SEVENTH FRAMEWORK PROGRAMME OF THE
EUROPEAN ATOMIC ENERGY COMMUNITY
Nuclear Fission and Radiation Protection

Project acronym: CHANDA
Project full title: Solving Challenges in Nuclear Data for the Safety of
European Nuclear Facilities
Grant Agreement no.:FP7 – 605203
Workpackage N°:9
Identification N°:
Type of document:Non-contractual report
Title:
UPM contribution to Task 9.4
“Processing of nuclear data”
Cross sections libraries processing methodology. Application to the JEFF-3.2 library.
Dissemination Level:PP
Reference:
Status:VERSION 0
Name / Partner / Date / Signature
written by: / F. Plaza / UPM
C.J. Díez / UPM
O. Cabellos / UPM
Task leader / J.Ch. Sublet / CCFE / Dec, 2016
WP leader: / A.J. Koning / NRG
IP Co-ordinator:

1

CONTENT

INTRODUCTION. REPORT CONTENTS

2. PROCESSING

2.1. NJOY CODE

2.2 CHANGES IN PROCESSING INPUTS PARAMETERS

2.2.1 RECONSTRUCTION TOLERANCE

2.2.2. PURR

2.2.3. THNMAX PARAMETER IN BROADR

3. VALIDATION OF JEFF-3.2

3.1. METHOD

3.2. RESULTS

4. CONCLUSIONS

5. REFERENCES

ANNEX 1. NJOY STEP-BY-STEP

Annex 2. Information on the JEFF-3.2 ACE library at 293.15K.

Annex 3. RESULTS TABLES

INTRODUCTION. REPORT CONTENTS

This report presents how to process the JEFF-3.2 library [1] with NJOY2012 [2], and to validate it with the benchmarks proposed by ICSBEP [3] using the MCNP code [4]. The goal of this work is to show how changes in input parameters for processing affect the benchmark results.

All materials included in JEFF-3.2 are processed to obtain cross section files (ACE format files) ready to be used with MCNP. Then, 123 cases of the benchmarks proposed in ICSBEP are calculated with MCNP and these processed files.

Different parameters are required as input by NJOY2012 for processing, e.g. the precision of the cross section reconstruction. The default values for a set of these parameters are modified and the effects of these changes on the calculated benchmarks are addressed.

The report is divided in two parts: in Section 2, the processing code and the changes in processing parameters are explained; and in Section 3, the validation results are shown. Finally, Annexes 1 to 3 are included to give more information about the method used.

2. PROCESSING

2.1. NJOY CODE

The processing code system to be used is NJOY2012. The modules used in this processing are: MODER, RECONR, BROADR, HEATR, GASPR, THERMR, PURR, ACER, GROUPR and MATXSR. This sequence is shown in the Figure 1:

In Figure 2 an example of an input of NJOY used to process material Fe56 (MAT 2631) at 293.15K is given.

