Helical wire-wrap spacers are a common structural element in liquid metal–cooled fast reactors (LMFRs), where they promote coolant mixing and improve heat transfer in densely packed fuel assemblies. Nevertheless, the wire spacers introduce additional thermal resistance which can result in localized peak temperature. An experiment was performed to measure local temperature distributions associated with wire contact in a simple rectangular channel under forced convection. Galinstan was employed as a working fluid to emulate the thermal–hydraulic behavior of low–Prandtl-number coolants. The results indicate that hotspot formation occurs when the Biot number exceeds unity, whereas for Biot numbers below one, a cold spot occurs due to the fin effect of the wire. The magnitude of the temperature peaks proportionally increases with applied heat flux. An empirical correlation was presented to estimate hotspot formation over a broad range of material properties and flow conditions and extensive parametric studies were performed using complementary computational fluid dynamics (CFD) simulations. The outcomes of this study provide valuable guidance for optimizing wire-wrap spacer designs to mitigate localized thermal hotspots in LMFR systems.

Sodium heat pipes, with their passive operation and high thermal transport efficiency, are well suited for integration into nuclear microreactor systems. A comprehensive understanding of their performance across a wide range of operating conditions is essential, especially under scenarios susceptible to evaporator dryout and localized overheating. In this study, these phenomena were investigated using temperature measurements complemented by high-resolution x-ray radiography. The effects of key parameters - specifically sodium filling ratio, inclination angle (including negative inclinations down to-15 degrees), and condenser cooling intensity (up to 360 W/m2 & sdot;K) - were examined to assess their influence on startup behavior, capillary limit performance, and operable regions. Results indicate that partial wetting of the evaporator facilitates startup, whereas complete dryout leads to severe overheating and potentially startup failure. Excessive cooling was found to promote sodium solidification in the condenser, which impedes liquid return and degrades overall heat transfer performance. Moreover, deviations from classical capillary-limit models were observed under conditions of strong axial temperature gradients, reflecting the influence of temperature-dependent sodium properties on capillary behavior. Under negative inclination, the opposing gravitational head further restricts liquid circulation, thereby intensifying evaporator overheating and limiting operable range. These findings provide new insight into the combined effects of parameters on overheating and dryout, contributing to the development of robust sodium heat pipe designs for advanced microreactors.

Alkali metal heat pipes are key components for high-temperature heat removal in microreactors, where multiple heat pipes are embedded directly within the reactor core. Their collective thermal behavior, including interactions under off-normal conditions, remains insufficiently characterized. This study investigates the performance of a ten-heat-pipe array under a range of operating conditions, including simulated failure scenarios. Under steady-state operation, the vertical configuration produced more uniform vapor temperatures among the heat pipes, albeit with slightly longer startup times and marginally higher average operating temperatures compared to the horizontal orientation. System uncertainties associated with the heat pipe array configuration were quantified to support future model validation. Failure conditions were emulated by replacing one or two heat pipes with empty stainless-steel tubes. In these cases, neighboring heat pipes redistributed the thermal load, resulting in increased local heat fluxes and a corresponding rise in system temperature. This localized overloading further amplified thermal stresses on adjacent heat pipes. The experimental results provide important guidance for the design of future experiments and the validation of numerical models for heat pipe arrays, as well as for the safe design and operation of microreactor systems.

