Viscosity is a fundamental physical property that is important in any system in which fluid movement occurs. In addition, most of the elements exist as ions in molten state in high-temperature molten salt, and electrical conductivity in such molten state is closely related to viscosity as a transport property. Molten salt reactor (MSR) and pyroprocess are representative processes dealing with high-temperature molten salts, actinide elements, and other radioactive materials. In MSR and pyroprocesses, the viscosity data must be provided as one of the fundamental physical property data required for safe process operations and countermeasures to severe accidents. In order to measure the viscosity of highly corrosive molten salt at high temperatures, we have built a in-house developed molten salt viscosity measurement system based on the Brookfield rotationary viscometer. We also developed a special correction technique to improve the accuracy of the viscosity measurement. In this study, the viscosity was measured at 500°C for NaCl-MgCl2 molten salt, which is selected as the base salt material of MSR system under development in Korea Atomic Energy Research Institute (KAERI), using our viscosity measurement system installed in a oxygen- and moisture-free Ar-atmosphere glovebox. Our viscosity measurement system was calibrated using a LiCl-KCl eutectic mixture with well-known viscosity value, and viscosity values obtained using our own correction methodology were compared with those of other conventional correction methods. In our further study, we plan to measure the NaCl-MgCl2-UCl3 system at various compositions and temperatures.

In this study, molten salt experiments were performed using a multi-purpose molten salt experimental loop to evaluate the corrosion and thermodynamic properties of the molten salt. The multi-purpose molten salt experimental loop is made of 1-inch austenitic 316 stainless steel, and 1/4-inch austenitic 316 stainless steel tubes were welded on the surface of a 1-inch pipe to measure temperatures of molten salt. During the experiment, the molten salt leaked due to corrosion of the welded part of the 1/4-inch tube connected to the 1-inch pipe. Therefore, the cause of corrosion damage of the leaked welded part was analyzed. The effect of NaCl-MgCl2 salt selected as the molten salt on corrosion failure was considered. And based on the operation data of molten salt experiments, the time of occurrence of the issue was estimated. Lastly, the cause of corrosion failure was estimated by comparing and analyzing the pipe shape before and after failure using SEM-EDS.

Plasma Arc Melter (MSO) system has been developed for the treatment and the stabilization of various kinds of hazardous and radioactive waste into the readily disposable solidification products. Molten salt oxidation system has been developed for the for the treatment of halogen- and sulfurbearing hazardous and radioactive waste without emissions of PCDD/Fs and acid gases. However, PAM system has showed some difficulty in the off-gas treatment system due to the volatilization of radionuclides and toxic metals at extremely high-temperature plasma arc melter and the emissions of acid gases. MSO system has also showed the difficulty in the treatment of spent molten salt into the disposable waste form. Present study discussed the results of organics destruction performance tests for the PAM-MSO combination system, which is proposed and developed to compensate the drawbacks of each system. The worst-case condition tests for the organics destruction were conducted at lowest temperatures and the worst-case condition tests for the retention of metals and radionuclides were conducted at highested temperatures under the range of normal operating condition. For the worst-case organic destruction test, C6H5Cl was selected as a POHCs (Principal Organic Hazardous Constituents) because of its high incinerability ranking and the property of generation of chlorine gases and PCDD/Fs when incompletely destroyed. Simulated concrete waste spiked with 1 L of C6H5Cl was treated and the emissions of 17 kinds of PCDD/Fs and other hazardous gases such as CO, THCs, NOx, SO2 and HCl/Cl2 were measured. For the worst-case condition tests for the retention of metals and radionuclides, Pb and Cs were selected because of its high volatility characteristics. The emissions of PCDD/Fs was extremely lowered than the emission limit and those of other hazardous constituents were below their emission limit. The results of performance tests on the organics destruction suggested that tested PAM-MSO combination system could readily treat PCBs-bearing spent insulation liquid, spent ion-exchange resins used for the treatment of spent decontamination liquid in the decommission process and the concreted debris bearing hazardous organic coating materials. The decontamination factor of Cs and Co were 1.4×105, 1.4×105, respectively. The emisison of Pb was 0.562 ppm. These results suggested that tested PAM-MSO system treated low-level radioactive and pb-bearing mixed waste.