moder / Extract/convert neutron evaluated data
1 21
'26-Fe-56 from JEFF32'/
20 2631
0/
reconr / Reconstruct XS for neutrons
21 22
'JEFF32 PENDF for 26-Fe-56'/
2631 2/
0.0010 0.0 0.0030 / err tempr errmax
'JEFF32: 26-Fe-56'/
'Processed by NJOY2012, UPM_July_2014'/
0/
broadr / Doppler broaden XS
21 22 23
2631 1 0 0 0./
0.0010 0.2E+07 0.0030 / errthn thnmax errmax
293.15
0/
heatr / Add heating kerma and damage energy
21 23 24/
2631 7 0 0 0 2/
302 303 304 318 402 443 444/
gaspr / Add gas production
21 24 25
thermr / Add thermal scattering data
0 25 61
0 2631 12 1 1 0 0 1 221 1/
293.15
0.0010 4.0
purr / Process Unresolved Resonance Range if any
21 61 26
2631 1 5 20 64/ matd ntemp nsigz nbin nladr
293.15
1.E+10 1.E+04 1.E+03 1.E+02 1.E+01
0/
acer / Prepare ACE files
21 26 0 27 28
1 0 1 .03 /
'26-Fe-56 from JEFF32, NJOY99.393, NEA_July2013'/
2631 293.15
1 1/
/
acer / Check ACE files
0 27 0 29 30
7 1 1 -1/
/
groupr / Prepare multigroup data for neutrons
21 26 0 31
2631 1 0 2 6 1 5 1/
'26-Fe-56 from JEFF32(JEFF32) NJOY 99.393 NEA JULY2011 '/
293.15
1.E+10 1.E+04 1.E+03 1.E+02 1.E+01
238 / Card 6A
1.0000E-05 1.0000E-04 5.0000E-04 7.5000E-04 1.0000E-03 1.2000E-03
1.5000E-03 2.0000E-03 2.5000E-03 3.0000E-03 4.0000E-03 5.0000E-03
7.5000E-03 1.0000E-02 2.5300E-02 3.0000E-02 4.0000E-02 5.0000E-02
6.0000E-02 7.0000E-02 8.0000E-02 9.0000E-02 1.0000E-01 1.2500E-01
1.5000E-01 1.7500E-01 2.0000E-01 2.2500E-01 2.5000E-01 2.7500E-01
3.0000E-01 3.2500E-01 3.5000E-01 3.7500E-01 4.0000E-01 4.5000E-01
5.0000E-01 5.5000E-01 6.0000E-01 6.2500E-01 6.5000E-01 7.0000E-01
7.5000E-01 8.0000E-01 8.5000E-01 9.0000E-01 9.2500E-01 9.5000E-01
9.7500E-01 1.0000E+00 1.0100E+00 1.0200E+00 1.0300E+00 1.0400E+00
1.0500E+00 1.0600E+00 1.0700E+00 1.0800E+00 1.0900E+00 1.1000E+00
1.1100E+00 1.1200E+00 1.1300E+00 1.1400E+00 1.1500E+00 1.1750E+00
1.2000E+00 1.2250E+00 1.2500E+00 1.3000E+00 1.3500E+00 1.4000E+00
1.4500E+00 1.5000E+00 1.5900E+00 1.6800E+00 1.7700E+00 1.8600E+00
1.9400E+00 2.0000E+00 2.1200E+00 2.2100E+00 2.3000E+00 2.3800E+00
2.4700E+00 2.5700E+00 2.6700E+00 2.7700E+00 2.8700E+00 2.9700E+00
3.0000E+00 3.0500E+00 3.1500E+00 3.5000E+00 3.7300E+00 4.0000E+00
4.7500E+00 5.0000E+00 5.4000E+00 6.0000E+00 6.2500E+00 6.5000E+00
6.7500E+00 7.0000E+00 7.1500E+00 8.1000E+00 9.1000E+00 1.0000E+01
1.1500E+01 1.1900E+01 1.2900E+01 1.3750E+01 1.4400E+01 1.5100E+01
1.6000E+01 1.7000E+01 1.8500E+01 1.9000E+01 2.0000E+01 2.1000E+01
2.2500E+01 2.5000E+01 2.7500E+01 3.0000E+01 3.1250E+01 3.1750E+01
3.3250E+01 3.3750E+01 3.4600E+01 3.5500E+01 3.7000E+01 3.8000E+01
3.9100E+01 3.9600E+01 4.1000E+01 4.2400E+01 4.4000E+01 4.5200E+01
4.7000E+01 4.8300E+01 4.9200E+01 5.0600E+01 5.2000E+01 5.3400E+01
5.9000E+01 6.1000E+01 6.5000E+01 6.7500E+01 7.2000E+01 7.6000E+01
8.0000E+01 8.2000E+01 9.0000E+01 1.0000E+02 1.0800E+02 1.1500E+02
1.1900E+02 1.2200E+02 1.8600E+02 1.9250E+02 2.0750E+02 2.1000E+02
2.4000E+02 2.8500E+02 3.0500E+02 5.5000E+02 6.7000E+02 6.8300E+02
9.5000E+02 1.1500E+03 1.5000E+03 1.5500E+03 1.8000E+03 2.2000E+03
2.2900E+03 2.5800E+03 3.0000E+03 3.7400E+03 3.9000E+03 6.0000E+03
8.0300E+03 9.5000E+03 1.3000E+04 1.7000E+04 2.5000E+04 3.0000E+04
4.5000E+04 5.0000E+04 5.2000E+04 6.0000E+04 7.3000E+04 7.5000E+04
8.2000E+04 8.5000E+04 1.0000E+05 1.2830E+05 1.5000E+05 2.0000E+05
2.7000E+05 3.3000E+05 4.0000E+05 4.2000E+05 4.4000E+05 4.7000E+05
4.9950E+05 5.5000E+05 5.7300E+05 6.0000E+05 6.7000E+05 6.7900E+05
7.5000E+05 8.2000E+05 8.6110E+05 8.7500E+05 9.0000E+05 9.2000E+05
1.0100E+06 1.1000E+06 1.2000E+06 1.2500E+06 1.3170E+06 1.3560E+06
1.4000E+06 1.5000E+06 1.8500E+06 2.3540E+06 2.4790E+06 3.0000E+06
4.3040E+06 4.8000E+06 6.4340E+06 8.1870E+06 1.0000E+07 1.2840E+07
1.3840E+07 1.4550E+07 1.5680E+07 1.7330E+07 2.0000E+07
3/
3 221 'Free Gas'/
3 251 'mubar'/
3 252 'xi'/
3 253 'gamma'/
3 259 '1/v'/
6/
6 221 'Free Gas'/
0/
0/
matxsr / Produce MATXS file
31 0 41/
1 'JEFF32:Fe56 '/
1 2 2 1
'JEFF32 26-Fe-56 from JEFF32'/
'Processed by NJOY-99.364 at NEA DEC2011 '/
'n' /
238
'nscat' 'ntherm' /
1 1/
1 1/
Fe56 2631/
Stop

