Processing methods

The NJOY modular system was used for processing the evaluated nuclear data files. The following modules were invoked in sequence for WIMS-D library generation: MODER-RECONR-BROADR-UNRESR or PURR-THERMR-GROUPR-WIMSR.

MODER module is simply a file format conversion utility and requires no additional explanation.

RECONR: Cross section reconstruction.

The input parameter of major interest is the resonance reconstruction tolerance criterion (ERR). A value of 0.1 % was chosen for all the isotopes. All the other options took the default values.

BROADR: Doppler broadening. A tolerance for thinning (ERRTHN) of 0.1 % was used for all isotopes.The temperatures were selected covering the range of typical spatial regions where the isotope is likely to be used. More details were given to temperature mesh for moderators and resonance absorbers. A temperature input list is also required by UNRESR, THERMR and GROUPR. The same mesh is taken for all modules.

PURR: Unresolved resonance data processing. PURR is a new updated code for unresolved resonance treatment of cross section data. This module was implemented in NJOY after UNRESR module. The principal input parameters are the temperature list and the Bondarenko background cross section mesh. The same values of these parameters used in GROUPR are taken for this module. This module was used for the most important resonance isotopes and where the UNRESR presented an anomalous behaviour.

UNRESR: Unresolved resonance data processing. This module was used for treating unresolved resonances of less important resonance materials. As for PURR the most important input parameters are the temperature and the Bondarenko cross section meshes. The same comments for PURR are applicable here.

THERMR: Thermal scattering law. The free gas model is used for all materials except for the main moderators:
H in H2O, H in ZrH, D in D2O, Be and C. For that materials, the thermal scattering laws are taken from the corresponding evaluated data file. A value of 0.1 % was used for the angular distribution tolerance criterion TOL. The selected value for NBIN, the number of equi-probable angles, is 12 for all the materials.The maximum energy for thermal treatment EMAX is set to 4.0 eV according to the WIMS-D library value.

GROUPR: Group averaged data preparation. The group structure parameter (IGN) is put to 1 (read in) for 172 group libraries and to 9 (epri-cpm) for 69 group libraries. The spectra parameter (IWT) is 1 for all cases, that means read in. These spectra has a 1/E form in the epithermal range. For resonant isotopes, the parameter IWT is put negative, that means the flux calculator for homogeneous mixtures is used to obtain the spectrum shape from 0.1 eV up to about the upper limit of the resolved resonance range. The narrow resonance approximation is used above this energy. The NJOY input instructions for the flux calculator require the following parameters: EHI, upper energy limit to which the flux calculator is applied [eV], SIGPOT, potential cross section of the resonant absorber [barns], NFLMAX, maximum number of points at which the neutron spectrum is generated with the flux calculator. These parameters are material dependent and can be read from the corresponding NJOY inputs. Legendre order (LORD) is 1 for all cases. The Bondarenko background cross sections were chosen according to dimensions and compositions at which the materials are likely to be used.

WIMSR: Formatting multigroup cross sections for WIMS. The relevant input options for this module are: the reference background cross section (SGREF), the potential cross section (SIGP), lambda parameter values (LAMBDA), current weighting spectrum (JP1) and burnup data (IBURN).
The WIMSD library format only allows self-shielding in the resonance energy range for the resonance isotopes which appear in the fuel. Even here the self-shielding effects are considered only for absorption and fission. If self-shielding is important for other data types, it should be incorporated by defining a reference value of the background cross section, so that appropriate cross sections can be picked when assembling the data for the WIMSD library. This selection must be exercised according to the conditions where the isotope in question is likely to be used. The potential cross section values from file 2 of evaluated data files is used for resonance materials. It is consistent with the WIMS approach. For non-resonance materials, the corresponding self-shielding scattering cross section is used because it is the correct value for the scattering cross section that is energy-dependent (SIGP=0). The LAMBDA values and burnup data are material dependent and a typical current spectrum for averaging the transport cross section in the fast and resonance groups was entered by input. Details of the selected values can be found on the Documentation page.

Special issues

Fission cross section of Am-241.

Burnup of Am-241 in the WIMS-D library with branching to Am-242g and Am-242m can not be represented accurately because the WIMS-D library allows a single capture product to be specified. The Am-242m nuclide is considered more important from the reactivity point of view, therefore its branch is treated explicitly. However, Am-242g is important due to the decay into Cm-242 which in turn decays into Pu-238. Am-242g production can only be dealt with by treating it as a fission product of Am-241 with a yield equal to the capture to fission ratio.
The fission channel in Am-241 has a threshold, therefore the capture to fission ratio is strongly spectrum dependent. The ratio may reach a value 124 for well thermalized lattices and may be as low as 42 for lattices with a high content of degraded plutonium. For a typical PWR lattice the ratio is about 92. To avoid spectrum dependence of the effective Am-242g production the shape of the fission cross section is forced to be proportional to the absorption cross section such that the capture to fission ratio of 92 is approximately conserved for an average lattice. This crude adjustment is tolerable because Am-241 is not very important from the reactivity point of view. A program, FIDLAM, to fiddle the Am-241 fission cross section was developed on the frame of WLUP.

