Development and Research of Non⁃Stirring Conveying Device for Waste Resin in Nuclear Power Plant

Blockage in the storage and transportation of waste resin is a difficult problem in the radioactive waste treatment process of nuclear power plants.In this study,in order to solve the problems of unstable resin transport concentration and easy blockage of conveying equipment and pipelines in nuclear power plants in China,a set of non⁃stirring conveying devices is developed,and theoretical calculations,simulation analysis and experimental verification are carried out.By transporting resin using the no stirring conveying device developed in this paper,it is not only to eliminate the risk of blockage and ensure the safety of transportation,but also to adjust the concentration of conveying resin to change the transport efficiency according to the operating conditions.The effective bearing rate of waste resin storage tank can be improved,so that the comprehensive performance of waste resin storage and transportation in nuclear power plants can be greatly improved.

Development of CONTHAC-3D and hydrogen distribution analysis of HPR1000

An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.

Research on Condenser Deterioration Evolution Trend Based on ANP-EWM Fusion Health Degree

This study presents a proposed method for assessing the condition and predicting the future status of condensers operating in seawater over an extended period.The aim is to address the problems of scaling and corrosion,which lead to increased loss of cold resources.The method involves utilising a set of multivariate feature parameters associated with the condenser as input for evaluation and trend prediction.This methodology offers a precise means of determining the optimal timing for condenser cleaning,with the ultimate goal of improving its overall performance.The proposed approach involves the integration of the analytic network process with subjective expert experience and the entropy weightmethod with objective big data analysis to develop a fusion health degreemodel.The mathematical model is constructed quantitatively using the improved Mahalanobis distance.Furthermore,a comprehensive prediction model is developed by integrating the improved Informer model and Markov error correction.This model takes into account the health status of the equipment and several influencing factors,includingmultivariate feature characteristics.This model facilitates the objective examination and prediction of the progression of equipment deterioration trends.The present study involves the computation and verification of the field time series data,which serves to demonstrate the accuracy of the condenser health-related models proposed in this research.These models effectively depict the real condition and temporal variations of the equipment,thus offering a valuable method for determining the precise cleaning time required for the condenser.

Fault prediction method for nuclear power machinery based on Bayesian PPCA recurrent neural network model

Early fault warning for nuclear power machinery is conducive to timely troubleshooting and reductions in safety risks and unnecessary costs. This paper presents a novel intelligent fault prediction method, integrated probabilistic principal component analysis(PPCA), multi-resolution wavelet analysis, Bayesian inference, and RNN model for nuclear power machinery that consider data uncertainty and chaotic time series. After denoising the source data, the Bayesian PPCA method is employed for dimensional reduction to obtain a refined data group. A recurrent neural network(RNN) prediction model is constructed, and a Bayesian statistical inference approach is developed to quantitatively assess the prediction reliability of the model. By modeling and analyzing the data collected on the steam turbine and components of a nuclear power plant, the results of the goodness of fit, mean square error distribution, and Bayesian confidence indicate that the proposed RNN model can implement early warning in the fault creep period. The accuracy and reliability of the proposed model are quantitatively verified.

k_(eff)uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gascooled reactors(HTGRs).In particular,the contribution of nuclear data to the k_(eff)uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for k_(eff)uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive information,and the "sandwich" method was used to quantify the k_(eff)uncertainty.We also compared the k_(eff)uncertainties to other typical reactors.Our results show that ^(235)U is the largest contributor to k_(eff)uncertainty for both the CZP and depletion conditions,while the contribution of ^(239)Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of ^(28)Si significantly contributes to the k_(eff)uncertainty owing to its specific fuel design.However,the k_(eff)uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12%owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty propagation and quantification study for small-sized HTGR.

The Setting of Sipping Test Devices for Irradiated Fuel in Nuclear Power Plant

The sipping test devices are used to identify the defective fuel. The defective fuel can be identified by detecting the occurrence of the fission products that are entrained by the medium rising around the fuel rods. This project is the setting of the two kinds of the three sipping test devices to inspect the tightness of the irradiated fuel assembly from two units NPP （nuclear power plant）. The in-mast sipping shared by two units is used for qualitative tightness test of each fuel assembly during refueling operation above the reactor. The two poolside sipping is used for quantitatively confirming the diagnosis of the in-mast sipping and identifying the tightness of the fuel at the side of the fuel storage pools of each unit after refueling. The project was implemented by research, design manufacture and calibration of these three devices and completed successfully with serious quality assurance and quality control. The performance of these devices is well demonstrated.

