Small Reactor Research Articles

Estimation of Near-Field and Far-Field Post-Accident Atmospheric Dispersion for Microreactors

The current study describes a simulation-based analysis of the atmospheric dispersion of radionuclide fission product particles in the near-field and far-field of a generic, conceptual microreactor, which is a small nuclear reactor with a power output typically ranging from 1 to 20 MW(thermal) and generally lower than 50 MW(electric). The near-field is a distance of up to 100 m from the microreactor while the far-field is a distance of 300 m or beyond from the microreactor. The generic microreactor operates at a pressure close to the ambient pressure. Therefore, in the event of a postulated accident that causes the leakage of radionuclide particles from the microreactor containment into the environment, the radionuclide particles are unlikely to travel too far from the reactor, as opposed to conventional nuclear reactors. The current paper provides estimates of average and 95th-percentile values of the normalized effluent concentration of the atmospheric radionuclide particle dispersion with respect to the source strength in the near-field and far-field of the conceptual microreactor. The computer code Atmospheric Relative CONcentrations in Building Wakes (ARCON96) was used to perform all simulations for the current study. It was observed that the 95th-percentile values of the normalized effluent concentration decrease by an order of magnitude as the receptor distance increases, i.e., from the near-field to the far-field. The dispersed aerosol concentration also decreases with time. A parametric study was performed to understand which input parameters affect the normalized effluent concentration values the most, and a definitive screening design was employed for this purpose. The atmospheric stability class and the distance between the reactor and the receptor were the parameters found to affect the aerosol dispersion characteristics by the greatest extent. The study recommends that the computer code RADTRAD (Radionuclide Transport and Removal And Dose Estimation) be used to estimate the actual dosage over distance using the outputs from ARCON96 as inputs, along with reactor-specific core inventories.

Dissolution of wood components during hot water extraction of spruce

The purpose of this study was to investigate the autohydrolysis of softwood, which is the main chemical operation in both hot water extraction and steam explosion. Control of the process and monitoring its course were ensured by the careful choice of experimental setup and conditions: a milled spruce material was extracted in a small flow-through reactor to minimize degradation of the dissolved material and to enable analysis of the resulting liquors extracted at selected time points. The obtained liquid and solid fractions were analyzed for sugar composition and acetic acid concentration. The results showed that partially degraded hemicelluloses were extracted; hemicelluloses side chains were cleaved off and detected as monomers, while deacetylation was limited. Chain scissions of cellulose were observed as a result of autohydrolysis.

Transient characteristics analysis of residual heat removal system for Helium-Xenon mixture cooled small reactor system

Helium-Xenon (He-Xe) gas mixtures are characterized by stable physical properties, easy compression, and excellent thermodynamic performance, making them suitable for thermal cycle coolant for underwater reactors. Due to the unattended deep sea, and harsh and complex operating environment, the residual heat removal characteristics are directly related to reactors’ safety characteristics and stable and reliable operation ability. Therefore, studying the He-Xe gas mixture residual heat removal characteristics can provide an essential guarantee for reactor safety analysis. This article optimizes the physical property calculation model of the He-Xe gas mixture in the Brayton Cycle Reactor System Analysis Program (BRESA) and obtains a system property analysis program suitable for calculating the natural circulation of the He-Xe gas mixture. And a natural circulation system for the He-Xe mixed gas cooling reactor was constructed using the modified BRESA. Numerical calculations and theoretical analysis were conducted on the He-Xe gas mixture’s natural circulation flow process and residual heat removal characteristics. The study of the transient factors of temperature, flow rate, and other related parameters in the system circuit found that when the reactor is shut down, relying on the natural circulation system can effectively output the residual heat of the reactor core to ensure reactor safety. The relevant research results have important theoretical and scientific significance for the safety assessment and evaluation of the Brayton cycle reactor system.

