bluuebunny/crossref_metadata_2025_split huggingface dataset:
Deuterization Process and Result Analysis of Resin in Heavy Water Purification System of a Reactor
Authors: Zhang Wen Juan , Xu Fengxia , Hua Xiao 
Published: November 2022

Abstract
This paper introduces in detail the whole process of resin replacement of heavy water purification system in a reactor reconstruction project in Algeria, including old resin discharge, pipeline cleaning, new resin filling, new resin deuterization, water quality detection, heavy water storage and continuous operation of the system. By sorting out deuterization data and tracking and analyzing the water quality detection results, it is considered that the resin deuterization process of the heavy water purification system of a reactor is smooth. After deuterization, the effluent concentrations of the three resin columns of the heavy water purification system meet the operation requirements. The whole deuterization process tries to save precious high concentration heavy water for Algeria. After deuterization, the resin column is connected to the heavy water purification system. After continuous operation and sampling detection, it is determined that the purification effect is ideal and the pH value is not affected. The resin of the heavy water purification system is successfully replaced and can be put into use. At the same time, aiming at some problems in the deuterization process, this paper analyzes the causes in detail, summarizes the experience and lessons, and gives optimization suggestions and precautions in the deuterization process.
Research on the Construction of the Advanced Nuclear Power Standards System Platform
Authors: Zhang Shaolin , Yang Wen , Dong Ruilin , Liang Yun , Liu Shangyuan 
Published: November 2022

Abstract
Advanced nuclear power standards system is a new type of nuclear power standards system composed of all kinds of standards in the nuclear energy field.
At present, the research results of the two stages of the subject and the achievements of nuclear power standard construction in recent years are scattered. It is necessary to establish a single nuclear power standard database to process and develop the standards in the field of nuclear power and various research results, reports, process documents, data and other databases, so as to provide a good data environment for the majority of nuclear engineering scientists and technicians.
In order to solidify the research results of advanced nuclear power standard system and nuclear power standard data, the application support platform of nuclear power standard system adopts innovative SaaS cloud service mode to build an Internet-based cloud service platform with process engine and knowledge map technology.
Through the integration of informatization and standardization, standardize the standardization process, we can provide convenient and fast information services, vigorously expand the scope of participants in nuclear power standardization and the depth of participation of all parties, and realize the standard system management mode of autonomy, industry participation and dynamic updating.
Analysis of Different Requirements of Dissimilar Metal Joint Welding Procedure Qualification Between RCC-M and ASME BPVC
Authors: Pei Wang , Dongxu Chen , Shumin Yang 
Published: November 2022

Abstract
In order to enhance the comprehension of the welding procedure qualification standards, RCC-M (Design and Construction Rules for Mechanical Components of PWR Nuclear Islands) and ASME BPVC (American Society of Mechanical Engineers Boiler and Pressure Vessel Code), used by the components fabrication for nuclear power industry in the People’s Republic of China, this paper takes the qualification of dissimilar metal joint between inlet-nozzle and safe end of steam generator in Generation III nuclear power unit as an example, taking a comparative approach, compares the requirements of different welding variables one by one between two standards and then lists the differences and analyzes the differences. It shows that RCC-M has more strict requirements than ASME BPVC on the qualification of dissimilar metal joint with buttering, the type of test couple qualified by RCC-M is more similar to product than that by ASME BPVC, and qualification by RCC-M may need more test couples than that by ASME BPVC, if choosing same value of variables which have overlap between these two standards, test coupon qualified by RCC-M can used for that qualified by ASME BPVC, but not vice versa. At last, the paper gives some implementation advises and guidance of subsequent qualification works during design and fabrication.
Analysis on the Nuclear Safety Supervision Mode of the World’s Major Nuclear Power Countries and Its Enlightenment to the Improvement of China’s Nuclear Safety Supervision Technical Support Ability
Authors: Wu Han , Yu Guoxin , Teng Keyan , Zheng Xiangyang 
Published: November 2022

