Dynamic simulation analysis of molten salt reactor-coupled air-steam combined cycle power generation system

A nonlinear dynamic simulation model based on coordinated control of speed and flow rate for the molten salt reactor and combined cycle systems is proposed here to ensure the coordination and stability between the molten salt reactor and power system.This model considers the impact of thermal properties of fluid variation on accuracy and has been validated with Simulink.This study reveals the capability of the control system to compensate for anomalous situations and maintain shaft stability in the event of perturbations occurring in high-temperature molten salt tank outlet parameters.Meanwhile,the control system’s impact on the system’s dynamic characteristics under molten salt disturbance is also analyzed.The results reveal that after the disturbance occurs,the controlled system benefits from the action of the control,and the overshoot and disturbance amplitude are positively correlated,while the system power and frequency eventually return to the initial values.This simulation model provides a basis for utilizing molten salt reactors for power generation and maintaining grid stability.

Induction System for a Fusion Reactor: Quantum Mechanics Chained up

In the quest for a sustainable and abundant energy source, nuclear fusion technology stands as a beacon of hope. This study introduces a groundbreaking quantum mechanically effective induction system designed for magnetic plasma confinement within fusion reactors. The pursuit of clean energy, essential to combat climate change, hinges on the ability to harness nuclear fusion efficiently. Traditional approaches have faced challenges in plasma stability and energy efficiency. The novel induction system presented here not only addresses these issues but also transforms fusion reactors into integrated construction systems. This innovation promises compact fusion reactors, marking a significant step toward a clean and limitless energy future, free from the constraints of traditional power sources. This revolutionary quantum induction system redefines plasma confinement in fusion reactors, unlocking clean, compact, and efficient energy production.

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.

CSTR处理低浓度豆腐废水的启动研究

为探究全混合式(CSTR)反应器处理豆腐废水的启动实验,在常温条件下对自行设计的CSTR反应器进行了相关实验。结果表明:在启动后的运行时间30 d后,进水COD为3500 mg·L^(-1),水力滞留时间为10 d,COD去除率达到了99.51%,池容产气率达到了0.46 L·L^(-1)d^(-1),甲烷的质量分数达到60.69%。所以,对CSTR高效厌氧反应器的低浓度废水启动方式是可行的,对原豆腐废料的降解方式也是可行的,为今后对CSTR高效厌氧反应器的设计应用提供了一种新型启动方式。

System Variables Design of Safety Analysis for Fast Reactors

This research aims to examine the risk in the technology design of fast breeder reactors while the development depends on safety considerations. The project explored the variables, which could affect positively the expected average fuel burn-up, breeding ratio, and decay heat removal. That is accomplished using features such as guard vessels and elevated pipe routing to prevent the cracked state of both core components and fuel cladding interface conditions. So, the cracked region of fuel was detected by thermal-hydraulic analysis. We used ZrFeCr alloys to estimating of the rise in fuel cladding and coolant that can be incorporated in the design ZrFeCr alloys to uniform corrosion in temperature and 10.3 Mpa pressure. Fast creep of the reactor vessel during the coolant heat-up transient is another issue to be considered corrosion resistance of structural material can be achieved by controlling oxygen content in steel alloy. In this trend, S4337 S5140 steels are wide and can be used in future fossil power plants because of their excellent high-temperature strength.

Proposal of a Deuterium-Deuterium Fusion Reactor Intended for a Large Power Plant

This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.

Application of the CatBoost Model for Stirred Reactor State Monitoring Based on Vibration Signals

Stirred reactors are key equipment in production,and unpredictable failures will result in significant economic losses and safety issues.Therefore,it is necessary to monitor its health state.To achieve this goal,in this study,five states of the stirred reactor were firstly preset:normal,shaft bending,blade eccentricity,bearing wear,and bolt looseness.Vibration signals along x,y and z axes were collected and analyzed in both the time domain and frequency domain.Secondly,93 statistical features were extracted and evaluated by ReliefF,Maximal Information Coefficient(MIC)and XGBoost.The above evaluation results were then fused by D-S evidence theory to extract the final 16 features that are most relevant to the state of the stirred reactor.Finally,the CatBoost algorithm was introduced to establish the stirred reactor health monitoring model.The validation results showed that the model achieves 100%accuracy in detecting the fault/normal state of the stirred reactor and 98%accuracy in diagnosing the type of fault.

