中广核研究院有限公司

As an important component of nuclear reactors,the pressurizer plays a crucial role in controlling the pressure of the reactor coolant system during normal operation,transient conditions,and accident scenarios.Currently,steam-type pressurizers are commonly used in pressurized water reactors(PWRs)both domestically and internationally.Its working principle relies on phase equilibrium and the compressibility of saturated steam for pressure control.Compared to steam-type pressurizers,nitrogen pressurizers,which utilize a mixture of nitrogen and steam or a high volume fraction of nitrogen for pressure stabilization,offer advantages such as compact size,flexible arrangement,and rapid response.Therefore,some compact small PWRs adopt nitrogen pressurizers for pressure control of the coolant system.Nitrogen,as a non-condensable gas,exhibits significant differences in physical properties compared to steam.It is necessary to conduct experimental research on the thermal response characteristics of nitrogen pressurization systems under high pressure.This paper focuses on the experimental investigation of the thermal response characteristics of a pure nitrogen-water nitrogen pressurization system under high pressure.The experimental setup includes a water tank,a surge line,and a nitrogen pressurizer.During the experiment,the water tank was heated or cooled to simulate the temperature rise and fall processes of the reactor.The numerical simulation analysis utilizes the LOCUST 1.0.2 thermal-hydraulic system analysis program,referencing the experimental circuit diagram to establish a one-dimensional nitrogen pressurization system calculation model.The modeled components include the surge line and the pressurizer,both of which are represented as tubular components.The inlet of the surge line is connected to a time-dependent control volume,with the input boundary conditions being the temperature of the water tank and the flow rate at the inlet of the surge line.The results indicate that,compared with the heating rates of+20,+40 and+60℃/h,the higher the heating rate,the faster the response of parameters such as pressure in the nitrogen pressurizer,and the smaller the difference in the response rate of the pressurizer under different heating rates.For the heating condition,when the temperature measurement point of the pressurizer is submerged in water,a sudden temperature rise occurs,with the maximum temperature increase being 9%.For the cooling condition,when the temperature measurement point of the pressurizer is exposed to the nitrogen space,a sudden temperature drop occurs,with the maximum temperature decrease being 3%.Compared with the calculated values of LOCUST 1.0.2 and the experimental values,the relative deviation of the pressure value of the pressurizer is 2.8%,which is within the acceptable range.

[Background]Labyrinth throttle can greatly reduce pressure in a short distance and reduce fluid noise and vibration,hence is widely used in engineering fields such as nuclear power to control and adjust the flow,pressure and velocity of coolant fluid,ensuring the safety and stability of nuclear power plant system.Because the thermal hydraulic parameters of the reactor are often adjusted during design and operation,the pressure drop required to be achieved through the labyrinth throttle will also be adjusted.The local resistance coefficient of labyrinth throttle is determined by many factors,including the number of baffles(n),the distance between baffles(d)and the ratio of baffle height to pipe diameter(C).At present,the relationship between the local resistance coefficient of labyrinth throttle and these factors is not clear.[Purpose]This study aims to investigate the specific impact of geometric parameters on the local resistance coefficient,offering insights for the design of labyrinth throttling components.[Methods]Firstly,the labyrinth throttling parts with baffle height of 60 mm,baffle spacing of 15 mm,thickness of 10 mm and number of 20 were taken as the research object for mesh independence analysis.Subsequently,several throttles with different parameters were designed,and the distribution of flow velocity and pressure and the change of pressure drop inside the labyrinth throttle were studied.Then,according to the central composite design,15 throttles with different parameters were designed and their local resistance coefficients were calculated using computational fluid dynamics software FLUENT.Finally,the independent effect and mutual effect of different parameter changes on the local resistance coefficient were calculated,and the specific influence of geometric parameters on the local resistance coefficient was obtained.[Results]The calculation results show that when the inlet mass flow rate is constant,the local resistance coefficient is mainly related to the ratio of baffle height to pipe diameter and the number of baffles.The ratio of baffle height to pipe diameter plays a major role,and the baffle spacing has little effect on the local resistance coefficient.The independent effect factors of the labyrinth throttle's geometric parameters are M(C)=67.275 03,M(n)=17.482 66,and M(d)=-0.592 72,while the mutual effect factors are I(n,C)=12.617 24,I(d,C)=7.769 52,and I(n,d)=0.102 1.[Conclusions]Simulation results of this study demonstrate that the primary considerations should be the influences of the ratio of baffle height to pipe diameter and the number of baffles when designing a labyrinth throttle for liquid lead-bismuth flow to adjust its local resistance coefficient,providing a certain reference value for the optimization design and engineering application of labyrinth throttling parts in lead bismuth circuit.

[Background]Lead-cooled Fast Reactors(LFR),as promising Generation-IV nuclear energy systems,feature excellent passive safety and economic efficiency.However,thermal stratification easily occurs in the upper plenum after emergency shutdown,seriously endangering reactor operation.Currently,relevant data are scattered in various literatures with inconsistent formats,limiting the development of numerical simulation and data-driven methods.[Purpose]This study aims to integrate and standardize existing data to establish a unified,reusable benchmark dataset,supporting turbulence model calibration,system code optimization,and data-driven thermal stratification prediction.[Methods]Firstly,publicly published relevant literatures available at home and abroad from 2004 to 2026 were systematically retrieved and after rigorous screening,experimental and numerical simulation data from 32 published literatures(covering typical reactors and experimental facilities developed in different countries and regions including China,Russia,South Korea and Europe,as well as various working conditions such as steady-state operation and typical accident transients)were collected.Then,curves in literatures were digitized via Origin software(with triple independent extractions for error reduction).Finally,a structured metadata framework was established,and data were standardized through unit unification and CSV format conversion.[Results]Triple independent extractions are verified the reliability and repeatability of the method.The established dataset integrates multi-source data including vertical temperature distribution,mass flow rate,and interface position,achieving high data consistency and reliability.[Conclusions]This standardized dataset proposed in this study resolves data fragmentation and heterogeneity,providing benchmark support for turbulence model evaluation and system code optimization.It also lays a foundation for machine learning-based rapid prediction of thermal stratification,facilitating LFR thermal-hydraulic safety analysis.