Figure2: NJOY input code for material Fe56

The list of NJOY modules used is briefly described:

  • MODER: Converts between ENDF-6 standard coded mode and the NJOY blocked binary mode.
  • RECONR: Reconstructs pointwise cross sections from ENDF resonance parameters and interpolation schemes.
  • BROADR: Doppler-broadens and thins pointwise cross sections.
  • HEATR: Generates pointwise heat production cross section (KERMA factors) and radiation damage production cross sections.
  • GASPR: Adds gas production to PENDF (Pointwise ENDF file).
  • THERMR: Generates neutron scattering cross sections and point-to-point scattering kernels in the thermal range for free or bound atoms.
  • PURR: Prepares unresolved resonance region probability tables for continuous energy Monte Carlo codes.
  • ACER: Prepares libraries in ACE format for MCNP.
  • GROUPR: Generates self-shielded multigroup cross sections and group-to-group scattering and photon production matrices.
  • MATXSR: Converts multigroup data into the comprehensive MATXS cross section interface format.

Each module and the options chosen in them are described in the Annex 1.

2.2CHANGES IN PROCESSING INPUTS PARAMETERS

Five changes are performed to the default sequence for processing with NJOY2012. Here, they are described, in conjunction with their modifications.

2.2.1 RECONSTRUCTION TOLERANCE

The first parameter to be studied is the reconstruction tolerance. This parameter sets the total number of points calculated in the pointwise reconstruction. The lower the reconstruction tolerance, the higher the number of points and the higher the processing time.

The tolerances chosen are 10%, 1% and 0.1%. This parameter is introduced in three modules: MODER, RECONR and THERMR. The values given are 0.1 for 10%, 0.01 for 1% and 0.001 for 0.1%. In the example given in Figure 2 the reconstruction tolerance is 0.1%.

In Figure 3, an example of the effect that this parameter causes on the reconstructed pointwise cross section of the 56Feis given:

Figure3: Effect of the reconstruction tolerance on the pointwise cross section, plotted with JANIS [5].

As shown in Figure 3, the number of points in the reconstructed pointwise cross section with a reconstruction tolerance of 10% is significantly lower than in the other two options. However, the cross section is worse described.

2.2.2. PURR

Another change is made to see the effect of the PURR module on the ACE format library. PURR module prepares the unresolved-region probability tables for the Monte Carlo codes. In this region, which ranges usually from 100 keV to 1 MeV, the value of the cross section cannot be measured with precision, so it is calculated using the probability tables.

