33         18         18          09437 1451
                                                                  9437 1  0
                                                                  9437 0  0
 9.42390E+4 2.36999E+2          0          0          0          09437 3 18
 1.99219E+8 1.99219E+8          0          0          1        6419437 3 18
        641          1                                            9437 3 18
 .000100000 11417.9620 .000105000 11151.7657 .000110000 10898.87749437 3 18
 .000115000 10665.8514 .000120000 10393.4353 .000127500 10091.19469437 3 18
 .000135000 9815.06416 .000142500 9561.30914 .000150000 9286.678129437 3 18
 .000160000 9000.58665 .000170000 8740.28134 .000180000 8501.343689437 3 18
 .000190000 8279.03421 .000200000 8076.06236 .000210000 7885.101369437 3 18
 .000220000 7707.73139 .000230000 7542.93171 .000240000 7349.490919437 3 18
 .000255000 7136.93344 .000270000 6972.91683 .000280000 6790.337609437 3 18
 .000300000 6567.67848 .000320000 6366.07015 .000340000 6181.431249437 3 18
 .000360000 6011.36969 .000380000 5856.12850 .000400000 5694.282379437 3 18
 .000425000 5528.66447 .000450000 5377.33630 .000475000 5238.609629437 3 18
 .000500000 5108.15369 .000525000 4988.92358 .000550000 4876.291389437 3 18
 .000575000 4771.53888 .000600000 4664.17685 .000630000 4553.731479437 3 18
 .000660000 4452.67626 .000690000 4355.73719 .000720000 4252.590649437 3 18
 .000760000 4141.99621 .000800000 4039.31515 .000840000 3945.218449437 3 18
 .000880000 3855.78668 .000920000 3773.80381 .000960000 3695.289949437 3 18
 .001000000 3613.90745 .001050000 3528.95703 .001100000 3449.325569437 3 18
 .001150000 3375.95333 .001200000 3289.40374 .001275000 3194.286399437 3 18
 .001350000 3107.33163 .001425000 3026.18775 .001500000 2939.993969437 3 18
 .001600000 2849.88622 .001700000 2767.64223 .001800000 2691.589279437 3 18
 .001900000 2622.12731 .002000000 2557.72693 .002100000 2497.297949437 3 18
 .002200000 2441.96483 .002300000 2389.05464 .002400000 2328.682139437 3 18
 .002550000 2261.58904 .002700000 2209.10387 .002800000 2152.066499437 3 18
 .003000000 2081.85104 .003200000 2017.96735 .003400000 1959.577349437 3 18
 .003600000 1906.37126 .003800000 1857.28629 .004000000 1806.170019437 3 18
 .004250000 1754.32598 .004500000 1707.02700 .004750000 1662.822959437 3 18
 .005000000 1622.24531 .005250000 1584.60383 .005500000 1549.000829437 3 18
 .005750000 1516.33952 .006000000 1482.02568 .006300000 1447.715219437 3 18
 .006600000 1415.36433 .006900000 1385.55510 .007200000 1353.066159437 3 18
 .007600000 1318.26127 .008000000 1286.45396 .008400000 1256.609289437 3 18
 .008800000 1228.84220 .009200000 1203.05200 .009600000 1178.520609437 3 18
 .010000000 1153.14397 .010500000 1126.33160 .011000000 1101.838519437 3 18
 .011500000 1078.49018 .012000000 1051.77203 .012750000 1022.246269437 3 18
 .013500000 994.946483 .014250000 969.957080 .015000000 943.3636809437 3 18
 .016000001 915.498777 .017000001 889.991693 .017999999 866.7475789437 3 18
 .018999999 845.428676 .020000000 825.467713 .021000000 807.2649719437 3 18
 .022000000 790.163425 .023000000 774.288231 .024000000 755.7022269437 3 18
 .025500000 735.781185 .027000001 720.045542 .028000001 703.0632959437 3 18
 .029999999 682.289356 .032000002 663.523948 .034000002 646.5597309437 3 18
 .035999998 631.248456 .037999999 617.287673 .039999999 602.8973619437 3 18
 .042500000 588.543054 .045000002 575.713193 .047499999 563.9898559437 3 18
 .050000001 553.473508 .052499998 543.973705 .055000000 535.1489319437 3 18
 .057500001 527.373235 .059999999 519.386396 .063000001 511.8185459437 3 18
 .066000000 505.024459 .068999998 499.108067 .071999997 493.1986959437 3 18
 .075999998 487.401245 .079999998 482.909720 .083999999 479.3131199437 3 18
 .088000000 476.719958 .092000000 475.123148 .096000001 474.2228099437 3 18
 .100000001 474.822582 .104999997 476.259529 .109999999 479.1538109437 3 18
 .115000002 483.522609 .119999997 490.951101 .127499998 502.9041239437 3 18
 .135000005 518.437086 .142499998 537.794995 .150000006 566.3252249437 3 18
 .159999996 607.478709 .170000002 659.867483 .180000007 726.5071849437 3 18
 .189999998 811.052002 .200000003 918.361412 .209999993 1055.264109437 3 18
 .219999999 1230.56672 .230000004 1453.80710 .239999995 1817.201079437 3 18
 .254999995 2361.74559 .270000011 2840.61081 .280000001 3190.837339437 3 18
 .300000012 2980.58752 .319999993 2225.17116 .340000004 1504.559999437 3 18
 .360000014 1018.55129 .379999995 717.551256 .400000006 512.0498319437 3 18
 .425000012 371.860678 .449999988 283.854365 .474999994 225.4229599437 3 18
 .500000000 184.796623 .524999976 155.324189 .550000012 133.2844959437 3 18