AN EFFICIENT ASSESSMENT METHOD FOR INTELLIGENT DESIGN RESULTS OF SHEAR WALL STRUCTURE BASED ON MECHANICAL PERFORMANCE,MATERIAL CONSUMPTION,AND EMPIRICAL RULES

Efficient methods for incorporating engineering experience into the intelligent generation and optimization of shear wall structures are lacking,hindering intelligent design performance assessment and enhancement.This study introduces an assessment method used in the intelligent design and optimization of shear wall structures that effectively combines mechanical analysis and formulaic encoding of empirical rules.First,the critical information about the structure was extracted through data structuring.Second,an empirical rule assessment method was developed based on the engineer's experience and design standards to complete a preliminary assessment and screening of the structure.Subsequently,an assessment method based on mechanical performance and material consumption was used to compare different structural schemes comprehensively.Finally,the assessment effectiveness was demonstrated using a typical case.Compared to traditional assessment methods,the proposed method is more comprehensive and significantly more efficient,promoting the intelligent transformation of structural design.

A method for establishing a bearing residual life prediction model for process enhancement equipment based on rotor imbalance response analysis

A rotating packed bed is a typical chemical process enhancement equipment that can strengthen micromixing and mass transfer.During the operation of the rotating packed bed,the nonreactants and products irregularly adhere to the wire mesh packing in the rotor,thus resulting in an imbalance in the vibration of the rotor,which may cause serious damage to the bearing and material leakage.This study proposes a model prediction for estimating the bearing residual life of a rotating packed bed based on rotor imbalance response analysis.This method is used to determine the influence of the mass on the imbalance in the vibration of the rotor on bearing damage.The major influence on rotor vibration was found to be exerted by the imbalanced mass and its distribution radius,as revealed by the results of orthogonal experiments.Through implementing finite element analysis,the imbalance response curve for the rotating packed bed rotor was obtained,and a correlation among rotor imbalance mass,distribution radius of imbalance mass,and bearing residue life was established via data fitting.The predicted value of the bearing life can be used as the reference basis for an early safety warning of a rotating packed bed to effectively avoid accidents.

Application of coupled analysis methods for prediction of blast-induced dominant vibration frequency

Blast-induced dominant vibration frequency （DVF） involves a complex, nonlinear and small sample system considering rock properties, blasting parameters and topography. In this study, a combination of grey relational analysis and dimensional analysis procedures for prediction of dominant vibration frequency are presented. Six factors are selected from extensive effect factor sequences based on grey relational analysis, and then a novel blast-induced dominant vibration frequency prediction is obtained by dimensional analysis. In addition, the prediction is simplified by sensitivity analysis with 195 experimental blast records. Validation is carried out for the proposed formula based on the site test database of the first- period blasting excavation in the Guangdong Lufeng Nuclear Power Plant （GLNPP）. The results show the proposed approach has a higher fitting degree and smaller mean error when compared with traditional predictions.

Compact Pneumatic Pulse-Jet Pump with Venturi-Like Reverse Flow Diverter

A compact pneumatic pulse-jet pump with a Venturi-like reverse flow diverter,which consists of a nozzle and diffuser,is designed for lifting and transporting a hazardous fluid through a narrow mounting hole.The pumping performance for a liquid mixture or a liquid-solid mixture is examined in terms of the effects of liquid viscosity,particle mass concentration,lifting height,and compression pressure.Results reveal that the pumping performance of the compact pneumatic pulse-jet pump is controlled by jet inertia and the flow resistance of the riser tube positioned after the diffuser.The capacity of the compact pneumatic pulse-jet pump increases with compression pressure and decreases with liquid viscosity.However,even for a liquid mixture with a high viscosity of 7.38 mPa·s,a pumping capacity of 170.7 L·h-1 was observed.For a liquid mixture,two dimensionless indices of performance were found to be the ratio of Euler numbers Euout/EuDV and the suction factor q.As the liquid-solid mixture was lifted to elevation of 6.74 m by the compact pump,the particle size distributions of the liquid-solid mixture in the tank and from the riser tube outlet were determined by a particle size analyzer and found to coincide well.

