Sodium-water Reaction Research Articles

Analysis of Sodium–Water Reaction as heat source for district heating and cooling

Finding new alternatives is a need to develop disruptive solutions that could be applied to the decarbonized, sustainable, circular systemic change for the energy system transition. In this case, a complete new approach, sodium as energy resource, is described. Traditionally, sodium has been considered a risky element even if it were proposed as coolant in many applications, as nuclear or solar thermal plants. Such applications have been concerned by the explosive reaction of alkali as Na with water. We analysed the alternative of profiting of such highly exothermic reaction for an energy use. Previously, we experimentally tested the utilization of sodium as propellant, as reported in Guerrero et al. (2019), what produced the correlation between the energy release and the sodium consumption in excess of water. We are now developing a boiler fuelled by sodium and water. Based on these previous and ongoing activities, this paper shows the proposal and the process analysis of a sodium–water boiler applied to heating and cooling networks,as one of the sectors to be decarbonized in the energy system. A tentative configuration for the design of such heater, and the heat exchangers to adapt heated water temperatures will be presented. We present the conceptual design of a 13.5 MW district heating plant for a ΔT=10°C heating water from 60 to 70 °C, consuming 1 kg/s Sodium, reaching an efficiency of 95%, comparable with existing boilers.

Fundamental evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR “Monju”. However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H2), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H2 bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

Impingement wastage experiment with SUS 316 in a printed circuit steam generator

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

Large leakage sodium-water reaction accident analysis in a sodium-cooled fast reactor with paralleling steam generators

In modern sodium-cooled fast reactor (SFR) designs, paralleling steam generators (SGs) are used to improve the economy. However, large leakage sodium-water reaction (SWR) models were developed only for designs using one SG. It is necessary to derive a pressure wave propagation model with parallel channels. Thus, a detailed pressure wave propagation model based on conversion equations was derived and verified with experimental data in this study. Using the developed model, a large leakage SWR accident in the SFR with four parallel SGs was simulated. The accident process was analyzed, and the design was verified to accommodate a large leakage SWR. Sensitivity analysis of critical parameters on large leakages was simulated and investigated. The SG pipe area, rupture disk bursting pressure, failure of the rupture disk, and release pipe area were found to have significant influences on the maximum pressure. Thus, the findings of this study are valuable for SG designs and protection from large leakage SWR accidents.

Bubble detection and identification based on the vibration response for the sodium water reaction

Real-time bubble behavior characterization and localization are critical in the chemical industry, especially for leakage of sodium-water reactions. This research investigates the properties of bubble signals under different gas injection rates and sodium-water reactions by employing a vibration-based approach. Specifically, the time-frequency and energy characteristics of bubble signals were studied by vibration response. The bubble signals excited by the two kinds of experiments have the same characteristic frequency band of 200–400 Hz, and the maximum energy is obtained at 250 Hz. The bubble signal from the sodium-water reaction can be effectively recognized by identifying the differences in the characteristics of multiple parameters when the gas injection rate is 0.07 m3/h with a signal-to-noise ratio of 11.8–17.1 dB. On the basis of energy analysis of the bubble signal, a bubble signal source location method based on time domain statistics was proposed, and the trajectory curve of the bubble signal excited by the sodium water reaction in the initial 600 ms was successfully obtained with a position error rate less than 5%. Comparative analysis indicates that kurtosis was more sensitive for bubble signal recognition, while the Root Mean Square (RMS) and Standard Deviation (SD) values were more suitable for bubble positioning. Thus, this study complemented the present techniques and knowledge for bubble identification and localization in multiphase flow engineering.