[Background]In some specific reactors,ammonia is added to regulate the pH of the primary coolant.The oxidizing species is effectively suppressed by radiolytic products of ammonia,thereby the coolant is maintained in a reductive chemical environment.[Purpose]This study aims to develop a predictive model capable of simulating water chemistry behavior under various control strategies to achieve coordinated regulation of both pH and dissolved hydrogen concentrations.[Methods]Firstly,a transport model for radiolytic species in reactor coolant was developed on the basis of the RETA reactor system analysis code,with applicability to a wide range of reactor types.The model demonstrated high predictive accuracy,with root-mean-square(RMS)errors of only 1.79×10-8 for NH₃ and 5.69×10-8 for H2 concentrations compared with experimental data.Then,the KLT-40S reactor was taken as a research object,three ammonia injection strategies were established and comparatively evaluated:initial dispersion,constant-rate injection,and constant-rate injection optimized with hydrogen removal.The simulations were conducted by defining initial coolant parameters and radiation field conditions,followed by stepwise adjustment of ammonia injection rates and the timing of hydrogen removal.Each strategy was simulated until the system reached a quasi-steady state(1.6×104 s),allowing assessment of its effectiveness in controlling pH and dissolved hydrogen levels.[Results]The results indicate that the initial dispersed injection strategy,while simple and effective in maintaining a reductive chemical environment,is only able to sustain pH regulation for less than 5 h.Conversely,the constant-rate ammonia injection strategy enables sustained pH control but results in excessive dissolved hydrogen concentrations,necessitating an appropriate hydrogen removal scheme.The hydrogen removal-optimized constant-rate ammonia injection strategy is capable of simultaneously regulating pH and maintaining dissolved hydrogen concentrations.The ammonia injection rate is 1.64 g·s⁻1,and the hydrogen removal system is initiated 1 200 s after the start of ammonia injection,with a controlled removal rate of 0.014 g·s⁻1.When the system is stable,the coolant pHT stabilizes at 6.9,with dissolved hydrogen concentrations ranging from 30 mL·kg⁻1 to 35 mL·kg⁻1(Standard Temperature and Pressure,STP).[Conclusions]This study is anticipated to provide valuable insights for the development of new reactor types and the optimization of water chemistry control strategies.

The application of high burnup fuel in commercial nuclear reactors is primarily driven by compelling economic imperatives.In the competitive landscape of nuclear power generation,increasing fuel burnup can significantly extend the operational cycle of reactor cores,reduce the frequency of refueling outages,and minimize the volume of spent nuclear fuel requiring long-term storage and disposal.These benefits collectively translate into substantial cost savings for nuclear power plants,enhancing their economic viability and competitiveness relative to other energy sources.However,as fuel burnup levels rise continuously,the fuel assemblies,particularly the uranium dioxide fuel pellets,undergo progressive performance degradation due to prolonged exposure to intense neutron irradiation,high temperatures,and mechanical stresses within the reactor core.This degradation manifests in various forms,with fuel pellet fragmentation and axial relocation emerging as two of the most critical issues that compromise fuel integrity.Fuel pellet axial relocation exerts multiple adverse effects on reactor safety,especially concerning accident prevention and mitigation.A key consequence is the formation of localized hotspots in the ballooned regions of the fuel cladding.When fragmented fuel pellets migrate axially and accumulate in specific areas of the cladding,they disrupt the uniform heat transfer between the fuel and the coolant.This uneven heat distribution leads to localized overheating of the cladding,which not only accelerates cladding corrosion and creep but also elevates the risk of cladding rupture under accident conditions such as loss-of-coolant accidents(LOCAs)or reactivity-initiated accidents(RIAs).Such a scenario would severely undermine the reactor's safety barriers and trigger radioactive material release,posing significant threats to both the environment and public health.To address this critical technical challenge,in this study,an axial relocation model was developed for high burnup fuel pellets based on the LOCUST code,a specialized tool widely utilized for fuel performance analysis in nuclear engineering applications.The development process was guided by fundamental theories of fuel mechanics,heat transfer,and irradiation-induced material behavior,ensuring the model's ability to capture the complex physical mechanisms underlying pellet relocation.Subsequently,rigorous verification of the model was conducted by comparing its predictions against analytical solutions derived from established theoretical frameworks.This verification step confirmed the model's accuracy in replicating key physical phenomena under idealized conditions.To further enhance the model's credibility and practical applicability,comprehensive validation was performed using data from the FR2 reactor test,which are renowned for providing high-quality,well-documented datasets on fuel performance under various operating conditions.The comparison between the simulation results generated by the LOCUST code's axial relocation model and the experimental data demonstrated excellent agreement across multiple key parameters,including pellet relocation distance,cladding temperature distribution,and hotspot characteristics.These verification and validation results collectively indicate that the axial relocation model developed in this study exhibits robust predictive capabilities,making it a valuable tool for assessing the safety of high burnup fuel in commercial nuclear reactors and supporting the optimal design and operation of next-generation nuclear fuel systems.