The leakage jet impact effect of steam generator heat exchanger piping causes the vibration response of adjacent piping and secondary sound radiation. This study establishes a three-dimensional numerical calculation model of a leakage jet impacting the pipe wall and studies in depth the action mechanism of the leakage jet on the adjacent pipe wall. It obtains the vibration characteristics of the pipe wall caused by the impact of the leakage jet. Calculations show that the leakage jet excites a complex flow field on the surface of the adjacent pipe wall, generating dynamic pressure loads and shear stresses. Dynamic pressure loads generate significant negative pressure wave signals on adjacent pipe walls, which in turn trigger pipe wall vibrations. With an increase in the leakage hole diameter and internal pressure, the dynamic load on the pipe wall increases significantly, leading to obvious changes in the vibration characteristics, revealing the influence law of the leakage jet on the dynamic response of the pipe wall.

In nuclear reactors, delayed neutron precursors (DNPs) are important for reactor safety and operation. In liquid nuclear fuels, DNPs are transported by the flow, and an advection-reaction balance equation for their concentration must be solved in addition to the neutron balance equation. This research paper applies the method of characteristics to solve the DNP equation using techniques previously developed in petroleum engineering and groundwater analysis. The calculation strategy incorporates a pathline generation algorithm covering all the mesh cells of the geometry for concentration calculations, and pathlines are reconstructed for both step and piecewise linear velocity fields. The analytical solutions are used on a Cartesian mesh to compute the DNP concentrations, assuming a step source of DNPs. The results obtained on the steady-state phases of a benchmark problem illustrate this new method’s capability to solve advection-dominated problems in liquid-fueled reactors. The method developed in this work is suited to smooth velocity fields (e.g. laminar or Reynolds-average Navier-Stokes flows) where pathlines can be tracked and where DNP diffusivity is negligible.

Simulated dissolved stainless steel–clad plutonium (Pu) and plutonium/uranium nuclear fuel in nitric acid (HNO3) was neutralized to a free hydroxide (OH−) concentration of 0.6 M. A thermal neutron poison, gadolinium (Gd), was added to the simulants at concentrations of either ~3 to 6 g/L or ~37 to 38 g/L. The supernate Pu concentration on the day of neutralization ranged from 0.48 to 8.75 mg/L. The supernate Pu concentration of a simplified simulant neutralized to 0.6 M OH− above precipitated solids containing Pu was demonstrated to decrease over 18 days. A significant portion of precipitated Pu was found to be insoluble in 8 M HNO3 at ambient temperature, but essentially quantitative Pu dissolution was achieved in 11.5 M HNO3/0.1 M potassium fluoride at 100°C. The difficulty in dissolving the Pu precipitate is believed to be due to the formation of refractory PuO2•xH2O during the neutralization process. Initial Gd concentrations of ~37 to 38 g/L were found to result in a greater aluminum precipitation when neutralized to 0.6 M OH− than initial Gd concentrations of ~3 to 6 g/L. Physical properties of the resultant slurries were measured and used to calculate limiting flow rates and slurry velocities by gravity only in transfer piping between the Savannah River Site’s H-Canyon Facility and the Concentration, Storage, and Transfer Facility. These results were compared to calculated deposition velocities to predict if solids would settle during the transfer. The Newtonian model was found to be reasonable for each diluted slurry evaluated. Deposition velocities of Pu-containing slurries are lower than nuclear fuel slurries primarily composed of uranium because of the high density of Pu solids. Dilution of slurries reduces the margin between the slurry and the deposition velocities due to the reduction in viscosity because higher viscous forces on the particles promote maintained suspension.