If PURR module is not run, the probability tables are not prepared. To see the impact of not using the probability tables, the PURR module is not run in the 4th processing inputs. The reconstruction tolerance used in this case is 1%.

2.2.3. THNMAX PARAMETER IN BROADR

The last change is made in BROADR module. This module Doppler-broadens and thins pointwise cross sections. The thnmax parameter sets the upper limit energy for Doppler broadening:no Doppler broadening is preformed above that energy. This energy is the lowest between the following values:

  1. The input value (thnmax).
  2. The upper limit of the resolved-resonance energy range.
  3. The lowest reaction threshold.
  4. The default value 1MeV.

An issue in regard with the value of thnmax has been detected, which stops the Doppler broadening at a lower energy than the given value for thnmax. To avoid this problem, the NJOY2012 manual says to force the Doppler broadening up to the thnmax energy value. That is achieved by setting thnmax to a negative value (with the same absolute desired energy). This issue only happens in the following materials:

Isotope / Max. energy for broadening and thinning
(eV) / Max. resonance range information
(eV)
Ag110M / 3 / 125
Ar38 / 228102 / 300000
Ba132 / 68 / 100000
C13 / 4131500 / 6950000
Ce144 / 50 / 100000
Cr52 / 1233310 / 1430000
Cs137 / 1700 / 100000
Eu156 / 1 / 100000
Fe54 / 399996 / 700000
Fe57 / 14668.2 / 200000
I127 / 3502.12 / 5200
I130 / 100 / 560
I131 / 30 / 100000
K41 / 117386 / 887400
Mo95 / 995.19 / 2141
Mo99 / 24 / 100000
Nb95 / 25 / 100000
Ne20 / 713153 / 2201000
Ni58 / 399996 / 812000
Ni62 / 444520 / 600000
Pb205 / 2340.44 / 100000
Pb206 / 807025 / 900000
Pm148 / 1.1 / 100000
Pm149 / 2.6 / 100000
S32 / 957397 / 1660000
Sc45 / 12678.1 / 100000
Sn113 / 100 / 510
Xe133 / 80.5 / 100000
Zr95 / 125 / 100000

Table 1: Materials in JEFF-3.2 with the issue in the PURR module

Therefore in the 5th processing input, the thnmax value used in the rest of cases, 2MeV is changed to -2MeV using a tolerance reconstruction of 1%. This change is performed in every isotope of the JEFF-3.2 library.As it is shown in table 1, the materials with the issue in the PURR module are not important in nuclear applications, so it is expected that this change will have negligible effects on the final results.

3. VALIDATION OF JEFF-3.2

3.1. METHOD

Benchmarks from International Handbook of Evaluated Safety Criticality Benchmark Experiments (ICSBEP handbook) are used to validate the processing of JEFF-3.2 presented in Section 2.

Once JEFF-3.2 is processed, pointwise cross section data files are generated, ready to be used with MCNP.For every material processed, two files are given:

  • Cross section data file in the ACE type 1 (ASCII) format: file.ace
  • Information required by xsdir file of MCNP code system: file.dir

Both files are prepared with the ACER module of NJOY2012. Also, probabilitytables (ptable) have been generated for those materials with unresolvedresonance data (see Annex 2).Only39K and 192Pt are not processed due to problems with the PURR processing. However, these materials are not included in any benchmark, so this problem is not important for the results.

Each benchmark will be run 5 times, one per change made in the processing inputs used. The number of benchmarks used is 123, divided in 5 groups, depending on the fissionable material used:

  • Uranium 233 (U233): 18 cases.
  • Highly enriched uranium (HEU): 40 cases.
  • Intermediate-enriched uranium (IEU): 17 cases.
  • Low-enriched uranium (LEU): 8 cases.
  • Plutonium (Pu): 36 cases.

For each fissionable material, the results can be classified by spectra, form, geometries and moderator/reflector.

The results are exposed in tables likeTable2, where the U233 benchmarks are shown. The rest of tables are provided in Annex 3.

