[en] RMC is a self-developed Monte Carlo software for nuclear reactor analysis by Reactor Engineering Analysis Lab(REAL), Tsinghua University. On the basis of the self-developed subchannel modular(RMC-TH) and Monte Carlo CellTally the inner coupling interface is developed, which combines both input files and realizes the fast mesh correspondence process using the cell expansion technology for repeat structure with thermal hydraulics feedback, and it breaks through the threshold of geometrical variability and extensibility for coupled code. On-The-Fly Doppler broaden method is adopted considering the temperature effect on micro cross section, which only needs the 0K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on single rod and 17 × 17 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology, so the technology roadmap and methodology foundation for the large scale and geometrically variable reactor steady-state as well as transient neutronics simulation with thermal hydraulic feedback and further neutronics-thermal hydraulics-depletion multi-physics simulation process are established. (authors)

[en] Accurate cross sections are of great importance to Monte Carlo particle transport calculation codes which use pointwise cross sections. It is well known that cross sections have much dependence on temperature on account of Doppler effect which is particularly vital for resonance absorption cross sections. In general, pre-prepared continuous-energy libraries are used in Monte Carlo codes, which is feasible in many applications. However, in multi-physics calculations where neutronics and thermal hydraulics are coupled, cross sections at a fine interval of temperatures are desired, which brings a great challenge to computer memory. Consequently, different methods about on the fly Doppler broadening for cross sections are proposed, which can generate cross sections at the specific temperature points during neutron simulations. These methods can be roughly classified into two categories: one is to calculate the required cross sections by different algorithms, such as interpolation using a high order function expansion, multipole representation, and different integration algorithms based on Doppler broadening equation. The other, for instance, target motion sampling (TMS), does not calculate the Doppler broadening cross sections but takes into account Doppler effect by rejection technique. In this paper, several on the fly Doppler broadening algorithms based on Doppler broadening equation are implemented in the Reactor Monte Carlo code RMC, and then the accuracy and efficiency are discussed and compared. In this paper, several algorithms for on the fly Doppler broadening based on Doppler broadening equation have been implemented in RMC and the focus falls on the comparison of the Gauss Legendre quadrature and Gauss Hermite quadrature. Gauss Legendre quadrature is more accurate than Gauss Hermite quadrature but also more time consuming. Base temperature has a significant impact on Gauss Hermite quadrature while increasing base temperature to room temperature (300 K) can give satisfactory results. In addition, the proportion of overhead of Doppler broadening treatment in total computational time tends to decrease with an increase number of nuclides in fuel. For depletion calculation, both algorithms are in good agreement with reference and the relative errors of K_inf are within 5 times of standard deviation. As for total computational time, the ratio between Gauss Legendre quadrature or Gauss Hermite quadrature and reference decreases and reaches 9.6 or 3.4 at last, respectively. Four important isotopes are chosen to compare the atom density and all relative errors are within 0.25%. After comparisons of different algorithms and different base temperatures, it can be found that Gauss Hermite quadrature broadening from 300 K can perform on the fly cross section Doppler broadening with least total computational time and acceptable precision. (authors)

[en] - This paper introduces a powerful and efficient method to improve the versatility of the hybrid coupling between the Monte Carlo code RMC and the sub-channel code CTF. This method is accomplished based on the HDF5 file which can be produced by CTF, by using the hierarchical data format to store data. The neutronics and thermal-hydraulics coupling between RMC and CTF has been achieved and applied to the BEAVRS benchmark previously by the REAL group of Tsinghua University. To broaden the scope of applications of the previous coupling codes, the method mentioned in this paper is proposed. The new versatile coupling code system that uses the HDF5 file is applied to the BEAVRS benchmark again to test and verify the versatility and high efficiency of the versatile coupling codes. This paper firstly introduces the HDF5 file and coupling codes RMC and CTF briefly, and then explains the versatile coupling method in details. Besides, the modeling of the BEAVRS benchmark is also presented in this paper. Next, this paper presents the results and analysis of coupling calculation. The conclusions and the future work are introduced finally. (authors)