Present study investigated the waste form integrity of melted products generated from PAM-MSO system, which is proposed and developed to compensate the drawbacks of each system. The disposal suitability of the melting solidification products generated from the plasma arc melting treatment of pulverized cement debris spiked by Pb, Cd and Cs, as indicators of typical hazardous metals and radionuclides existed in the low-level mixed waste in the KHNPPs. The final waste form obtained by the test was evaluated for suitability for disposal. The compressive strength was 261.10 MPa, showing much higher values when compared to other waste form products. The compressive strength of both the sample after irradiation with 107 Gy radiation and that after long-term submersion test (90 days) satisfied the disposal criteria. As a result of the leaching test conducted according to the ANS 16.1 test method, it was confirmed that the leaching index satisfies the disposal criteria.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, Offgas system should be properly designed since the fission products in off-gas accelerates the corrosion in reactor structure materials and deteriorates the purity of liquid fuel. The design of off-gas system therefore requires the preliminary study of the behavior of evolved fission products in off-gas units and the development of off-gas model is crucial in developing such system. In this study, we corrected the off-gas illustrative model proposed by ORNL (Nuclear Engineering and Design, vol 385(15) 111529, 2021) by employing physically consistent concept of capture rate of fission product and holdup. For the application of the corrected off-gas model to Chloride-based 6 MW Molten Salt Reactor, major fission products were firstly determined from OpenMC based neutronics calculation and chain reaction related to the major fission products were defined. Based on these data, the holdup behavior of fission products in off-gas units (decay tank, caustic scrubber, Halide trap, H2O trap and charcoal bad) were investigated.

Based on the results of a review for various precipitation methods phosphorylation (phosphate precipitation) of metal chlorides considered as a proper treatment method for recovering of the fission products in a molten salt. In previous precipitation tests, the powder of lithium phosphate (Li3PO4) added into LiCl-KCl molten salt containing metal chlorides as a precipitation agent. The reaction of metal chlorides containing actinides and rare earths to recover with lithium phosphate in a molten salt known as solid-liquid reaction. The powder of lithium phosphate disperse in a molten salt by stirring thoroughly in order to enhance the precipitation reaction. As a result, metal phosphates as the reaction products precipitate on the bottom of the vessel and cutting at the lower part of the salt ingot considered as one of the recovery method of the precipitates. Recently, the vacuum distillation of upper part of the salt proposed as another recovering method. Cutting method of precipitate at the lower part of the salt ingot would be difficult to handle the increased size of the salt ingot produced from the practical scale equipment. In this presentation, a new method for collecting the precipitates of phosphorylation reaction into a small vessel is introduced with test results in a molten salt containing uranium and rare earths such as Nd, Ce, and La. As the first step of a series of test lithium phosphate ingot was prepared by melting the powder at a temperature 1,300°C, and the ingot put into LiCl-KCl molten salt at 500°C for more than three hours to examine the shape of ingot to be deformed or not. The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible.

Under the circumstance of energy transition policy of the previous government in which nuclear energy portion will be gradually reduced, some R&D study looking for alternatives other than Pyro- SFR recycling could be very valuable and timely suitable. New alternative study started to evaluate the possibility of it if there are some advantages in terms of waste burden in case that the spent fuel are appropriately treated and disposed of in a disposal site, instead of recycling of spent nuclear fuels (SNF). The alternative study separate the fission products (minor actinides and rare earths) from SNF in a molten salt medium. The molten salt coming from the alternative study is radioactive and heat generating because it contains the fission products chlorides. It is necessary to collect the fission products from the waste molten salt for minimization of the high-level waste volume and to generate a final waste form containing the fission products compatible to the disposal site. Based on the results of a review for various precipitation methods, phosphorylation (phosphate precipitation) of metal chlorides selected as a proper treatment method for recovering of the fission products in a molten salt. Phosphate precipitation has the potential for removing most of fission product elements from a molten salt arising from the treatment of spent nuclear fuel. The performance of phosphate precipitation method evaluated using a salt mixture with the actinide and rare earth chlorides. The molten salt containing uranium as surrogate of the actinides and three rare earths (Nd, Ce, La) chloride was used for testing a phosphate precipitation method at experimental condition (temperature 500°C, salt stirring 200~300 rpm, and 1~1.2 eq. of phosphorylation agent). A cyclic voltammetry (CV) method monitored in-situ phosphate precipitation progress for determining the precipitation rate and conversion ratio evaluated. The phosphorylation reaction increased greatly at a salt stirring 300 rpm.