Figures 4 - 8 present the results to facilitate the comparison between changes in the input. There, vertical axis shows the ratio between the calculated value and the experimental one, so the experimental value is always 1.The grey band represents the uncertainty of the experiments.The horizontal axis shows each benchmark name.The results are named according to the change made in each case, so, 0.1, 0.01 and 0.001 are the cases where the reconstruction tolerance is changed, 0.01_No-PURR is the case where the PURR module is not included. And 0.01_Neg-BROADR is the case where the thnmax parameter is changed.

1

3.2. RESULTS

Difference to benchmark in pcm
Benchmark
Keff / PURR ON / PURR OFF
Spectrum / Form / Shape / Moderator and/or Reflector / ICSBEP Benchmark name / 0.1 / 0.01 / 0.001 / 0.01_Neg-BROADR / 0.01_No-PURR
Fast / Metal / Sphere / Unreflected / u233-met-fast-001 / 1.0000±0.0010 / -28 / 60 / 10 / -18 / -13
HEU / u233-met-fast-002-CASE_1 / 1.0000±0.0010 / -69 / -73 / -149 / -99 / -118
u233-met-fast-002-CASE_2 / 1.0000±0.0011 / 58 / 16 / 43 / 86 / -5
Normal Uranium / u233-met-fast-003-CASE_1 / 1.0000±0.0010 / -73 / -87 / -65 / -86 / -84
u233-met-fast-003-CASE_2 / 1.0000±0.0010 / -66 / -76 / -49 / -30 / -20
u233-met-fast-006 / 1.0000±0.0014 / 95 / 61 / 37 / 30 / 39
Tungsten / u233-met-fast-004-CASE_1 / 1.0000±0.0007 / -68 / -106 / -96 / -81 / -41
u233-met-fast-004-CASE_2 / 1.0000±0.0008 / -317 / -298 / -256 / -301 / -197
Berylium / u-233-met-fast-005-CASE_1 / 1.0000±0.0030 / -459 / -364 / -406 / -392 / -414
u-233-met-fast-005-CASE_2 / 1.0000±0.0030 / -521 / -507 / -474 / -509 / -488
Intermediate / Solution / Sphere / Berylium / u233-sol-inter-001-case1 / 1.0000±0.0083 / -1509 / -1432 / -1360 / -1438 / -1428
Thermal / UO2+ZrO2 / Lattice / Water / u233-comp-therm-001-case3 / 1.0000±0.0024 / 217 / 297 / 475 / 222 / 337
Solution / Sphere / Unreflected / u233-sol-therm-001-case1 / 1.0000±0.0031 / 60 / 217 / 152 / 181 / 175
u233-sol-therm-001-case2 / 1.0000±0.0033 / 123 / 157 / 158 / 127 / 150
u233-sol-therm-001-case3 / 1.0000±0.0033 / 69 / 121 / 131 / 103 / 135
u233-sol-therm-001-case4 / 1.0000±0.0033 / 108 / 106 / 61 / 64 / 139
u233-sol-therm-001-case5 / 1.0000±0.0033 / -29 / 61 / 43 / 111 / 63
u233-sol-therm-008 / 1.0000±0.0029 / 37 / 169 / 129 / 130 / 134

Table 2: U233 Benchmarks

Figure4: U233 Benchmarks

Figure5: HEU Benchmarks

Figure6: IEU Benchmarks

Figure7: LEU Benchmarks

Figure8: Pu Benchmarks

1

After comparing the results shown in Figures 4 to 8 there are three points to discuss:

First of all, comparing the effects that the changes in the reconstruction tolerance make on the final values, it is easy to see that the error obtained using the processed library with 10% of reconstruction tolerance is significantly higher than the others when the neutron spectrum is thermal. The difference is smaller in the U233 benchmarks than in the rest of them, and in the Pu cases, this error appears only in the benchmarks where the plutonium is mixed with uranium.In conclusion it is recommended to not use the 10% reconstruction tolerance in applications where the neutron spectrum is thermal.