[en] As the computational accuracy required for design and safety analysis of current and future nuclear reactors is continuously increasing, high fidelity numerical reactor simulations including different physics raise wide concern of researchers. The interplay between neutronic and thermal-hydraulic (N-TH) effects of a nuclear reactor core plays an important role in reactor design, safety analysis and core performance. 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. Armed with the massively parallel algorithms, the computational cost of Monte Carlo codes can be greatly reduced. Compared to single channel thermal process, sub-channel codes have enormous advantages of detailed modeling of full-core reactor or single assembly and high accuracy of calculation results. Detailed sub-channel modeling of a full-core also needs running in parallel, otherwise the computational cost will be relatively high. In this paper, a new hybrid coupling with continuous-energy Monte Carlo code RMC and the sub-channel code CTF has been developed. The coupling codes are applied to steady-state simulations of BEAVRS benchmark in hot full power condition. In order to deal with the temperature dependence of cross sections for Monte Carlo code, the on-the-fly cross sections treatment has been developed in RMC. The on-the-fly method was applied to the N-TH coupling in this work. (authors)

[en] In reactor analog simulation, the Monte Carlo method has relatively high fidelity since it can reduce calculation errors resulting from using approximation methods by describing complex geometry and neutron energy spectra. The Monte Carlo method is, in essence, a stochastic simulation of the neutron transport process, and statistical uncertainty of all the computed results must be provided. In large-scale reactor computation, this method can accurately estimate integral quantities such as the effective multiplication factor (K_eff), and it can limit the uncertainty within an acceptable range via reasonably selecting simulation particles. Nevertheless, it faces great challenges in estimating local quantities like neutron flux and fission power, with uncertainty of estimators beyond acceptance. To decrease the uncertainty of local quantities by increasing simulation particles may cost much time and reduce efficiency. The main reasons are as follows. Firstly, the scale of reactors is so large that common pressurized water reactors have tens of thousands of lattice cells; secondly, uneven distribution of neutron flux and fission power causes uneven distribution of uncertainty. Smith K. puts forward the 95/95 principle to estimate whether uncertainty calculated with Monte Carlo Code satisfies the requirement. According to the principle, with a confidence level of 95%, statistical errors of power density in more than 95% of the regions should be less than 1%. To meet the requirement of the principle, the uniform fission site (UFS) method is raised currently. UFS aims at redistributing fission sources to uniformize the fission source sampling, which can make statistical variance distribution more uniform. This method, which is easy to use, will not obviously increase computation time. Many studies show that UFS has a good variance reduction effect in Monte Carlo criticality calculation, but less attention is paid to its performance in burnup calculation and thermal-hydraulic coupled calculation. Uncertainty arising from the Monte Carlo method has a large impact on burnup calculation and thermal-hydraulic coupled calculation. Variance propagates between burnup steps and between iterative steps of coupled calculation, which causes asymmetry of results of the physical symmetrical regions and instability of coupled iterative convergence errors. To further verify the effect of UFS, this study uses it in the RMC (the Reactor Monte Carlo code) burnup calculation module and the neutronics/thermohydraulics coupled calculation module, and conducts two-dimensional pressurized water reactor burnup calculation and BEAVRS benchmark thermalhydraulic coupled calculation. In this way, it verifies the improvement effect of UFS on variance propagation. (authors)

[en] RMC (Reactor Monte Carlo) is a self-developed Monte Carlo code for nuclear reactor analysis by Reactor Engineering Analysis Lab (REAL), Tsinghua University. On the basis of the self-developed subchannel module (RMC-TH) and Monte Carlo Cell Tally, the internal coupling interface is developed, which combines both input files to one and realizes the fast mesh correspondence process using the cell expansion technology for repeated structure with thermal-hydraulics feedback. It breaks through the bottleneck of geometrical extensibility for coupled code. On-the-fly Doppler broadening method is adopted as the way to consider the temperature effect on microscopic cross section, which only needs the 0 K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on PWR fuel pin and 17x17 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology. Therefore, a promising technology road map for the large-scale and geometrically universal nuclear reactor in both steady-state and transient conditions with thermal-hydraulic feedback are established. The road map can be further applied to neutronics-thermal-hydraulics-depletion coupling in multi-physics simulation process. (authors)

