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This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO2 and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations. [ABSTRACT FROM AUTHOR]

Monitoring the radial power distribution during the operation of a pressurized light water reactor (PWR) is crucial for ensuring safe operating conditions and achieving high levels of fuel burnup. This paper introduces a methodology utilizing Artificial Neural Networks (ANN) for reconstructing the radial power distribution in the core of a Pressurized Water Reactor (PWR) with a hot full power of 1876 MWth, such as the Angra 1 reactor. This approach uses measurements from Self-Powered Neutron Detectors (SPND), simulated through the SERPENT code. The use of ANN demonstrated good accuracy in predicting the radial power distribution with an average relative error of 1.27%, considering 36 active detectors, with maximum relative error of 6.99%. Moreover, the proposed process demonstrated robust performance, even when measurements from one, two, or three SPND detectors were unavailable, resulting in errors of 1.24%, 1.13 %, and 1.09%, respectively. Therefore, the methodology ensures a reliable reconstruction of the radial power distribution, even when SPND detector measurements are unavailable, enabling the optimization of detector use and contributing to the increase of operational safety margins.

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The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. The three-dimensional numerical model consisting of the reactor pressure vessel with core internals was developed using available geometrical and material data as well as the reactor's operating conditions. The reactor core was divided into 11 axial and 22 radial burnup zones in order to recreate the spatial distribution of the fuel burnup. The isotopic changes due to the nuclear transmutations and decays were calculated in each burnup zone until the desired average burnup of 17.571 GWd/tHMint was reached. The calculations include changes in the boric acid concentration at defined time steps and the burnout of the gadolinia burnable absorber embedded in the nuclear fuel. This study shows the spatial distribution of minor and major actinides at the end of the reactor cycle as well as the depletion of uranium, the build-up of plutonium, and the formation of neptunium, americium, and curium during the reactor's operation. [ABSTRACT FROM AUTHOR]

A new method for the generation of axial power distributions using artificial neural network (ANN) technique and real-time measured planar radial peaking factor, Fxy (hereinafter 'Live Fxy') method for digital core protection system of pressurized water reactors is presented. ANN with Simulated Annealing (SA) technique to find global optimum solution was used to calculate core average axial power distribution. In Live Fxy method, axial node-wise Fxys were calculated by multiplying pin/box factors to the measured assembly powers inferred from in-core detector signals without calculating detailed 3D power distributions. The hot pin power distribution for DNBR and LPD was then obtained by multiplying the core average axial power distribution and Fxys at axial nodes. The validation of the method was performed for various core conditions (e.g. core power levels, control rod positions, etc.) of Korean OPR1000 power plant. The result showed a decrease in RMS errors by 2.2% ~ 7.0% (3.6% on average), and the minimum thermal margins for DNBR and LPD were increased by 6.4% and 15.6%, respectively. The application of the method would improve the accuracy of axial power distribution and thermal margin, contributing to the operability and safety of digital core protection system. [ABSTRACT FROM AUTHOR]

In order to achieve a full mechanistic understanding of stress corrosion cracking (SCC), the key operating mechanisms need to be identified but also quantified. In this study, we summarize and rationalize key findings from the last 15 years of high-resolution characterization of SCC in our group. A comprehensive characterization of a set of austenitic alloys with different Ni content and constant Cr level, tested under simulated pressurized water reactor (PWR) primary water conditions at various temperatures, has revealed evidence for at least two operating mechanisms: one diffusion-related and the other deformation-related. For their relevance to the nuclear industry, two additional alloys with increased Cr content were also studied (A800 and A690). Key precursors for SCC initiation and propagation are identified and their effect on alloy degradation discussed. A list of key materials' properties that ensure low SCC susceptibility is proposed. [ABSTRACT FROM AUTHOR]

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0–300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

AbstractThis research presents the development of a one-dimensional analytical model to investigate the impact of pressure variations in the primary loop on natural circulation (NC). The model takes into account a sinusoidal input heat distribution and derives equations for the parameters of NC. The model covers a broad spectrum of NC patterns, spanning from fully single-phase to fully two-phase flow. The research demonstrates a smooth and continuous transition between various kinds of NC. Moreover, the research demonstrates that NC is capable of efficiently dissipating the decay heat generated inside the core of a pressurized water reactor, encompassing a range from 100% to 60% of the total inventory present within the primary loop. The findings of this study are compared to prior research outcomes and demonstrate a reasonable level of consistency. Additionally, comparisons are made with uniform input power distribution to demonstrate that there are no significant differences in the NC parameters when using sinusoidal heat input. [ABSTRACT FROM AUTHOR]