The effect caused by no including the PURR module on the results is of importance for 5 benchmarks where using or not PURR lead to differences higher than 150 pcm: heu-met-fast-003-case2, heu-met-fast-003-case3, heu-met-fast-003-case8, ieu-met-fast-007-case1 and mix-met-fast-008-case7. In the two last ones, the usage of probability tables should be mandatory since their effect could lead to underestimations of 400-1000 pcm of reactivity. The other set of benchmarks (HEU – High Enriched Uranium), the effect of using probability tables could induce overestimations of 150 to 200 pcm.

Finally, the effect caused by changing the thnmax parameter from 2MeV to -2MeV in the BROADR module is negligible. There are only two cases with a difference between 100 and 150 pcm (leu-sol-therm-007-CASE30 and pu-met-fast-022). That was expected because the materials where the error (Doppler-broadening stopped before reaching selected energy) happened were not important in these types of calculations.

4.CONCLUSIONS

In this report, a methodology for processing cross section libraries is presented and applied to the JEFF-3.2 cross section library. In the first part, how to use the NJOY processing code is described with an input file example. Different parameters/options have been used for processing JEFF-3.2, and their effects in the ICSBP benchmarks are assessed. First of all, three different reconstruction tolerance values are used. Next, the usage of the PURR module is switched off for addressing the impact of not using probability tables on the unresolved-region. Finally, the thnmax parameter of the BROADR module, which controls the Doppler broadening, is set in order to force the Doppler broadening up to the energy wanted. Cross section files for MCNP are generated for all this five different processing inputs. 123 benchmarks from ICSBP are run with these files, comparing the results with the experimental values. Comparisons between the effect of the different processing inputs on these criticality calculations are performed.

These comparisons show that the tolerance reconstruction value is very important in benchmarks with thermal neutron spectra. On the other hand, not including PURR (probability tables for cross sections in unresolved resonance region) or force the Doppler-broadening to a desired value are not of relevance for most of the benchmarks studied. There are very few exceptions where large deviations are found: 5 for not using probability tables (from 150 to 1000 pcm) and 2 for not forcing Doppler-broadening (up to 150 pcm). Such cases have been identified.

5. REFERENCES

[1].JEFF-3.2 Evaluated Data Library - Neutron data.OECD/NEA Data Bank (2014).

[2].Methods forprocessing ENDF/B-VII with NJOY. Nuclear Data Sheets, 111(12):2739 – 2890.MacFarlane, R. E. and Kahler, A. C, 2010.

[3].International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03/I, Nuclear Energy Agency. Paris, September 2010 Edition.

[4].MCNP – A General Monte Carlo N-Particle transport code. Technical Report LA-CP-03-0245, LANL. Version 5.X-5 Monte Carlo Team, 2003.

[5].JANIS 4: An Improved Version of the NEA Java-based Nuclear Data Information System. N. Soppera, M. Bossant, E. Dupont, Nuclear Data Sheets, Volume 120, June 2014, Pages 294-296, ISSN 0090-3752,

[6].Processing of the JEFF-3.2T3 Cross Section Library with the NJOY Code System into Various Formats for Testing Purposes. O. Cabellos. NEA, 2013.

ANNEX 1. NJOY STEP-BY-STEP

In this section, a summary of the most important processing options used by NJOY in a typical input file (Fe-56 at 293.15K) is presented below.

MODER

Convert ENDF “tapes” back and forth between formatted (that is, ASCII, EBCDIC, etc.) and blocked binary modes.

moder / Extract/convert neutron evaluated data
1 21
'26-Fe-56 from JEFF32'/
20 2631
0/
  • Input unit: 20 (tape20, JEFF-3.1 cross section library)
  • Material: 2631 (U-235)

MODER reads the JEFF-3.2T3 file (tape20) which contains the material (for Fe-56 the material number is 2631), extract it and put the ENDF file in binary mode in tape21.

RECONR

Reconstruct pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes.

reconr / Reconstruct XS for neutrons
21 22
'JEFF32 PENDF for 26-Fe-56'/
2631 2/
0.0010 0.0 0.0030 / err tempr errmax
'JEFF32: 26-Fe-56'/
'Processed by NJOY2012, UPM_July_2014'/
0/
  • Reconstruction tolerance: 0.1%
  • Resonance-integral-check tolerance: 0.3%
  • Reconstruction temperature: 0.0 K

RECONR read tape21, extract material 2631 and reconstruct cross section at 0K with a reconstruction tolerance of 0.1%. The resulting pointwise cross section is written in tape22.