Design and Thermal-Hydraulic Evaluation of Integrated Steam Generator with Serpentine-Tube Configuration for Prevention of Sodium-Water Reaction

Although sodium-cooled fast reactors (SFRs) can have inherently enhanced safety features, there is a concern regarding unexpected events arising from a sodium-water reaction (SWR) in a Rankine cycle steam generator. Innovative concepts of integrated steam generators (ISGs), including an intermediate heat exchanger, have been suggested for eliminating an SWR by employing an intermediate fluid between liquid sodium and steam/water system. In this study, we propose an ISG designed with a novel tube configuration: a serpentine tube-type integrated steam generator (S-ISG). Furthermore, we developed a dedicated one-dimensional thermal-hydraulic analysis code for the specific design evaluation of the S-ISGs. The evaluation results are reported with reasonable explanations and compared with the results obtained using a conventional CFD tool to analyze local thermal-hydraulic characteristics that cannot be shown in the one-dimensional analyses. The proposed S-ISG has an inherent advantage for modularization as well as prevention of SWR; therefore, it is characterized by its ability to improve operational efficiency even in the case of repair, despite the low space efficiency. The results demonstrate that the component developed in this study can serve as an option to maintain cost efficiency while enhancing the public acceptability of an SFR.

Experimental study on interface characteristics of liquid sodium-water vapor interaction in SWR

When the sodium-water boundary in the sodium-cooled fast reactor (SFR) fails, water vapor will enter the liquid sodium and cause the sodium water reaction (SWR) accident. SWR may lead to many serious consequences for power plant. Thus, it is a non-negligible factor to be considered in the design and operation of SFR. In this study, experiments on the interaction between liquid sodium and water vapor were carried out. The reaction phenomena, temperature of liquid sodium, morphology of reaction product and other characteristics of the chemical were summarized and analyzed in this paper. The experiment was conducted under different initial sodium temperatures (300–450 °C) and different mass fraction of water vapor (0.1–0.9 at 200 °C) conditions. Visualization scheme was used in this experiment and the reaction process was recorded by a high-speed camera. The results indicate that both mass fraction of water vapor and initial temperature of liquid sodium are important factors in the reaction process. The mass fraction of water vapor affects the actual reaction rate. The initial temperature determines the form of reaction products (liquid NaOH) on the surface of liquid sodium. For the experiments with initial temperature below 400 °C, the reaction products appear as bubbles and foams. While the initial experiment temperature exceeds 400 °C (including 400 °C), bubbles and foams gone, replaced by a thin film. By analyzing the conditions of bubble generation, the surface tension of liquid NaOH is an important cause, which can be affected by the temperature and bounding water molecules. Furthermore, according to the measurement results of sodium temperature, the reaction products in the form of a large concentration of foams have a strong blocking effect on mass transfer and greatly slow down the actual reaction rate. This research has important reference significance for the study of the mechanism of liquid sodium and water vapor reaction.

Vibration method for the characterization of bubble behavior in sodium flows

Real-time bubble behavior characterization is critical in the chemical industry, especially for leakage in sodium–water reactions. In this study, the injection of gas into sodium flows was conducted to evaluate bubble characterization methods. Simulation research was performed to dynamically characterize bubble generation, growth and bursting in flowing sodium, and experimental research was conducted to explore the time-frequency and time domain statistical characteristics of bubble signals. After gas injection, significant low-amplitude signals were observed in the time domain, the power spectral density fluctuated in the 0–20 kHz band; its spectral components were more abundant above 8 kHz, and the short-time Fourier transform of the energy spectrum exhibited a nonlinear intermittent distribution in the 8–20 kHz band. Based on these differences, the bubble signal was effectively detected when the gas injection rate was 0.11 m3/h with a signal-to-noise ratio less than 0.5 dB. Thus, this study complemented the present techniques and knowledge in the field of chemical engineering.

An analysis of the pressure wave propagation by the sodium-water reaction in the printed circuit steam generator using the SWAAM-II code

The Sodium-Water Reaction (SWR) is the biggest problem for the development of the steam generator (SG) of the Sodium-cooled Fast Reactor (SFR). The pressure wave generation by the SWR has effects on the structural integrity of the Intermediate Heat Transport System (IHTS). However, when the Printed Circuit Steam Generator (PCSG) is used, the safety of SFR is expected to be improved. Since the PCSG is composed of a millimeter size of flow channels, the SWR zone is limited to the small channel. Therefore, it is necessary to evaluate the pressure spike and propagation in the PCSG. In this study, the SWAAM-II is used to analyze the pressure wave transfer by the SWR when the PCSG is used. The pressure in the sodium channel suddenly increases when the SWR occurs and the high pressure wave is transferred towards the headers. The peak pressure in the sodium channel is 16 MPa which is 21 % of the allowable stress based on the design by rule. However, the pressure wave is attenuated in the header due to the large difference in the cross-section area between the sodium channel and header. When it is assumed that the SWR occurs in the header, the pressure in the IHTS is increased. However, the increase of pressure in the IHTS is much less than the case of in the shell and tube type SG.