During electrorefining, fission products, such as Sr and Cs, accumulate in a eutectic LiCl-KCl molten salt and degrade the efficiency of the separation process by generating high heat and decreasing uranium capture. Thus, the removal of the fission products from the molten salt bath is essential for reusing the bath, thereby reducing the additional nuclear waste. While many studies focus on techniques for selective separation of fission products, there are few studies on processing monitoring of those techniques. In-situ monitoring can be used to evaluate separation techniques and determine the integrity of the bath. In this study, laser-induced breakdown spectroscopy (LIBS) was selected as the monitoring technique to measure concentrations of Sr and Cs in 550°C LiCl-KCl molten salt. A laser spectroscopic setup for analyzing high-temperature molten salts in an inert atmosphere was established by coupling an optical path with a glove box. An air blower was installed between the sample and lenses to avoid liquid splashes on surrounding optical products caused by laser-liquid interaction. Before LIBS measurements, experimental parameters such as laser pulse energy, delay time, and gate width were optimized for each element to get the highest signal-to-noise ratio of characteristic elemental peaks. LIBS spectra were recorded with the optimized conditions from LiCl-KCl samples, including individual elements in a wide concentration range. Then, the limit of detections (LODs) for Sr and Cs were calculated using calibration curves, which have high linearity with low errors. In addition to the univariate analysis, partial least-squares regression (PLSR) was employed on the data plots to obtain calibration models for better quantitative analysis. The developed models show high performances with the regression coefficient R2 close to one and root-mean-square error close to zero. After the individual element analysis, the same process was performed on samples where Sr and Cs were dissolved in molten salt simultaneously. The results also show low-ppm LODs and an excellent fitted regression model. This study illustrates the feasibility of applying LIBS to process monitoring in pyroprocessing to minimize nuclear waste. Furthermore, this high-sensitive spectroscopic system is expected to be used for coolant monitoring in advanced reactors such as molten salt reactors.

One of the promising candidates for heat transfer fluid is molten chloride salts. They have been studied in various fields such as the electrolyte of pyroprocessing, the molten salt reactor coolant, and the energy storage system media. Main considerations for utilizing molten chloride salts are the compatibility of salts with structural materials. The corrosion behavior of structural materials in molten chloride salts must be understood to identify suitable materials against the corrosive environment. In this study, the corrosion behavior of a candidate structural material, Hastelloy N, in molten LiCl- KCl salt at 500°C were investigated by the electrochemical impedance spectroscopy (EIS) method. The sheet type of Hastelloy N was utilized as the working electrode in LiCl-KCl to measure the EIS data for 100 hours with 5 hours of time intervals. The EIS data were measured in the frequency range from 104 Hz to 10-2 Hz with the AC signal (amplitude = 20 mV) at open circuit potential. The capacitance semicircle observed in Nyquist plots for all periods indicates that charge-transfer controlled reactions occur. As the immersion time progresses, the radius of the semicircle in Nyquist plots and the impedance and phase angle in Bode plots decrease. These behaviors suggest a decreasing reaction resistance and the corrosion reactions are accelerated with the immersion time. The EIS data were fitted using the equivalent circuit to achieve quantitative results. Two capacitor-resistor components were considered due to the overlapped shape of two valleys in phase angle. The depressed shape of the semicircle in Nyquist plots led to the use of the constant phase element(Q) instead of the capacitor. Therefore, R(Q(R(QR))) circuit was selected to fit the EIS data. Fitting results show that the charge transfer resistance decreases dramatically within 1 day and then converges. The film resistance shows no clear trends, but the increase of the film admittance value indicates the decreased film thickness. Consequently, the film appears to exist like the oxide layer but it does not act as a protective layer. The real-time EIS data were measured in molten salt and provides the corrosion behavior over time. The corrosion mitigation strategy should consider that the corrosion of Hastelloy N accelerates over time and its intrinsic film cannot act as the protective layer. The next steps of this study are to evaluate other candidate structural materials and to demonstrate the presence of the film.