The fatigue life of Type 316 stainless steel in the pressurized water reactor (PWR) water environment and its dependency on that in air were investigated. Specimens from two heats of the material were prepared for fatigue tests, and additionally, previous test findings obtained using another heat were applied. The fatigue test results obtained for various strain ranges and strain rates revealed that the fatigue life in the water environment Nf(env) did not depend on that in air Nf(air) when the strain rate was 0.4%/s. The variation in the ratio of Nf(env) to Nf(air) was larger when Nf(air) obtained for each heat of the material was used than that when a general Nf(air) was used. The changes in the penalty factor Fen with applied strain range and strain rate could be predicted by assuming Nf(env) depended not on strain rate but on rise time. Highlights: Fatigue life in PWR water environment did not depend on that in air.Fen should be determined using a common life in air regardless of material heat.Rise time better to represented test speed to predict the environmental effect.Changes in Fen with strain range and rate could be predicted using rise time. [ABSTRACT FROM AUTHOR]

A major subject for evaluating the corrosive conditions in the pressurized water reactor (PWR) primary coolant is to determine the optimal hydrogen concentration for mitigating primary water stress corrosion cracking (PWSCC) without adverse effects on major structural materials. An analytical method combining water radiolysis and electrochemical corrosion potential (ECP) analyses was proposed for evaluating the corrosive conditions in PWR primary coolant. These procedures originally developed for boiling water reactors (BWRs) were successfully extended to PWRs with different water chemistry parameters, including pH, temperature, and radiation qualities, after minor changes in the original procedures and major input parameters, such as the inclusion of the effects of alpha radiolysis and Li+ (Na+) and H+ effects for the anodic polarization curve. This study discusses the results of water radiolysis analysis for PWR primary coolant, and the characteristic behavior of hydrogen peroxide (H2O2) as a function of hydrogen inlet concentrations in PWR primary coolant conditions (higher pH and α-ray irradiation). A possible reaction scheme involving eaq– was proposed for H2O2 suppression under alkaline conditions. The corrosive conditions were discussed in the following publication, using ECP as the major index for PWR corrosive conditions. [ABSTRACT FROM AUTHOR]

Background The optimization of core loading patterns in nuclear reactors is one of the most studied optimization problems in nuclear engineering due to the enormous economical and safety benefits. Various algorithms such as Genetic Algorithms (GA), Simulated Annealing (SA), and Parallel Simulated Annealing (PSA) have been used in the past for such problems. Methods In this work, a PSA algorithm was developed and integrated into the Modularly Implemented Design Assistance Suite (MIDAS), a framework developed at North Carolina State University to solve nuclear engineering problems. The effectiveness of PSA was compared against the GA and SA algorithms available in MIDAS for a Pressurized Water Reactor first cycle core loading pattern optimization problem. Results PSA consistently generates more optimal solutions than SA and GA by having the higher average fitness, and showing less variance in its performance and thus being more robust. This provides confidence in the PSA implementation within MIDAS. The obtained loading pattern positions high reactive fuel in peripheral locations and low reactive fuel towards the centers in a strategy resembling both Out-In-Checkboard and L3P loading pattern approaches. Conclusions Future studies will involve applying the PSA algorithm to other optimization studies in larger combinatorial spaces, such as in multi-cycle optimization problems.

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

The load control of a nuclear reactor is important due to the nonlinear nature of its dynamics and the dependence of some parameters on the output power. the Proportional-Integral-Derivative controller (PID) is commonly regarded as an easy choice for reliable control. In this research, the relative neutron density in the point kinetics model of a Pressurized Water Reactor (PWR) is controlled by an optimized PID with the meta-heuristic Differential Evolution (DE) algorithm. The Integral of Time-Absolute Error (ITAE) performance index has been used for optimization with this algorithm. The simulation results show that the optimized control system with the DE algorithm has the appropriate efficiency and accuracy in response to power demand.

Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %. [ABSTRACT FROM AUTHOR]

In the field of nuclear reactor design, precise calculation of reactor dimensions is crucial for operational efficiency and safety. This study focuses on the critical dimension prediction for a spherical nuclear thermal reactor, employing a computational approach to address this challenge. Utilizing the two-group neutron diffusion equation, the research aims to establish a mathematical model for neutron distribution within the reactor's geometry. A key aspect of this model involves the prediction of neutrons' spatial distribution, essential for understanding the reactor's behavior under operational conditions. The methodology adopted in this investigation involves using a MATLABbased program specifically developed for solving the two-group diffusion equation in spherical reactor geometry. This approach facilitates the determination of exact dimensions and optimal fuel mass for the reactor. The study's findings indicate a critical core radius of 21.7 cm, with a water mass of 40.5 kg and a U235 fuel mass of 1.12 kg. Additionally, the ratio of fast flux to slow flux was approximately 1.5. These results not only align closely with prior research in this domain but also enhance the understanding of spherical reactor design. Crucially, the study's outcomes demonstrate a high degree of compatibility with existing literature, thereby reinforcing the validity of the computational model used. This research contributes significantly to the nuclear engineering field by providing a robust method for determining the critical dimensions and mass of fuel required for the efficient and safe operation of spherical nuclear thermal reactors. Which still regarded nuclear energy as an essential partner in friendly energy production in the world to reduce CO2 emissions. The implications of these findings are substantial, offering a pathway to optimized reactor design and a deeper understanding of neutron behavior in complex geometrical configurations. [ABSTRACT FROM AUTHOR]