Existing Condition Analysis of Dry Spent Fuel Storage Technology

As in some domestic nuclear power plants,spent fuel pools near capacity,away-from-reactor type storage should be arranged to transfer spent fuel before the pool capacity is full and the plants can operate in safety.This study compares the features of wet and dry storage technology,analyzes the actualities of foreign dry storage facilities and then introduces structural characteristics of some foreign dry storage cask.Finally,a glance will be cast on the failure of away-from-reactor storage facilities of Pressurized Water Reacto(rPWR)to meet the need of spent-fuel storage.Therefore,this study believes dry storage will be a feasible solution to the problem.

Efficient Terahertz Photoconductive Emitters with Improved Electrode Structures

We present the design, fabrication, and characterization of two new types of terahertz photoconductive emitters. One has an asymmetric four-contact electrode structure and the other has an arc-shaped electrode structure, which are all modified from a traditional strip line antenna. Numerical simulations and real experiments confirm the good performance of the proposed antennas. An amplitude increase of about 40% is experimentally observed for the terahertz signals generated from the new structures. The special electrode structure and its induced local bias field enhancement are responsible for this radiation efficiency improvement. Our work demonstrates the feasibility of developing highly efficient terahertz photoconductive emitters by optimizing the electrode structure.

The Implementation of Radiological Characterization for Reactor Decommissioning

The radiological characterization is to determine the type, the distribution and the radioactivity of the radionuclide in the reactor by investigation of the information, instrumentation detection, sampling analyses and theoretical calculation. It is used for determination of the decommissioning implementation option and radiological protection measure of the workers. After completion of the decommissioning it is used for determination of the site release. The radiological characterization should go through whole decommissioning process, which includes： the collection and review of historical file, the performing calculation of radionuclide inventory in the reactor, in situ measurement, sampling analyses, the review and evaluation of the data obtained and the comparison of calculated result with measured date etc.. The special attention should be paid to the radiological characterization information from the key part of reactor in decommissioning end state. The sampling from the ＂hot spot＂ should not be lost; the number of the sampling should be reasonable based on reliable statistics. The radioactivity density for site release should comply with the guide, standard and regulation of International Atomic Energy Agency and China.

Performance of real‑time neutron/gamma discrimination methods

Nuclear security usually requires the simultaneous detection of neutrons and gamma rays.With the development of crystalline materials in recent years,Cs2LiLaBr6(CLLB)dual-readout detectors have attracted extensive attention from researchers,where real-time neutron/gamma pulse discrimination is the critical factor among detector performance parameters.This study investigated the discrimination performance of the charge comparison,amplitude comparison,time comparison,and pulse gradient_(m)ethods and the effects of a Sallen–Key filter on their performance.Experimental results show that the figure of merit(FOM)of all four methods is improved by proper filtering.Among them,the charge comparison method exhibits excellent noise resistance;moreover,it is the most_(s)uitable method of real-time discrimination for CLLB detectors.However,its discrimination performance depends on the parameters t_(s),t_(m),and t_(e).When t_(s)corresponds to the moment at which the pulse is at 10%of its peak value,t_(e)requires a delay of only 640–740 ns compared to t_(s),at which time the potentially optimal FOM of the charge comparison method at 3.1–3.3 MeV is greater than 1.46.The FOM obtained using the t_(m)value calculated by a proposed maximized discrimination difference model(MDDM)and the potentially optimal FOM differ by less than 3.9%,indicating that the model can provide good guidance for parameter selection in the charge comparison method.

Mesh‑free semi‑quantitative variance underestimation elimination method in Monte Caro algorithm

The inter-cycle correlation of fission source distributions(FSDs)in the Monte Carlo power iteration process results in variance underestimation of tallied physical quantities,especially in large local tallies.This study provides a mesh-free semiquantitative variance underestimation elimination method to obtain a credible confidence interval for the tallied results.This method comprises two procedures:Estimation and Elimination.The FSD inter-cycle correlation length is estimated in the Estimation procedure using the Sliced Wasserstein distance algorithm.The batch method was then used in the elimination procedure.The FSD inter-cycle correlation length was proved to be the optimum batch length to eliminate the variance underestimation problem.We exemplified this method using the OECD sphere array model and 3D PWR BEAVRS model.The results showed that the average variance underestimation ratios of local tallies declined from 37 to 87%to within±5%in these models.