 .574999988 116.307646 .600000024 101.773185 .629999995 89.36090219437 3 18
 .660000026 79.5626784 .689999998 71.6606229 .720000029 64.24306199437 3 18
 .759999990 57.4175488 .800000012 51.9416016 .839999974 47.43805919437 3 18
 .879999995 43.6819360 .920000017 40.5067606 .959999979 37.79521469437 3 18
 1.00000000 35.1919394 1.04999995 32.6995076 1.10000002 30.57096249437 3 18
 1.14999998 28.7219728 1.20000005 26.7441226 1.27499998 24.72048819437 3 18
 1.35000002 23.0219483 1.42499995 21.5594338 1.50000000 20.10110599437 3 18
 1.60000002 18.6803581 1.70000005 17.4691784 1.79999995 16.42722979437 3 18
 1.89999998 15.5120732 2.00000000 14.7124153 2.09999990 14.00077169437 3 18
 2.20000005 13.3633376 2.29999995 12.7905016 2.40000010 12.15160549437 3 18
 2.54999995 11.4763503 2.70000005 10.9776666 2.79999995 10.45086159437 3 18
 3.00000000 9.83794056 3.20000005 9.31370858 3.40000010 8.864508019437 3 18
 3.59999990 8.47962122 3.79999995 8.14254875 4.00000000 7.826542379437 3 18
 4.25000000 7.53615788 4.50000000 7.30727947 4.75000000 7.137375899437 3 18
 5.00000000 7.02514311 5.25000000 6.97966955 5.50000000 7.013222689437 3 18
 5.75000000 7.14644542 6.00000000 7.45452877 6.30000019 8.102649819437 3 18
 6.59999990 9.40646772 6.90000010 12.4226492 7.19999981 29.59755039437 3 18
 7.59999990 367.574182 8.00000000 17.2892425 8.39999962 4.309537539437 3 18
 8.80000019 3.85693634 9.19999981 4.78786972 9.60000038 7.569513829437 3 18
 10.0000000 22.6400356 10.5000000 475.866785 11.0000000 274.3945919437 3 18
 11.5000000 209.272688 12.0000000 11.1515458 12.7500000 12.85650919437 3 18
 13.5000000 55.9508997 14.2500000 304.738093 15.0000000 86.14984099437 3 18
 16.0000000 17.9291733 17.0000000 146.797252 18.0000000 3.579175879437 3 18
 19.0000000 1.00688576 20.0000000 1.15654924 21.0000000 9.262832479437 3 18
 22.0000000 196.043550 23.0000000 11.4103951 24.0000000 3.148274099437 3 18
 25.5000000 59.9961160 27.0000000 2.19378375 28.0000000 .8160675629437 3 18
 30.0000000 1.13292690 32.0000000 10.7885237 34.0000000 2.279400119437 3 18
 36.0000000 1.23322715 38.0000000 1.24006515 40.0000000 27.17570199437 3 18
 42.5000000 16.1198564 45.0000000 10.2630847 47.5000000 51.42448259437 3 18
 50.0000000 26.8688800 52.5000000 32.0651320 55.0000000 57.76124979437 3 18
 57.5000000 183.143727 60.0000000 43.8941711 63.0000000 102.3385239437 3 18
 66.0000000 36.1098363 69.0000000 8.96265222 72.0000000 149.5954389437 3 18
 76.0000000 6.56611001 80.0000000 45.1179049 84.0000000 115.5168529437 3 18
 88.0000000 27.4882308 92.0000000 22.6634900 96.0000000 33.22585189437 3 18
 100.000000 13.9130139 105.000000 24.2609932 110.000000 3.647983829437 3 18
 115.000000 57.7872580 120.000000 11.6310258 127.500000 28.03524109437 3 18
 135.000000 11.2537904 142.500000 19.6672234 150.000000 19.08359889437 3 18
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 220.000000 10.1820368 230.000000 10.8567713 240.000000 12.15820879437 3 18
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 300.000000 10.4535178 320.000000 9.25415388 340.000000 9.279422379437 3 18
 360.000000 7.00919877 380.000000 5.55506439 400.000000 12.76524959437 3 18
 425.000000 6.03518620 450.000000 11.7330973 475.000000 7.812129229437 3 18
 500.000000 21.5770880 525.000000 15.8445952 550.000000 19.23575839437 3 18
 575.000000 4.96655734 600.000000 3.43493327 630.000000 6.139049169437 3 18
 660.000000 3.89531636 690.000000 3.10873388 720.000000 6.574240269437 3 18
 760.000000 5.68757768 800.000000 4.50031369 840.000000 5.843087579437 3 18
 880.000000 5.41416850 920.000000 5.47578655 960.000000 11.99075619437 3 18
 1000.00000 6.70737619 1050.00000 4.39687108 1100.00000 7.151840429437 3 18
 1150.00000 4.80583395 1200.00000 4.61323475 1275.00000 6.381521809437 3 18
 1350.00000 5.68490872 1425.00000 4.14856557 1500.00000 2.560602469437 3 18
 1600.00000 3.94877408 1700.00000 3.39585663 1800.00000 5.172265319437 3 18
 1900.00000 2.15622616 2000.00000 2.03226243 2100.00000 2.945164939437 3 18
 2200.00000 2.35456787 2300.00000 3.64157538 2400.00000 3.780709339437 3 18
 2550.00000 3.51284758 2700.00000 3.62104701 2800.00000 3.601374859437 3 18
 3000.00000 3.49644336 3200.00000 3.34231604 3400.00000 3.191645089437 3 18
 3600.00000 3.00762249 3800.00000 2.82759550 4000.00000 2.664083569437 3 18
 4250.00000 2.51438488 4500.00000 2.42419996 4750.00000 2.392268739437 3 18
 5000.00000 2.37053298 5250.00000 2.35899271 5500.00000 2.328055329437 3 18
 5750.00000 2.27898824 6000.00000 2.20961595 6300.00000 2.122590689437 3 18
 6600.00000 2.11484269 6900.00000 2.13733868 7200.00000 2.184675969437 3 18
 7600.00000 2.24353389 8000.00000 2.28536525 8400.00000 2.285844109437 3 18
