Design of a ball screw for the electric control rod drive mechanism in nuclear reactors

Neutronics calculation and analysis of a small lead-bismuth fast reactor core with control rods integrated within fuel assemblies

3-D numerical studies of single control rod withdrawal transients in an LBE cooled critical reactor

Study on non-uniform control rod with BeO reflectors

Development of cross-section model of two-step code system considering control rod depletion for soluble boron free operation of SMR

Tightly coupled multi-physics algorithm for control rod follow mode based on Newton-Krylov method

Characterization and visualization of eutectic melt in B4C powder and pellet-based/304 stainless steel composite control rods under radiative heating

Modeling and analysis of control rod drop in a floating nuclear power plant under ocean conditions

Compared to traditional large-scale reactors, the more heterogeneous, boron-free SMR cores create additional challenges for accurate multiphysics simulations. Therefore, advanced modelling and simulation tools should be used to create high-fidelity, high-accuracy, and computationally efficient multiphysics and multiscale solvers. These solvers can evaluate the safety and performance of SMRs and could be attractive for industrial applications if the computational power requirements were reasonably low. The first crucial step in building a computationally efficient simulation model is to define an SMR benchmark model. This model is a reference for validating the simulation results. In this paper, the benchmark model is a NuScale-like SMR, where the Serpent code has been utilized to run the neutronic simulation. The neutronic simulation was then performed again in the benchmark model, this time utilizing OpenMC code. The results of the Serpent and OpenMC codes were compared in terms of the reactivity coefficient, control rod worth and radial and axial power distribution. By comparing two different codes to validate the simulation of the NuScale-like benchmark, OpenMC will be utilized for future work, such as generating the nuclear material cross-section data for core simulators.

FDA Oversight and the Magnetic Expansion Control Rod

Abstract The development of a 10 MW research reactor core using high-density U3Si2–Al plate fuel is presented, focusing on enhancing neutronic safety through optimized core configuration and control element design. This study employs MCNP and CITATION codes to evaluate neutronic parameters, including Power Peaking Factor (PPF) and Stuck Rod Shutdown Margin (SRSM). By incorporating four symmetrical irradiation boxes in the core center, the PPF is reduced by approximately 5 %, distributing neutron flux more uniformly. A novel control assembly design with a thin D2O layer increases control rod worth ensuring SRSM <−1,000 pcm without additional absorber material. Reactivity feedback analysis confirms safe operation with negative coefficients for fuel temperature and moderator conditions (−1.56 pcm/°C for D2O −20.11 pcm/°C for combined moderator temperature and density with D2O). This approach enhances safety margins and extends cycle length, offering a novel solution for high-density fuel research reactors.

In this study, the input data required to model and perform neutronic calculations for the Bushehr Nuclear Power Plant (BNPP) with the MCNPX code were prepared. A comprehensive workflow was developed and implemented to solve the statistical transport equations, from which the neutron flux was obtained. Subsequently, key parameters, including the thermal and fast fluxes, were evaluated at Hot Zero Power (HZP) under three core conditions: a clean core, insertion of control-rod group~10, and insertion of all control-rod groups. The results indicate that, owing to the relatively large spacing and arrangement of control-rod groups~9 and~10, they do not shadow one another; instead, an anti-shadowing effect (ASE) is observed. Consequently, inserting these rods increases the neutron flux in the core, thereby enhancing the overall worth of the control rods.

The objective of this work is to determine the temperature distribution in the IRT-4M type fuel assembly with fuel 19.75% enriched in 235U in the core of the WWR-SM reactor for the case of a square tube with rounded edges and a round hole in the center and the case of a round tube. In the case of installing a round tube inside the fuel assembly instead of a square tube with rounded edges and a round hole in the center, the volume of water in this space increases. On the one hand, this leads to improved heat removal, since the volume of cooling water increases, and on the other hand, an increase in the volume of water leads to an increase in thermal neutrons on this side of the fuel element, and this, in turn, leads to an increase in energy release. To determine these changes, we performed neutron-physical and thermal-hydraulic calculations for a channel with a square tube with rounded edges and a round hole in the center and for a round tube. It has been determined that replacing a square tube with rounded edges and a round hole in the center with a round tube as a guide for installing a compensating control rod will not affect the nuclear safety of the WWR-SM reactor operation.

This paper addresses the current challenge of developing and refining a mathematical model of a nuclear power plant with a VVER-1000 reactor, enabling accurate investigation of power changes and associated technological processes. An approach to automated power control of the NPP using a three-loop regulation system is proposed, combining changes in boric acid concentration, control rod movement, and main turbine valve positioning. Special attention is given to the distributed core model, which provides a more realistic representation of transient processes, particularly the influence of poisons. A step-by-step algorithm for implementing the method in practice is presented, and its effectiveness is analyzed in comparison with traditional reactor control methods.

Corrigendum to “Neutronic analysis for withdrawal of TRIGA MARK II reactor control rods using OpenMC program” [Ann. Nucl. Energy 217 (2025) 111330

Application of DeepONet to predict transient drop motion of the control rod in real-time

Research on radiological consequences analysis of reactor control rod ejection accident based on peak radial average fuel enthalpy rise

Digital Twin Model and Platform Based on a Dual System for Control Rod Drive Mechanism Safety

Corrosion behavior of rare earth hafnate for neutron control rod in subscritical water

This paper presents the implementation and verification of the developed OpenNode nodal diffusion code for the static neutron analysis of liquid-metal fast breeder reactors (LMFBRs). OpenNode, initially implemented and tested on pressurized water reactor benchmarks, is based on the nodal expansion method and features a graphical user interface with Blender integration for advanced three-dimensional (3D) geometric modeling and results visualization. In this work, OpenNode is applied to fast spectrum reactor problems to assess its accuracy and flexibility in modeling LMFBR cores. Several two-dimensional and 3D LMFBR benchmark cases are analyzed, including configurations with four neutron energy groups, heterogeneous fuel assemblies, radial and axial blankets, and partial control rod insertion scenarios. Numerical results are compared with reference solutions from the KOMODO code, which is also based on the semi-analytical nodal methodology, to assess performance in terms of effective multiplication factor k eff , neutron flux distributions per group, and power profiles. This study also includes a mesh refinement analysis to investigate the impact of axial discretization on the stability and spatial accuracy of the solution. To reinforce the evaluation, relative error distribution plots are provided for flux and power comparisons, offering a quantitative insight into the behavior of the solver. The results show strong agreement with reference data, with reactivity deviations generally below a few hundred pcm and spatial errors generally below 2%. These results confirm that OpenNode can be reliably extended to fast reactor simulations, and support its use as an accessible tool for research, code verification, and educational purposes in the field of reactor physics.