Sodium-water Reaction Research Articles

Rapid Heating Tube Rupture Simulation Experiments in case of Sodium-water Reaction in Steam Generator of Sodium-cooled Fast Reactor (2<sup>nd</sup> report)

Rapid Heating Tube Rupture Simulation Experiments in case of Sodium-water Reaction in Steam Generator of Sodium-cooled Fast Reactor (2<sup>nd</sup> report)

Application of unstructured mesh-based numerical method to sodium-water reaction phenomenon analysis code SERAPHIM

Application of unstructured mesh-based numerical method to sodium-water reaction phenomenon analysis code SERAPHIM

Investigation of the threshold temperatures of sodium-carbon dioxide reaction for SFR system design

In a sodium-cooled fast reactor, a sodium-water reaction (SWR) could be occurred at pressure boundaries between the two separated systems, the steam Rankine cycle and the sodium coolant system, which results in a serious problem in the system. Since it has been a design challenge for researchers, as an alternative system, the supercritical CO2 (S-CO2) Brayton cycle was suggested. To utilize the S-CO2 Brayton cycle to a SFR, a series of studies on the coupled system have been conducted at Korea Atomic Energy Research Institute (KAERI). The studies were mainly focused on the pressure boundary failure events and their consequences including Na-CO2 reactions. Prior to the experimental studies, the authors classified the reaction behaviors of Na-CO2 reaction based on the possible scenarios that could take place at pressure boundaries. Through experimental studies, it was found that the reaction follows typical chemical formulas and it is essentially dependent on the reaction temperature. More importantly, the reaction rate changed drastically at the specific temperatures and the reaction became flammable exceeding a certain temperature. To specify the design conditions of the SFR system coupled with the S-CO2 Brayton cycle, by means of a dedicated set-up the self-ignition temperature has been evaluated. Exceeding this threshold temperature a significant increase in the reaction speed is observed. We setup an experimental apparatus and a test matrix based on one of possible scenarios where the flammable Na-CO2 reaction takes place and we installed a visualized chemical reactor to watch reaction phenomenon around the self-ignition temperature. The experimental apparatus was designed to be easy to control heaters, to regulate gas supply, and to measure reaction temperature. Through the experiment, it was found that the flammable temperature of Na-CO2 reaction lies between 598.3˚C and 599.9˚C and how liquid sodium reacts with gaseous carbon dioxide at these temperatures. The results will be help to set design criteria for the pressure boundary and it will take long time but an appropriate safety system would be introduced.

Development of unstructured mesh-based numerical method for sodium–water reaction phenomenon in steam generators of sodium-cooled fast reactors

ABSTRACTWhen pressurized water or vapor leaks from a failed heat transfer tube in a steam generator of sodium-cooled fast reactors, a high-velocity and high-temperature jet with sodium–water chemical reaction may cause wastage on the adjacent tubes. For safety assessment of the steam generator, a computational fluid dynamics code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium–water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method for the SERAPHIM code was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based SERAPHIM code. The calculated pressure profile showed good agreement with the experimental data. To investigate the effect of the introduction of the unstructured mesh and to confirm applicability of the numerical method for the actual situation, water vapor discharging into liquid sodium was analyzed. The calculated behavior of the reacting jet agreed with the previous experimental knowledge. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium–water reaction phenomenon.

Thermodynamic analysis and preliminary design of closed Brayton cycle using nitrogen as working fluid and coupled to small modular Sodium-cooled fast reactor (SM-SFR)

Sodium-cooled fast reactor (SFR) is considered the most promising of the Generation IV reactors for their near-term demonstration of power generation. Small modular SFRs (SM-SFRs) have less investment risk, can be deployed more quickly, are easier to operate and are more flexible in comparison to large nuclear reactor. Currently, SFRs use the proven Rankine steam cycle as the power conversion system. However, a key challenge is to prevent dangerous sodium-water reaction that could happen in SFR coupled to steam cycle. Nitrogen gas is inert and does not react with sodium. Hence, intercooled closed Brayton cycle (CBC) using nitrogen as working fluid and with a single shaft configuration has been one common power conversion system option for possible near-term demonstration of SFR. In this work, a new two shaft nitrogen CBC with parallel turbines was proposed to further simplify the design of the turbomachinery and reduce turbomachinery size without compromising the cycle efficiency. Furthermore, thermodynamic performance analysis and preliminary design of components were carried out in comparison with a reference single shaft nitrogen cycle. Mathematical models in Matlab were developed for steady state thermodynamic analysis of the cycles and for preliminary design of the heat exchangers, turbines and compressors. Studies were performed to investigate the impact of the recuperator minimum terminal temperature difference (TTD) on the overall cycle efficiency and recuperator size. The effect of turbomachinery efficiencies on the overall cycle efficiency was examined. The results showed that the cycle efficiency of the proposed configuration was comparable to the 39.44% efficiency of the reference cycle. In addition, the study indicated that the new configuration has the potential to simplify the design of turbomachinery, reduce the size of turbomachinery and provide opportunity for improving the efficiency of the turbomachinery. The findings so far revealed that the proposed two-shaft CBC with nitrogen as working fluid could be a promising power conversion system for SM-SFRs near-term demonstration.

