ENDF6 Cross Sections with many data points does not extract all data points from endf file

[sasqueechie]

Hi, I am trying to extract all the datapoints from cross section files for things like U-235 and I am running into an issue where it seems like some part of the reader/extraction method in your code stops processing information after a certain amount of data points OR its like its missing a part of the file when reading and only getting a chunk of it.

The following is what I am running with:

import sandy import pandas as pd import plotly.graph_objects as go import os from scripts import functions as fn

Read ENDF-6 data for a specific nuclide (e.g., U-235)

nuclide = fn.format_input(input("Input nuclide of interest\n"))

fy = sandy.get_endf6_file("endfb_80", "xs", 922350)

fyr = fy.get_records()

fydict = fy.read_section(mat=9228, mf=3, mt=18)

here you can see that it gets most of the upper energy datapoints in detail but then jumps to 1e-5 all of a sudden.

print(fydict['E'], fydict['XS'])

Create a scatter plot with log-log scale

fig = go.Figure()

fig.add_trace(go.Scatter(x=fydict['E'], y=fydict['XS'], mode="lines", name=f"{nuclide}", text="test"))

fig.show()

---

[sasqueechie]

[sasqueechie]

Looking at the ENDF file itself, it looks like it may actually be extracting all the data available, it is just not plotting/processing

the resonance/capture regions of the graph as seen here:

Retrieved by E4-Web: 2024/06/29,01:27:57 1 0 0 0 9.223500+4 2.330248+2 1 1 0 19228 1451 1 0.000000+0 0.000000+0 0 0 0 69228 1451 2 1.000000+0 3.000000+7 0 0 10 89228 1451 3 0.000000+0 0.000000+0 0 0 656 29228 1451 4 92-U -235 IAEA EVAL-NOV17 IAEA CIELO Collaboration 9228 1451 5 NDS 148, 1 (2018) DIST-FEB18 9228 1451 6 ----ENDF/B-VIII.0 MATERIAL 9228 9228 1451 7 -----INCIDENT NEUTRON DATA 9228 1451 8 ------ENDF-6 FORMAT 9228 1451 9 9228 1451 10 Source data included in the evaluation 9228 1451 11 9228 1451 12 Version u235ib46o28t6DNcnu5ef0STzc 9228 1451 13 Source Roberto\ib46\u235.out 9228 1451 14 Resonance Marco\u235-res-28t+-stan.endf 9228 1451 15 Delayed Resonance\U235_dlyd_e71.end 9228 1451 16 Fission Resonance\U235_fissp1_gbDNnu5jd.end 9228 1451 17 Fission xs Resonance\U235_fisxs_STD-2017-final-11Feb17-vp-at.end 9228 1451 18 Mod.Fis.xs 9228 1451 19 Fiss.E Resonance\fis_energy.endf 9228 1451 20 Fiss.gamma Fission_gamma\u5-mf6mt18-pnu-ng_Ionel20171116.endf 9228 1451 21 Covariance Covariance\u235_cov_zc.endf 9228 1451 22 Inel.Cov. Covariance\u235_cov-inel.endf 9228 1451 23 9228 1451 24 9228 1451 25 General information for Version u235ib46o28t6DNcnu5ef0STzc 9228 1451 26 (with corrections for missing data above 20 MeV, 19-Apr-2016) 9228 1451 27 9228 1451 28