Pseudo fission product to simulate (n,2n) reaction for U-238, U-233 and Pa-231.

To model (n, 2n) production from U-238, U-233 and Pa-231 the approach used in the WIMKAL-88 library was applied. It basically consists in introducing a pseudo fission product with an effective fission yield that represent the (n,2n) to fission reaction rate ratio of the precursor nuclide and with null absorption cross section data. It decays with a constant of unity. That means that in normal burnup calculations this fission product goes immediately to the equilibrium concentration and appears as an effective (n, 2n) source in the corresponding burnup equation.

Processing methods for dosimetry materials

The dosimetry materials of the WIMS-D library are special materials from the formatting point of view. They are non-burnable, without resonance table and scattering data. The WIMSD formatted files for these materials contain the corresponding dosimetry reaction instead of absorption and transport cross sections.

These simplifying features allow the use of the PREPRO-2000 code system for the nuclear data processing. Besides, the code WIMSIE was developed in the frame of the project to perform the WIMS-D formatting and eventually to produce the special dosimetry files as the positive and negative 1/v materials, the resonant part of the 1/v absorber, the constant absorber and the inverse lethargy interval cross section.

The sequence MERGER-LINEAR-SIGMA1-GROUPIE-DICTION-WIMSIE was used for the evaluated nuclear data processing.MERGER was called to extract the needed dosimetry reaction and the general information section (MF=1/MT=451) from the source evaluated nuclear data files. LINEAR was used because some point-wise dosimetry files from the JENDL/D-99 library were not linearly interpolable over the resonance range. The reconstruction tolerance was set up to 0.1% in all cases. It is worthy to note that the use of RECENT was skipped because the resonance contribution (MF=2) was always included on the point-wise cross section data (MF=3) of the source evaluated file. SIGMA1 Doppler broadened the cross section at 300 K and GROUPIE performed the multi-group data processing into 69 and 172 energy group structures. The averaging spectrum applied for most the materials in the frame of the project was converted to linearly interpolable data and used for generating infinite diluted multi-group cross sections. Finally, the index on the general information section was updated by DICTION and the WIMS-D formatting was performed by WIMSIE.

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Processing methods

The NJOY modular system was used for processing the evaluated nuclear data files. The following modules were invoked in sequence for WIMS-D library generation: MODER-RECONR-BROADR-UNRESR or PURR-THERMR-GROUPR-WIMSR. MODER module is simply a file format conversion utility and requires no additional explanation. RECONR: Cross section reconstruction. The input parameter of major interest is the resonance reconstruction tolerance criterion (ERR). A value of 0.1 % was chosen for all the isotopes. All the other options took the default values. BROADR: Doppler broadening. A tolerance for thinning (ERRTHN) of 0.1 % was used for all isotopes.The temperatures were selected covering the range of typical spatial regions where the isotope is likely to be used. More details were given to temperature mesh for moderators and resonance absorbers. A temperature input list is also required by UNRESR, THERMR and GROUPR. The same mesh is taken for all modules. PURR: Unresolved resonance data processing. PURR is a new updated code for unresolved resonance treatment of cross section data. This module was implemented in NJOY after UNRESR module. The principal input parameters are the temperature list and the Bondarenko background cross section mesh. The same values of these parameters used in GROUPR are taken for this module. This module was used for the most important resonance isotopes and where the UNRESR presented an anomalous behaviour. UNRESR: Unresolved resonance data processing. This module was used for treating unresolved resonances of less important resonance materials. As for PURR the most important input parameters are the temperature and the Bondarenko cross section meshes. The same comments for PURR are applicable here. THERMR: Thermal scattering law. The free gas model is used for all materials except for the main moderators: H in H2O, H in ZrH, D in D2O, Be and C. For that materials, the thermal scattering laws are taken from the corresponding evaluated data file. A value of 0.1 % was used for the angular distribution tolerance criterion TOL. The selected value for NBIN, the number of equi-probable angles, is 12 for all the materials.The maximum energy for thermal treatment EMAX is set to 4.0 eV according to the WIMS-D library value. GROUPR: Group averaged data preparation. The group structure parameter (IGN) is put to 1 (read in) for 172 group libraries and to 9 (epri-cpm) for 69 group libraries. The spectra parameter (IWT) is 1 for all cases, that means read in. These spectra has a 1/E form in the epithermal range. For resonant isotopes, the parameter IWT is put negative, that means the flux calculator for homogeneous mixtures is used to obtain the spectrum shape from 0.1 eV up to about the upper limit of the resolved resonance range. The narrow resonance approximation is used above this energy. The NJOY input instructions for the flux calculator require the following parameters: EHI, upper energy limit to which the flux calculator is applied [eV], SIGPOT, potential cross section of the resonant absorber [barns], NFLMAX, maximum number of points at which the neutron spectrum is generated with the flux calculator. These parameters are material dependent and can be read from the corresponding NJOY inputs. Legendre order (LORD) is 1 for all cases. The Bondarenko background cross sections were chosen according to dimensions and compositions at which the materials are likely to be used. WIMSR: Formatting multigroup cross sections for WIMS. The relevant input options for this module are: the reference background cross section (SGREF), the potential cross section (SIGP), lambda parameter values (LAMBDA), current weighting spectrum (JP1) and burnup data (IBURN). The WIMSD library format only allows self-shielding in the resonance energy range for the resonance isotopes which appear in the fuel. Even here the self-shielding effects are considered only for absorption and fission. If self-shielding is important for other data types, it should be incorporated by defining a reference value of the background cross section, so that appropriate cross sections can be picked when assembling the data for the WIMSD library. This selection must be exercised according to the conditions where the isotope in question is likely to be used. The potential cross section values from file 2 of evaluated data files is used for resonance materials. It is consistent with the WIMS approach. For non-resonance materials, the corresponding self-shielding scattering cross section is used because it is the correct value for the scattering cross section that is energy-dependent (SIGP=0). The LAMBDA values and burnup data are material dependent and a typical current spectrum for averaging the transport cross section in the fast and resonance groups was entered by input. Details of the selected values can be found on the Documentation page.