The occurrence of a steam explosion resulting from fuel coolant interaction poses a significant threat to the integrity of nuclear power plants when extremely high-temperature molten corium is released into the preflooded reactor cavity. The present study establishes and verifies a computational fluid dynamics (CFD) model by simulating the TROI steam explosion experiment. The suggested model uses the Lagrangian method to simulate particles and adopts a secondary breakup model by which the particles are fragmented based on the critical Weber number. The increased number of fine particles, surface area growth, and the propagation of the explosion pressure wave following the triggering of the steam explosion are effectively simulated with the established model. The formation of steam flow and the subsequent breakup of particles are basically governed by the heat transfer between the corium particles and the cooling fluids. The mass distribution of particle sizes after breakup is obtained by modifying the main terms of the error function, which determines the diameter of child particles to be comparable with experimentally measured distributions. With this modeling, the maximum pressure obtained by the simulation approaches the measured peak pressure. This suggests that the established CFD model is successful in describing the overall thermal-hydraulic phenomena during a steam explosion. In the future, the steam explosion CFD model will be further enhanced to obtain a more sophisticated model to minimize the uncertainty in steam explosion predictions.

In this article, we evaluate how the use of threshold fission chambers (FCs) affects the performance of the current-to-flux (CTF) technique being considered for accelerator-driven system (ADS) reactivity monitoring. These detectors are envisaged because of their longer service life in the lead-bismuth eutectic reactor of the MYRRHA ADS. For this purpose, we study an extract of the new CTF experiments carried out at the GUINEVERE facility, equipped with FCs featuring 235U or 238U deposits. The extract selected for its representativeness is about the measurement of the change in subcriticality of the VENUS-F reactor triggered by the removal of one of the control rods (CRs) from its core. The results show that raw CTF reactivity values depend on the composition of the FC used and on its relative position to the CR moved. The space-energy effects, which distort the reactivity values, are understood with the help of simulations computed with the Monte Carlo neutron transport code MCNP. Based on simulated fluxes and detector count rates at all assembly positions, we investigate the physics mechanisms of absorption and diffusion by the boron carbide filling the CRs, which wear away the flux around the CRs. The amplitude of the phenomenon depends on the neutron energy and the position relative to the CR and consequently affects differently the FCs with or without a fission threshold placed close to the neutron absorber.

This study investigates the rewetting behavior of an Advanced Heavy Water Reactor (AHWR) fuel rod bundle during a loss-of-coolant accident using computational fluid dynamics simulations with ANSYS CFX. The analysis focuses on the cooling effectiveness of radial jet impingement at varying flow rates and its impact on rewetting temperature and wetting delay. Simulations were conducted by maintaining a constant initial wall temperature, with cooling curves and contour profiles extracted from various angular positions along the axial rod surfaces. The results reveal that rewetting is faster near the jet sections due to enhanced coolant interaction, while areas farther from the jets exhibit delayed wetting and elevated wall temperatures, where vapor accumulation hinders heat dissipation. Higher flow rates minimize wetting delays and improve cooling by promoting transition and nucleate boiling. However, irregular coolant splashing and vapor dominance disrupt the uniformity of rewetting across the bundle. The study highlights the limited impact of increased flow rates on achieving consistent rewetting along the entire rod length, with substantial fluctuations observed in cooling performance at different vertical positions. The findings emphasize the need for further research under high-temperature steam conditions to better understand boiling mechanisms and improve the stability of emergency cooling systems in nuclear reactors.

The master-slave robotics system was invented at Argonne National Laboratory in 1948 to protect an operator from the harmful effects of radiation. Worker safety has been the highest priority at the U.S. Department of Energy (DOE) as it continues its cleanup mission. Safety is achieved by separating the human from the nuclear source and shielding that source to a manageable level. Robotics with remote control is a natural solution for many of DOE’s cleanup tasks. Unlike conventional robotics, a remote-handling system always involves a human being within the control process. The main handling device is a manipulator because the majority of remote-handling tasks need the intuition and intelligence of a human operator. In this paper, we take a historical perspective to the technology when exploring robotic systems from the past, on what we are currently doing, and what technologies would enable new capabilities for the future. We look at past successes and failures, and we glean lessons learned that can be applied to current and upcoming robotic activities within DOE. We also take a look at some new and emerging technologies that when matured could have a positive impact on where robotics within DOE will be needed.