ナトリウム－水反応現象解析コードSERAPHIMに対する非構造格子用解析手法の適用

For safety assessment of a steam generator in sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for a complex-shaped domain including multiple heat transfer tubes. The multiphase flow under the tube failure accident is calculated by the multi-fluid model considering compressibility. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under-expanded jet experiment, which is a key phenomenon in the tube failure accident. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. The calculated temperature field agreed with the existing experimental knowledge. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.

高速炉蒸気発生器伝熱管周囲に形成されるナトリウム－水反応環境を考慮したターゲットウェステージ評価（第三報）

高速炉蒸気発生器伝熱管周囲に形成されるナトリウム－水反応環境を考慮したターゲットウェステージ評価（第三報）

Reaction path and product analysis of sodium-water chemical reactions using laser diagnostics

In a sodium-cooled fast reactor (SFR), liquid sodium is used as a coolant because of its excellent heat transport capability and a large safety margin to the boiling point (1153K) at the atmospheric pressure. On the other hand, it has strong chemical reactivity with water vapor. One of the design basis accidents of the SFR is the water leakage into the liquid sodium flow caused by a breach of heat transfer tubes. Therefore, the study on sodium-water chemical reactions is of importance for safety of SFR. The purpose of this study aims to clarify the gas phase sodium-water reaction path and reaction products quantitatively. The counter-flow diffusion experiment device was employed to analyze the reaction path and reaction products using laser diagnostics. sodium (Na), sodium dimer (Na2), water (H2O), sodium hydroxide (NaOH), sodium oxides (Na2O and NaO) and hydroxyl radical (OH) were measured using Raman spectroscopy, absorption spectroscopy, photo-fragmentation spectroscopy and laser induced fluorescence in the reaction field for the sodium-water vapor, sodium-oxygen and sodium-hydrogen reaction. Based on the measurement results by absorption spectroscopy, it was revealed that the sodium-oxygen reaction was faster than that of sodium-water. It was indicated that sodium-hydrogen reaction did not react noticeably with sodium from the measurement result as well. The main product of sodium-water reaction was determined to be NaOH and the sodium oxide was not notably measured compared with NaOH from the results by photo-fragmentation spectroscopy. The sodium oxide concentration ratio to NaOH was determined 2% with standard deviation of 3%. Its reaction path was also discussed using sodium-water, sodium-oxygen and sodium-hydrogen reaction analyses.

전산유체역학을 이용한 소듐-소듐 열교환기 설계코드의 검증

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the <TEX>$4^{th}$</TEX> generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Abstract The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Impingement wastage experiments with 9Cr 1Mo steel

Steam Generator (SG) is one of the vital components of sodium cooled fast reactor (SFR). The main safety concern with SG is a probable sodium–water reaction. In case, one of its water/steam carrying tubes leaks, water/steam gets into contact with sodium causing sodium-water reaction, which is highly exothermic and producing corrosive NaOH and hydrogen. The ejecting reaction products at high temperature, impinges upon adjacent tubes by a process called impingement wastage. It could damage one of the neighboring tubes in a short time, if the detection and protection systems are failing. IGCAR and CEA carried out a collaborative study on impingement wastage of 9Cr 1Mo steel, which is one of the candidate materials for SFR SG tubes. The studies comprise of experimental works at IGCAR and simulation works with PROPANA code at CEA. This paper brings out the data and experience gained through this cooperative work.

Experimental study and kinetic analysis on sodium oxide–silica reaction

ABSTRACTIn a sodium-cooled fast reactor (SFR), if considering hypothetical severe accidental condition such as the steel liner failure of structural concrete caused by intensive leakage of liquid sodium (Na) coolant, the liquid sodium–concrete reaction (SCR) may take place. The major consequences of SCR are hydrogen release, energy release and concrete ablation. Thus, it is important to understand the phenomenology and kinetic behavior of SCR. As a part of a series of studies on SCR, this study focused on the reaction between sodium oxide (Na2O) and silica (SiO2), which is dealt with possible formation of Na2O as the result of sodium–water reaction in the initial stage of SCR. Through thermoanalytical and X-ray diffraction measurements, it was revealed that Na2O–SiO2 reaction to form sodium orthosilicate (Na4SiO4) occurs at significantly lower temperature in comparison with Na–SiO2 reaction. The reaction is kinetically characterized as a largely overlapping multistep reaction, which is composed of at least three reaction steps. On the basis of the observations, the impact of Na2O–SiO2 reaction in the overall SCR and the significance of the conventional kinetic analysis using the Kissinger method are discussed.