Abstract
In the view of the International Atomic Energy Agency (IAEA), the decision-making of safety supervision of civil nuclear facilities needs independence. In order to deeply evaluate and review a series of issues related to the safety of civil nuclear facilities, national regulatory authorities usually need to consult independent institutions or external committees with professional capabilities. Such organizations providing technical services are called Technical Support Organizations (TSO), which can be national laboratories, research institutions or consulting units. In some cases, support from professional organizations in other countries may also be required.
Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. All along, China has continuously strengthened the construction of nuclear safety supervision capacity, and the safety supervision capacity has been gradually improved.
China has established a nuclear and radiation safety supervision system, including National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional nuclear and radiation safety supervision stations, Nuclear and Radiation Safety Center (NSC) and other units. China has established a nuclear safety expert committee, and mainly relies on the National Nuclear and Radiation Safety Supervision Technology R & D Base (R & D Base) as the central level test, verification and R & D platform to provide consultation and technical support.
The R & D base has completed the construction of some test, verification and R & D functions. China has built a series of important laboratories in the R & D base, it still has not formed R & D system. In the future, China need to improve the technical support capacity of nuclear safety supervision.
In short, the independence of nuclear safety regulatory decision-making requires to enhance from institutional independence to capability independence. The independence of China’s nuclear safety regulatory authority has been basically established. In the future, it is more important to improve the regulatory capacity. Only by establishing a perfect technology R & D system can nuclear safety decision-making be based on sufficient scientific basis.
Fire Probability Safety Analysis for a Compact Small Modular Reactor
Authors: Chen Yanzhu 
Published: November 2022

Abstract
Traditional fire protection design applied in the second generation nuclear power plant,was based on determinism and reinforced by Fire Probability Safety Analysis,in which Problems such as difficult selection of fire protection-related technical solutions, low efficiency of fire protection layout, and large rework workload for later technical solutions update.For highly compact SMRs with tight design schedule, it’s impossible to acceptable such shortcomings.
According to the requirements of SMRs fire protection design,this paper based on NUREG/CR-6850 fire probability safety analysis methodology, combined the design depth (the concept design phase)of SMRs, develop the SMRs fire PSA model, calculated quantitatively Contribution of core damage frequency (CDF) of fire PSA model.At the same time, based on the results of internal fire PSA analysis, the design weakness is found, and the design/improvement proposal based on fire PSA analysis is given.
Review on Reactor Coolant Pumps Trip Criteria After Small Break LOCA in PWR
Authors: Jiao Zhenying , Wang Zhenying , Huang Yu , Liu Liu , Wang Taike 
Published: November 2022

Abstract
The post-accident operation status of reactor coolant pumps (RCP) play an important rule in accident mitigation, especially for the loss of coolant accident (LOCA), the prolonged RCP operation after medium and small LOCA will aggravate the deterioration process of RCS water inventory. This paper first analyzes the adverse influence of delayed trip of RCPs after medium and small break on RCS, then the principle of determining the trip criteria of RCPs is suggested, the trip criteria of RCPs from the worldwide nuclear steam supply system venders are investigated and summarized. From the perspective of prolonging the operator ‘s non-intervention time, reasonable measures to alleviate the consequences of medium and small LOCA, such as the RCP automatic trip logic and RCS fast depressurization by cooling, etc., are of great significance to improving the safety level of the plant.
Modeling and Simulation of a Pressurized Water Reactor Nuclear Power Plant and Power System
Authors: Wang Haomin , Wei Xinyu 
Published: November 2022

Abstract
As the proportion of nuclear power in the power grid continues to increase, for the safe, stable, and economical operation of the power system, nuclear power plants must participate in the peak shaving operation of the power system. Therefore, it is very important to study the impact of the power system on nuclear power plants. In this paper, a nonlinear dynamic model of a pressurized water reactor (PWR) is built on the MATLAB/Simulink simulation platform. The model includes the point kinetics reactor core model, the thermal-hydraulics model, the piping & plenum model, the steam generator module, the reactivity module, the sensor model and the turbine module, in addition to the governor module and the simulation model of the power system. Based on the established dynamic model of nuclear power plant and an island power system module, the response of the model was tested under load variations. There are often some short circuit faults in the power system. To verify the stability of the model, the impact of the single-phase-to-ground fault in the power system on the nuclear power plant is analyzed. In this paper, the nuclear power plant can maintain stable operation in the event of a transient failure in the power system.
Development of an MPS Code for Corium Behavior Analysis
Authors: Jian Lijun , Lu Hongfan , Yu Peng , Yuan Yidan 
Published: November 2022

Abstract
After fukushima nuclear accident in Japan, phenomenology of serious accident has been paid more and more attention. And particle methods have a natural advantage in computational simulations involving deformation and free surfaces. The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: a heat transfer model for thermal conductivity and a phase change model for Solid-liquid conversion, an enthalpy-based viscosity model to realize a smooth transition of viscosity at the solid-liquid interface and a surface tension model for better simulation of surface shape. Validation of the developed simulation approach is carried out with a classical water column collapse example. The heat transfer model is validated by the example of a one-dimensional plate. The surface tension model is validated by a case of droplet oscillation. A comprehensive example of the melting of “Wood’s alloy” is carried out to verify the capacity of MPS method in the simulation of melting and expansion. The simulation results are in good agreement with the experimental results, which indicates that MPS method promises well in the field of severe accidents.
A Statistical Approach to Quantify Taylor Microscale for Turbulent Flow Surrogate Model
Authors: Ross Molly , Matulis John , Bindra Hitesh 
Published: November 2022