Numerical analysis of the Sodium–Water Reaction in a minichannel to evaluate the safety of a Printed Circuit Steam Generator

ABSTRACT A printed circuit heat exchanger is expected to improve the safety of a sodium-cooled Fast reactor when it is used as the steam generator. To use such a Printed Circuit Steam Generator (PCSG), the Sodium–Water Reaction (SWR) should be evaluated first. Because the diameter and length of a sodium channel in the PCSG are 4 mm and 2 m, respectively, a numerical approach is useful to understand the SWR characteristics in a minichannel. CFD modeling is conducted using commercial code and the SWR in a single sodium channel is analyzed under operating conditions. H2O at 16.8 MPa and 683 K is injected into the sodium channel through the cracks of different sizes (0.1–1.0 mm). Although the SWR is extreme in the early stage, most of the Na(l) is pushed towards the inlet and outlet of the channel within 15.3–35.3 ms as H2O inflowed. The peak temperature is around 1200–1600 K in the early stage of the SWR and the temperature reaches around 683 K. Moreover, the calculated depth of target wastage is around 0.62–3.34 μm. Therefore, the effect of the SWR is negligible in the PCSG core under operating conditions.

Behavior Estimation Focusing on the Existing Form of Hydrogen in Sodium in Sodium-Cooled Fast Reactors

In the cooling system of sodium-cooled fast reactors, it is required to identify and rapidly detect hydrogen explosively produced by the sodium-water reaction when a water leak occurs due to damage of a heat exchanger tube in a steam generator (SG), in contrast to low-concentration background hydrogen permeating through SG tubes during normal operation. In the present study, we focus on the difference between the background hydrogen and the hydrogen generated by the sodium-water reaction, and theoretically estimate the hydrogen behavior in liquid sodium by using computational chemistry methods. We find that dissolved H or NaH, rather than H2, is the predominant form of the background hydrogen in liquid sodium, and that hydrogen produced in large amounts by the sodium-water reaction can exist stably as fine bubbles with a NaH layer on their surface.

Validation of Liquid Droplet Entrainment and Transport Model in Sodium-water Reaction Analysis Code, SERAPHIM

In order to establish a safety evaluation method of a steam generator of sodium-cooled fast reactors, a computer program called SERAPHIM has been developed to calculate compressible multicomponent multiphase flow with sodium-water chemical reaction under tube failure accident. In this study, a numerical model for liquid droplet entrainment from an interface of the gaseous jet and its transport was validated through the analysis of experiment on vertical discharging of high-pressure air in a water pool. This model considers the mass, momentum, and energy conservation equations including the correlation of liquid droplet entrainment velocity. The numerical result showed the entrained liquid droplet phase was accelerated by the air jet. The calculated velocity of the liquid droplet falls within the dispersion of the measured data.

Study on sodium-water reaction jet evaluation model based on engineering approaches with particle method

Study on sodium-water reaction jet evaluation model based on engineering approaches with particle method

Conceptual design of a copper-bonded steam generator for SFR and the development of its thermal-hydraulic analyzing code

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

Large-leak sodium–water reaction accident analysis of sodium-cooled fast reactor

A sodium-cooled fast reactor (SFR) is one of Generation IV nuclear systems. The once-through steam generator is important for heat transfer and acts as a barrier to separate the sodium and water. If the heat transfer tube breaks, a sodium–water reaction (SWR) occurs, and the integrity of the secondary loop may be destroyed. Thus, a dynamic model to analyze the SWR must be developed, especially for a large-leak SWR. A large-leak SWR analysis model was developed for an SFR, consisting of four modules: water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and SWR protection system action. A large-leak SWR accident was simulated using this model, and the process was investigated. Sensitivity analysis was performed for the critical parameters, and the parameters with a key influence on the large-leak SWR were determined. The results are significant for the design of the steam generator and the SWR protection system.