[en] The information of primary knock-on atom of material under neutron radiation in nuclear reactors is the inputs for multi-scale material computational simulations. Two different computational methods for primary knock-on atom (PKA) simulation under neutron radiation, including scattering matrix transformation and Monte Carlo simulation, were investigated in this paper. The PKAs of Zr, Fe, W and SiC under the neutron radiation in pressurized water reactor (PWR) were simulated using these two methods. The energy spectrums of PKAs were attained and compared, which shows the good agreements between these two methods. Moreover, the Monte Carlo simulation method has advantages for considering the effects such as the thermal motion of target nuclides and the chemical bonds between nuclides. The research in this paper provides foundations of the follow-up molecular dynamics and multi-scale material simulations. (authors)

[en] CTF (Coolant Boiling in Rod Arrays-Two Fluid) is a new sub-channel thermal/hydraulic simulation code developed by CASL (The Consortium for Advanced Simulation of Light Water Reactors) and PSU (Pennsylvania State University). It can solve steady or transient-state problems efficiently for both single assembly and full-core reactor. Thus, this code solves the computational efficiency and memory consumption problems effectively. First, CTF computes the BEAVRS benchmark in parallel with the domain decomposition technology. The power data which CTF uses are calculated by RMC(a Monte Carlo code for reactor core analysis). After calculating for 268 s, we get detailed fuel pin temperature, water temp and density output results. Accordingly, the efficiency and reliability of CTF is verified. On this basic work of CTF calculation for BEAVRS benchmark,the coupling between RMC and CTS for full-core problem will achieved soon. (authors)

[en] A novel uranium-thorium mixed fuel assembly for SWCR is introduced in this paper, based on the neutron spectrum of SCWR and neutronic characteristics of thorium fuel. Neutronic characteristics of the introduced fuel assembly have been investigated using the Dragon codes. The parameters in different working conditions, such as infinite multiplication factors, reactivity temperature coefficients, fissile inventory ratio (FIR) and its relation with the operation period have been assessed by comparing with conventional uranium assembly. Moreover, the Moderator-to-Fuel Ratio (MFR) was changed in order to investigate its influence on the neutronic characteristics of fuel assembly. Results show that the harder neutron spectrum of SCWR is favorable for breeding the fissile nuclides, and the new uranium-thorium mixed fuel assembly has advantages on both efficient fuel utilization and lower minor actinide generation. (authors)

[en] In conventional transient analysis, thermal hydraulics calculations are generally coupled to deterministic neutronics solvers. Due to the limited accuracy of deterministic methods and the rapid development of computer power, Monte Carlo neutronics calculations have received increasing attention nowadays. One of challenges for Monte Carlo methods coupled to thermal hydraulics calculations is the temperature dependent cross sections. Taking the detailed temperature distribution feedback into account, each nuclide may be in a large range of temperatures, pre-generated libraries at every temperature point are no longer applicable owing to unaffordable memory footprint. Different solutions have been proposed to deal with this problem. TMS, accounting for Doppler broadening effect by the explicit treatment of thermal motion, was used to model fast transients in Serpent 2. Stochastic mixing, using two sets of libraries for each nuclide and mixing them to get the correct temperature, was used in transient analysis by Sjenitzer and Hoogenboom. Besides, there are other on the fly Doppler broadening methods being developed, including series expansion of cross sections, multipole representation, different integration algorithms based on Doppler broadening equation. In this paper, on the fly Doppler broadening treatment based on integration algorithms was applied to transient calculation in Reactor Monte Carlo code RMC . In the following section, the integration algorithms implemented and dynamic simulation method in RMC is briefly introduced. Finally, a pure physical transient fuel assembly model was chosen for numerical validation. (authors)