Abstract
Non-equilibrium statistical mechanics models can be used to construct reduced order models from the time-dynamics data such as numerical or physical fluid mechanics experiments. One of the well-established statistical projection methods is the Kramers-Moyal expansion (KM) method. The first two terms of the KM expansion result can be used to construct a non-linear Langevin equation, which can serve as the statistically-trained reduced-order model. This non-linear Langevin equation can be approximated to the Fokker-Planck equation, which is similar to Advection-Diffusion equation, thereby preserving some characteristics of fluctuations associated with fluid mechanics. The KM method captures continuous-time dynamics, however, any data obtained through measurement is discrete. In order to accurately capture the time dynamics of the discrete data, the method for calculating the KM coefficients must be carefully chosen and implemented. To better represent the solution from discrete data, the drift and diffusion coefficients can be calculated at multiple time scales and then extrapolated to a time scale of zero, assuming a linear correlation. One challenge in using this method is that the calculated KM coefficients are only accurate for time scales greater than the Taylor microscale. This means that the extrapolation must use only the KM coefficients calculated for time scales greater than the Taylor microscale, however, this value is not always provided from the data nor simple to calculate. This work presents a method of approximating the Taylor microscale from the data through the relationship between the Markov property and the Taylor microscale and implementing this method to find the extrapolated KM coefficients. The KM method implementing the Taylor microscale estimation was applied to existing DNS turbulent channel flow data to model a time series. This generated time series was then compared to the DNS data using a statistical analysis including probability density function, autocorrelation, and power spectral density.
Analysis of Double Heterogeneous Effect Under Double Dispersion of Fuel and Burnable Poison Particles
Authors: Yuan Anmin , Wu Hongchun , Li Yunzhao , Cao Liangzhi , Wang Cong , Jiang Chaofei , Yu Shengzhi , Wang Qianglong , Qiu Jinrong 
Published: November 2022

Abstract
Dispersed particle fuel is an advanced form of fuel. Due to the need for reactivity control, burnable poison particles and fuel particles are often dispersed in the matrix together. In order to prove the double dispersion mechanism of fuel and burnable poison particles with high calculation accuracy, a random dispersion model of burnable poison particles is constructed in this paper under the conditions of fuel homogenization and random dispersion of fuel particles. Compared with the random dispersion model, direct homogenization of fuel or burnable poison can cause large deviation in the calculation of system eigenvalues according to the MCX calculation. Under the same fuel phase volume, fuel enrichment and burnable poison loading, the maximum deviation can reach 1494pcm. The deviation increases with the increase of the loading mass of the burnable poison in the system, subject to fuel particle diameter of the order ∼100 μm. At a relatively high content of the burnable poison, the deviation decreases with the change of the content of the burnable poison. The calculation results show that the double heterogeneity effect of the system composed of fuel and burnable poison particles cannot be ignored, and the traditional homogenization calculation method must be corrected according to the actual situation. This paper is of valuable reference for the calculation and correction of dispersion fuel homogenization under the condition of strong absorber, and the program development and verification of using a new method to deal with the double heterogeneity effect.
Study on Maintenance Method of Design Basis in Operating Nuclear Power Plants
Authors: Wu Guangjun , Xiao Wei 
Published: November 2022

Abstract
How to maintain the design basis to maintain or improve the safety level of nuclear power plants in their lifetime, the major nuclear power operation countries led by the United States and France have made a beneficial attempt. There is no systematic research and Application on the maintenance of design basis in China. With the accumulation of operation time of domestic nuclear power plants, we will soon face the challenge of design basis and the resulting pressure of safe operation. Design basis maintenance concept was raised from international Nuclear Power Plants (NPPs) operating experience, due to the safety and operating performance being challenged by the missing or violating of design bases. Investigation in the international experience of design basis maintenance is carried out and basis and key performance factors are identified base on the research in World Association of Nuclear Operators (WANO) and International Atomic Energy Agency (IAEA) suggestion and criteria. Methodology framework is proposed based on the basis. The framework clarifies the goal and tasks of design basis maintenance, raises a systematic method that carrying out margin management during NPP daily operation and maintenance, controlling the modification and carrying out periodic safety review to maintain the design bases. It can provide a useful reference for how to carry out design basis maintenance in operational nuclear power plants.
Implementation Challenges and Proposed Suggestions for Nuclear Material Accountancy Management in Spent Fuel Reprocessing Plant
Authors: He Lixia , Bai Lei , Miao Qiang , Yang Qun 
Published: November 2022