 8800.00000 2.20463535 9200.00000 1.99148915 9600.00000 1.904747879437 3 18
 10000.0000 1.88624839 10500.0000 1.87566285 11000.0000 1.867047379437 3 18
 11500.0000 1.86014402 12000.0000 1.85302761 12750.0000 1.839203249437 3 18
 13500.0000 1.82593941 14250.0000 1.81519147 15000.0000 1.805703939437 3 18
 16000.0000 1.79814520 17000.0000 1.79134140 18000.0000 1.776920929437 3 18
 19000.0000 1.76254138 20000.0000 1.74575301 21000.0000 1.726366579437 3 18
 22000.0000 1.70879476 23000.0000 1.69692448 24000.0000 1.683257279437 3 18
 25500.0000 1.66142665 27000.0000 1.64492621 28000.0000 1.651188729437 3 18
 30000.0000 1.65012352 32000.0000 1.64267796 34000.0000 1.647742849437 3 18
 36000.0000 1.63999510 38000.0000 1.63057252 40000.0000 1.623534919437 3 18
 42500.0000 1.61659407 45000.0000 1.61083609 47500.0000 1.608999869437 3 18
 50000.0000 1.60699685 52500.0000 1.60099001 55000.0000 1.579709229437 3 18
 57500.0000 1.54457760 60000.0000 1.52677381 63000.0000 1.525740949437 3 18
 66000.0000 1.53099915 69000.0000 1.53547431 72000.0000 1.530272479437 3 18
 76000.0000 1.52062947 80000.0000 1.50827194 84000.0000 1.497137759437 3 18
 88000.0000 1.49289278 92000.0000 1.50050247 96000.0000 1.509998429437 3 18
 100000.000 1.51692002 105000.000 1.52077196 110000.000 1.518924959437 3 18
 115000.000 1.51138301 120000.000 1.50088571 127500.000 1.485379229437 3 18
 135000.000 1.47319397 142500.000 1.46371951 150000.000 1.456080389437 3 18
 160000.000 1.44976138 170000.000 1.44637046 180000.000 1.446984899437 3 18
 190000.000 1.44975281 200000.000 1.45454866 210000.000 1.461639519437 3 18
 220000.000 1.46935923 230000.000 1.47884945 240000.000 1.490846189437 3 18
 255000.000 1.50448517 270000.000 1.51255492 280000.000 1.518771849437 3 18
 300000.000 1.52244293 320000.000 1.52283288 340000.000 1.520100139437 3 18
 360000.000 1.51736737 380000.000 1.51881560 400000.000 1.525900049437 3 18
 425000.000 1.53796584 450000.000 1.55070731 475000.000 1.562162689437 3 18
 500000.000 1.56908977 525000.000 1.57330123 550000.000 1.575601719437 3 18
 575000.000 1.57790220 600000.000 1.58043273 630000.000 1.583535499437 3 18
 660000.000 1.59144516 690000.000 1.60957953 720000.000 1.639983359437 3 18
 760000.000 1.67482742 800000.000 1.68921990 840000.000 1.690850069437 3 18
 880000.000 1.69181419 920000.000 1.70354441 960000.000 1.720411149437 3 18
 1000000.00 1.75518022 1100000.00 1.80501424 1200000.00 1.856524509437 3 18
 1300000.00 1.90259926 1400000.00 1.93784935 1500000.00 1.963949439437 3 18
 1600000.00 1.98224948 1700000.00 1.99364950 1800000.00 1.998649479437 3 18
 1900000.00 1.99652322 2000000.00 1.99018236 2100000.00 1.979777849437 3 18
 2200000.00 1.96772652 2300000.00 1.95567520 2400000.00 1.943623899437 3 18
 2500000.00 1.93157257 2600000.00 1.91952125 2700000.00 1.907595139437 3 18
 2800000.00 1.89632982 2900000.00 1.88516085 3000000.00 1.873991889437 3 18
 3100000.00 1.86282290 3200000.00 1.85165393 3300000.00 1.840484969437 3 18
 3400000.00 1.83132538 3500000.00 1.82417521 3600000.00 1.817025049437 3 18
 3700000.00 1.80987486 3800000.00 1.80272469 3900000.00 1.795574529437 3 18
 4000000.00 1.78611012 4100000.00 1.77433150 4200000.00 1.762552759437 3 18
 4300000.00 1.74978707 4400000.00 1.73607895 4500000.00 1.722370849437 3 18
 4600000.00 1.70866272 4700000.00 1.69495460 4800000.00 1.683325589437 3 18
 4900000.00 1.67377567 5000000.00 1.66692569 5100000.00 1.662775659437 3 18
 5200000.00 1.66097566 5300000.00 1.66152571 5400000.00 1.664750889437 3 18
 5500000.00 1.67065117 5600000.00 1.68130132 5700000.00 1.696701339437 3 18
 5800000.00 1.71680138 5900000.00 1.74160147 6000000.00 1.771621549437 3 18
 6100000.00 1.80686159 6200000.00 1.84349906 6300000.00 1.888520959437 3 18
 6400000.00 1.93494025 6500000.00 1.98135955 6600000.00 2.027778859437 3 18
 6700000.00 2.07268574 6800000.00 2.10851816 6900000.00 2.142838179437 3 18
 7000000.00 2.17129817 7100000.00 2.19389818 7200000.00 2.215073239437 3 18
 7300000.00 2.22769849 7400000.00 2.23889880 7500000.00 2.246859079437 3 18
 7600000.00 2.25157930 7700000.00 2.25594449 7800000.00 2.258179449437 3 18
 7900000.00 2.26005935 8000000.00 2.25979936 8100000.00 2.257399479437 3 18
 8200000.00 2.25499958 8300000.00 2.25259969 8400000.00 2.250199809437 3 18
 8500000.00 2.24779991 8600000.00 2.24540002 8700000.00 2.243000139437 3 18
 8800000.00 2.24060024 8900000.00 2.23820035 9000000.00 2.236700379437 3 18
 9100000.00 2.23610032 9200000.00 2.23550026 9300000.00 2.234900209437 3 18
 9400000.00 2.23430015 9500000.00 2.23370009 9600000.00 2.233100049437 3 18
 9700000.00 2.23249998 9800000.00 2.23189992 9900000.00 2.231299879437 3 18
 10000000.0 2.23026487 10100000.0 2.22879492 10200000.0 2.227324989437 3 18