• The basis is the Empire calculation 'ib46' (with capture at the 9228 1451 29 upper limit of Jandel data, tuned to better fit Wallner and 9228 1451 30 Allmond data). The model parameters are tuned to improve 9228 1451 31 agreement with Kammerdiener ddx at 14 MeV. 9228 1451 32 Non-threshold reactions e.g. (n,a), (n,x) are extrapolated 9228 1451 33 down to thermal energies as 1/v. 9228 1451 34 9228 1451 35
• Resonance parameters are ORNL version 'o28t'. 9228 1451 36 9228 1451 37
• Prompt nu-bar near thermal energies is evaluated by Pronyaev, 9228 1451 38 The prompt nu-bar (nu_p) values follow Reed data taken from the 9228 1451 39 EXFOR extry 10427003 up to 26.4 eV. Simon and Frehaut data 9228 1451 40 shapes are identical; they were normalized to Reed data in the 9228 1451 41 interval 10-27 eV; Simon data were adopted from 30 eV to 75 eV. 9228 1451 42 Subsequently, all nubar data were renormalized to the thermal 9228 1451 43 value of 2.41161, which agrees with Standards-2017 within 9228 1451 44 quoted uncertainties. 9228 1451 45 The energy grid of Reed data is modified slightly by shifting 9228 1451 46 the energies at 1.135 eV to 1.5 eV and 2.370 eV to 2.25 eV 9228 1451 47 without changing the nu_p values, thus narrowing the dip at 9228 1451 48 1 eV in order to increase the reactivity of thermal systems. 9228 1451 49 In addition, the dip at 2.04 eV is made shallower by 0.02, 9228 1451 50 which is within the experimental uncertainty. Above the range 9228 1451 51 of Reed data, nu_p approximately follows the data of Frehaut. 9228 1451 52 A point is added at 30 MeV (approximately extrapolating Frehaut 9228 1451 53 data) for continuity. 9228 1451 54 The nu_prompt in the range 30-70 keV is reduced, similarly 9228 1451 55 like in the JENDL-4.0 evaluation. 9228 1451 56 Additionally, the revised interpretation of Standards-2017 9228 1451 57 (Jul.2017) caused an increase in the fission cross section 9228 1451 58 in the URR, which is compensated by a decrease in nu_prompt 9228 1451 59 starting at 100 eV, reaching the minimum value at 20 keV. 9228 1451 60 The nu_prompt is also reduced by up to 0.28 in the energy 9228 1451 61 range 1.2-2.4 MeV to improve the performance in integral 9228 1451 62 benchmarks. This is thought to counteract the adjustment done 9228 1451 63 to nu_p in ENDF/B-VII.1, but the details are not known. 9228 1451 64 At higher energies, nu_p is equal to the ENDF/B-VII.1 9228 1451 65 evaluation. A point is added at at 30 MeV (approximately 9228 1451 66 extrapolating Frehaut data) for continuity. 9228 1451 67 9228 1451 68
• Fission cross section is from Standards (February 2017) 9228 1451 69 In the unresolved resonance range the shape is taken from 9228 1451 70 ENDF/B-VII.1, rescaled such that consistency with 9228 1451 71 group-averaged Standards is preserved. Above the unresolved 9228 1451 72 resonance range up to 75 keV the shape is from the linearised 9228 1451 73 Weston data, rescaled in a similar manner. Above 75 keV the 9228 1451 74 data are taken from Standards-2017 directly. 9228 1451 75 9228 1451 76
• Other fission data (except PFNS and cross sections) are 9228 1451 77 from ENDF/B-VII.1 9228 1451 78 9228 1451 79
• Delayed nu-bar and spectra are taken from ENDF/B-VII.1 9228 1451 80 9228 1451 81
• PFNS is as provided by P. Talou, consisting of:(10-Feb-2917) 9228 1451 82
  • IAEA (Capote) GMA evaluation (Standards-2017) at thermal 9228 1451 83
  • LANL (Neudecker) evaluation without chi-nu data below 5 MeV 9228 1451 84
  • LANL (Neudecker) 12/2016 evaluation above 5 MeV (using more 9228 1451 85 recent Chi-Nu experimental data) 9228 1451 86
  • PFNS at 20 MeV is duplicated to 30 MeV for completeness. 9228 1451 87 9228 1451 88
  • The prompt-gamma multiplicity is set to the Oberstedt [Ob2013] 9228 1451 89 value at thermal neutron incident energy, and at higher 9228 1451 90 incident energies is adjusted to reproduce the available 9228 1451 91 total gamma production data by Dake [Dr1970,Dr1974] 9228 1451 92 at several incident neutron energies. 9228 1451 93 9228 1451 94 [Dr1970] Darrell M. Drake, John C. Hopkins, C. S. Young, 9228 1451 95 and H. Conde, Nucl. Sci. Eng. 40, 294 (1970) 9228 1451 96 [Dr1974] DM Drake, Nucl. Sci. Eng. 55, 427 (1974) 9228 1451 97 9228 1451 98
  • The evaluation of the prompt gamma spectrum is based on CGMF 9228 1451 99 calculations [St14], which agree rasonably well with 9228 1451 100 experimental data by Obserstdt et al. [Ob2017]. At higher 9228 1451 101 energies, the evaluation is based on the (unpublished data) 9228 1451 102 by Nishio et al. [Ni2016]. The same spectrum was used for 9228 1451 103 the whole incident neutron energy range, which is consitent 9228 1451 104 with the good reproduction at several neutron incident 9228 1451 105 energies of the total production data by Drake. 9228 1451 106 9228 1451 107 [Ob2017] Oberstedt et. al., Phys. Rev. C 87, 051602(R) (2013). 9228 1451 108 [St14] I. Stetcu, P. Talou, T. Kawano, and M. Jandel, 9228 1451 109 Phys. Rev. C 90, 024617 (2014) 9228 1451 110 [Ni2016] Nishio et. al., private communication. 9228 1451 111 9228 1451 112
  • Fission neutron and gamma photon probabilities in the new 9228 1451 113 MF 6 format are included. 9228 1451 114 9228 1451 115 9228 1451 116