Abstract
Nuclear material license holder should develop a nuclear material accounting management system and achieve closed balance accountancy. According to regulations requirements, nuclear material accountancy implementation is based on physical inventory and material measurement. Generally, the spent fuel reprocessing plant operates uninterrupted, its annual throughput of nuclear materials is huge, there are different kinds of measurement and analysis techniques configured combing the reprocessing process. So the closed balance approach of nuclear materials accountancy is a great challenge for reprocessing plant holders. In order to improve the accuracy and reliability of nuclear material, enhance the timeliness of abnormal detection, near-real time accounting prospect of spent fuel reprocessing plant was researched. In this paper, the crucial aspects affecting closed balance accountancy was discussed, the factors such as the head-end receiver-shipper inventory difference, the on-line process monitoring accuracy, the applicability of the international target value of nuclear material measurement uncertainty, and the nuclear material balance model of reprocessing plants. As summarized, proposed suggestions and solutions for nuclear material balance in spent fuel reprocessing plant was put forward on the end of the paper.
Experimental Research on Drag Characteristic of Air Flow Path for ACP-100 Passive Containment Air Cooling System
Authors: Yu Mingrui , Wang Hongliang , Liu Zhuo , Feng Yu , Han Xu , Yuan Yidan 
Published: November 2022

Abstract
Passive containment air cooling system (PAS) is an important special safety system for the small modular pressurized water reactor ACP-100. The heat is transferred to the environment by the natural draft air cooling under accident conditions. The flow loss of the air flow paths has an important influence on the heat transfer power of PAS. A one-sixth scale model of PAS air flow path was constructed which was a 90° wedge angle sector of the cylindrical symmetric passages. The pressure distributions at different sections of the model were measured, the drag coefficients were analyzed. The experimental results showed that the total drag coefficients of the four different air inlet paths were almost the same, and they showed a decreasing trend with the increase of Reynolds number. The sleeves had the largest drag coefficient, accounting for about 26%, followed by the air inlet, chimney outlet, the turning and the corridor. The other components had little effect on the pressure loss. Afterward, the aerodynamic modification was carried out, the modification for the turning could reduce the total drag coefficient by 4.5 %. The experimental results of this paper provide necessary data support for the ACP 100 engineering design.
Event Tree Analysis for Material Relocation on Core Catcher in a Sodium-Cooled Fast Reactor
Authors: Yamano Hidemasa , Kubo Shigenobu , Kan Taro , Shibata Akihiro , Hourcade Edouard , Dirat Jean François 
Published: November 2022

Abstract
For an in-vessel retention strategy, severe accident analysis needs to be input the condition of loading on a core catcher based on physically compatible assumptions. This study has conducted event tree analyses emphasizing key phenomena in the lower plenum based on available experimental knowledge as well as design measure effectiveness. Criticality assessment has also been performed for the event tree analysis. In this paper, the approach to event tree development and the scope of the event tree analysis are described with key points on the loading conditions on the core catcher. For the analytical conditions, two loading conditions on core catcher were given as bounding and conservative cases. For the important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability of molten core materials on the core catcher. In this study, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the loading condition on the core catcher and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.
High Power Electron Beam Dump at DALS
Authors: Tao Kai , Huang Liming , Ding Yadong , Zou Zhenhai , E Dejun 
Published: November 2022

Abstract
The electron beam dump for Dalian Advanced Light Source (DALS) is designed to absorb 15 kW of electron beam power at beam energy up to 120 MeV. The DALS accelerator produces an electron beam with very small beam size of up to 100 μA average current. The resulting beam power, up to 15 kW at 120 MeV, and the very high beam power density, pose challenging problems for beam dump design.
High power dump with water cooled has been developed for DALS. In the dump, most of the beam power is finally absorbed in water and taken out from the dump.
The core of the high power electron beam dump is designed to be constructed from an aluminum alloy using a cylindrical geometry with fins arranged around to promote the heat transfer. The cooling water is forced by a cooling pump, to cool the core of the dump through the cooling channels.
The beam is stopped in the dump involving a high production of neutron and gamma radiation and activation of its surface. A shield has been designed to attenuate both the radiation produced during accelerator operation and the residual radiation.
Design details for the dump, including radiation shielding calculations, thermal analysis are presented.