Review on synergistic damage effect of irradiation and corrosion on reactor structural alloys

The synergistic damage effect of irradiation and corrosion of reactor structural materials has been a prominent research focus.This paper provides a comprehensive review of the synergistic effects on the third-and fourth-generation fission nuclear energy structural materials used in pressurized water reactors and molten salt reactors.The competitive mechanisms of multiple influencing factors,such as the irradiation dose,corrosion type,and environmental temperature,are summarized in this paper.Conceptual approaches are proposed to alleviate the synergistic damage caused by irradiation and corrosion,thereby promoting in-depth research in the future and solving this key challenge for the structural materials used in reactors.

无挡板CSTR内部流场及颗粒分布的CFD模拟研究

为探究无挡板连续搅拌式反应釜(CSTR)内部流场分布及颗粒运动情况,采用计算流体力学(CFD),在欧拉-拉格朗日模型的基础上对三种搅拌釜工况进行了研究分析。同时,探讨了粒径为100,300和500μm的颗粒在各个工况下的混合情况。研究表明:在时间为0~100 s时,初始搅拌釜内颗粒的平均速度相较于初始时增加了39.22%,达到了1.4797 m/s,搅拌釜内的流场平均速度为1.7372 m/s,颗粒的平均速度与流场速度呈现相近的趋势;搅拌桨位于罐内中间位置时,颗粒分布最为均匀;采用双层搅拌桨时,颗粒速度得到了显著的提升,最大颗粒速度可达到4.2383 m/s。

Introducing the nth-Order Features Adjoint Sensitivity Analysis Methodology for Nonlinear Systems (nth-FASAM-N): II. Illustrative Example

This work highlights the unparalleled efficiency of the “nth-Order Function/ Feature Adjoint Sensitivity Analysis Methodology for Nonlinear Systems” (nth-FASAM-N) by considering the well-known Nordheim-Fuchs reactor dynamics/safety model. This model describes a short-time self-limiting power excursion in a nuclear reactor system having a negative temperature coefficient in which a large amount of reactivity is suddenly inserted, either intentionally or by accident. This nonlinear paradigm model is sufficiently complex to model realistically self-limiting power excursions for short times yet admits closed-form exact expressions for the time-dependent neutron flux, temperature distribution and energy released during the transient power burst. The nth-FASAM-N methodology is compared to the extant “nth-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Nonlinear Systems” (nth-CASAM-N) showing that: (i) the 1st-FASAM-N and the 1st-CASAM-N methodologies are equally efficient for computing the first-order sensitivities;each methodology requires a single large-scale computation for solving the “First-Level Adjoint Sensitivity System” (1st-LASS);(ii) the 2nd-FASAM-N methodology is considerably more efficient than the 2nd-CASAM-N methodology for computing the second-order sensitivities since the number of feature-functions is much smaller than the number of primary parameters;specifically for the Nordheim-Fuchs model, the 2nd-FASAM-N methodology requires 2 large-scale computations to obtain all of the exact expressions of the 28 distinct second-order response sensitivities with respect to the model parameters while the 2nd-CASAM-N methodology requires 7 large-scale computations for obtaining these 28 second-order sensitivities;(iii) the 3rd-FASAM-N methodology is even more efficient than the 3rd-CASAM-N methodology: only 2 large-scale computations are needed to obtain the exact expressions of the 84 distinct third-order response sensitivities with respect to the Nordheim-Fuchs model’s parameters when applying the 3rd-FASAM-N methodology, while the application of the 3rd-CASAM-N methodology requires at least 22 large-scale computations for computing the same 84 distinct third-order sensitivities. Together, the nth-FASAM-N and the nth-CASAM-N methodologies are the most practical methodologies for computing response sensitivities of any order comprehensively and accurately, overcoming the curse of dimensionality in sensitivity analysis.

Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors

Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it is proposed to develop and deploy(1)an enhanced Zrbased alloy or coated zircaloy for the fuel cladding,(2)alternative cladding materials with better accident tolerance,and(3)alternative fuels with enhanced accident tolerance and/or a higher U density.This review presents the features of the current UO2-zircaloy system.Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized.The features of several selected alternative fuels and cladding materials are reviewed and discussed.The neutronic evaluations of alternative fuel-cladding systems are analyzed.It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.

Temperature fault-tolerant control system of CSTR with coil and jacket heat exchanger based on dual control and fault diagnosis

For the characteristics of the continuous stirred-tank reactor(CSTR) with coil and jacket cooling system,a CSTR temperature dual control solution based on the analysis of the CSTR exothermic reaction control characteristic was proposed for an organic material polymerization production.The control solution has passive fault-tolerant ability for the jacket cooling water cutting off fault and active fault-tolerant potential for the coil cooling water cutting off fault,and it has good control ability,high saving energy and reducing consumption performance.Fault detection and diagnosis and fault-tolerant control strategy are designed for the coil cooling fault to achieve the active fault-tolerant control function.The CSTR temperature dual control,process fault detection and diagnosis and active fault-tolerant control were full integrated into the CSTR temperature fault-tolerant control system,which achieve fault tolerance control of CSTR temperature for any severe malfunction of jacket cooling or coil cooling cutting off,and the security for CSTR exothermic reaction is improved.Finally,the effectiveness of this system was validated by semi-physical simulation experiment.

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

Treatment of Practical Dye Wastewater Using a Double Dielectric Barrier Discharge System Combined with a Sequencing Batch Reactor System

Synthetic dyes are substances that are relatively stable and difficult to degrade in wastewater treatment plants using normal physical,chemical or / and biological treatment. The present work explored the synergistic effect of non-thermal plasma( NTP) and biological wastewater treatment technologies on practical dye wastewater degradation by establishing a double dielectric barrier discharge( DDBD) system combined with a sequencing batch reactor( SBR) system. The biodegradation and degradation efficiency of the DDBD-SBR system was investigated. The investigation results indicated that the DDBD technology was effective in treating the practical dye wastewater as a pre-treatment process. After a 10-min treatment,although the total organic carbon( TOC) removal efficiency was not so significant, the decolouration and the biodegradation were improved greatly. The microbial toxicity test revealed that the sample after degradation became less toxic than the original dye,which demonstrated the treatment had a significant effect on the reduction of toxicity. In addition,the SBR technology remedied the defects of DDBD treatment and improved TOC removal efficiency noticeably. The hybrid DDBD-SBR system made full use of the advantages of the individual technologies and exhibited an efficient capability for practical dye wastewater treatment.

Shunt Reactor Switching Characteristics and Maintenance Planning in 161 kV System

The high frequency transient recovery voltage caused by usually switching operation of the circuit breakers, used on shunt reactor switching, have become a noticeable problem recently. For extension the service life time and normal operation of the circuit breakers, a well modified maintenance strategy is proposed. The field testing and experimental measurement showed the maintenance strategy proposed had been proved effectively and adopted in Taiwan Power Company.

基于Hammerstein-Wiener模型的CSTR反应器辨识

针对化工过程中广泛应用的连续搅拌反应釜(CSTR)反应器,提出一种新的基于极限学习机的Hammerstein-Wiener模型的辨识建模方法。其中,Hammerstein-Wiener模型的两个非线性环节采用两个不同的极限学习机逼近,线性环节采用自回归ARX模型。因极限学习机的特殊结构,此模型可以表示成线性回归的形式,最终利用广义最小二乘法求解模型的参数。此方法辨识过程简单,辨识过程的计算量较小。最后对CSTR的辨识结果表明,在相同条件下与基于多项式的Hammerstein模型和ARX-LSSVM Hammerstein模型相比,该方法具有较高辨识精度,表明了该方法的有效性。