This research article explores the potential of optical fibers as sensors, highlighting their ability to measure various parameters such as temperature, pressure, stress, and radiation dose. The study focuses on investigating the material compatibility of optical fibers in challenging sensing environments like Gen II/II+ and advance nuclear reactors, as well as concentrated solar power (CSP) plants. Material compatibility tests were conducted to determine the feasibility of using fluorine and germanium optical fiber sensors in these environments. The study found that raw fibers were corrosion-resistant to lead bismuth eutectic at 600 °C, regardless of the coating. In molten salt environments, raw fibers were incompatible with FLiNaK but showed corrosion resistance to MgCl₂-NaCl-KCl. However, the survivability of raw fiber optics improved with a gold coating in FLiNaK. Raw fiber optics were found to be incompatible in high-temperature steam at 1200 °C and in a pressurized water reactor (PWR) at 300 °C.

In thiswork, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4–5 w/o UO2 fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er2O3) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd2O3) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO2 in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

This study proposes an effective repair technology using arc additive manufacturing for pressurized water reactors (PWRs) in nuclear power plants (NPP) aimed at avoiding complete replacements and post-weld-heat treatments (PWHTs) of component parts while ensuring safety and reliability. Effective repair technology is defined as economic and process efficiency, because of maintenance costs and radiation exposure, and it is critical in related industries. The technology is designed to relieve the hardness and martensite fraction of the welding heat affected zone (HAZ) of low alloy steels (SA508) in PWRs penetration/nozzles by heat source generated in the WAAM process, thus ensuring structural integrity. In the first layer of wire arc additive manufacturing processes, 89.6% of the martensite phase was formed in the HAZ of SA508, which was significantly reduced to 45.7% due to repetitive thermal behaviors at the third layer. The resulting process dramatically reduced hardness from 450Hv in the initial layer additive manufacturing to 320Hv in 3 layers without additional heat treatments. Moreover, the study quantitatively investigated the martensite starting temperature (Ms) and bainite starting temperature (Bs) and analyzed the microstructure and mechanical behavior of the developed process using thermodynamic calculation (CALPHAD), finite element method (FEM) simulation, and microstructure quantitative analysis by electron backs-catter diffraction (EBSD). The proposed technologies and their quantitative analysis results can be a substantial alternative to the repair technology for penetration/nozzles in nuclear primary water cooling reactor applications, complying with ASME Sec.IX Qw-462.12 and ISO 15614-1.

This research article explores the potential of optical fibers as sensors, highlighting their ability to measure various parameters such as temperature, pressure, stress, and radiation dose. The study focuses on investigating the material compatibility of optical fibers in challenging sensing environments like Gen II/II+ and advance nuclear reactors, as well as concentrated solar power (CSP) plants. Material compatibility tests were conducted to determine the feasibility of using fluorine and germanium optical fiber sensors in these environments. The study found that raw fibers were corrosion-resistant to lead bismuth eutectic at 600 °C, regardless of the coating. In molten salt environments, raw fibers were incompatible with FLiNaK but showed corrosion resistance to MgCl₂-NaCl-KCl. However, the survivability of raw fiber optics improved with a gold coating in FLiNaK. Raw fiber optics were found to be incompatible in high-temperature steam at 1200 °C and in a pressurized water reactor (PWR) at 300 °C. [ABSTRACT FROM AUTHOR]

The investigation of thermal-hydraulic parameters during steam-line break (SLB) accidents is performed by applying the personal computer transient analyzer (PCTRAN) simulator model of the VVER-1200 nuclear power plant (NPP). Five cases, namely, 0.005 m2 (Case-1), 0.01 m2 break (Case-2), 0.02 m2 break (Case-3), 0.04 m2 (Case-4), and 0.08 m2 (case-5) of SLB accident inside containment with the concurrent loss of AC power have been simulated. There was no variation in the timing of the trip of the reactor coolant pumps, the main feedwater pumps, or the turbine in any of the five SLB accidents. However, the reactor scram's onset time varies slightly between the five scenarios. Pressure and temperature in the reactor coolant system (RCS) quickly reached a peak following the start of the SLB accident, fell shortly after the reactor scram, and eventually stabilized in all cases. In comparison to the larger breaks in the SLB accident, the smaller breaks result in a higher RCS temperature and pressure. After the SLB accident, the pressurizer's liquid level rises and then quickly drops in all cases. The break mass flow rate from the steam line rapidly increases until the occurrence of the reactor scram and then decreases to a stabilized value. Steam generator A has a faster rate of heat removal rate than steam generator B, and its pressure and liquid level decrease more quickly than those of steam generator B. The thermal power of the reactor, peak cladding temperature, and fuel temperature showed a rapid drop after the initiation of the SLB accident. There was no increase in these parameters from the initial state of the simulation. The radiation in the air of the reactor building and steam line was very low during the simulation period. Therefore, there was no violation of the safety aspects of the SLB accident of the PCTRAN simulation of the VVER-1200 NPP model.

