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[en] Highlights: • Data decomposition techniques are proposed for memory reduction. • New strategies are put forward and implemented in RMC code to improve efficiency and accuracy for sensitivity calculations. • A capability to compute region-specific sensitivity coefficients is developed in RMC code. - Abstract: The iterated fission probability (IFP) method has been demonstrated to be an accurate alternative for estimating the adjoint-weighted parameters in continuous-energy Monte Carlo forward calculations. However, the memory requirements of this method are huge especially when a large number of sensitivity coefficients are desired. Therefore, data decomposition techniques are proposed in this work. Two parallel strategies based on the neutron production rate (NPR) estimator and the fission neutron population (FNP) estimator for adjoint fluxes, as well as a more efficient algorithm which has multiple overlapping blocks (MOB) in a cycle, are investigated and implemented in the continuous-energy Reactor Monte Carlo code RMC for sensitivity analysis. Furthermore, a region-specific sensitivity analysis capability is developed in RMC. These new strategies, algorithms and capabilities are verified against analytic solutions of a multi-group infinite-medium problem and against results from other software packages including MCNP6, TSUANAMI-1D and multi-group TSUNAMI-3D. While the results generated by the NPR and FNP strategies agree within 0.1% of the analytic sensitivity coefficients, the MOB strategy surprisingly produces sensitivity coefficients exactly equal to the analytic ones. Meanwhile, the results generated by the three strategies in RMC are in agreement with those produced by other codes within a few percent. Moreover, the MOB strategy performs the most efficient sensitivity coefficient calculations (offering as much as an order of magnitude gain in FoMs over MCNP6), followed by the NPR and FNP strategies, and then MCNP6. The results also reveal that these three strategies employed by RMC maintain high parallel efficiency of approximately 96–98% within the observed 600 processors, and the memory requirements per processor decrease almost linearly as the number of processors increases from 120 to 600. To conclude, RMC is capable of performing sensitivity analysis with sufficient accuracy and high efficiency
[en] It is one of the efficient approach to reduce the memory consumption in Monte Carlo based reactor physical simulations by using the On-the-fly Doppler broadening for temperature dependent nuclear cross sections. RXSP is a nuclear cross sections processing code being developed by REAL team in Department of Engineering Physics in Tsinghua University, which has an excellent performance in Doppler broadening the temperature dependent continuous energy neutron cross sections. To meet the dual requirements of both accuracy and efficiency during the Monte Carlo simulations with many materials and many temperatures in it, this work enables the capability of on-the-fly pre-Doppler broadening cross sections during the neutron transport by coupling the Fast Doppler Broaden module in RXSP code embedded in the RMC code also being developed by REAL team in Tsinghua University. Additionally, the original OpenMP-based parallelism has been successfully converted into the MPI-based framework, being fully compatible with neutron transport in RMC code, which has achieved a vast parallel efficiency improvement. This work also provides a flexible approach to solve Monte Carlo based full core depletion calculation with many temperatures feedback in many isotopes. (author)
[en] Methods suitable for sensitivity analysis in continuous-energy Monte Carlo codes become a research hotspot in the field of reactor physics. In this work, the formulas of sensitivity coefficients of five different reaction types were established. Then, the theoretical basis and the algorithm of the iterated fission probability method which was used widely currently were discussed. Furthermore, two Monte Carlo codes, RMC and MCNP6, were used to compute eigenvalue sensitivity coefficients to nuclear data. The agreement between RMC and MCNP6 is well. The results indicate that RMC is capable to perform sensitivity analysis preliminarily. (authors)
[en] In order to improve the computational efficiency and reduce the memory usage of the function in the Reactor Monte Carlo code RMC to compute the effective multiplication factor (k_e_f_f) sensitivity coefficients with regard to nuclear data, two parallel strategies which are based on two ways to estimate the adjoint flux, namely, the next generation neutron number estimator and the neutron production rate estimator, are implemented in RMC. A multi-group infinite-medium test and a continuous-energy test are used to verify the new strategies. Results show that the sensitivity coefficients computed by the two parallel strategies agree well with the analytic solutions and those computed by MCNP6. Furthermore, the new strategies run 3 times as fast as MCNP6 and the figure of merits of the new strategies to compute the sensitivity coefficients to total cross section of different isotopes are 4 to 8 times as high as MCNP6. (authors)
[en] The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.