• Covariances of the cross sections in the resonance region 9228 1451 117 were determined from the covariances of the resonance 9228 1451 118 parameters because the latter is impractical due to the 9228 1451 119 large size (1.7 MB). The reactions cross sections with 9228 1451 120 covariances include elastic, fission, capture and the 9228 1451 121 corresponding cross-reaction covariances. 9228 1451 122 The total is defined as the sum of the partials. 9228 1451 123 9228 1451 124
• The covariance matrix prior for the cross sections and 9228 1451 125 Legendre coefficients of the angular distributions above 9228 1451 126 the resolved resonance range is generated using the Monte 9228 1451 127 Carlo technique by statistical sampling of the model 9228 1451 128 parameter in the Empire calculations. About 3000 ENDF 9228 1451 129 files were generated in the process. The first discrete 9228 1451 130 inelastic cross section is treated explicitly while all 9228 1451 131 the others are lumped into MT 851. Similarly, absorption 9228 1451 132 reactions MT649 (m,p) and the MT800 series discrete level 9228 1451 133 (n,alpha) are lumped into MT 852. This is necessary to 9228 1451 134 preserve unitarity without defining the covariances for 9228 1451 135 each reaction channel separately. 9228 1451 136 9228 1451 137
• The final covariance matrices for all reactions (including 9228 1451 138 cross reaction correlations) were obtained by the generalised 9228 1451 139 least-squares method in the GANDR system using measured 9228 1451 140 data from the EXFOR data base, except for fission, 9228 1451 141 in which case the Standards-2017 was input to GANDR 9228 1451 142 instead of the experimental data. A fully-correlated 9228 1451 143 uncertainty of 1.2 percent was included to account for the 9228 1451 144 unrecognised sources of uncertainty. 9228 1451 145 9228 1451 146 Experimental data marked with * were renormalized to the 9228 1451 147 IRDFF-1.05 cross sections @ https://www-nds.iaea.org/IRDFF/. 9228 1451 148 9228 1451 149 Total cross section 9228 1451 150 Selected data for fitting the total were the following: 9228 1451 151 Author Year EXFOR entry & subentry 9228 1451 152
  1. C.A.UTTLEY,ET.AL. (66) X4 21088 4 (60<E<960 keV) 9228 1451 153
  2. V.P.VERTEBNYY,ET.AL. (80) X4 40609 2 E=24.5 keV 9228 1451 154
  3. W.P.POENITZ,ET.AL. (81) X4 10935 6 E>200 keV 9228 1451 155
  4. J.Cabe,ET.AL. (73) X4 20480 17 9228 1451 156
  5. R.B.Schwartz,ET.AL. (74) X4 10280 5 *2 9228 1451 157
  6. F.L.Green,ET.AL. (73) X4 10588 2 9228 1451 158
  7. J.Cabe,ET.AL. (73) X4 20480 18 9228 1451 159
  8. W.P.POENITZ,ET.AL. (83) X4 12853 54 9228 1451 160
  9. D.G.FOSTER JR,ET.AL. (71) X4 10047 96 9228 1451 161
  10. J.M.PETERSON,ET.AL. (60) X4 11108 44 9228 1451 162 9228 1451 163 Capture cross section 9228 1451 164 Capture cross sections in EXFOR were measured mainly as 9228 1451 165 ratios. At present we could not include them in the 9228 1451 166 analysis. the Gwin data (EXFOR entry 14149003) are 9228 1451 167 labelled as "derived" and do not have the uncertainties. 9228 1451 168 Selected data for fitting were only: 9228 1451 169 M.Jandel,ET.AL. (12) X4 14149 3 2<E<60 keV 9228 1451 170 9228 1451 171 (n,2n) 9228 1451 172 The EXFOR entry #21568004 by J. Frehaut, which includes the 9228 1451 173 normalisation correction for new Standards, was selected. 9228 1451 174 Furthermore the data points below 8.5 MeV were rejected 9228 1451 175 because of experimental uncertainties due to the difficulty 9228 1451 176 in discriminating between neutrons from the (n,2n) and the 9228 1451 177 fission reaction. 9228 1451 178 The measurement #20795010 by D.S. Mather was rejected due 9228 1451 179 to unreasonably large discrepancy with the Frehaut data 9228 1451 180 and the shape, which was in contradiction with the model 9228 1451 181 calculation. 9228 1451 182 Hence, selected data for fitting (n,2n) were only: 9228 1451 183 J.Frehaut (2016) X4 entry 21568004 E>8 MeV 9228 1451 184 9228 1451 185 (n,3n) 9228 1451 186 There are only two data sets in EXFOR. Only the data in the 9228 1451 187 EXFOR entry 10795002 by Veeser were adopted, while the data 9228 1451 188 in the EXFOR entry #20795011 by Mather were rejected. 9228 1451 189 Hence, selected data for fitting (n,3n) were only: 9228 1451 190 L.R.Veeser,ET.AL. (78) X4 10795 2 9228 1451 191 9228 1451 192
• Cross-material covariances were prepared manually from the 9228 1451 193 Standards-2017 file. Unrecognised sources of uncertainties 9228 1451 194 (USU) of 1.2 percent were added to the cross-covariances of 9228 1451 195 the fission cross sections between different materials. 9228 1451 196 9228 1451 197
• The uncertainties of the delayed neutron multiplicities were 9228 1451 198 taken from Keepin rather than Tuttle, but taken from the 9228 1451 199 publication by Tuttle, which includes a comparative analysis: 9228 1451 200 R. J. Tuttle (1975) Delayed-Neutron Data for Reactor-Physics 9228 1451 201 Analysis, Nuclear Science and Engineering, 56:1, 37-71, 9228 1451 202 DOI: 10.13182/NSE75-A26620 9228 1451 203 9228 1451 204
• The initial covariances for the prompt fission neutron 9228 1451 205 multiplicities were taken from ENDF/B-VII.1, which originated 9228 1451 206 from P. Young least-square fit of available nubar data. 9228 1451 207 The following modifications were made: 9228 1451 208
  • Below 0.7 eV an additional uncertainty component was added 9228 1451 209 so as to make the uncertainty match the raw value 9228 1451 210 recommended by Standards-2017. 9228 1451 211