This work provided a review of three techniques—(1) spectrochemical, (2) electrochemical, and (3) spectroelectrochemical– for molten salt medias. A spectroelectrochemical system was designed by utilizing this information. Here, we designed a spectroelectrochemical cell (SEC) and calibrated temperature controllers, and performed initial tests to explore the system’s capability limit. There were several issues and a redesign of the cell was accomplished. The modification of the design allowed us to assemble, align the system with the light sources, and successfully transferred the setup inside a controlled environment. A preliminary run was executed to obtain transmission and absorption background of NaCl-CaCl2 salt at 600°C. It shows that the quartz cuvette has high transmittance effects across all wavelengths and there were lower transmittance effects at the lower wavelength in the molten salt media. Despite a successful initial run, the quartz vessel was mated to the inner cavity of the SEC body. Moreover, there was shearing in the patch cord which resulted in damage to the fiber optic cable, deterioration of the SEC, corrosion in the connection of the cell body, and fiber optic damage. The next generation of the SEC should attach a high temperature fiber optic patch cords without introducing internal mechanical stress to the patch cord body. In addition, MACOR should be used as the cell body materials to prevent corrosion of the surface and avoid the mating issue and a use of an adapter from a manufacturer that combines the free beam to a fiber optic cable should be incorporated in the future design.

Electroanalytical study for the rotating cylinder electrode in molten LiCl-KCl eutectic salt (58– 42mol%) containing MgCl2 (0.1wt%) at 600°C is conducted. The researches of rotating cylinder electrode have been widely conducted for the century. The advantage of the electrode is that it can mitigate the unintended natural convection by providing a controlled diffusion boundary layer thickness. However, the experimental data for the high temperature molten salts is barely existed. The study adopts the electrochemical techniques such as cyclic voltammetry for the static cell and linear sweep voltammetry for the dynamic cell to calculate the diffusion coefficient. The peak current density and limiting current density are measured according to the scan rate. In order to evaluate the mass transfer under hydrodynamic flow condition, the revolution speeds of cylindrical electrode are varied from 10 rpm to 500 rpm which are corresponded to the Reynolds number of 4 and 185 respectively. The flow regime covers from the laminar to semi-turbulent regime (transient) as the critical Reynolds number Recrit is 200. The limiting current density shows a linear trend with the revolution speed and agrees well with the existing mass transfer correlations. For the extended flow regime, a new mass transfer correlation is suggested as the relation of non-dimensional numbers (Sh = aRebScc) based on the dimensionless analysis.

At high temperatures, molten salt has heat transfer properties like water. Molten salt has the characteristics of a strong natural circulation tendency, large heat capacity, and low thermal conductivity. Unlike sodium, molten salt does not react explosively exothermically with air. However, molten salt has a strong tendency to corrode materials, and its properties are easily changed by a sensitive reaction to oxygen and moisture. Therefore, it is necessary to study material corrosion properties and chemical control methods for nuclear fuel salts, which are eutectic mixtures. In this study, the optimal operation method of the thermal convection loop is established to perform the experiments on the molten salt. The process describes briefly as follows. The operation step consists of preparation, purification, transportation, and operation. In the preparation, the step checks the entire structure and equipment (TC, blower, vacuum pump, etc.). And melt the salt mixture at a high temperature (670°C) slowly in the purification step. Before injecting the molten salt, the surface temperature of the entire loop must retain temperature (about 500°C) constantly. Completely melted molten salt in the melting pot is flow along the pipe of the thermal convection loop in the transportation step. Lastly, the convection of molten salt goes to keep by the temperature difference. The thermal convection loop can be utilized for various experiments such as corrosion tests, component analyses, chemistry control, etc.