 10300000.0 2.22585504 10400000.0 2.22438509 10500000.0 2.222915159437 3 18
 10600000.0 2.22144520 10700000.0 2.21997526 10800000.0 2.218505329437 3 18
 10900000.0 2.21703537 11000000.0 2.21658542 11100000.0 2.217155469437 3 18
 11200000.0 2.21772550 11300000.0 2.21829554 11400000.0 2.218865589437 3 18
 11500000.0 2.21943562 11600000.0 2.22000566 11700000.0 2.220575709437 3 18
 11800000.0 2.22114574 11900000.0 2.22171578 12000000.0 2.224250799437 3 18
 12100000.0 2.22875077 12200000.0 2.23325075 12300000.0 2.237750739437 3 18
 12400000.0 2.24225071 12500000.0 2.24675068 12600000.0 2.251250669437 3 18
 12700000.0 2.25575064 12800000.0 2.26025062 12900000.0 2.264750609437 3 18
 13000000.0 2.27067552 13100000.0 2.27802539 13200000.0 2.285375269437 3 18
 13300000.0 2.29272512 13400000.0 2.30007499 13500000.0 2.307424869437 3 18
 13600000.0 2.31477472 13700000.0 2.32212459 13800000.0 2.329474469437 3 18
 13900000.0 2.33682432 14000000.0 2.34417419 14100000.0 2.351524059437 3 18
 14200000.0 2.35887392 14300000.0 2.36622379 14400000.0 2.373573659437 3 18
 14500000.0 2.38092352 14600000.0 2.38827339 14700000.0 2.395623259437 3 18
 14800000.0 2.40297312 14900000.0 2.41032299 15000000.0 2.413197989437 3 18
 15100000.0 2.41159810 15200000.0 2.40999822 15300000.0 2.408398339437 3 18
 15400000.0 2.40679845 15500000.0 2.40519857 15600000.0 2.403598699437 3 18
 15700000.0 2.40199881 15800000.0 2.40039892 15900000.0 2.398799049437 3 18
 16000000.0 2.39529910 16100000.0 2.38989911 16200000.0 2.384499119437 3 18
 16300000.0 2.37909912 16400000.0 2.37369912 16500000.0 2.368299139437 3 18
 16600000.0 2.36289913 16700000.0 2.35749914 16800000.0 2.352099149437 3 18
 16900000.0 2.34669915 17000000.0 2.33954928 17100000.0 2.330649539437 3 18
 17200000.0 2.32174978 17300000.0 2.31285003 17400000.0 2.303950289437 3 18
 17500000.0 2.29505054 17600000.0 2.28615079 17700000.0 2.277251049437 3 18
 17800000.0 2.26835129 17900000.0 2.25945154 18000000.0 2.253901649437 3 18
 18100000.0 2.25170158 18200000.0 2.24950153 18300000.0 2.247301479437 3 18
 18400000.0 2.24510141 18500000.0 2.24290136 18600000.0 2.240701309437 3 18
 18700000.0 2.23850124 18800000.0 2.23630119 18900000.0 2.234101139437 3 18
 19000000.0 2.23405117 19100000.0 2.23615130 19200000.0 2.238251439437 3 18
 19300000.0 2.24035156 19400000.0 2.24245169 19500000.0 2.244551839437 3 18
 19600000.0 2.24665196 19700000.0 2.24875209 19800000.0 2.250852229437 3 18
 19900000.0 2.25295235 20000000.0 0.0                             9437 3 18
                                                                  9437 3  0
                                                                  9437 0  0
 9.42390E+4 2.36999E+2          0          0          0          1943733 18
 0.000000+0 0.000000+0          0         18          0          4943733 18
 0.000000+0 0.000000+0          0          1         16          8943733 18
 1.000000-5 0.000000+0 1.000000+0 2.500000-3 3.000000+2 3.600000-3943733 18
 3.000000+4 4.900000-3 1.000000+5 6.400000-3 2.000000+5 0.000000+0943733 18
 1.500000+7 1.000000-2 2.000000+7 0.000000+0                      943733 18
 0.000000+0 0.000000+0          0          1          8          4943733 18
 1.000000-5 0.000000+0 1.000000+0 4.000000-4 1.500000+7 0.000000+0943733 18
 2.000000+7 0.000000+0                                            943733 18
 0.000000+0 0.000000+0          0          1         18          9943733 18
 1.000000-5 0.000000+0 2.000000+5 4.900000-5 1.000000+6 4.900000-5943733 18
 2.000000+6 4.900000-5 3.000000+6 1.400000-4 6.000000+6 1.000000-4943733 18
 8.000000+6 2.250000-4 1.500000+7 0.000000+0 2.000000+7 0.000000+0943733 18
 0.000000+0 0.000000+0          1          5         21          6943733 18
 1.000000-5 2.530000-2 9.000000-2 2.500000-1 1.000000+0 2.000000+7943733 18
 1.000000-4 3.230000-5 3.420000-5 3.230000-5 0.000000+0 3.600000-5943733 18
 3.390000-5 3.200000-5 0.000000+0 6.400000-5 3.380000-5 0.000000+0943733 18
 2.560000-4 0.000000+0 0.000000+0                                 943733 18
                                                                  943733  0
                                                                  9437 0  0
                                                                     0 0  0
 9.52410E+4 2.38986E+2          0          1         34         109543 1451
 0.0        1.00000E+0          0          0          0          69543 1451
 1.00000E+0 2.00000E+7          0          0         10       20029543 1451
 3.00000E+2 0.0                 1          0        110          39543 1451
 95-Am-241 JAERI      EVAL-MAR88 T.NAKAGAWA                       9543 1451