Dow nuclear reactor set for Texas

Dow has selected its petrochemical complex in Seadrift, Texas, as the site to install a small modular nuclear reactor system designed by X-Energy Reactor. The reactor will supply electricity and steam for the complex and will save about 440,000 metric tons of carbon dioxide emissions per year. The companies are now preparing permits for the Nuclear Regulatory Commission. They expect to begin construction in 2026 and complete the project by the end of the decade.

Evaluating the performance of Indonesia’s nuclear energy program using INPRO methodology

Abstract A project was initiated to assess the sustainability of Indonesia’s planned nuclear energy system using the IAEA INPRO Methodology to develop an awareness of sustainability issues to support nuclear energy development strategic planning and decision making. Accordingly, for a nuclear energy system to be sustainable, fulfilment of current needs should not compromise the ability of future generations to meet theirs. The sustainability of the planned nuclear energy system was assessed based on the basic principles, user requirements and criteria in the areas of economics, infrastructure, waste management, proliferation resistance, physical protection, environment, and safety. The assessments covers the sustainability of a large reactor completed in 2014 to support the Pre-Feasibility Study in Bangka Belitung, and of a small and medium-sized reactor (SMR) currently ongoing to support the study in West Kalimantan. The results indicate that evidence of achieving sustainability exists in many respects, but there remain gaps to be addressed in due time on aspects such as economics, investment climate, availability of technology, and global development on non-renewable resources. A preliminary study on fuel cycle arrangements is being performed in parallel, focussing on collecting data including on fuel cycle services abroad. Collecting data has been the main challenge for a newcomer country that has yet to decide on the exact reactor technology. The paper is intended to evaluate the performance of Indonesia’s nuclear energy program using the INPRO Methodology to address the sustainability of its planned NES, i.e., reactors and fuel cycle facilities, and to evaluate the effectiveness of the efforts to achieve the purpose. Awareness of sustainability issues plays a key role in nuclear energy policy, especially for determining the energy mix by 2040, to achieve the Net Zero Emission policy target by 2060 or earlier.

Ensuring safety of new, advanced small modular reactors for fundamental safety and with an optimal main heat transport systems configuration

Abstract Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.

SMR, 3D source term simulation for exact shielding design based on genetic algorithm

One of the most important issue in the safe design of a small modular nuclear power reactor (SMR) is utilizing an efficient, compact, and lightweight radiation shielding capable of meeting the principal requirements of radiation protection. The radiation shielding design is a complicated multi-objective problem considering the current limitations. In this study, the three-dimensional and accurate modeling of the source term considering the real conditions of the reactor core has been performed and uses the genetic algorithm (GA) to improve the reactor radiation protection shielding system of SMR power reactors. The method upgraded in this research aims to reach an efficient design for the radiation shielding of the SMR power reactors regarding the present requirements and limitations. The thickness of different layers against the mixed-radiation fields of neutrons and gamma emitted from a nuclear reactor has been optimized to obtain a compact and lightweight shield design. The genetic algorithm coupled with the Monte Carlo Nth-Particle (MCNP) has been used for optimization. The shielding of a small modular reactor has been considered as a practical example and the reference for the calculations to evaluate and validate the developed technique and the thickness of different layers of radiation shielding has been optimized using the developed intelligent method. The shielding materials are identical to the original design, and optimization merely involves layer thickness values. Finally, considering the requirements and limitations, our design has been proposed, with a weight and volume comparison to the original design.The calculations indicate that the overall thickness of the shielding is suggested as 128.6 cm, compared to 153.7 cm the thickness of the reference to obtain the total dose of neutrons and Gamma less than 10µSv/h. The results imply a 20.9% reduction in volume and 19.35% in weight of the radiation shielding compared to the reference reactor design. It should be noted that unlike previous research, the present study used an optimization based on Monte Carlo calculations considering the real 3-D geometry of the reactor and its shield.