  • Between 0.7 eV and 73 eV the uncertainties were essentially 9228 1451 212 taken from the experimental data, assuming they were 9228 1451 213 statistical. In the range 5.6 eV to 75 eV an additional 9228 1451 214 fully correlated component of 0.5 percent was assigned 9228 1451 215 to account for the systematic uncertainty. 9228 1451 216
  • Above 73 eV up to 20 keV no data experimental exist, hence 9228 1451 217 a 2 percent uncertainty was assigned over energy subregions 9228 1451 218 73, 100, 200, 400, 1000, 2000, 4000, 10000 and 20000 eV. 9228 1451 219
  • An additional 0.4 percent fully correlated uncertainty was 9228 1451 220 added over the entire energy range from 1.E-5 eV to 30 MeV 9228 1451 221 to account for the unrecognized systematic uncertaities of 9228 1451 222 nubar measurements as recommended in Standards 2017. 9228 1451 223
  • Cross-material covariances are given for the nu-bar of 9228 1451 224 U-233, Pu-239 and Pu-241 as evaluated with Standards-2017. 9228 1451 225 The covariances are assumed energy-independent, although 9228 1451 226 the values in Standards-2017 are applicable to the thermal 9228 1451 227 energy region only. 9228 1451 228 9228 1451 229
• The covariances of the Legendre elastic scattering component 9228 1451 230 up to order P2 with cross-correlations as obtained in the 9228 1451 231 prior are available in MF34. 9228 1451 232 9228 1451 233
• The PFNS covariances were evaluated with the spectra. In the 9228 1451 234 thermal range they correspond to Standards-2017. At higher 9228 1451 235 energies the covariance evaluation was done at LANL together 9228 1451 236 with the evaluation of the spectra. 9228 1451 237 9228 1451 238 9228 1451 239 General references to the U-235 evaluation 9228 1451 240 1) A.Trkov, R. Capote, M. T. Pigni, V.G. Pronyaev, M. Sin, 9228 1451 241 E.S. Soukhovitski,'Evaluation of the neutron induced reactions 9228 1451 242 on U-235 from 2.25 keV up to 30 MeV', Proc. ND2016, 9228 1451 243 EPJ Web of Conf. 146, 02029 (2017). 9228 1451 244 9228 1451 245 2) R. Capote, A. Trkov, M. Sin, M.T. Pigni, et al., 'IAEA CIELO 9228 1451 246 Evaluation of Neutron-induced Reactions on U-235 and U-238 9228 1451 247 targets up to 30~MeV', Nucl. Data Sheets}, accepted. 9228 1451 248 9228 1451 249 References to RR and URR: 9228 1451 250 3) V.G. Pronyaev, R. Capote, A. Trkov, A. Wallner, 9228 1451 251 'New fit of neutron thermal constants for U-233,235,Pu-239,241 9228 1451 252 and Cf-252: microscopic vs integral data', Proc. ND2016, 9228 1451 253 EPJ Web of Conf. 146, 02011 (2017). 9228 1451 254 9228 1451 255 4) M.T. Pigni, R. Capote, A. Trkov, V. G. Pronyaev, 9228 1451 256 'U-235 Resonance Parameters and Neutron Multiplicities in the 9228 1451 257 Neutron Energy Region below 100 eV', Proc. ND2016, 9228 1451 258 EPJ Web of Conf. 146, 02011 (2017). 9228 1451 259 9228 1451 260 References to the fast energy region: 9228 1451 261
• OMP: RIPL 2408 potential with extended coupling scheme. 9228 1451 262 5) R. Capote, E.Sh. Soukhovitskii, J.M. Quesada, S. Chiba, 9228 1451 263 Phys. Rev. C72, 064610 (2005) 9228 1451 264 9228 1451 265 6) R. Capote, S. Chiba, E.Sh. Soukhovitski, J.M. Quesada, E.Bauge 9228 1451 266 J. Nucl. Sci. Tech. 45, 333-340 (2008). 9228 1451 267 9228 1451 268 7) F.S. Dietrich, I.J. Thompson, and T. Kawano, 9228 1451 269 Phys. Rev. C85, 044611 (2012). 9228 1451 270 9228 1451 271
• Cross section calculation: 9228 1451 272 8) M. Sin, R. Capote, M. W. Herman, and A. Trkov, Phys. Rev. C 93,9228 1451 273 034605 (2016), 'Extended optical model for fission' 9228 1451 274 9228 1451 275 9) M. Sin, R. Capote, M.W. Herman, and A. Trkov, 'Modelling 9228 1451 276 Neutron-induced Reactions on {232-237}U from 10 keV up to 9228 1451 277 30 MeV', Nucl. Data Sheets 139, 138-170 (2017). 9228 1451 278 9228 1451 279 References to the PFNS evaluation 9228 1451 280 10) A.Trkov, R. Capote, V.G. Pronyaev, Nucl. Data Sheets 123 9228 1451 281 (2015) 8-15, 'Current Issues in Nuclear Data Evaluation 9228 1451 282 Methodology: U-235 Prompt Fission Neutron Spectra and 9228 1451 283 Multiplicity for Thermal Neutrons' 9228 1451 284 9228 1451 285 11) A.Trkov, R. Capote, 'Evaluation of the Prompt Fission Neutron 9228 1451 286 Spectrum of Thermal-neutron Induced Fission in U-235', 9228 1451 287 Phys. Proc. 64, 48-54 (2015). 9228 1451 288 9228 1451 289 12) M.E.Rising, P.Talou, T.Kawano amd A.K.Prinja, 9228 1451 290 Nucl. Sci. Eng. 175, 81-93 (2013), 'Evaluation and Uncertainty 9228 1451 291 Quantification of Prompt Fission Neutron Spectra of Uranium 9228 1451 292 and Plutonium Isotopes'. 9228 1451 293 9228 1451 294 13) D. Neudecker, P. Talou, A.C. Kahler, M.C. White and T. Kawano,9228 1451 295 including a detailed analysis of experimental data and 9228 1451 296 EPJ Web of Conf. 146, 04012 (2017). 9228 1451 297 9228 1451 298 *** 9228 1451 299 FAST ENERGY REGION 9228 1451 300 9228 1451 301 M.Sin, R.Capote, A.Trkov, b46~b36 (beta4), new sources, HR grid 9228 1451 302 ___ 9228 1451 303 9228 1451 304 Nuclear reaction model code EMPIRE-3.2.3 Malta 9228 1451 305 (SVN rev. 4929; source rev. 4929) 9228 1451 306 by M.W.Herman et al [EMP,EMP-man]. 9228 1451 307 9228 1451 308 EMPIRE dimensions (dimension.h): 9228 1451 309 NDNUC= 50 NDEXCLUS= 10 NDEX= 601 NDLW= 50 9228 1451 310 LEVCC= 30 NDLV= 40 NDBR= 40 NDMSCS= 4 9228 1451 311 9228 1451 312 CROSS-SECTION EVALUATION PROCEDURE 9228 1451 313 9228 1451 314 Adopted procedure is based on careful theoretical 9228 1451 315 analysis utilizing available experimental data and 9228 1451 316 nuclear reaction model calculations. 9228 1451 317 9228 1451 318 This code integrates into a single system a number 9228 1451 319 of important modules and features: 9228 1451 320 9228 1451 321