[en] Most of Monte Carlo (MC) simulations typically employ the continuous energy representation of the neutron cross-section (XS) for all isotopes in the nuclear system. Therefore, in order to consider temperature feedback, i.e., Doppler broadening effect, during MC transport simulation, a large number ACE formatted XSs at various temperatures, i.e., 10~50K intervals, need to be pre-generated. This increases tremendous memory burden associated with storing the XSs during the neutron transport calculation. In order to overcome the memory burden, various onthe- fly Doppler broadening techniques were proposed. One of the most novel approaches is the windowed multipole (WMP) representation which was proposed by Computational Reactor Physics Group (CRPG) in MIT. This paper presents a brief review of the methodology of WMP library generation and its application and verification in the on-the-fly Doppler broadening routine of MCS code. The conversion of conventional formalism resonance parameters into rigorous multipole parameters was performed. The reduction of the overall number of multipole parameters was achieved through the window concept. The verification of the generated WMP library was demonstrated by applying to VERA benchmark suite pin-cell problem using the Monte Carlo code MCS, in terms of multiplication factor and isotopic number density. The performance of the generated WMP library was evaluated. The XS generation time increases by 50 % and the overall Monte Carlo simulation performance loss is 30 %.
[en] The cutting of single-walled carbon nanotubes (SWNTs) with γ radiation in the presence of sulfuric acid was researched systemically. The photoacoustic effect of SWNTs caused by laser and γ radiation in aqueous solution was initially investigated. Moreover, the proposed mechanism for the cutting process of SWNTs was also analyzed. The short SWNTs were characterized and analyzed with transmission electron microscopy (TEM), Fourier transform infrared spectroscopy (FT-IR), ultraviolet-visible spectroscopy (UV-Vis) and Raman spectroscopy. Results indicate a synergetic effect between the γ radiation and sulfuric acid oxidation on the cutting of SWNTs. The resulted short nanotubes have a length distribution of 200-400 nm. Moreover, it can be uniformly dispersed in water for one week. The cutting approach induces the production of a certain number of -OH, -COOH functional groups on the short SWNTs. In addition, SWNTs irradiated by Q switch pulsed laser will occur a large photoacoustic effect in the solution, while SWNTs irradiated by γ exist a much weaker photoacoustic effect. (authors)
[en] UNIST developed Monte Carlo code – MCS has been fully coupled with sub-channel thermal/hydraulic code – CTF, together with fuel performance prediction code – FRAPCON recently to implement the neutronic, thermal-hydraulic, and fuel performance simulation capability during Monte Carlo neutron transport based depletion analysis. The BEAVRS Benchmark cycle 1 depletion with multi-physics coupling feedback has been performed by the fully coupled MCS/CTF/FRAPCON code system. Firstly, the CBC letdown with 100% power level was compared with the measured values shown in Table 24 and Table 25 of the manual, which shows a very good agreement between MCS results and measurement. Besides, the axially integrated power distribution at EOC calculated from the real power level, control rod position, inlet coolant temperature was compared with measured data, which also shows the good accuracy of this coupling code system.
[en] The development of advanced thorium-based nuclear system raises new requirements on nuclear data. The multi-group data file of critical nuclides in the thorium-uranium recycle is the foundation of physical design, analysis and calculation of the reactor core. Based on authoritative nuclear data processing code NJOY, this paper obtains a WIMS format multi-group cross section data files through processing the ENDF/B-VII.1 evaluation nuclear data file, uses the specific update maintenance procedure WILLIE to get a WIMS format data file, and conducts a series of critical benchmarks on the data file using the multi-group reactor core calculation code WIMSD5B. The results show that the computed results of the WIMS file based on the processing of ENDF/B-VII.1 are basically the same as those of the latest WIMS-D file published on the websites of the 'WIMS-D' library updating project (WLUP) with higher accuracy and reliability than those of the shipped WIMS-D file of the WIMSD5B code. Furthermore, the average deviation of the new WIMS file performing in the validation of 16 thorium-uranium cycle benchmarks is 0.225 3% smaller than that of the old WIMS file. (authors)
[en] Monte Carlo method can provide high fidelity neutronics analysis of different nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. It can also be coupled with thermal-hydraulics codes to consider the feedback. For thermal reactors such as PWR and HTGR, two main temperature effects should be considered. The first is the effect of thermal motion of target nuclei in resolved resonance energy regions, which is known as Doppler effect. The second is the thermal scattering and bound effect in thermal energy regions. The traditional approach of pre-generated cross sections has difficulty in memory footprint for detailed temperature modelling in multi-physics calculations. Recently, on-the-fly (OTF) technique has been proposed in order to reduce the memory consumption for both resolved resonance energy and thermal energy. For fast reactors calculations, the temperature dependence of cross sections in the unresolved resonance energy regions is also important. However, less attention was paid to the unresolved resonance regions compared with resolved resonance energy and thermal energy. In this paper, the on-the-fly temperature treatment of cross sections in unresolved resonance regions was proposed and developed in RMC codes. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy. The formats of libraries were changed, in which the probability tables were separated from the cross sections. Then the interpolation of probability table was performed on-the-fly during the transport routine. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy, showing that TMS method is effective and accurate for cross sections treatment in resolved resonance energy regions. The memory saving of the on-the-fly treatment is also considerable. For the high-fidelity 'transport-burnup-thermal-hydraulics' coupling calculation, the memory requirement is 60 Mb. (authors)