Current plans and efforts of reactor operators and vendors to include extended-enrichment (EE) fuel and accident-tolerant fuel (ATF) in current reactor fleets motivate the study of these changes in reactor physics analysis. This work uses the U.S. Nuclear Regulatory Commission's core simulator PARCS to do the core calculation and the SCALE Polaris lattice physics code to generate the homogenized, few-group constants. In this work, both pressurized water reactor and boiling water reactor (BWR) colorset models are used to verify the proposed approach. The accuracy presented in the colorset models verified the capability of the PARCS/Polaris procedures for the transition core analysis in light water reactors. For the whole-core calculation, the ATF and EE-ATF transition core models were incorporated, in addition to the nominal core model. The BWR model was chosen to represent the entire core calculation due to its challenging design. The core parameters studied are the core power distribution, power peaking factor, Doppler temperature coefficients, and control rod worth at cold zero power and hot full power. When the core parameters of the transition cores are compared with those of the nominal core in PARCS, the results suggest that there is no drastic change in the core parameters for the implementation of ATF and EE fuels. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

58Co and 60Co are the primary radionuclides and contributors to collective dose of PWRs, which may exist in ionic, particle and colloidal state in the primary loop. The pore diameter of the most advanced filter used in chemical and volume control system (CVCS) in domestic PWRs is 0.1 μm, the removal efficiency of Co colloids with the diameter of less than 0.1 μm is very low. Activated corrosion products such as Co colloids may deposit on the surface of the primary loop and auxillary systems during shutdown, contributing to collective dose. The chemical conditions of primary loop during the oxidation operation process greatly affect the removal efficiency of radionuclides such as Co colloids. To achieve the best dissolution and removal effect of Co colloids, the dissolution behavior of Co colloids need to be studied. In this paper, simulated Co colloids were first synthesized in the laboratory. The average particle size of simulated Co colloids was characterized by TEM to be 40-50 nm. The composition of simulated Co colloids was characterized by TEM, SEM and EDS as CoO and Co. Then the effects of temperature (60-80 ℃), Li concentration (pH, 0.1-3 mg/kg), H2O2 concentration (0.20 mg/kg) and reaction time (0-6 h) on the dissolution behavior of Co colloids were investigated on simulated chemical condition of the oxidation operation process. The results show that the dissolution of Co colloids is quite fast, reaching reaction equilibrium in 0.5-1 h. The dissolution behavior of Co colloids is affected by temperature, Li concentration and H2O2 concentration. The dissolution of Co colloids is promoted by addition of H2O2 due to the weak acidity and strong oxidizing power of H2O2. However, too much H2O2 is unfavourable for the dissolution of Co colloids, which may be due to the possible formation of passive film on Co colloids. Low Li concentration or low pH is favourable for the dissolution of Co colloids. The oxidizing power of H2O2 is stronger at higher temperature promoting the dissolution of Co colloids. Yet high temperature is not favorable for the dissolution of Co colloids, as the dissolution of Co colloids is an exothermal process. Therefore a moderate temperature is the best choice for the dissolution of Co colloids. To summarize, the best chemical condition for the dissolution and removal of Co colloids during the oxidation operation process is at 70 ℃ with as low Li concentration as possible and about 10 mg/kg H2O2. The results in this paper provide important references for the improvement of the oxidation operation process of PWRs.

Abstract Striving to peak carbon emissions and achieve carbon neutrality (known as the "Dual-Carbon" goal) is an inevitable requirement for elevating the environmental resource constraints and realizing harmonious coexistence between the mankind and the earth. In the energy system, nuclear energy offers various advantages, such as high energy density, low carbon emission, strong environmental adaptability and large potential for energy co-generation and co-supply. It is one of the supporting energy sources for the transformation and upgradation of the energy system to a clean, efficient and low-carbon way. In this paper, the opportunities and challenges for innovation-driven nuclear energy development in the fields of electricity generation, hydrogen production, heat supply and seawater desalination under the goal of "Dual-Carbon" are discussed and analyzed. Besides, the relevant research on improving the safety and economy of the pressurized water reactor (PWR) and sodium-cooled fast reactor (SFR) conducted by the Nuclear THermal–hydraulic research Lab (NuTHeL) of Xi’an Jiaotong University is briefly introduced.