  • Spherical and deformed Optical Model including 9228 1451 322 coupled-channels code ECIS06 by J. Raynal 9228 1451 323
  • Soft-rotator deformed Optical Model including 9228 1451 324 coupled-channels code OPTMAN by E.Soukhovitskii 9228 1451 325 and coworkers 9228 1451 326
  • Hauser-Feshbach statistical model including 9228 1451 327 width fluctuation correction, and the 9228 1451 328 optical model for fission with partial damping 9228 1451 329
  • Quantum-mechanical MSD TUL model (codes ORION & 9228 1451 330 TRISTAN by H.Lenske), and MSC NVWY model 9228 1451 331
  • Exciton model with Iwamoto-Harada cluster 9228 1451 332 emission and Kalbach systematic angular distr. 9228 1451 333 (code PCROSS by R.Capote et al) 9228 1451 334
  • Hybrid Monte Carlo preequilibrium model(M.Blann) 9228 1451 335 (code DDHMS by M.Chadwick, mod. by B.V.Carlson) 9228 1451 336
  • Complete gamma-ray cascade after emission of 9228 1451 337 each particle, including realistic treatment of 9228 1451 338 discrete transitions 9228 1451 339
  • Access to OM segment of the RIPL library [RIPL] 9228 1451 340
  • Built-in input parameter files, such as masses, 9228 1451 341 level density, discrete levels, fission barriers 9228 1451 342 and gamma strength functions based on the RIPL 9228 1451 343 library [RIPL] 9228 1451 344
  • Automatic retrieval of experimental data from the 9228 1451 345 EXFOR library 9228 1451 346
  • ENDF-6 formatting (code EMPEND by A.Trkov) 9228 1451 347 coupled to graphical presentation capabilities 9228 1451 348 (code ZVView by V. Zerkin) through the chain of 9228 1451 349 PrePro codes by D. Cullen 9228 1451 350
  • ENDF checking codes (CHECKR, FIZCON, PSYCHE) 9228 1451 351
  • Highly automated fitting of optical model 9228 1451 352 parameters (by B.V. Carlson) 9228 1451 353
  • KALMAN code for fitting experimental data and 9228 1451 354
  • generating covariances 9228 1451 355
  • Support for NJOY 9228 1451 356 9228 1451 357 9228 1451 358 PARAMETERIZATIONS 9228 1451 359 9228 1451 360 Following models and parameters were used in the 9228 1451 361 current evaluation: 9228 1451 362 9228 1451 363 Discrete levels were taken from the RIPL-3+ 9228 1451 364 library based on the 2015 version of ENSDF. 9228 1451 365 Gamma production cross sections not printed 9228 1451 366 ENDF formatting enabled 9228 1451 367 Exclusive emission from CN and target enabled 9228 1451 368 Exclusive spectra available for residues distant 9228 1451 369 from CN up to 3 neut. or 3 prot. 9228 1451 370 Coupled Channels Method used for Tl calcul. in outgoing ch. 9228 1451 371 Exciton model calculations with code PCROSS 9228 1451 372 Cluster emission in terms of the Iwamoto-Harada model 9228 1451 373 Kalbach systematics angular distributions (see RIPL-1) 9228 1451 374 Mean free path parameter in PCROSS set to 1.9 (Recomm:~1.5) 9228 1451 375 PE emission width of ejectile 1 from 236U multiplied by 1.400 9228 1451 376 PE emission width of ejectile 0 from 236U multiplied by 1.150 9228 1451 377 HRTW width fluctuation correction calculated up to 4.84 MeV 9228 1451 378 Energy continuum for nucleus 236U starts at 1.17 9228 1451 379 Energy continuum for nucleus 234U starts at 1.20 9228 1451 380 Energy continuum for nucleus 233U starts at 0.55 9228 1451 381 Energy continuum for nucleus 235Pa starts at 0.25 9228 1451 382 Energy continuum for nucleus 233Pa starts at 0.30 9228 1451 383 Energy continuum for nucleus 231Pa starts at 0.30 9228 1451 384 Energy continuum for nucleus 231Th starts at 0.33 9228 1451 385 Energy continuum for nucleus 230Th starts at 0.40 9228 1451 386 Energy continuum for nucleus 229Th starts at 0.35 9228 1451 387 EMPIRE-specific level densities (J>>K aprox.) 9228 1451 388 LDSHIFT in 235U set to 1.000 9228 1451 389 For all nuclei: fission ignored 9228 1451 390 For 236 -U fission induced by light particles 9228 1451 391 For 235 -U fission induced by light particles 9228 1451 392 For 234 -U fission induced by light particles 9228 1451 393 For 233 -U fission induced by light particles 9228 1451 394 For 232 -U fission induced by light particles 9228 1451 395 EGSM (J>>K approx.) used for fission LD for all nuclei 9228 1451 396 CN anisotropy calculated using Blatt-Biedenharn coefficients 9228 1451 397 CN decay and Direct cross sections added incoherently 9228 1451 398 Traditional Hauser-Feschbach Tlj{CC}(E)=Tlj{GS}(E-Egs) 9228 1451 399 E1 strength function set to RIPL MLO1 9228 1451 400 GDR parameters from RIPL/Exp.data+Plujko systematics 9228 1451 401 Gamma-transition multipolarity set to 2 9228 1451 402 9228 1451 403 ============================================================= 9228 1451 404 Reaction 1n +235U at incident energy 6.500D-05 MeV (LAB) 9228 1451 405 ============================================================= 9228 1451 406 9228 1451 407 9228 1451 408 RIPL GDR parameters used 9228 1451 409 9228 1451 410 Use of direct and preequilibrium models 9228 1451 411 --------------------------------------- 9228 1451 412 9228 1451 413 Exit channel ECIS MSD MSC HMS PCROSS 9228 1451 414 9228 1451 415 neut. disc. 1 0 0 0 0 9228 1451 416 neut. cont. 0 0 0 0 1 9228 1451 417 prot. disc. 0 0 0 0 1 9228 1451 418 prot. cont. 0 0 0 0 1 9228 1451 419 gammas 0 0 0 0 1 9228 1451 420 alpha cont. 0 0 0 0 1 9228 1451 421 deut. cont. 0 0 0 0 1 9228 1451 422 trit. cont. 0 0 0 0 1 9228 1451 423 He-3 cont. 0 0 0 0 1 9228 1451 424 LI cont. 0 0 0 0 0 9228 1451 425 alpha disc. 0 0 0 0 1 9228 1451 426 deut. disc. 0 0 0 0 1 9228 1451 427 trit. disc. 0 0 0 0 1 9228 1451 428 He-3 disc. 0 0 0 0 1 9228 1451 429 9228 1451 430 Collective levels used in direct calculations 9228 1451 431 N < 30 for coupled levels in CC calculation 9228 1451 432 92 235 nucleus is treated as deformed 9228 1451 433 9228 1451 434 Ncoll Lmax IDef Kgs (Def(1,j),j=2,IDef,2) 9228 1451 435 43 8 6 3.5 0.2110E+00 0.1070E+00 -.2100E-01 9228 1451 436 9228 1451 437 N E[MeV] J pi Nph L K Dyn.Def. 9228 1451 438 1 0.00000 3.5 -1. 