Production of gasoline octane booster aromatics benzene, toluene and ethylbenzene from multiphase catalytic pyrolysis of mixed waste expanded polystyrene and high density polyethylene

The main objective of this study is to enhance the valuable octane booster aromatics benzene, toluene, and ethylbenzene (BTE) and reduction of styrene in the fuel oil obtained from the pyrolysis of equally mixed waste expanded polystyrene (WEPS) and waste high density polyethylene (WHDPE). This research article, demonstrated the pyrolysis of equally mixed (50 wt.%:50 wt.%) waste expanded polystyrene (WEPS) and waste high-density polyethylene (WHDPE) followed by in situ-hydrogenation and aromatization in a small laboratory scale reactor for the production of value added commodity and gasoline octane boosters mainly benzene, toluene, and ethylbenzene (BTE). Three reactor arrangements i.e., liquid phase/L-type, vapour phase/V-type, and multiphase/M-type were used for conducting the catalytic pyrolysis of feedstock using ZSM-5, ammonium powder as a catalyst. The maximum liquid yield of 87.41 wt.% was obtained for thermal pyrolysis at a temperature of 650°C and at a heating rate of 15°C/min. Whereas, the maximum liquid yield of 82.65 wt.% was obtained for liquid phase/L-type catalytic pyrolysis at a temperature of 600°C, and at a heating rate of 15°C/min. On the other side, vapour phase/V-type and multiphase/M-type catalytic pyrolysis produced a maximum liquid yield of 78.62 wt.% and 74.07 wt.%, respectively at a reaction temperature of 550°C and at the same heating rate. The pyrolysis oil obtained from thermal pyrolysis contained lowest BTE content of 10.43 wt.% and highest styrene content of 66.25 wt.%. Whereas, the highest BTE content of 30.16 wt.% and lowest styrene content of 53.32 wt.% was found in pyrolysis oil obtained from multiphase/M-type catalytic pyrolysis. The quantitative measurement of BTE and styrene content in the pyrolysis oil was made by the gas chromatography (GC) in a flame ionization detector (FID) mode.

Analysis of Small Modular Nuclear Reactor Construction on the Moon

The current goals of lunar exploration include the return of humans to the Moon for scientific activities and the ability to implement lunar in-situ resource utilization (ISRU) for propellant production and construction. Both goals will consume power and one option to meet these power requirements are small nuclear reactors (Kilopower or Pylon). Launching these might not be beneficial in the future as the lunar infrastructure develops and the power requirements grow. The possibility of lunar independence from Earth in terms of ISRU must be considered. This paper discusses the materials used in each nuclear reactor including the fuel, reflector, working fluid, etc. along with lunar abundancy of each element used in these materials. Since no lunar ore mining or material production concepts were found, a review of Earth-based mining and production methods for each material were described. As a result, furnaces, mills/crushers, and fluidized bed reactors can be utilized for various processes in material production. Other processes including leaching, flotation, etc. will need their own machines. In the future, designing multifunctional furnaces and ball mills or crushers can be considered since this will simplify the lunar infrastructure and can be advantageous in the long run decreasing Earth dependency and cost. Keywords: Lunar Resource Utilization, Fission Surface Power, Material Production

Numerical simulation of natural convection around the dome in the passive containment air-cooling system

The Passive containment Air-cooling System (PAS) can effectively remove the decay heat of the modular small nuclear reactor after an accident. The details of natural convection around the dome, which is a key part of PAS, were investigated numerically in the present study. The thermal dynamics around the dome were studied through the temperature, pressure and velocity contours and the streamlines. Additionally, the formation of the buoyant plume at the top of the dome was investigated. The results show that with the increase of Ra, the lift-off point moves toward the bottom of the dome, and the eddy under the buoyant plume grows larger gradually, which enhances the heat transfer. And the heat transfer along the dome surface with different truncation angles was investigated. As the angle increases, the heat transfer coefficient becomes stronger as well. Consequently, a newly developed heat transfer correlation considering the influence of truncation angle for the dome is proposed based on the simulated results. This study could provide a better understanding of natural convection around the dome of PAS and the proposed correlation could also offer more predictive value in the improvement of nuclear safety.

Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

Owing to their high miniaturization potential and strong natural circulation capability, lead-bismuth reactors have been widely employed for nuclear energy utilization as well as military and civilian applications. However, there is a need to optimize the design of these reactors. To solve the complex high-dimensional nonlinear single-/multiobjective optimization problem of lead-bismuth reactors, we proposed a single-/multiobjective optimization method (termed as RBF-OLHS-NGA/NSGA II method) for optimizing the reactor core, which is based on the radial basis function (RBF) surrogate model (for prediction), orthogonal Latin hypercube sampling, niche genetic algorithm (for single-objective optimization), and nondominated sorting genetic algorithm (for multiobjective optimization). In view of this approach, the design optimization procedure of the lead-bismuth reactor based on RBF (DOPPLER-R) was developed, which combined the reactor Monte Carlo (RMC) code and the steady-state thermo-hydraulic analysis calculation (STAC) code to sample, predict, and optimize reactor core parameters. Furthermore, the single-/multiobjective optimization approach was verified by considering the core fuel loading and active zone volume of SPALLER-4 as the optimization objectives. The results show that the RBF surrogate model can accurately and rapidly predict the core characteristic parameters of lead-bismuth reactors. When compared to the value calculated using the RMC code, the relative error associated with the predicted effective multiplication factor ( <math xmlns="http://www.w3.org/1998/Math/MathML" id="M1"> <msub> <mrow> <mi>k</mi> </mrow> <mrow> <mi>e</mi> <mi>f</mi> <mi>f</mi> </mrow> </msub> </math> ) is within ±0.1%. Compared to the unoptimized values, the single-objective optimized fuel loading reduced by 400 kg, multiobjective optimized fuel loading decreased by 455–493 kg (optimization ratio of 78%–84%), and active zone volume reduced by 166362–182888 cm3 (optimization ratio of 72%–79%). These results prove that the proposed optimization method is feasible, exhibits high efficacy, and can provide new technical ideas for single-/multiobjective optimization of multiphysics, multivariable, and multiconstraint lead-bismuth reactors.

Development and verification of helium–xenon mixture cooled small reaction system

Although Helium–Xenon mixture cooled small reactor has been widely concerned, there is no mature thermal calculation software that can be applied to Helium–Xenon mixture cooled small reactor in China. In this paper, a semi empirical formula used to predict the physical properties of binary inert gas mixtures is used to establish a physical property calculation model of Helium–Xenon mixture gas mixtures, and the flow and heat transfer calculation model, turbine model and compressor model are modified and improved, so as to achieve the calculation of the Brayton cycle system. The function of the original system analysis program is expanded and improved, and the simulation program for he xenon mixed gas reactor system is developed. At the same time, the closed Brayton cycle system is established by using the system analysis program developed in this paper, and the calculation results are compared with the design values. The maximum relative error is 6.9%. The steady-state operation results of the model are in good agreement with the design operation parameters, which proves the applicability of the improved simulation model. The steady-state response characteristics analysis, the introduction of positive and negative reactivity accidents and the calculation of transient conditions such as step load drop of the reactor were carried out. The overall trend is the same as the literature survey results. The results show that the system analysis program developed in this paper is suitable for the analysis of the Brayton cycle system of Helium–Xenon mixture cooled small reactor.

Thermotolerant methanotrophs belonging to the Methylocaldum genus dominate the methanotroph communities in biogas slurry and cattle dung: A culture-based study from India

