How to generate the precise broad group cross section is important for the fast reactor design. In this study, a fast reactor multi-group cross-section generation code MGGC2.0 are developed in-house ...for processing ultrafine group MATXS format library. Validation and verification are performed for MGGC2.0 code by applying the benchmarks of ICSBEP handbook, and the results of MGGC2.0 agree well with that of MCNP. The consistent PN method with critical buckling search is in good agreement that condensed with TWODANT flux and flux moment for the inner core and outer core region. For the radial blanket and reflector, two region approximation method has been applied in MGGC2.0 by using collision Probability Method neutron flux solver. The RBEC-M benchmark was used to verify the power distribution calculation, and the relative error of power distribution comparison with the reference are less than 0.8% in the fuel region and the maximum relative error is 5.58% in the reflector region. Therefore, the precise broad cross section can be generated by MGGC2.0 for fast reactor.
To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ...ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.
The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating ...libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark.
The advanced reactor design needs an accurate cross-section generation code. In this study, a new nuclear data processing code AXSP is developed, and the method and performance of which are ...described. Compared with the NJOY program, the precision of the unresolved resonance processing module UnresXS has been significantly improved due to the adoption of a more accurate solution method and the consideration of in-sequence overlap integrals. The time consumption of PUnresXS has been decreased significantly due to an optimized sorting algorithm. At the same time, other modules of AXSP are relatively comprehensive. The function of resolved resonance cross-section reconstruction and linearization is the ReconXS module. The Doppler broadening module is BroadXS by using Gauss–Hermite quadrature and Gauss–Legendre quadrature from 0 K temperature pointwise cross section to any temperature which is defined by the user. The shielding factor in the unresolved resonance energy region is calculated by the UnresXS or the PUnresXS module, which are developed based on the Bondarenko method and the probability table method, respectively. The ACE formatted cross sections for the Monte Carlo code is processed by the ACEXS module, and the multigroup cross sections are generated by the GroupXS module. The cross sections processed by different modules were verified by the NJOY2016 code, and the multigroup cross sections were also verified by using the critical benchmarks. The multiplication factor difference between AXSP and NJOY2016 is less than 20 pcm. In addition to this, the ZPR6/7 fast reactor is used for ACE format library verification. The results show that the criticality calculated by AXSP has a good agreement with that of NJOY2016.
Recently, three successful antineutrino experiments (Daya Bay, Double Chooz, and RENO) measured the neutrino mixing angle
θ
13
; however, significant discrepancies were found, both in the absolute ...flux and spectral shape. Much effort has been expended investigating the possible reasons for the discrepancies. In this paper, the change of neutrino energy spectrum with burnup is analyzed from the point of view of the change of neutrino energy spectrum with burnup. An accurate method for calculating neutrino energy spectrum is proposed. The non-equilibrium correction is studied by using this method. It is found that the non-equilibrium correction contributes not only to the energy region less than 4.0 MeV, but also to the energy region greater than 4.0 MeV, with a maximum correction of about 3%.
•A new WIMS-D library based on SHEM 281 energy structures is developed.•The method for calculating the lambda factor is illustrated and parameters are discussed.•The results show the improvements of ...this library compared with other libraries.
The WIMS-D library based on WIMS 69 or XMAS 172 energy group structures is widely used in thermal reactor research. Otherwise, the resonance overlap effect is not taken into account in the two energy group structure, which limits the accuracy of resonance treatment. The SHEM 281 group structure is designed by the French to avoid the resonance overlap effect. In this study, a new WIMS-D library with SHEM 281 mesh is developed by using the NJOY nuclear data processing system based on the latest Evaluated Nuclear Data Library ENDF/B-VII.1. The parameters such as the thermal cut-off energy and lambda factor that depend on group structure are discussed. The lambda factor is calculated by Neutron Resonance Spectrum Calculation System and the effect of this factor is analyzed. The new library is verified through the analysis of various criticality benchmarks by using DRAGON code. The values of multiplication factor are consistent with the experiment data and the results also are improved in comparison with other WIMS libraries.
Compared to traditional transverse integration methods, the variational nodal method, with its unique advantages, is more suitable for high-fidelity calculations of reactor physics in reactors with ...complex geometries and finer detail descriptions. In this study, the basic theory of the variational nodal method was derived and the VINUS code is developed. The neutron solver based on this method is adaptable to various geometric models, and showcased the code's fundamental framework. On this basis, a set of self-designed macroscopic cross-section benchmarks, actual macroscopic cross-section benchmark VVER-440, and few-group microscopic cross-section benchmark RBEC-M for fast spectrum reactors were used to verify different functionalities of VINUS. The results were shown that VINUS code maintains good computational accuracy and convergence trends. For the VVER-440 benchmarks, the deviation of keff of VINUS from reference is less than 100 pcm, and the maximum power deviation is less than 4 %. For the RBEC-M, the deviation of keff is 125 pcm, and the maximum power deviation is less than 5 %. These outcomes collectively demonstrate the solver's potential for engineering applications in future advanced reactor designs.
To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ...ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.
In the high-fidelity reactor physics design of fast reactors, the influence of photon effect needs to be explicitly simulated with the neutron-photon coupled transport method. The neutron-photon ...coupled transport-burnup method was researched based on the fast reactor physics design software package MOSASAUR. Photon library, photon cross-section production, photon source and photon power calculation methods were researched and the related calculation modules were developed and verified with the RBEC-M benchmark and the MET-1000 benchmark. Numerical results showed the good accuracy of the newly-developed neutron-photon coupled transport ability with MOSASAUR. The photon effect introduced the 80% difference for the assembly power in the RBEC-M benchmark, which showed the importance of the neutron-photon coupled transport in the fast reactor design.
In order to accurately calculate the photon heating distribution of fast reactor, the photon cross section processing code NGAMMA has been developed. NGAMMA can process MATXS format library and ...generate multi-group forms of photo-atomic cross sections, prompt photon production cross sections, delayed photon production cross sections, and neutron-photon KERMA factors for neutron-photon coupled heating calculations. An improved 21-groups photon library containing both prompt and delayed photon data is generated using the NGAMMA code based on the 1968n × 94γ MATXS format database. The photon library was verified using the fast reactor benchmark problem ZPR-6/7 and MET-1000. The results of the calculations show that the use of the 21-groups photon library generated by energy group collapse using 94-groups photon fluxes results in a significant improvement in the accuracy of the prompt photon heating calculations, especially in the non-fuel region, resulting in a relative error of less than 1% between the photon heating calculations and the MCNP calculations, compared to an error of 7% using the previous photon library. For the ZPR-6/7 benchmark problem, consideration of the delayed photons results in a 30% improvement in photon heating in the fuel region. For the MET1000 benchmark problem, the total power results of the neutron-photon coupling calculations based on the improved photon library are in good agreement with the MCNP, with a maximum relative error of no more than 3% in the total power of the fuel assembly. For reflector assemblies where photon power is a major contribution to heat release, the maximum relative error in prompt photon power is only −5.96%. Considering delayed photon significantly affects the photon heating, and for the MET1000 benchmark problem, considering delayed photon results in an overall increase of more than 30% in the photon power of the fuel assembly.
•Developed a high-precision photon library processing code, NGAMMA.•The delayed photon heating was calculated distribution for fast reactors accurately.•Neutron and photon coupled heating calculations were performed for the MET1000 benchmark problem.
PDF