                      DIST-Feb2004                                9543 1451
----IRDF-2002         MATERIAL 9543                               9543 1451
-----INCIDENT NEUTRON DATA                                        9543 1451
------ENDF-6 FORMAT                                               9543 1451
****************************************************************  9543 1451
 95-AM-241 JAERI      EVAL-MAR88 T.NAKAGAWA                       9543 1451
JAERI-M 89-008        DIST-SEP89                                  9543 1451
----JENDL-3.2         MATERIAL 9543                               9543 1451
 <DOSIMETRY CROSS SECTION>                                        9543 1451
      AM-241 FISSION                                              9543 1451
****************************************************************  9543 1451
HISTORY                                                           9543 1451
82-03 EVALUATION FOR JENDL-2 WAS MADE BY Y.KIKUCHI (JAERI) /1/.   9543 1451
88-03 RE-EVALUATION FOR JENDL-3 WAS MADE BY T.NAKAGAWA (JAERI)    9543 1451
      /2/.                                                        9543 1451
                                                                  9543 1451
MF=1  GENERAL INFORMATION                                         9543 1451
  MT=451   COMMENT AND DICTIONARY                                 9543 1451
                                                                  9543 1451
MF=2,MT=151   RESONANCE PARAMETERS                                9543 1451
  RESOLVED RESONANCES FOR MLBW FORMULA : 1.0E-5 - 150 EV          9543 1451
        DATA OF DERRIEN AND LUCAS /5/ WERE ADOPTED AND 5 NEGATIVE 9543 1451
        RESONANCES WERE ADDED.  VALUES OF TOTAL SPIN J WERE       9543 1451
        REPLACED WITH ARBITRARILY ASSUMED VALUES.                 9543 1451
                                                                  9543 1451
  UNRESOLVED RESONANCES : 150 EV - 30 KEV                         9543 1451
        PARAMETERS WERE DETERMINED BY USING ASREP/6/ SO AS TO     9543 1451
        REPRODUCE THE CAPTURE CROSS SECTION MEASURED BY VANPRAET  9543 1451
        ET AL. /7/ AND THE FISSION CROSS SECTION BY DABBS ET AL.  9543 1451
        /8/.                                                      9543 1451
          ENERGY INDEPENDENT PARAMETERS:                          9543 1451
            R=9.37 FM, GAM-G=0.044 EV,  DOBS=0.4 EV               9543 1451
          ENERGY DEPENDENT PARAMETERS:                            9543 1451
            AT 150 EV: S0= 1.08E-4, S1=2.72E-4, WF=0.24 MILLI-EV  9543 1451
            AT 30 KEV: S0= 0.79E-4, S1=1.99E-4, WF=0.30 MILLI-EV  9543 1451
                                                                  9543 1451
  CALCULATED 2200-M/S CROSS SECTIONS AND RESONANCE INTEGRALS      9543 1451
                      2200 M/S VALUE       RES. INT.              9543 1451
        TOTAL           614.6   B            -                    9543 1451
        ELASTIC          11.13  B            -                    9543 1451
        FISSION           3.018 B           13.9  B               9543 1451
        CAPTURE         600.4   B           1300  B               9543 1451
                                                                  9543 1451
MF=3  NEUTRON CROSS SECTIONS                                      9543 1451
                                                                  9543 1451
  MT=1,2  TOTAL AND ELASTIC SCATTERING CROSS SECTIONS             9543 1451
        CALCULATED WITH OPTICAL AND STATISTICAL MODELS BY USING   9543 1451
        CASTHY/9/.  OPTICAL POTENTIAL PARAMETERS/10/ WERE OBTAINED9543 1451
        BY FITTING THE DATA OF PHILLIPS AND HOWE /11/ :           9543 1451
           V = 43.4 - 0.107*EN                      (MEV)         9543 1451
           WS= 6.95 - 0.339*EN + 0.0531*EN**2       (MEV)         9543 1451
           WV= 0             , VSO = 7.0            (MEV)         9543 1451
           R = RSO = 1.282   , RS = 1.29            (FM)          9543 1451
           A = ASO = 0.60    , B  = 0.5             (FM)          9543 1451
                                                                  9543 1451
                                                                  9543 1451
  MT=18      FISSION CROSS SECTION                                9543 1451
        EVALUATED ON THE BASIS OF THE DATA BY DABBS ET AL./8/     9543 1451
                                                                  9543 1451
MF=33  COVARIANCES OF NEUTRON CROSS SECTIONS FROM JENDL/D-99      9543 1451
 MT=18                                                            9543 1451
     ADOPTED FROM ENDF/B-VI/12/.                                  9543 1451