Biogas reactors run on various types of waste, with cattle dung and agricultural wastes being the primary sources in India. As biogas contains 50–60% methane, there is a possibility that the reactors harbour methanotrophs or methane-oxidizing bacteria. We set up serial endpoint dilution enrichments for the cultivation of methanotrophs using slurry from a small biogas reactor and cattle dung samples and obtained cultures of Methylocaldum gracile, a thermotolerant methanotroph. The study was expanded by sampling reactors of another small reactor of 20 L capacity and two 1000 L reactors. Dung samples were obtained from two Indian cattle breeds (Tharparkar and Gir). Pulverized rice straw used for feeding the biogas was also used for experiments. All the enrichment bottles were incubated at 39 °C, the reactors’ in-situ temperature, and the rumen gut temperature. Our study isolated four pure cultures most related to Methylocaldum gracile VKM-14LT, two strains from cattle dung samples, and two from reactors. The study also resulted in the cultivation of four additional cultures of Methylocaldum gracile and Methylocaldum tepidum, which were non-axenic and identified by pmoA gene sequencing. Pure cultures Methylocaldum gracile RS-9 and CDP-2 were studied for optimum temperature and oxygen. Both the strains were thermotolerant and grew in the temperature range of 25–45 °C with the optimum between 37 and 45 °C. The cultures could grow with minimal oxygen (0.5%–1%) in the headspace, with growth up to 10% oxygen. To summarize, we report the cultivation and isolation of methanotrophs from biogas slurries and cattle dung samples. Methylocaldum was the dominant methanotroph cultured, probably due to its thermotolerant nature and the ability to grow under variable oxygen conditions. The present study also expands the existing knowledge about habitats known for the genus Methylocaldum. An analysis of the isolated cultures would help us design strategies for methane mitigation from ruminants.

Construction schedule and cost risk for large and small light water reactors

Capital cost and construction schedule estimates for nuclear projects often under-estimate the realized costs upon project completion. Uncertainty quantification of nuclear construction project cost estimates have seldom been used in open-literature and have not shown to effectively bound the realized schedules and costs. Building on recent studies of advanced light water reactor costs and construction schedules, this paper explored the risk of different construction delay drivers and the implication on total installed cost. Using data from previous nuclear projects, as well as other megaprojects and analyses, we modeled the impact of supply chain delays, human error, change orders, and productivity on four different reactor architectures: a large passively safe PWR, a large modular BWR, small modular BWR, and a multi-module PWR. The probability and impact of supply chain delays and change orders were modeled based on the AP1000 construction experience at Vogtle Units 3 & 4, while the human error and productivity models came from non-nuclear sources. The estimated risk to cost overrun and schedule delays were found to be significant. The multi-module plant was the most susceptible to construction delays, and the small modular BWR was the least susceptible. However, the risk mitigation by a small reactor only offset the lost economy of scale in a few scenarios relative to the large reactors, because there was still significant onsite work despite modularization. Access to large labor markets mitigated the upper bound cost and schedule risks significantly for all reactor types. Finally, design changes and the resulting productivity hits were the largest driver of construction delays, and human error was the second largest driver, but supply chain delays only accounted for a small fraction of the delays for the assessed water-cooled concepts.

Scaling challenges in small modular reactor

The small modular reactor (SMR) has become a popular reactor design concept in recent years. Many regulatory authorities are evaluating a plethora of license applications related to SMRs. One important step of the reactor design is the prototype computer model assessment, and scaling is indispensable toward developing meaningful assessment data. In SMR scaling, new challenges may rise due to compact layout and unique thermal hydraulic phenomena. These phenomena may be greatly distorted or may not even be reproducible at all in the test facility. Therefore, investigation of these phenomena is needed to ensure figures of merit are adequately preserved with minimum distortion. In this paper, some phenomena were highlighted to understand the complexity. An early SMR design - the Multi-Application Small Light Water Reactor (MASLWR) is used to illustrate the scaling of these phenomena.

Shaping of Porous CeO2 Powders into Highly Active Catalyst Carriers.

CeO2 is attracting more and more attention because of its outstanding performance in heterogeneous catalysis, as an active support and a reaction promoter in reactions of industrial interest. We herein describe a novel and scalable manufacturing process of mm-sized CeO2 spheres by a combination of extrusion and spheronization of CeO2 porous powders. In this study, wet paste formulation and fabrication procedures were optimized, and as a result methylcellulose was identified as the best plasticizer for paste extrusion to provide well-defined spherical shapes and smooth surfaces, as well as reproducible batches. After nickel impregnation (10 wt %), the catalytic performance of CeO2 supports was evaluated in the CO2 methanation reaction (T = 250-350 °C, P = 5 bar·g) and compared with that of commercial Al2O3 spheres doped or not with CeO2. These novel CeO2-based catalysts are easily reduced at a moderate temperature and more active than the Al2O3 analogues, particularly at low reaction temperatures and small reactor volumes, properties that make their implementation in emerging reactor configurations very promising.