A condensation inside a vertical tube was experimentally studied under both pure steam and air-steam mixture conditions ranging 100-750 kPa of the pressure, 7.5-50 kg/m2s of the inlet steam mass flux, and the inlet air mass fraction up to 0.5. The detailed distributions of the local heat flux, condensation heat transfer coefficient, and air mass fraction were measured and analyzed along the radial and axial directions according to the axial change in steam quality. The effect of the inlet steam mass flux, pressure, and air mass fraction on the local heat transfer coefficient was comprehensively discussed in relation to the accumulation of the condensate film and air layer. Based on the present data, a new prediction model for the film condensation heat transfer coefficient was proposed with the empirical correlation factors accounting for the entrance effect, condensate convection, and degradation of condensation by the non-condensable gas. The proposed model showed 4.0 % and 6.6 % of relative mean absolute errors for pure steam and air-steam mixture condition, respectively. It also provided a satisfactory accuracy with overall 28-37 % of relative mean absolute errors in the evaluation including the existing experimental data (D = 7.5-47.5 mm, P in = 0.031-7.46 MPa, G s.in = 1.3-456 kg/m2s, w nc.in = 0-0.42), showing a noticeable improvement over the existing empirical models.

Stratified flow in horizontal tubes is frequently observed in gas-liquid two-phase flow system. In the two-fluid modeling, it is important to define the interface shape in solving the balance equations to determine the key parameters such as the interfacial transfer terms, void fraction, and pressure drop. A double-circle model is usually introduced to depict the concave-down interface in a horizontal circular tube under the stratified-wavy flow condition. However, calculation of the central angle in the double-circle model, which represents the interfacial curvature, requires an appropriate iterative numerical root-finding scheme to solve the implicit transcendental equation. In this study, an explicit approximate equation has been proposed without requirement of the iterative scheme and numerical instability, which is expected to improve the coding process and computation efficiency in the analysis code with the two-fluid model.

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

Microreactor technologies are required to provide reliable carbon-free power generation in remote applications. The heat pipe–cooled microreactor concept, in particular, offers notable advantages due to the passive operation of heat pipes, enabling increased reliability and simplicity in a more compact form factor. There is a significant need for experimental work to aid and expedite the deployment of heat pipe microreactors due to their unique technological characteristics. Thus, there has been increased interest in heat pipe experiments by numerous institutions in order to support these efforts.

Alkali metal heat pipes are considered as a promising primary cooling system of special-purpose nuclear reactors. For the deployment, it is important to understand the heat pipe's behaviors under different operation conditions. In this paper, the startup characteristics and operation performance of three sodium heat pipes with 67%, 102% and 172% sodium filling ratios, respectively, were investigated, and the two-phase sodium flow was visualized using high-speed, high-resolution x-ray radiography. The boiling flow regimes in the evaporator region were categorized into stagnant sodium liquid pool, geyser boiling and developed boiling. The effect of key parameters, such as the boundary conditions of the evaporator and condenser, the sodium filling ratio, and the inclination angle were investigated. It was found that the heater power and condenser cooling conditions have a great influence on the operation temperature within the achievable range of heat transfer rate. The filling ratio and inclination angle affect the boiling phenomena and the heat pipe operational range. The experiments can provide valuable data for heat pipe model validation as well as the design of the special-purpose nuclear reactor and other high-temperature heat pipe systems.

High-resolution two-phase flow data in the rod bundle are important in the development and validation of high-fidelity models for computational fluid dynamics and subchannel codes, in particular, those pertaining to light water reactor cooling systems. The Michigan Adiabatic Rod Bundle Flow Experiment (MARBLE) has been constructed as a modular assembly of an 8 x 8 lattice rod bundle to simulate scaled pressurized water reactor and boiling water reactor subchannel assemblies. To establish a high-spatial resolution database of the void fraction in the reactor fuel assembly geometries, tomographic measurements were performed with the High-Resolution Gamma-ray Tomography System, which was designed and built in house; the detector system has a spatial resolution of less than 1.0 mm using 240 LYSO (Lu1.8Y0.2SiO5) scintillators with a fan-beam array. In the present study, the local void fraction was measured with the MARBLE facility under various air-water flow conditions (j(g) = 0.04 to 0.85 m/s and j(l) = 0.12 to 0.77 m/s) covering from bubbly to cap-turbulent flows. The local void fraction was also successfully measured under nonuniform and asymmetric air bubble distribution conditions with an investigation of the effect of spacer grids and mixing vanes on void drift across subchannels.