0 0 0 0.21100 9228 1451 439 4 0.04621 4.5 -1. 0 0 0 0.21500 9228 1451 440 7 0.10304 5.5 -1. 0 0 0 0.21500 9228 1451 441 10 0.17071 6.5 -1. 0 0 0 0.21500 9228 1451 442 14 0.24913 7.5 -1. 0 0 0 0.21500 9228 1451 443 19 0.33852 8.5 -1. 0 0 0 0.21500 9228 1451 444 26 0.43860 9.5 -1. 0 0 0 0.21500 9228 1451 445 51 0.60100 1.0 -1. 0 0 0 0.00550 9228 1451 446 51 0.68010 1.0 -1. 0 0 0 0.00550 cont 9228 1451 447 51 0.73190 3.0 -1. 0 0 0 0.00550 cont 9228 1451 448 51 0.83190 2.0 1. 0 0 0 0.00800 cont 9228 1451 449 51 0.92720 0.0 1. 0 0 0 0.00800 cont 9228 1451 450 51 0.93050 1.0 -1. 0 0 0 0.00800 cont 9228 1451 451 53 0.96610 2.0 1. 0 0 0 0.00800 cont 9228 1451 452 55 0.99720 0.0 1. 0 0 0 0.01000 cont 9228 1451 453 56 0.99760 3.0 -1. 0 0 0 0.00880 cont 9228 1451 454 58 1.03730 2.0 1. 0 0 0 0.01000 cont 9228 1451 455 59 1.05640 4.0 1. 0 0 0 0.00600 cont 9228 1451 456 51 1.05770 3.0 1. 0 0 0 0.00600 cont 9228 1451 457 51 1.05970 3.0 1. 0 0 0 0.00600 cont 9228 1451 458 52 1.06030 2.0 1. 0 0 0 0.02000 cont 9228 1451 459 54 1.10570 3.0 1. 0 0 0 0.02000 cont 9228 1451 460 55 1.12880 2.0 -1. 0 0 0 0.00800 cont 9228 1451 461 56 1.13070 4.0 1. 0 0 0 0.00500 cont 9228 1451 462 57 1.13570 1.0 1. 0 0 0 0.00500 cont 9228 1451 463 51 1.16300 4.0 1. 0 0 0 0.00500 cont 9228 1451 464 51 1.16800 4.0 1. 0 0 0 0.00500 cont 9228 1451 465 52 1.17000 3.0 -1. 0 0 0 0.03809 cont 9228 1451 466 53 1.25000 2.0 1. 0 0 0 0.03017 cont 9228 1451 467 54 1.44000 3.0 -1. 0 0 0 0.05600 cont 9228 1451 468 55 1.59000 3.0 -1. 0 0 0 0.03811 cont 9228 1451 469 56 1.75000 3.0 -1. 0 0 0 0.03946 cont 9228 1451 470 57 1.85000 3.0 -1. 0 0 0 0.03526 cont 9228 1451 471 58 1.95000 3.0 -1. 0 0 0 0.04075 cont 9228 1451 472 59 2.15000 3.0 -1. 0 0 0 0.04740 cont 9228 1451 473 51 2.30000 3.0 -1. 0 0 0 0.05300 cont 9228 1451 474 51 2.39000 4.0 1. 0 0 0 0.00881 cont 9228 1451 475 52 2.49000 2.0 1. 0 0 0 0.02512 cont 9228 1451 476 53 2.94000 2.0 1. 0 0 0 0.02715 cont 9228 1451 477 54 3.18900 2.0 1. 0 0 0 0.02529 cont 9228 1451 478 55 3.38800 2.0 1. 0 0 0 0.02507 cont 9228 1451 479 56 3.53800 2.0 1. 0 0 0 0.01539 cont 9228 1451 480 57 3.63700 2.0 1. 0 0 0 0.01612 cont 9228 1451 481 9228 1451 482 9228 1451 483 Total number of nuclei considered : 42 9228 1451 484 Number of exclusive nuclei : 6 9228 1451 485 Nuclei marked with < below produces inclusive spectra only 9228 1451 486 9228 1451 487 Inelastic o. m. parameters: RIPL catalog number 2408 9228 1451 488 Neutron o. m. parameters: RIPL catalog number 2408 9228 1451 489 Proton o. m. parameters: RIPL catalog number 5408 9228 1451 490 Alpha o. m. parameters: RIPL catalog number 9600 9228 1451 491 Deuteron o. m. parameters: RIPL catalog number 6200 9228 1451 492 Triton o. m. parameters: RIPL catalog number 7100 9228 1451 493 He-3 o. m. parameters: RIPL catalog number 8100 9228 1451 494 9228 1451 495 E j e c t i l e m a s s e s [amu] 9228 1451 496 9228 1451 497 1n 1p 4He 2d 3t 3h 9228 1451 498 9228 1451 499 1.008665 1.007825 4.002602 2.014101 3.016049 3.0160289228 1451 500 9228 1451 501 S e p a r a t i o n e n e r g i e s [MeV] 9228 1451 502 9228 1451 503 Nucleus 1n 1p 4He 2d 3t 3h 9228 1451 504 9228 1451 505 92-U -236 6.544 7.172 -4.572 11.031 9.993 11.219 9228 1451 506 92-U -235 5.298 6.711 -4.679 9.706 9.975 9.461 9228 1451 507 92-U -234 6.843 6.632 -4.859 10.934 10.230 10.600 9228 1451 508 92-U -233 5.760 6.315 -4.908 9.644 10.204 8.875 9228 1451 509 92-U -232 < 7.272 6.108 -5.414 10.701 10.238 9.909 9228 1451 510 91-Pa-235 6.083 5.574 -3.989 9.541 8.070 11.751 9228 1451 511 91-Pa-234 < 5.219 5.682 -4.354 8.244 8.424 10.505 9228 1451 512 91-Pa-233 < 6.526 5.249 -4.384 9.462 8.323 11.004 9228 1451 513 91-Pa-232 < 5.553 5.160 -4.624 8.054 8.591 9.667 9228 1451 514 91-Pa-231 < 6.817 4.725 -5.150 9.295 8.292 10.400 9228 1451 515 90-Th-232 < 6.437 7.755 99.000 11.249 10.181 11.922 9228 1451 516 90-Th-231 < 5.118 7.036 -4.213 10.001 10.030 10.057 9228 1451 517 90-Th-230 < 6.794 7.107 -4.770 11.169 9.939 11.246 9228 1451 518 90-Th-229 < 5.254 6.599 -5.168 9.402 9.673 9.013 9228 1451 519 90-Th-228 < 7.109 6.372 -5.520 10.676 9.817 10.155 9228 1451 520 89-Ac-231 < 5.718 5.923 -3.833 9.661 7.976 11.415 9228 1451 521 89-Ac-230 < 5.189 6.167 -3.799 8.515 8.565 11.026 9228 1451 522 89-Ac-229 < 6.286 5.550 -4.397 9.633 7.937 11.589 9228 1451 523 89-Ac-228 < 5.027 5.571 -4.821 7.908 8.047 9.894 9228 1451 524 89-Ac-227 < 6.528 5.105 -5.042 9.277 7.922 10.729 9228 1451 525 9228 1451 526 < indicates inclusive spectra only 9228 1451 527 9228 1451 528 RESULTS: 9228 1451 529 MF=1 9228 1451 530 MT=456 nubar adopted from evaluated libraries 9228 1451 531 9228 1451 532 MF=3 Cross sections 