KINETIC REGULARITIES OF THE CATALYTIC OXIDATION OF ISOBUTYLENE TO METHYLACROLEIN ON AN ANTIMONY-TIN CATALYST

The parameters that determine the choice of the reactor are the reactor power, catalyst activity (intensity per unit volume), the total required amount of catalyst, the value of the adiabatic change in the temperature of the reaction mixture at complete conversion, the nature of the change in the reaction rate and selectivity with the depth of conversion, etc. the results of the theoretical optimization showed the need for the implementation of the process of partial oxidation of isobutylene to methylacrolein on an antimony-tin oxide catalyst in an isothermal reactor. To determine the optimal conditions for conducting the process on an antimony-tin catalyst, the choice of the type and main structural dimensions of the reactor for the partial catalytic oxidation of isobutylene to methylacrolein, we used the method of mathematical modeling, which is based on data on the rates of chemical transformation. To obtain quantitative characteristics, a series of experiments was carried out by influence of the composition of the reaction mixture and the rate of the oxidation of isobutylene to methylacrolein. The kinetic regularities of the process of catalytic oxidation of isobutylene to methylacrolein on an antimony-tin catalyst were recorded on a flow-type unit with a small volume reactor without a gradient. Reaching the stationary regime in this reactor occurs in 15–25 minutes

Uncertainty quantification of once-through steam generator for nuclear steam supply system using latin hypercube sampling method

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

Adsorption of CO2 on Activated Carbon, Fe‐Based Metal Organic Framework, ZnO, and CaO for Carbon Capture and Storage Application

AbstractActivated carbon (AC) from coconut shell, iron‐based metal organic framework (Fe‐MOF), zinc oxide (ZnO), and calcium oxide (CaO) were prepared as adsorbents for the CO2 adsorption process. The adsorption experiments were performed in a self‐designed fixed‐bed small reactor to study the influence of reaction conditions on the CO2 adsorption capacity of adsorbents. According to the findings, Fe‐MOF showed better performance in CO2 adsorption compared to the other selected adsorbents. The results were supported by the high surface area of Fe‐MOF, which was greater than that of AC, ZnO, and CaO. The optimal parameters for CO2 adsorption of all adsorbents were determined in terms of adsorbent dosage, operating pressure, and CO2 adsorption time.

Modeling and program development of supercritical CO2 sub-channel based on Modelica language

The application of supercritical carbon dioxide (S-CO2) in small nuclear reactors is a new research direction with promising prospects. Most of the research on S-CO2 cooled reactors focuses on the design and optimization of core parameters, but few on detailed thermal and hydraulic analysis of the reactor interior using the sub-channel method. To supplement the lack of research in S-CO2 small nuclear reactors, this paper developed an S-CO2 cooled reactor two-dimensional sub-channel analysis model based on Modelica language. To verify the correctness of the model and program, the calculated results of the sub-channel program are compared with the results of KAIST MMR in Korea, the power deviation is less than 4%, the coolant temperature deviation is less than 7%, and the fuel rod temperature deviation is less than 10%. Based on the developed sub-channel model, the transient analysis is carried out, including reactor power increase, reactivity insertion, and coolant flow loss accidents, and then the safety performance of the reactor is evaluated. The transient response of the reactor under accident conditions is obtained, and the self-stability and self-adjustment characteristics are proved. This paper relies on theoretical modeling and program development methods, studying the sub-channel thermal-hydraulic characteristics of steady-state and transient conditions, which provides theoretical reference and program support for the study of the S-CO2 small nuclear reactor.