                                                                  9543 1451
                                                                  9543 1451
REFERENCES                                                        9543 1451
 1) KIKUCHI Y.: JAERI-M 82-096 (1982).                            9543 1451
 2) NAKAGAWA T.: JAERI-M 88-008 (1989).                           9543 1451
 5) DERRIEN H. AND LUCAS B.: 1975 WASHINGTON, P.637,              9543 1451
    NBS-SP-425 (1975).                                            9543 1451
 6) KIKUCHI Y.: PRIVATE COMMUNICATION.                            9543 1451
 7) VANPRAET G. ET AL.: 1985 SANTA FE, 1, 493 (1986).             9543 1451
 8) DABBS J.W.T. ET AL.: NUCL. SCI. ENG., 83, 22 (1983).          9543 1451
 9) IGARASI S.: J.NUCL.SCI.TECHNOL.,12,67 (1975).                 9543 1451
10) IGARASI S. AND NAKAGAWA T.: JAERI-M 8342 (1979).              9543 1451
11) PHILLIPS T.W. AND HOWE R.E.: NUCL. SCI. ENG., 69, 375 (1979). 9543 1451
12) ZHOU DELIN, GU FUHUA ET AL.: ENDF/B-VI (MAT=9543) (1988).     9543 1451
****************************************************************  9543 1451
                                                                  9543 1451
                                                                  9543 1451
                                                                  9543 1451
                                                                  9543 1451
 ***************** Program LINEAR (VERSION 2002-1) ***************9543 1451
 For All Data Greater than 1.0000E-10 barns in Absolute Value     9543 1451
 Data Linearized to Within an Accuracy of .100000000 per-cent     9543 1451
 ***************** Program RECENT (VERSION 2002-1) ***************9543 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9543 1451
 Data Linearized to within an Accuracy of .100000000 per-cent     9543 1451
 ***************** Program SIGMA1 (VERSION 2002-1) ***************9543 1451
 Data Doppler Broadened to 300.000000 Kelvin                      9543 1451
 for All Data Greater than 1.0000E-10 barns in Absolute Value     9543 1451
 Data Linearized to Within an Accuracy pf .100000000 per-cent     9543 1451
 ***************** Program FIXUP (Version 2002-1) ****************9543 1451
 Corrected ZA/AWR in All Sections-----------------------------Yes 9543 1451
 Corrected Thresholds-----------------------------------------Yes 9543 1451
 Extended Cross Sections to 20 MeV----------------------------No  9543 1451
 Allow Cross Section Deletion---------------------------------No  9543 1451
 Allow Cross Section Reconstruction---------------------------No  9543 1451
 Make All Cross Sections Non-Negative-------------------------Yes 9543 1451
 Delete Energies Not in Ascending Order-----------------------Yes 9543 1451
 Deleted Duplicate Points-------------------------------------Yes 9543 1451
 Check for Ascending MAT/MF/MT Order--------------------------Yes 9543 1451
 Check for Legal MF/MT Numbers--------------------------------Yes 9543 1451
 Allow Creation of Missing Sections---------------------------No  9543 1451
 Allow Insertion of Energy Points-----------------------------No  9543 1451
 Create Uniform Energy Grid-----------------------------------No  9543 1451
 Delete Section if Cross Section =0 at All Energies-----------Yes 9543 1451
 ***************** Program GROUPIE (VERSION 2002-1) **************9543 1451
 Unshielded Group Averages Using  640 Groups                      9543 1451
 Weighting Spectrum: Flat (Constant) Spectrum                     9543 1451
                                1        451        117          19543 1451
                                3         18        217          19543 1451
                               33         18         16          09543 1451
                                                                  9543 1  0
                                                                  9543 0  0
 9.52410E+4 2.38986E+2          0          0          0          09543 3 18
 1.99999E+8 1.99999E+8          0          0          1        6419543 3 18
        641          1                                            9543 3 18
 .000100000 51.3052773 .000105000 50.1079487 .000110000 48.97045709543 3 18
 .000115000 47.9222736 .000120000 46.6968760 .000127500 45.33727459543 3 18
 .000135000 44.0950790 .000142500 42.9534963 .000150000 41.71795199543 3 18
 .000160000 40.4307858 .000170000 39.2595742 .000180000 38.18445049543 3 18
 .000190000 37.1841024 .000200000 36.2707160 .000210000 35.41134409543 3 18
 .000220000 34.6130933 .000230000 33.8713750 .000240000 33.00070879543 3 18
 .000255000 32.0439328 .000270000 31.3056049 .000280000 30.48366259543 3 18