Molten salt used in the multipurpose molten salt experiment must be of high purity. Depending on the purpose of the experiment, only the base component of the molten salt be used, or a component simulating a nuclear fission product be added to the base component and used. In all cases, an increase in the concentration of impurities such as oxygen and moisture may lead to an erroneous interpretation when analyzing the experimental results. Therefore, molten salt should be purified before use. In this study, the purification of molten salt is described for multi-purpose molten salt experiments. The salt mixture is selected as MgCl2-NaCl and is quantified at a mixing ratio of 43mol%:57mol%. The salt mixture is treated in a glove box environment because of must minimize the reaction of adsorbing oxygen and moisture when the salt mixture is exposed to the atmosphere. MgCl2 is more likely to contain water than NaCl, the purification of the NaCl-MgCl2 mixture is established according to the purification process for removing water from MgCl2. A process for purifying the salt mixture briefly consists as follows: drying moisture, melting salts, purification, removing HCl, and stabilization. Through the process be able to obtain high-purity molten salt and more accurate experiment results.

A molten salt reactor (MSR) has considerably attracted attention due to its several advantages for the safety and efficiency over the light water reactors. Because the structural material in MSR is contacted with high-temperature liquid fuel during long-term, the excellent material for corrosion resistance is required to be applied in MSR. In this study, we evaluated the corrosion resistance for alloy 600 and 617, which are the nickel-based materials, in KCl molten salt at 800ºC for 100 h under Ar atmosphere containing less than 1 ppm of moisture and oxygen. After the corrosion experiments of alloy 600 and 617, the amount of the weight loss for them caused by the KCl molten salt were determined. In addition, the variation in the crystal structure, surface morphology, and elemental distribution was examined using X-ray diffraction and scanning electron microscopy equipped with energy dispersive X-ray spectroscopy.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

Molten Salt Reactor (MSR) is one of the generation-IV advanced nuclear reactors in which hightemperature molten salt mixture is used as the primary coolant, or even the fuel itself unlike most nuclear reactors that adopt solid fuels. The MSR has received a great attention because of its excellent thermal efficiency, high power density, and structural simplicity. In particular, since the MSR uses molten salts with boiling points higher than the exit temperature of the reactor core, there is no severe accident such as a core melt-down which leads to a hydrogen explosion. In addition, it is possible to remove the residual heat through a completely passive way and when the fuel salt leaks to the outside, it solidifies at room-temperature without releasing radioactive fission products such as cesium, which make the MSR inherently safe. Both fluoride and chloride mixtures can be used as liquid fuel salts by adding actinide halides for MSRs. However, the MSRs using chloride-based salt fuels can be operated for a long time without adding nuclear fuel or online reprocessing because the actinide solubility in chloride salts is about six times higher than that in fluoride salts. Therefore, the chloride-based MSRs are more effective for the transmutation of long-lived radionuclides such as transuranic elements than the fluoride-based MSRs, which is beneficial to resolve the high radioactive spent nuclear fuel generated from light water reactors (LWRs). This paper examines liquid fuel fabrication using an improved U chlorination process for the chloride-based MSRs and presents the strategy for the management of gaseous fission products generated during the operation of MSR.

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl2 and CrCl3, other metal chloride species such as FeCl2, FeCl3, MoCl2, MnCl2 and NiCl2 will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of < 1 ppm per day. During the incubation period, metal concentration show 1st order kinetics and increasing linearly with time1/2. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi3+ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl3 in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi3+ on inert W electrode and the shift of cathodic peak potentials of Li+ and Bi3+ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.

High-performance carbon materials were prepared via a one-step molten salt carbonization of tobacco waste used as electrode materials for supercapacitors. Carbon material prepared by carbonization for 3 h in molten CaCl2 at 850 °C exhibits hierarchically porous structure and ideal capacitive behavior. In a three-electrode configuration with 1 mol L− 1 H2SO4 aqueous solution, it delivers specific capacitance of 196.5 F g− 1 at 0.2 A g− 1, energy density of 27.2 Wh kg− 1 at 0.2 A g− 1, power density of 983.5 W kg− 1 at 2 A g− 1, and excellent cyclic stability with 94% capacitance retention after 5000 charge–discharge cycles at 1 A g− 1. Moreover, in a symmetrical two-electrode configuration with 6 mol L− 1 KOH aqueous solution, it delivers specific capacitance of 111.1 F g− 1 at 0.2 A g− 1, energy density of 3.8 Wh kg− 1 at 0.2 A g− 1, and power density of 482.0 W kg− 1 at 2 A g− 1. The relationship between hierarchically porous structure and capacitive performance is also discussed.