Heat pipe microreactors are reactor designs that primarily use liquid-metal heat pipes to cool the core. The main interest in heat pipes is the fact that they can remove heat passively. This, along with the use of liquid metal, allows the reactor to operate at higher temperatures. Although the use of heat pipes in nuclear reactors is new, liquid-metal heat pipe technology is mature. Nevertheless, experimental data on heat pipes are scarce, and very little is known about their behavior during abnormal operations and close to their thermal limits. Therefore, new experiments and accurate heat pipe simulations are needed to develop reliable closure models. This work describes a joint experimental and numerical investigation into heat pipes that attempts an initial closure of this gap. The numerical and experimental efforts are currently proceeding in parallel, aimed at different aspects of heat pipes. The numerical part is focused on gaps in local closures, and the experiments capture the overall heat pipe behavior.

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS).There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an in-containment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance.In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-con-densable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.(c) 2022 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. The hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature.

An experiment for steam condensation in the presence of a non-condensable gas was conducted in a nearly horizontal tube with an inner diameter of 40 mm and a downward slope of 3 degrees. The multi-dimensional distributions of the local air-steam mixture temperature and heat transfer coefficient were measured in the stratified and annular flow regimes. The influence of the non-condensable gas on the local heat transfer coefficient was investigated by varying the inlet steam mass flux, air mass fraction, and pressure. It was found that the film-wise condensation appearing in the upper region of the tube was significantly deteriorated by the air layer near the wall, but little effect was observed in the con-vective heat transfer by the condensate accumulated in the lower region. The degradation effect on the condensation increased further downstream due to the air accumulation near the cold wall. In addition, an empirical correlation for the degradation factor of the condensation heat transfer coefficient was de-veloped, and it showed a mean absolute error of 10.4% against the circumferential average heat transfer coefficient of the present experimental data. This correlation is applicable to separated flow covering inlet air mass fractions of 0.01-0.56, pressures of 1-5 bar, and inlet air-steam mixture Reynolds numbers of 7,50 0-166,0 0 0.(c) 2022 Elsevier Ltd. All rights reserved.

SMART (system-integrated modular advanced reactor), an integrated pressurized light water reactor developed by the Korea Atomic Energy Research Institute, adopted a new design concept called containment pressure and radioactivity suppression system (CPRSS) to cool containment buildings and reduce radioactive materials. In this study, assuming long-term cooling after 72 hours of small break loss-of-coolant-accident (SBLOCA), a vertical tube condensation heat transfer test with vapor and non-condensable gas was carried out under natural circulation flow path conditions. When a non-condensing gas is additionally injected into an experimental device where a certain amount of steam is being injected to maintain a steady state pressure, transient state characteristics in CPRSS were investigated.

Flow transient critical heat flux (CHF) in a narrow rectangular channel has not been studied adequately, and the same is currently predicted using steady-state CHF correlations. To address this gap in research, experiments were performed in this study to investigate the flow transient CHF in a narrow rectangular channel of a research reactor under downward flow. The experiments cover a wide range of flow conditions, including mass fluxes in the 950-4,700 kg/m(2)s range, inlet temperature of 37 degrees C, outlet pressure of 170 kPa, and transient time parameters in the 8.7-485 range. Evaluation of previous correlations applicable to flow transient CHF in a tube revealed they could not reasonably predict experimental data obtained in this study. Consequently, the present experimental data and Celata R-12 database were used to deduce a new correlation. The CHF ratio (the ratio of transient CHF to steady-state CHF under same mass flux) was observed to increase up to 1.6 for water and 2.7 for R-12 under the flow transient condition. As observed, the proposed correlation is applicable to both downward flow of water in a narrow rectangular channel and upward flow of R-12 in a tube. The correlation shows average and root-mean-square error values 0.04% and 3.46%, respectively, under a flow reduction rate of 0.1-2.5 s(-1). (C) 2020 Elsevier Ltd. All rights reserved.