 .000300000 29.4812057 .000320000 28.5734378 .000340000 27.74200399543 3 18
 .000360000 26.9761460 .000380000 26.2769608 .000400000 25.54796499543 3 18
 .000425000 24.8019302 .000450000 24.1202708 .000475000 23.49517349543 3 18
 .000500000 22.9072419 .000525000 22.3699039 .000550000 21.86226549543 3 18
 .000575000 21.3900875 .000600000 20.9060918 .000630000 20.40814609543 3 18
 .000660000 19.9524634 .000690000 19.5153070 .000720000 19.05007149543 3 18
 .000760000 18.5511751 .000800000 18.0878986 .000840000 17.66326379543 3 18
 .000880000 17.2596280 .000920000 16.8895239 .000960000 16.53503649543 3 18
 .001000000 16.1675054 .001050000 15.7837856 .001100000 15.42401349543 3 18
 .001150000 15.0924242 .001200000 14.7011985 .001275000 14.27109619543 3 18
 .001350000 13.8777583 .001425000 13.5105958 .001500000 13.12041729543 3 18
 .001600000 12.7123489 .001700000 12.3397233 .001800000 11.99500599543 3 18
 .001900000 11.6800064 .002000000 11.3878334 .002100000 11.11357199543 3 18
 .002200000 10.8622996 .002300000 10.6219594 .002400000 10.34754989543 3 18
 .002550000 10.0424243 .002700000 9.80361723 .002800000 9.543892829543 3 18
 .003000000 9.22394878 .003200000 8.93261626 .003400000 8.666118549543 3 18
 .003600000 8.42306741 .003800000 8.19866358 .004000000 7.964785439543 3 18
 .004250000 7.72734253 .004500000 7.51048463 .004750000 7.307647269543 3 18
 .005000000 7.12124209 .005250000 6.94814792 .005500000 6.784296339543 3 18
 .005750000 6.63379085 .006000000 6.47555913 .006300000 6.317125229543 3 18
 .006600000 6.16762235 .006900000 6.02971171 .007200000 5.879208639543 3 18
 .007600000 5.71777295 .008000000 5.56999636 .008400000 5.431170789543 3 18
 .008800000 5.30182317 .009200000 5.18150547 .009600000 5.066938199543 3 18
 .010000000 4.94821064 .010500000 4.82262453 .011000000 4.707663779543 3 18
 .011500000 4.59795348 .012000000 4.47213784 .012750000 4.332816099543 3 18
 .013500000 4.20373606 .014250000 4.08529824 .015000000 3.958938599543 3 18
 .016000001 3.82620741 .017000001 3.70438526 .017999999 3.593050599543 3 18
 .018999999 3.49065480 .020000000 3.39456853 .021000000 3.306660139543 3 18
 .022000000 3.22388318 .023000000 3.14682311 .024000000 3.056392539543 3 18
 .025500000 2.95907786 .027000001 2.88205910 .028000001 2.798584939543 3 18
 .029999999 2.69607102 .032000002 2.60300703 .034000002 2.518435849543 3 18
 .035999998 2.44168652 .037999999 2.37134032 .039999999 2.298442269543 3 18
 .042500000 2.22520397 .045000002 2.15915714 .047499999 2.098291019543 3 18
 .050000001 2.04303612 .052499998 1.99258453 .055000000 1.945325219543 3 18
 .057500001 1.90302922 .059999999 1.85914432 .063000001 1.816529909543 3 18
 .066000000 1.77732441 .068999998 1.74195361 .071999997 1.705121189543 3 18
 .075999998 1.66726015 .079999998 1.63503985 .083999999 1.606556369543 3 18
 .088000000 1.58222929 .092000000 1.56226398 .096000001 1.544740649543 3 18
 .100000001 1.52947423 .104999997 1.51748803 .109999999 1.510657309543 3 18
 .115000002 1.50769305 .119999997 1.51101174 .127499998 1.524423069543 3 18
 .135000005 1.54870410 .142499998 1.58430580 .150000006 1.642207309543 3 18
 .159999996 1.73260227 .170000002 1.85412976 .180000007 2.015956329543 3 18
 .189999998 2.22915507 .200000003 2.51245839 .209999993 2.894467449543 3 18
 .219999999 3.41860254 .230000004 4.16063103 .239999995 5.626778769543 3 18
 .254999995 8.82815710 .270000011 13.8569548 .280000001 24.68428309543 3 18
 .300000012 32.9036091 .319999993 20.0317145 .340000004 8.912000729543 3 18
 .360000014 4.55430795 .379999995 2.78968183 .400000006 1.902517629543 3 18
 .425000012 1.45804726 .449999988 1.32153894 .474999994 1.452995389543 3 18
 .500000000 2.06823238 .524999976 4.16458177 .550000012 8.955286699543 3 18
 .574999988 9.04637758 .600000024 3.80727810 .629999995 1.488845779543 3 18
 .660000026 .835810765 .689999998 .583407230 .720000029 .4509664349543 3 18
 .759999990 .380905725 .800000012 .353868563 .839999974 .3527661479543 3 18
 .879999995 .372648705 .920000017 .414908308 .959999979 .4864857529543 3 18
 1.00000000 .623933129 1.04999995 .908768459 1.10000002 1.544761279543 3 18
 1.14999998 3.46717767 1.20000005 21.5180722 1.27499998 21.98491769543 3 18
 1.35000002 2.79636200 1.42499995 .910301031 1.50000000 .4298381579543 3 18
 1.60000002 .249932347 1.70000005 .205604116 1.79999995 .5592888289543 3 18
 1.89999998 1.47115486 2.00000000 .219907486 2.09999990 .1377515599543 3 18
 2.20000005 .252101211 2.29999995 1.80450792 2.40000010 .8271200229543 3 18
 2.54999995 2.20098818 2.70000005 .199515711 2.79999995 .0688983929543 3 18
 3.00000000 .039982739 3.20000005 .034907434 3.40000010 .0396874759543 3 18