Thermal-hydraulic modeling and simulation are heavily dependent on the knowledge of flow regimes, which sort multiphase flow patterns into classifications with static boundaries that display characteristic features of flow patterns and phase distributions. Flow regime maps heretofore have been primarily developed mainly for vertical or horizontal pipe orientations by capturing images of flows at varying conditions and assigning them to different flow regimes based on visualization studies. As such, this method is inherently subjective. Furthermore, classifying two-phase flows into regimes with static boundaries is not physical, as two-phase flows undergo gradual changes as they transition from one regime to another. Therefore, such a static approach is not only unphysical but also inaccurate, resulting in the erroneous analysis of two-phase flows. Additionally, subjective observations at such flow conditions are prone to disagreement and inconsistencies in labeling. These difficulties are exacerbated in inclined two-phase flows. In this study, a novel two-phase flow pattern map employing the dynamic red, green, and blue (RGB) color model is presented. The model contains the embedded information of three normalized time-averaged statistics from signals obtained by a dual-ring impedance meter. A database of impedance meter signals from a 25.4-mm-inner-diameter adiabatic air-water inclinable test facility is used to compute statistics for each experimental flow condition. Multiple statistical measures are evaluated, and a principal component analysis is performed for feature selection. By mapping each parameter to intensities of red, green, and blue, a dynamic RGB flow pattern map can be produced in which groups of similar colors are representative of various flow regimes and the dynamic transition characteristics between the flow regimes are characterized physically. Therefore, this new method makes viable the observation of how flow patterns gradually change with respect to angle, as well as the fuzzification or quantification of the uncertainty surrounding experimental transition boundaries.

A series of simulated experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency to simulate a light water reactor core with a tight lattice cell containing highly enriched mixed-oxide fuel with a fissile plutonium (Pu) ratio >15%. The prediction accuracy of the neutron computation codes and nuclear data libraries in a wide range of neutron spectra was evaluated by constructing three experimental configurations of the FCA-XXII-1 assembly with different void fractions (45%, 65%, and 95%) of the moderator material (foamed polystyrene). The hydrogen-to–nuclear fuel atomic ratio was systematically varied from 0.1 to 0.8. In a previous paper, we reported the criticality and reactivity worths measured in these experiments. This technical note provides the experimental results for the central reaction rate ratios and fission distributions as follows. The fission rate ratios of uranium (U) (238U) and 239Pu relative to 235U were measured at the core centers using three calibrated fission chambers, and the 238U capture reaction rate ratio relative to 235U fission was measured using depleted U foils. The reaction rate distributions were also obtained by traversing four micro fission chambers of highly enriched U, natural U, Pu, and neptunium through each core region in the radial and axial directions. The experimental analyses were performed using detailed models of the Monte Carlo code MVP3 with the Japanese evaluated nuclear data library of JENDL-4.0. Most calculation results agreed well with the experiments, whereas those for the fission rate ratio of 239Pu to 235U were underestimated by up to 6% with the softening neutron spectrum.

The quasidiffusion (QD) method is an established and efficient iterative technique for solving particle transport problems. Each QD iteration consists of a high-order SN sweep, followed by a low-order QD calculation. QD has two defining characteristics: (1) its iterations converge rapidly for any spatial grid and (2) the converged scalar fluxes from the high-order SN sweep and the low-order QD calculation differ, by spatial truncation errors, from each other and from the scalar flux solution of the SN equations. In this paper, we show that by including a transport consistency factor in the low-order equation, the converged high-order and low-order scalar fluxes become equal to each other and to the converged SN scalar flux. However, the inclusion of the transport consistency factor has a negative impact on the convergence rate. We present numerical results that demonstrate the effect of the transport consistency factor on stability.