9228 1451 533 EMPIRE calculations were adopted for: 9228 1451 534 9228 1451 535 MT=1 Total 9228 1451 536 MT=2 Elastic scattering 9228 1451 537 MT=4, 51-91 Inelastic scattering 9228 1451 538 MT=102 Capture 9228 1451 539 MT=16 (n,2n) 9228 1451 540 MT=17 (n,3n) 9228 1451 541 MT=18 (n,f) 9228 1451 542 MT=28 (n,np+pn) 9228 1451 543 MT=37 (n,4n) 9228 1451 544 MT=103, 600-649 (n,p) 9228 1451 545 MT=104, (n,d) 9228 1451 546 MT=105, (n,t) 9228 1451 547 MT=106, (n,He-3) 9228 1451 548 MT=107, 800-849 (n,a) 9228 1451 549 9228 1451 550 MF=4 Angular distributions 9228 1451 551 EMPIRE calculations were adopted 9228 1451 552 9228 1451 553 MF=5 Energy distributions 9228 1451 554 MT=18 PFNS: EMPIRE calculations using 9228 1451 555 Kornilov et al parameterization [KO] 9228 1451 556 9228 1451 557 MF=6 Energy-angle distributions of reaction 9228 1451 558 products; EMPIRE calculations were adopted 9228 1451 559 9228 1451 560 MF=12 Transition probability arrays for photon 9228 1451 561 production; taken from the RIPL library 9228 1451 562 9228 1451 563 MF=14 Photon angular distributions 9228 1451 564 isotropic distributions were assumed 9228 1451 565 9228 1451 566 9228 1451 567 REFERENCES 9228 1451 568 9228 1451 569 [EMP] 9228 1451 570 M.Herman, R.Capote, B.Carlson, P.Oblozinsky, 9228 1451 571 M.Sin, A.Trkov, H.Wienke and V.Zerkin 9228 1451 572 9228 1451 573 "EMPIRE: Nuclear Reaction Model Code System 9228 1451 574 for data evaluation" 9228 1451 575 Nuclear Data Sheets 108 (2007) 2655-2715 9228 1451 576 9228 1451 577 [EMP-man] 9228 1451 578 M.Herman, R.Capote, A.Trkov, M.Sin, B.Carlson, 9228 1451 579 P.Oblozinsky, C.Mattoon, H.Wienke, S. Hoblit, 9228 1451 580 Young-Sik Cho, G.P.A. Nobre, V.A. Plujko and 9228 1451 581 V.Zerkin 9228 1451 582 9228 1451 583 "EMPIRE-3.2 Malta: Modular system for nuclear 9228 1451 584 reaction calculations and data evaluation", 9228 1451 585 EMPIRE users' manual, report INDC(NDS)-0603 9228 1451 586 (IAEA, Vienna, 2013) and BNL-101378-2013 9228 1451 587 (BNL, Upton, NY, USA, 2013) 9228 1451 588 9228 1451 589 [RIPL] 9228 1451 590 R.Capote, M.Herman, P.Oblozinsky, P.G.Young, 9228 1451 591 S.Goriely, T.Belgya, A.V.Ignatyuk, A.J.Koning, 9228 1451 592 S.Hilaire, V.A.Plujko, M.Avrigeanu, 9228 1451 593 Zhigang Ge, Yinlu Han, S.Kailas, J.Kopecky, 9228 1451 594 V.M.Maslov, G.Reffo, M.Sin, 9228 1451 595 E.Sh.Soukhovitskii and P. Talou 9228 1451 596 9228 1451 597 "RIPL - Reference Input Parameter Library for 9228 1451 598 Calculation of Nuclear Reactions and 9228 1451 599 Nuclear Data Evaluations", 9228 1451 600 9228 1451 601 Nuclear Data Sheets 110 (2009) 3107-3214 9228 1451 602 9228 1451 603 Data available online at 9228 1451 604 http://www-nds.iaea.org/RIPL-3/ 9228 1451 605 9228 1451 606 [MN] D.G.Madland and J.R.Nix, 9228 1451 607 Nuc. Sci. Eng. 81, (1982) 213 9228 1451 608 9228 1451 609 [KK] N.V.Kornilov, A.B.Kagalenko, F.-J.Hambsch 9228 1451 610 Phys. At. Nuclei 62 (1999) pp 173-185 9228 1451 611 9228 1451 612 **** 9228 1451 613 9228 1451 614 9228 1451 615 9228 1451 616 9228 1451 617 * ENDRESS - Merging source and resonance ENDF files 9228 1451 618 Source ENDF u235.e0 9228 1451 619 Resonance ENDF Marco\u235-res-28t+-stan.endf 9228 1451 620 RESONANCE PARAMETER 9228 1451 621 9228 1451 622 Major feature : 9228 1451 623 9228 1451 624 Description of the o28t resonance file for 235U 9228 1451 625 ----------------------------------------------- 9228 1451 626 ORNL-v28t (M. T. Pigni 03-Feb-2017) 9228 1451 627 9228 1451 628 Fiss(1) Capt(1) Fiss(2) Capt(2) 9228 1451 629 9228 1451 630 0.00001 - 0.0206 eV : 1313.68 239.11 27.062 4.9258 9228 1451 631 0.0206 - 0.0623 eV : 463.00 76.85 19.312 3.2050 9228 1451 632 0.0623 - 0.6 eV : 155.56 29.90 83.647 16.082 9228 1451 633 0.6 - 7.8 eV : 23.54 21.47 169.44 154.59 9228 1451 634 7.8 - 11 eV : 77.08 30.38 246.67 97.234 9228 1451 635 11 - 17.5 eV : 41.54 40.84 9228 1451 636 17.5 - 22 eV : 73.01 51.52 9228 1451 637 22 - 30 eV : 35.82 18.05 9228 1451 638 30 - 45 eV : 48.70 31.89 9228 1451 639 45 - 100 eV : 31.57 15.41 9228 1451 640 100 - 200 eV : 21.14 12.19 9228 1451 641 200 - 300 eV : 20.75 8.94 9228 1451 642 300 - 400 eV : 13.18 6.12 9228 1451 643 400 - 500 eV : 13.80 4.73 9228 1451 644 500 - 600 eV : 15.22 4.71 9228 1451 645 600 - 700 eV : 11.53 4.41 9228 1451 646 700 - 800 eV : 11.12 4.74 9228 1451 647 800 - 900 eV : 8.24 3.98 9228 1451 648 900 - 1000 eV : 7.52 4.77 9228 1451 649 1000 - 1500 eV : 7.89 3.14 9228 1451 650 1500 - 2000 eV : 6.42 2.44 9228 1451 651 2000 - 2200 eV : 5.13 2.52 9228 1451 652 9228 1451 653 (1) b 9228 1451 654 (2) b eV 9228 1451 655 9228 1451 656 thermal fission xcs ( T = 0 K ) : 5.867870+2 b 9228 1451 657 thermal capture xcs ( T = 0 K ) : 9.939090+1 b 9228 1451 658 **** 9228 1451 659 **** Program DICTIN (VERSION 2017-1) ****9228 1451 660 1 451 1083 19228 1451 661 3 18 283 19228 1451 662 0.000000+0 0.000000+0 0 0 0 09228 1 099999 0.000000+0 0.000000+0 0 0 0 09228 0 0 0 9.223500+4 2.330248+2 0 0 0 09228 3 18 1 1.934054+8 1.934054+8 0 0 1 8399228 3 18 2 839 2 9228 3 18 3 1.000000-5 0.000000+0 2.250000+3 0.000000+0 2.250000+3 2.634378+09228 3 18 4 2.250014+3 2.668097+0 2.250056+3 2.769988+0 2.250112+3 2.907176+09228 3 18 5 2.250251+3 3.252747+0 2.250307+3 3.389935+0 2.250363+3 3.525523+09228 3 18 6 2.250419+3 3.658711+0 2.250470+3 3.778100+0 2.250517+3 3.884190+09228 3 18 7 2.250563+3 3.987480+0 2.250598+3 4.063173+0 2.250633+3 4.136965+09228 3 18 8 2.250668+3 4.209058+0 