A Brief Review on He Ion Irradiation Research of Steel and Iron‑Based Alloys in Nuclear Power Plants

Nuclear power plays a key role as renewable energy in alleviating the worldwide energy shortage.The material degradation caused by high-temperature and high-flux neutron irradiation is the most concerning issue for nuclear reactor safety.A large number of He atoms produced through the(n,α)transmutation reaction diffuse and migrate in metals and accumulate to form He bubbles because of the extremely low solubility of He atoms in metal materials.The helium bubbles gather at the grain boundary or grain to cause swelling,hardening,embrittlement,and other damages to the in-core structural components.This paper mainly summarizes the research progress on He irradiation in steel and iron-based alloys,including the diffusion and accumulation of He atoms,the nucleation and growth of He bubbles,and the microstructure and macroscopic degradation of material performance caused by He irradiation.The mechanism of helium irradiation-induced corrosion in steel and iron-based alloys in recent years is reviewed as well.Moreover,the investigations on irradiation performance in additive manufactured stainless steels are summarized,and the mechanism of irradiation resistance is prospected.

先进压水堆技术——“华龙一号”及福清5号机组

1.Introduction HPR1000 is an advanced mega-kilowatt-class third-generation pressurized water reactor(PWR)nuclear power technology developed by China,based on more than 30 years of experience in the scientific research,design,equipment manufacturing,construction,and operation of nuclear power.

Magnetic Co-doped 1D/2D structuredγ-Fe_(2)O_(3)/MoS_(2) effectively activated peroxymonosulfate for efficient abatement of bisphenol A via both radical and non-radical pathways

Iron-based catalysts have been widely used to treat refractory organic pollutants in wastewater.In this paper,magnetic Co-γ-Fe_(2)O_(3)was synthesized by a facile tartaric acid-assisted hydrothermal method,and Co-γ-Fe_(2)O_(3)/MoS_(2) nanocomposite catalyst was obtained via in situ growth of MoS_(2) nanosheets on Co-γ-Fe_(2)O_(3)nanoparticles.The nanocomposite catalysts were used to decompose bisphenol A(BPA)by activating peroxymonosulfate(PMS).It was shown that only 0.15 g/L catalyst and 0.5 mmol/L PMS degraded 10 mg/L of BPA(99.3%within 10 min)in the pH range of 3–9.PMS was activated due to redox cycling among the pairs Co(III)/Co(II),Fe(III)/Fe(II),and Mo(VI)/Mo(IV).Quenching experiments and electron paramagnetic resonance spectroscopy demonstrated that both radical and non-radical pathways were involved in BPA degradation,in which active radical sulfate radical and non-radical singlet oxygen were the main reactive oxygen species.Ten intermediates were identified by liquid chromatography-coupled mass spectrometry,and three possible BPA degradation pathways were proposed.The toxicity of several degradation intermediates was lower,and Co-γ-Fe_(2)O_(3)/MoS_(2) exhibited excellent reusability and could be magnetically recovered.

In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

After the first concrete was poured on December 9, 2012 at the Shidao Bay site in Rongcheng, Shandong Province, China, the construction of the reactor building for the world＇s first high-temperature gas- cooled reactor pebble-bed module （HTR-PM） demonstration power plant was completed in June, 2015. Installation of the main equipment then began, and the power plant is currently progressing well to- ward connecting to the grid at the end of 2017. The thermal power of a single HTR-PM reactor module is 250 MW,h, the helium temperatures at the reactor core inlet/outlet are 250/750 ℃, and a steam of 13.25 MPa/567 ~C is produced at the steam generator outlet. Two HTR-PM reactor modules are connect- ed to a steam turbine to form a 210 MW nuclear power plant. Due to China＇s industrial capability, we were able to overcome great difficulties, manufacture first-of-a-kind equipment, and realize series ma- jor technological innovations. We have achieved successful results in many aspects, including planning and implementing R＆D, establishing an industrial partnership, manufacturing equipment, fuel produc- tion, licensing, site preparation, and balancing safety and economics; these obtained experiences may also be referenced by the global nuclear community.

HPR1000： Advanced Pressurized Water Reactor with Active and Passive Safety

HPR1000 is an advanced nuclear power plant(NPP)with the significant feature of an active and passive safety design philosophy,developed by the China National Nuclear Corporation.On one hand,it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP;on the other hand,it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies,active and passive safety systems,comprehensive severe accident prevention and mitigation measures,enhanced protection against external events,and improved emergency response capability.Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems,the reactor core,and the main equipment.The design of HPR1000fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements,and addresses the safety issues relevant to the Fukushima accident.Along with its outstanding safety and economy,HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.