BROADR

Doppler-broadens and thins pointwise cross sections.

broadr / Doppler broaden XS
21 22 23
2631 1 0 0 0./
0.0010 0.2E+07 0.0030 / errthn thnmax errmax
293.15
0/
  • Thinning tolerance : 0.1%
  • Integral criterion tolerance: 0.3%
  • Thnmax value: 2MeV (The upper limit for broadening and thinning is selected asthe lowest among: input values THNAMX, the lowest reaction threshold and the start of the resolved range.)
  • Temperature: 293.15 K

BROADR reads tape21-tape22 and Doppler-broadens at 293.15K for material 2631.The tolerance used is 0.1%. The results are written in tape23.

HEATR

Generate pointwise heat production cross sections (KERMA factors) and radiation-damage-production cross sections.

heatr / Add heating kerma and damage energy
21 23 24/
2631 7 0 0 0 2/
302 303 304 318 402 443 444/
  • MT=302: Elastic
  • MT=303:Non-elastic
  • MT=304: Inelastic
  • MT=318: Fission
  • MT=402: Capture
  • MT=443: Total kinematic kerma
  • MT=444: Total damage energy production

HEATR computes energy-balance heating and damage energy for material 2631 at 293.15K. Input files are tape21-tape23. The output file is tape24.

GASPR

Add gas production (mt203-207) to PENDF.

gaspr / Add gas production
21 24 25

GASPR add gas production reactions to PENDF tape24. Output file is tape25

THERMR

Produce cross sections and energy-to-energy matrices for free or bound scatterers in the thermal energy range.

thermr / Add thermal scattering data
0 25 61
0 2631 12 1 1 0 0 1 221 1/
293.15
0.0010 4.0
  • Number of angles bins: 12
  • Tolerance: 0.1%
  • Max. Energy: 4 eV
  • Scattering option: 221 (free gas)
  • Temperature: 293.15 K

THERMR generates neutron scattering cross sections and point-to-point scattering kernels in the thermal range. The input file is only tape25, the output file is tape61. The generation of neutron scattering cross sections is donefor material 2631 with 12 equi-probable angles at 293.15K. Free gas option is chosen and no elastic cross sections in the thermal range. A tolerance of 0.1% and the maximum energy for thermal treatment is 4 eV.

PURR

Prepare unresolved-region probability tables for the MCNP continuous-energy Monte Carlo code.

purr / Process Unresolved Resonance Range if any
21 61 26
2631 1 5 20 64/ matd ntemp nsigz nbin nladr
293.15
1.E+10 1.E+04 1.E+03 1.E+02 1.E+01
0/
  • Number of probability bins: 15
  • Number of resonance ladders: 32
  • Temperature: 293.15 K
  • Bondarenko 0 values: 1.E+10 1.E+04 1.E+03 1.E+02 1.E+01

PURR calculates probability tables for treating unresolved-resonance self-shielding for Monte Carlo codes. Inputs tape21-tape61, output in tape 26. Calculation performed for material 2631 at 293.15K with five 0 values (1.0E+10, infinite dilute cross section). A total number of 15 probability bins and 32 resonance ladders were selected.

ACER (cross section libraries)

Prepare libraries in ACE format for the Los Alamos continuous-energy Monte Carlo code MCNP.

acer / Prepare ACE files
21 26 0 27 28
1 0 1 .03 /
'26-Fe-56 from JEFF32, NJOY99.393, NEA_July2013'/
2631 293.15
1 1/
/
acer / Check ACE files
0 27 0 29 30
7 1 1 -1/
/
  • Type of ACE file: 1 (ascii)
  • ZAID suffix: .03
  • No thinning
  • Temperature: 293.15 K

ACER prepares a data library for MCNP (in ASCII). Input files: tape21-26, output files: tape27 (ACE file) and tape28 (XSDIR file). The material processed is 2631 at 293.15K. The id suffix for this material is ”.03”.