2.250703+3 4.279151+0 2.250738+3 4.347343+09228 3 18 9 2.250772+3 4.413436+0 2.250807+3 4.477529+0 2.250842+3 4.539621+09228 3 18 10 2.250877+3 4.599614+0 2.250912+3 4.657406+0 2.250947+3 4.713099+09228 3 18 11 2.250982+3 4.766491+0 2.251017+3 4.817884+0 2.251052+3 4.867076+09228 3 18 12 2.251087+3 4.914169+0 2.251122+3 4.959062+0 2.251169+3 5.015851+09228 3 18 13 2.251211+3 5.063843+0 2.251248+3 5.104135+0 2.251285+3 5.142327+09228 3 18 14 2.251322+3 5.178319+0 2.251359+3 5.212311+0 2.251416+3 5.259699+09228 3 18 15 2.251471+3 5.302887+0 2.251527+3 5.342175+0 2.251583+3 5.377764+09228 3 18 16 2.251639+3 5.409952+0 2.251695+3 5.439040+0 2.251751+3 5.465128+09228 3 18 17 2.251807+3 5.488516+0 2.251863+3 5.509404+0 2.251937+3 5.533789+09228 3 18 18 2.252012+3 5.554673+0 2.252086+3 5.572357+0 2.252161+3 5.587341+09228 3 18 19 2.252235+3 5.599925+0 2.252347+3 5.615201+0 2.252459+3 5.626678+09228 3 18 20 2.252533+3 5.632762+0 2.252683+3 5.641830+0 2.252832+3 5.647799+09228 3 18 21 2.252981+3 5.651767+0 2.253279+3 5.655903+0 2.253875+3 5.658277+09228 3 18 22 2.299301+3 5.692828+0 2.301737+3 5.698211+0 2.498795+3 6.271256+09228 3 18 23 2.499699+3 6.271964+0 2.500662+3 6.270059+0 2.502649+3 6.260037+09228 3 18 24 2.645374+3 5.437922+0 2.649456+3 5.415256+0 2.650860+3 5.411057+09228 3 18 25 2.688970+3 5.379663+0 2.898884+3 5.209777+0 2.902250+3 5.204962+09228 3 18 26 3.000000+3 5.032400+0 3.145885+3 4.766898+0 3.149096+3 4.761966+09228 3 18 27 3.150111+3 4.762639+0 3.150945+3 4.764671+0 3.152614+3 4.771335+09228 3 18 28 3.248475+3 5.209915+0 3.250000+3 5.214225+0 3.250646+3 5.214617+09228 3 18 29 3.254090+3 5.210110+0 3.500000+3 4.775450+0 3.697810+3 4.408978+09228 3 18 30 3.699178+3 4.408236+0 3.700158+3 4.410548+0 3.701138+3 4.415761+09228 3 18 31 3.703097+3 4.431086+0 3.797841+3 5.239572+0 3.798920+3 5.246696+09228 3 18 32 3.799640+3 5.249078+0 3.800000+3 5.249320+0 3.800755+3 5.247607+09228 3 18 33 3.801510+3 5.243293+0 3.804026+3 5.221982+0 3.897656+3 4.373041+09228 3 18 34 3.898729+3 4.364664+0 3.899802+3 4.358987+0 3.900517+3 4.357470+09228 3 18 35 3.901550+3 4.358488+0 3.903616+3 4.365325+0 3.998016+3 4.721072+09228 3 18 36 3.999479+3 4.724640+0 4.000000+3 4.724900+0 4.001060+3 4.723625+09228 3 18 37 4.003179+3 4.716376+0 4.147780+3 4.152061+0 4.148987+3 4.149291+09228 3 18 38 4.149252+3 4.149197+0 4.150067+3 4.150416+0 4.150617+3 4.152730+09228 3 18 39 4.151716+3 4.160457+0 4.153914+3 4.180609+0 4.247936+3 5.094376+09228 3 18 40 4.249010+3 5.102601+0 4.249857+3 5.106222+0 4.250281+3 5.106732+09228 3 18 41 4.251125+3 5.105452+0 4.252532+3 5.098786+0 4.295465+3 4.823421+09228 3 18 42 4.344982+3 4.505214+0 4.348854+3 4.481107+0 4.350006+3 4.476235+09228 3 18 43 4.351158+3 4.473663+0 4.353461+3 4.472818+0 4.499099+3 4.466828+09228 3 18 44 4.500886+3 4.464807+0 4.512802+3 4.440321+0 4.797816+3 3.840327+09228 3 18 45 4.799695+3 3.837435+0 4.800954+3 3.836473+0 4.862252+3 3.847363+09228 3 18 46 4.996802+3 3.872057+0 5.000000+3 3.871900+0 5.042493+3 3.826890+09228 3 18 47 5.148656+3 3.712721+0 5.150000+3 3.712310+0 5.152728+3 3.713885+09228 3 18 48 5.349793+3 3.921007+0 5.429582+3 3.934417+0 5.498906+3 3.944990+09228 3 18 49 5.557364+3 3.989147+0 5.597741+3 4.019878+0 5.599200+3 4.018661+09228 3 18 50 5.600312+3 4.014899+0 5.601424+3 4.008337+0 5.602908+3 3.996722+09228 3 18 51 5.647392+3 3.611565+0 5.648497+3 3.603203+0 5.649462+3 3.597823+09228 3 18 52 5.650427+3 3.594843+0 5.651123+3 3.594430+0 5.652246+3 3.595870+09228 3 18 53 5.656736+3 3.608429+0 5.797984+3 4.031129+0 5.799496+3 4.033817+09228 3 18 54 5.800000+3 4.034180+0 5.801536+3 4.033072+0 5.806145+3 4.022746+09228 3 18 55 5.999585+3 3.555148+0 6.138205+3 3.236710+0 6.148612+3 3.213269+09228 3 18 56 6.150264+3 3.211517+0 6.150815+3 3.211500+0 6.154073+3 3.213588+09228 3 18 57 6.398574+3 3.456826+0 6.400000+3 3.457340+0 6.403390+3 3.456148+09228 3 18 58 6.595748+3 3.284463+0 6.600000+3 3.280490+0 6.744682+3 3.231679+09228 3 18 59 6.748321+3 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1.738300+09228 3 18 274 8.000000+6 1.785300+0 8.500000+6 1.792700+0 9.000000+6 1.779300+09228 3 18 275 1.000000+7 1.758900+0 1.100000+7 1.735900+0 1.150000+7 1.703700+09228 3 18 276 1.200000+7 1.720600+0 1.300000+7 1.886400+0 1.400000+7 2.079000+09228 3 18 277 1.450000+7 2.089900+0 1.500000+7 2.126500+0 1.600001+7 2.153900+09228 3 18 278 1.700000+7 2.098300+0 1.800000+7 2.059000+0 1.900001+7 2.032500+09228 3 18 279 2.000000+7 2.021600+0 2.100000+7 2.088100+0 2.200000+7 2.115500+09228 3 18 280 2.300000+7 2.144500+0 2.400000+7 2.089100+0 2.500000+7 2.126800+09228 3 18 281 2.600001+7 2.128200+0 2.700001+7 2.108500+0 2.800000+7 2.134600+09228 3 18 282 2.900001+7 2.105300+0 3.000000+7 2.155000+0 9228 3 18 283 0.000000+0 0.000000+0 0 0 0 09228 3 099999 0.000000+0 0.000000+0 0 0 0 09228 0 0 0 0.000000+0 0.000000+0 0 0 0 0 0 0 0 0 0.000000+0 0.000000+0 0 0 0 0 -1 0 0 0

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So the question now would be, without resorting to outputting a gendf/pendf file (which I've seen works to remedy this issue of now showing those regions), how can I or what functions need to be added to capture that part of the endf file?

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After some digging, it might seem that this is a doomed effort short of recreating NJOY or preprocessing some pendf files if I want to pull them up faster at a later time.