 3.59999990 .072453611 3.79999995 1.31740811 4.00000000 .4545130579543 3 18
 4.25000000 .072191967 4.50000000 .118200170 4.75000000 2.114117579543 3 18
 5.00000000 1.19108663 5.25000000 13.7127105 5.50000000 1.312201169543 3 18
 5.75000000 .244578528 6.00000000 1.24055456 6.30000019 .0703622439543 3 18
 6.59999990 .156676942 6.90000010 .024090727 7.19999981 .0253565279543 3 18
 7.59999990 .064327211 8.00000000 .192074768 8.39999962 .0303044469543 3 18
 8.80000019 .786315512 9.19999981 .179092954 9.60000038 4.106449809543 3 18
 10.0000000 .356873590 10.5000000 .252712792 11.0000000 .2142557149543 3 18
 11.5000000 .036851057 12.0000000 .018356860 12.7500000 .0459599409543 3 18
 13.5000000 .052559219 14.2500000 2.89463552 15.0000000 .1253619489543 3 18
 16.0000000 .965078754 17.0000000 .385138061 18.0000000 .0207186779543 3 18
 19.0000000 .018943345 20.0000000 .061033820 21.0000000 .0530467249543 3 18
 22.0000000 .082007664 23.0000000 .355252688 24.0000000 .2939772919543 3 18
 25.5000000 .560448508 27.0000000 .183892344 28.0000000 .1891116289543 3 18
 30.0000000 .272887127 32.0000000 .085314866 34.0000000 .2306432359543 3 18
 36.0000000 .807807646 38.0000000 .551630875 40.0000000 .1508855449543 3 18
 42.5000000 .271310533 45.0000000 .165348949 47.5000000 .1528098829543 3 18
 50.0000000 .296220510 52.5000000 .060686151 55.0000000 .1331005929543 3 18
 57.5000000 .040280714 60.0000000 .043411491 63.0000000 .1992882169543 3 18
 66.0000000 .068171197 69.0000000 .310106001 72.0000000 .0561975089543 3 18
 76.0000000 .260455098 80.0000000 .175890724 84.0000000 .1405369599543 3 18
 88.0000000 .078253970 92.0000000 .258987041 96.0000000 .1515834449543 3 18
 100.000000 .251279864 105.000000 .293826556 110.000000 .1412264799543 3 18
 115.000000 .126536451 120.000000 .174290158 127.500000 .1521780509543 3 18
 135.000000 .282033747 142.500000 .159172258 150.000000 .1800625889543 3 18
 160.000000 .170428121 170.000000 .161564109 180.000000 .1533616509543 3 18
 190.000000 .145733788 200.000000 .140323995 210.000000 .1370180329543 3 18
 220.000000 .133936250 230.000000 .131052574 240.000000 .1280735459543 3 18
 255.000000 .129565957 270.000000 .132107517 280.000000 .1351273289543 3 18
 300.000000 .139126207 320.000000 .143089220 340.000000 .1453491219543 3 18
 360.000000 .137595202 380.000000 .128697674 400.000000 .1193435079543 3 18
 425.000000 .109672615 450.000000 .103294662 475.000000 .0998990509543 3 18
 500.000000 .096778133 525.000000 .093855034 550.000000 .0911159329543 3 18
 575.000000 .088568598 600.000000 .086145871 630.000000 .0838297369543 3 18
 660.000000 .081694916 690.000000 .079662060 720.000000 .0774746649543 3 18
 760.000000 .075146614 800.000000 .072809837 840.000000 .0704543959543 3 18
 880.000000 .068255689 920.000000 .066195938 960.000000 .0642602369543 3 18
 1000.00000 .062210457 1050.00000 .060053350 1100.00000 .0580342259543 3 18
 1150.00000 .056137361 1200.00000 .053921502 1275.00000 .0514492349543 3 18
 1350.00000 .049160922 1425.00000 .047032897 1500.00000 .0451896189543 3 18
 1600.00000 .043598478 1700.00000 .042151847 1800.00000 .0408282849543 3 18
 1900.00000 .039610752 2000.00000 .038898213 2100.00000 .0386550219543 3 18
 2200.00000 .038422564 2300.00000 .038254526 2400.00000 .0380639089543 3 18
 2550.00000 .037914118 2700.00000 .037820646 2800.00000 .0377258699543 3 18
 3000.00000 .037612621 3200.00000 .037514037 3400.00000 .0372424449543 3 18
 3600.00000 .035884975 3800.00000 .034439064 4000.00000 .0329269199543 3 18
 4250.00000 .031371299 4500.00000 .030103309 4750.00000 .0290752989543 3 18
 5000.00000 .028126982 5250.00000 .027235710 5500.00000 .0263973269543 3 18
 5750.00000 .025613688 6000.00000 .024934246 6300.00000 .0243536219543 3 18
 6600.00000 .023821664 6900.00000 .023316969 7200.00000 .0227771749543 3 18
 7600.00000 .022205621 8000.00000 .021788948 8400.00000 .0215181129543 3 18
 8800.00000 .021270344 9200.00000 .021042443 9600.00000 .0208317459543 3 18
 10000.0000 .020546983 10500.0000 .020192328 11000.0000 .0198674259543 3 18
 11500.0000 .019576077 12000.0000 .019222437 12750.0000 .0188465129543 3 18
 13500.0000 .018497288 14250.0000 .018185644 15000.0000 .0178422659543 3 18
 16000.0000 .017476674 17000.0000 .017145879 18000.0000 .0168440749543 3 18
 19000.0000 .016566722 20000.0000 .016388102 21000.0000 .0163000329543 3 18
 22000.0000 .016214043 23000.0000 .016140548 24000.0000 .0160465039543 3 18
 25500.0000 .015872482 27000.0000 .015713719 28000.0000 .0155446019543 3 18
 30000.0000 .015288990 32000.0000 .015067719 34000.0000 .0148486819543 3 18
 36000.0000 .014654185 38000.0000 .014466451 40000.0000 .0143010409543 3 18
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