Numerical approach for quantification of selfwastage phenomena in sodium-cooled fast reactor

Abstract Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium–water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called “self-wastage phenomena.” A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium–water chemical reaction: s odiumwat er r eaction a nalysis ph ysics of i nterdisciplinary m ultiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

Interparticle potential of 10 nanometer titanium nanoparticles in liquid sodium: theoretical approach

Abstract A suspension of titanium nanoparticles (Ti NPs) in liquid sodium (Na) has been proposed as a method to mitigate the violent sodium–water reaction (SWR). The interparticle potential between Ti NPs in liquid Na may play a significant role in the agglomeration of NPs on the reaction surface and in the bulk liquid Na, since the potential contributes to a reduction in the long-term dispersion stability. For the effective control of the SWR with NPs, a physical understanding of the molecular dynamics of NPs in liquid Na is key. Therefore in this study, the nonretarded Van der Waals model and the solvation potential model are employed to analyze the interparticle potential. The ab initio calculation reveals that a strong repulsive force driven by the solvation potential exceeds the interparticle attraction and predicts the agglomeration energy required for two 10-nm Ti NPs to be 4 × 10−17 J. The collision theory suggests that Ti NPs can be effective suppressors of the SWR due to the high energy barrier that prevents significant agglomeration of Ti NPs in quiescent liquid Na.

Reaction rate of Na-based titanium nanofluid with water

Suppression of chemical reactivity in reaction between sodium (Na) and water by Na-based titanium nanofluid (NaTiNF) was verified by measuring the difference between the H2 generation rates of NaTiNF–water reaction and sodium–water reaction (SWR). H2 generation at the beginning of the reaction was slower in NaTiNF–water reaction than in SWR. To determine the effect of temperature on the reaction rate, NaTiNF–water reactions were conducted at 104 °C, 120 °C and 150 °C. At 104 °C, the reaction rate of SWR was twice as fast as NaTiNF–water reaction. This result demonstrates that the chemical reactivity of liquid Na is suppressed by the presence of Ti nanoparticles in liquid Na. The reaction rate of NaTiNF–water reaction increased with reaction temperature. The activation energy of NaTiNF–water reaction was obtained using the Arrhenius equation.

Numerical Analysis of Self-wastage Phenomena Caused by Sodium-Water Reaction in Sodium-Cooled Fast Reactor through Simulant Experiment

Numerical Analysis of Self-wastage Phenomena Caused by Sodium-Water Reaction in Sodium-Cooled Fast Reactor through Simulant Experiment

Water experiment on phased array acoustic leak detection system for sodium-heated steam generator

A phased array acoustic leak detection system for sodium heated steam generator has been proposed. The major advantage of the new system is it could provide information of acoustic source direction. An acoustic source of a sodium–water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the new system could separate the target leak source from steam generator background noise. In the previous study, the methodology was proposed and basic performance was confirmed by numerical analysis. However, in the numerical analysis, acoustic transportation through the SG tube bundle was not modeled. In the present study, performance the proposed system has been confirmed in water experiments with mockup tube bundles.

Experimental Study of the Tritium Distribution in the Effluents Resulting from the Sodium Hydrolysis

Within the framework of the dismantling of fast breeder reactors in France (PHENIX, SUPERPHENIX, RAPSODIE), several processes are under investigation regarding sodium disposal. One of them, called ELA (radioactive sodium waste treatment process), is based on the implementation of the sodium-water reaction, in a controlled and progressive way, to remove residual sodium mainly from the sodium purification systems called cold traps. This sodium contains impurities such as sodium hydride, sodium oxide and tritiated sodium hydride. The hydrolysis of these various chemical species leads to the production of a liquid effluent, mainly composed of an aqueous solution of sodium hydroxide, and a gaseous effluent, mainly composed of nitrogen (inert gas), hydrogen and steam. The tritium is distributed between these effluents, and, within the gaseous effluent, according to its forms HT and HTO. HTO being 10,000 times more radiotoxic than HT, a precise knowledge of the mechanisms governing the phase distribution of tritium is necessary. Indeed, it will help to design the process needed to optimize the treatment of the off-gas before its release into the environment. This paper presents the first experimental results from a parametric study on the tritium distribution between the various effluents generated during hydrolysis operations. This parametric study has been performed in a laboratory scale hydrolysis process designed at the CEA Cadarache.

Experimental and numerical reaction analysis on sodium-water chemical reaction field

Experimental and numerical reaction analysis on sodium-water chemical reaction field