Bayesian belief-based model for reliability improvement of the digital reactor protection system

The digital reactor protection system(RPS)is one of the most important digital instrumentation and control(I&C)systems utilized in nuclear power plants(NPPs).It ensures a safe reactor trip when the safety-related parameters violate the operational limits and conditions of the reactor.Achieving high reliability and availability of digital RPS is essential to maintaining a high degree of reactor safety and cost savings.The main objective of this study is to develop a general methodology for improving the reliability of the RPS in NPP,based on a Bayesian Belief Network(BBN)model.The structure of BBN models is based on the incorporation of failure probability and downtime of the RPS I&C components.Various architectures with dual-state nodes for the I&C components were developed for reliability-sensitive analysis and availability optimization of the RPS and to demonstrate the effect of I&C components on the failure of the entire system.A reliability framework clarified as a reliability block diagram transformed into a BBN representation was constructed for each architecture to identify which one will fit the required reliability.The results showed that the highest availability obtained using the proposed method was 0.9999998.There are 120 experiments using two common component importance measures that are applied to define the impact of I&C modules,which revealed that some modules are more risky than others and have a larger effect on the failure of the digital RPS.

Systematic research on the performance of self-designed microwave plasma reactor for CVD high quality diamond

Optical grade diamond, in particular, high-end products, are mainly used for aerospace and defense purposes. How to manufacture them with higher quality and lower cost was critical technology. Systematic research on the performance of Self-designed plasma reactor with higher power for synthesizing diamond film was carried out. Microwave input power can reach up to 10 kW, the plasma reactor has good adaptability for the deviation of microwave frequency f at 2.45 GHz ± 20 MHz. The secondary plasma could be eliminated by adjusting the working gas pressure and the cylinder adjustment structure,it is helpful to improve the deposition rate with less energy loss. The concentrated and steady plasma could be obtained device with higher gas pressure and relative lower microwave input power. Gas supply modes and inlet gas flow rate were optimized, which would be beneficial to synthesize the film with good quality.

Scale Effects on Magnet Systems of Heliotron-Type Reactors

For power plants heliotron-type reactors have attractive advantages, such asno current-disruptions, no current-drive, and wide space between helical coils for the maintenanceof in-vessel components. However, one disadvantage is that a major radius has to be large enough toobtain large Q-value or to produce sufficient space for blankets. Although the larger radius isconsidered to increase the construction cost, the influence has not been understood clearly, yet.Scale effects on superconducting magnet systems have been estimated under the conditions of aconstant energy confinement time and similar geometrical parameters. Since the necessary magneticfield with a larger radius becomes lower, the increase rate of the weight of the coil support to themajor radius is less than the square root. The necessary major radius will be determined mainly bythe blanket space. The appropriate major radius will be around 13 m for a reactor similar to theLarge Helical Device (LHD).

Development of multi-group Monte-Carlo transport and depletion coupling calculation method and verification with metal-fueled fast reactor

The accurate modeling of depletion,intricately tied to the solution of the neutron transport equation,is crucial for the design,analysis,and licensing of nuclear reactors and their fuel cycles.This paper introduces a novel multi-group Monte-Carlo depletion calculation approach.Multi-group cross-sections(MGXS)are derived from both 3D whole-core model and 2D fuel subassembly model using the continuous-energy Monte-Carlo method.Core calculations employ the multi-group Monte-Carlo method,accommodating both homogeneous and specific local heterogeneous geometries.The proposed method has been validated against the MET-1000 metal-fueled fast reactors,using both the OECD/NEA benchmark and a new refueling benchmark introduced in this paper.Our findings suggest that microscopic MGXS,produced via the Monte-Carlo method,are viable for fast reactor depletion analyses.Furthermore,the locally heterogeneous model with angular-dependent MGXS offers robust predictions for core reactivity,control rod value,sodium void value,Doppler constants,power distribution,and concentration levels.