Thermal-hydraulic design of natural circulation integrated steam generator using design map

A natural circulation integrated steam generator (NCISG) is a new steam generator concept that prevents the sodium-water reaction, which is a major safety issue in sodium heat transfer systems using the water Rankine cycle. To estimate and enhance the heat transfer performance of NCISG according to the design parameters, a design parameter evaluation method using the design map is suggested for the natural circulation system. The design map evaluates the effect of the temperature distribution of the heat transfer system in advance, allowing a simpler evaluation of the design parameters of a natural circulation system. The design map of NCISG is presented, and the effects of the tube diameter and pitch-to-diameter ratio (P/D) on the heat transfer area and natural circulation flow rate were evaluated. The heat transfer area is mostly affected by the tube diameter and the natural circulation flow rate is mostly affected by the P/D. The tendency of the total heat transfer area to be small in a specific natural circulation mass flow rate, regardless of the tube diameter or P/D, was identified and analyzed as a trend of the total UA value according to the natural circulation flow rate and optimal temperature. Finally, the optimal design of the NCISG under the evaluated design variable conditions is presented.

Performance and size comparison of two-stage and three-stage axial turbines for a nitrogen gas Brayton cycle coupled with a sodium-cooled fast reactor

A nitrogen (N2) Brayton cycle is considered as an alternative to the conventional steam Rankine cycle to overcome essential risk during pressure boundary rupture accidents with sodium-water reaction in a steam-generator of a sodium-cooled fast reactor (SFR). The objectives of this study are to estimate the turbine performances after proposing three-dimensional (3-D) blade designs and to suggest an option of detailed turbine design for N2 Brayton cycle-SFR applications. A free-vortex design method was applied in the 3-D design. Three candidates, including two- and three-stage designs, were selected from the Smith charts. Aerodynamic losses were investigated and compared through computational fluid dynamics analyses for the cases selected above. According to the results, rotor blade with milder curvature increased proportions of secondary and tip leakage losses among the total aerodynamic loss. When comprehensively considering the turbomachinery size and cycle thermal efficiency, the size of the two-stage turbine could be reduced to under 55% of the three-stage turbine size, resulting in no significant decrement of the cycle efficiency.

Thermal-hydraulic performance of a PCHE with sodium and sCO2 as working fluids

The Printed Circuit Heat Exchanger (PCHE) is one of the most attractive options for heat exchanger in compact facilities due to the high area concentration and superior heat transfer efficiency. A PCHE with sodium and supercritical carbon dioxide (sCO2) as working media is used as the intermediate heat exchangers in sodium-cooled fast reactor to reduce facility size and mitigate the consequences of sodium-water reaction by replacing steam with sCO2. The thermal-hydraulic performance of PCHE was analyzed by CFD with SST k-ω and Abid low Reynolds k-ε turbulence models for sCO2 and sodium domains, respectively. Besides, a segregated coupling strategy for the conjugate heat transfer was proposed based on the CORBA interface. The overall and localized heat transfer coefficient and friction factor were obtained and compared with empirical correlations. The mass flux and inlet temperature have significant impacts on the local thermal and hydraulic characteristics.

Computed Torque Control Method for Two-Axis Planar Serial Robotic Arm

The heat generated in reactor core is extracted using the sodium water heat exchanger system in fast breeder reactor. In the heat exchanger system there are 8 steam generators. Steam generator uses sodium and water to generate steam. Sodium–water reaction is exothermic in nature, so SG tubes need inspection to check healthiness of the tube. Due to space constraints, in-service inspection of SG needs remote-controlled robotic arm to locate SG tubes. The tube sheet is planar, so revolute joint two-link robotic arm is adapted. The movement of two-axis robotic arm needs control on position as well as velocity. The dynamic equation of two-axis robotic arm is coupled in nature which makes the control of end effector very difficult. The un-modeled phenomena, as that of friction and errors in mass distribution, will drastically affect the tracking of the end effector in the desired path. The Coriolis Effect is also accounted in the model. To precisely control the robotic arm end effector, the computed torque control technique is proposed and used in this paper.

Development of Unstructured-mesh Sodium-water Reaction Analysis Code, SERAPHIM

Development of Unstructured-mesh Sodium-water Reaction Analysis Code, SERAPHIM