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The development of fast reactors with complex designs and operation status requires more accurate and effective simulation. The Monte-Carlo method can generate multi-group cross-sections in arbitrary geometry without approximation on resonances treatment and leads to good results in combination with diffusion codes. However, in previous studies, the coupling of Monte-Carlo generated multi-group cross-sections (MC-MGXS) and transport solvers has shown relatively large biases in fast reactor problems. In this paper, the main contribution to the biases is proved to be the neglect of the angle-dependence of the total cross-sections. The flux-moment homogenization technique (MHT) is proposed to take into account this dependence. In this method, the angular dependence is attributed to the transfer cross-sections, keeping an independent form for the total sections. For the MET-1000 benchmark, the multi-group transport simulation results with MC-MGXS generated with MHT are improved by 700 pcm and an additional 120 pcm with higher order scattering. The factors that cause the residual bias are discussed. The core power distribution bias is also significantly reduced when MHT is used. It proves that the MC-MGXS with MHT can be applicable with transport solvers in fast reactor analysis.

With the development of nuclear energy, microreactors have received increasing interest among researchers in recent years. In this paper, a megawatt heat pipe reactor with metal hydride moderators and burnable poisons is proposed. The hydrogen stability of the reactor under accident conditions, including reactivity insertion accidents, loss of power conversion unit heat sink accidents, and heat pipe failure accidents are analyzed. In this work, Gd2O3 is introduced as a burnable poison in the form of mixing with the UO2 fuel. According to the results of the burnable poison design, the 0.1% mass fraction of Gd2O3 is selected as the burnable poison loaded in the core. Safety analysis indicates that the introduction of burnable poisons can be beneficial during the positive reactivity insertion accident as it can reduce the excessive reactivity at BOL, thus reducing power peak and preventing hydrogen dissociation in ZrH1.4 rods. However, during the loss of PCU heat sink accident, ZrH1.4 rods will dissociate regardless of the presence of burnable poisons, whereas YH1.7 rods show better hydrogen stability. Moreover, in the event of the heat pipe failure accident, ZrH1.4 rods are more susceptible to dissociation than YH1.7 rods. As a result, the YH1.7+BP core is a better choice compared with other designs proposed in this paper as it provides a relatively high temperature margin.

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on “Neutronics Benchmark of CEFR Start-Up Tests” offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.