Special issues

Fission cross section of Am-241. Burnup of Am-241 in the WIMS-D library with branching to Am-242g and Am-242m can not be represented accurately because the WIMS-D library allows a single capture product to be specified. The Am-242m nuclide is considered more important from the reactivity point of view, therefore its branch is treated explicitly. However, Am-242g is important due to the decay into Cm-242 which in turn decays into Pu-238. Am-242g production can only be dealt with by treating it as a fission product of Am-241 with a yield equal to the capture to fission ratio. The fission channel in Am-241 has a threshold, therefore the capture to fission ratio is strongly spectrum dependent. The ratio may reach a value 124 for well thermalized lattices and may be as low as 42 for lattices with a high content of degraded plutonium. For a typical PWR lattice the ratio is about 92. To avoid spectrum dependence of the effective Am-242g production the shape of the fission cross section is forced to be proportional to the absorption cross section such that the capture to fission ratio of 92 is approximately conserved for an average lattice. This crude adjustment is tolerable because Am-241 is not very important from the reactivity point of view. A program, FIDLAM, to fiddle the Am-241 fission cross section was developed on the frame of WLUP. Pseudo fission product to simulate (n,2n) reaction for U-238, U-233 and Pa-231. To model (n, 2n) production from U-238, U-233 and Pa-231 the approach used in the WIMKAL-88 library was applied. It basically consists in introducing a pseudo fission product with an effective fission yield that represent the (n,2n) to fission reaction rate ratio of the precursor nuclide and with null absorption cross section data. It decays with a constant of unity. That means that in normal burnup calculations this fission product goes immediately to the equilibrium concentration and appears as an effective (n, 2n) source in the corresponding burnup equation.

Processing methods for dosimetry materials

The dosimetry materials of the WIMS-D library are special materials from the formatting point of view. They are non-burnable, without resonance table and scattering data. The WIMSD formatted files for these materials contain the corresponding dosimetry reaction instead of absorption and transport cross sections. These simplifying features allow the use of the PREPRO-2000 code system for the nuclear data processing. Besides, the code WIMSIE was developed in the frame of the project to perform the WIMS-D formatting and eventually to produce the special dosimetry files as the positive and negative 1/v materials, the resonant part of the 1/v absorber, the constant absorber and the inverse lethargy interval cross section. The sequence MERGER-LINEAR-SIGMA1-GROUPIE-DICTION-WIMSIE was used for the evaluated nuclear data processing.MERGER was called to extract the needed dosimetry reaction and the general information section (MF=1/MT=451) from the source evaluated nuclear data files. LINEAR was used because some point-wise dosimetry files from the JENDL/D-99 library were not linearly interpolable over the resonance range. The reconstruction tolerance was set up to 0.1% in all cases. It is worthy to note that the use of RECENT was skipped because the resonance contribution (MF=2) was always included on the point-wise cross section data (MF=3) of the source evaluated file. SIGMA1 Doppler broadened the cross section at 300 K and GROUPIE performed the multi-group data processing into 69 and 172 energy group structures. The averaging spectrum applied for most the materials in the frame of the project was converted to linearly interpolable data and used for generating infinite diluted multi-group cross sections. Finally, the index on the general information section was updated by DICTION and the WIMS-D formatting was performed by WIMSIE.

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