bluuebunny/crossref_metadata_2025_split huggingface dataset:
Coupled OpenMC/CTF to VERA Core Physics Benchmark Problem 6
Authors: Zhang Chang , Wang Ruixiang , Guo Hui , Gu Hanyang , Xiao Yao 
Published: November 2022

Abstract
The multiple-physics modeling has been demonstrated to be a high-fidelity and effective method for the analysis of reactor core physics and thermal-hydraulics. OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. In this work, OpenMC and CTF are coupled for the analysis of light water reactor fuel assembly. OpenMC provides axial and radial fuel pin normalized power distribution to CTF, and CTF gives the fuel temperatures and coolant properties to the neutronics simulation of OpenMC. The windowed multipole temperature method is used in OpenMC to match the accurate temperature distribution of fuel rods and coolants obtained from CTF. In this study, the OpenMC/CTF is applied and validated using the Virtual Environment for Reactor Applications (VERA) core physics benchmark problem 6, from the Consortium for Advanced Simulation of Light Water Reactors. The problem involves a three-dimensional fuel assembly in Hot Full Power conditions. The Keff eigenvalue, pin power distribution, fuel temperatures, and coolant properties are obtained and compared with VERA’s reference results (MPACT/CTF). Our converged results showed good consistency with the reference solution, eigenvalue differences agreed within 183 pcm, axially-integrated normalized radial fission distribution difference agreed within +1.4%/−1.9%, local volume-averaged fuel pin temperatures agreed within +26.3°C, and local subchannel exit coolant temperatures agreed within +3.6°C/−0.4°C. These preliminary solutions prove that OpenMC coupled to CTF method has shown high-fidelity results in the three-dimensional fuel assembly neutronics and thermal-hydraulics coupled problems.
Application of Genetic Algorithm in Pressure Control of Pressurizer
Authors: Shan Yuzhong , Xing Kunpeng 
Published: November 2022

Abstract
The pressure control function of Pressurizer in the PWR nuclear power plant is one of the safety functions of the nuclear power plant. The traditional pressure control adopts the PID control method and manually adjusts the three parameters (Kp, Ki and Kd) to minimize the difference between the actual pressure value and the set value, thereby achieving the purpose of setting the three parameters (Kp, Ki and Kd) and determining the control model. The method of manually adjusting the three parameters (Kp, Ki and Kd) has low work efficiency, and generally requires several attempts to find parameters that meet the expectations, and depends on the work skills of the personnel.
In this study, Genetic Agorithm was used to convert the three parameters (Kp, Ki and Kd) into the three gene coding sequences, and the three parameters (Kp, Ki and Kd) were quickly determined through individual operation operators (selection, crossover and mutation). The research shows that the efficiency of parameter adjustment by genetic algorithm is much higher than that of manual parameter adjustment, and the optimization result is better, the control system optimized by the genetic algorithm can reach the steady state faster, has less overshoot and shorter rise time. This achievement can be considered for subsequent application in the nuclear power plants affiliated to CGNPC.
Preparation and Characterization of Radioactive Aerosol Protective Nanofiber Membranes
Authors: Li Jiawen , Chen Mingyi , Li Li , Liu Feng , Gao Yufei , Zhu Jian , Zhan Jie , Chen Jian , Zeng Yuan , Chen Rouxi , Wang Hsing-Lin 
Published: November 2022

Abstract
Radioactive aerosols are highly hazardous aerosols containing radionuclides. Inhalation of radioactive aerosols can lead to serious internal exposure hazards to human body. If discharged without proper treatment, it will also harm the working environment and even the ecological environment. Therefore, radioactive aerosol protection is a significant part of environmental protection and personnel protection in the nuclear field. However, majority of existing protective fabrics for radioactive aerosol filtration always meet the trade-off among filtration capacity, mechanical properties and air permeability. In this study, nanofiber layers were prepared by electrospinning technology using TPU, PVDF, PVA polymer materials and electret materials SiO2 as spinning materials. Composite membranes, prepared by coating different nanofiber layers on the PET non-woven fabrics substrate, were investigated. The results show that the 12wt% TPU nanofiber membrane has a three-dimensional spatial hierarchical structure. Its ultra-fine fiber diameter with small pore size greatly enhances the PM capture ability (PM0.3 filtration efficiency 99.99%); and the beaded spatial structure is beneficial to reduce the air resistance to 299 Pa (flow rate 95 L/min). Meanwhile, TPU nanofiber membrane has high extensibility, and it is superior to PVDF and PVA composite membranes in mechanical properties after thermal compounding. Appropriate content of SiO2 can improve filtration performance. The study shed light on developing electrospun nanofiber for radioactive aerosol protection, which can be used in the purification of ambient air in nuclear facilities, or as a high-performance fabric for radioactive aerosol protective clothing.
Development of Containment Spray Removal Model for Integrated Severe Accident Analysis Code ISAA
Authors: Li Jishen 
Published: November 2022

Abstract
The fission products released from the core fuel in severe accidents are mainly transported in the form of aerosols and exist in the gas phase in the containment. Once the containment fails, it will cause radioactive leakage. As an important mitigation measure for severe accidents, containment spray can effectively reduce the containment pressure and aerosol concentration, as well as the release of radioactivity to the environment. In order to analyze the thermal-hydraulic and aerosol behavior in the containment after accidents, a containment spray removal model was added to the integrated severe accident analysis code ISAA. The removal mechanism considers spray droplets washing aerosols by inertial impaction, interception collection, diffusion, thermophoresis, and diffusiophoresis. The containment spray removal model was coupled with the containment thermal-hydraulic module in ISAA to calculate the rate constant for spray removal. Benchmark experiments were selected for assessment of the improved code. The comparison shows that the results of the thermal-hydraulic response and aerosol mass distribution simulated by ISAA are consistent with the experimental data trends. The improved code can accurately simulate the thermal-hydraulic response and aerosol mass distribution during aerosol removal by spray droplets. Implementation of the containment spray removal model in ISAA satisfies the analysis of aerosol behavior in the containment.
The Research on the Interception Engineering Layout of Active Intercepting and Guiding in Water Intake Open Channel of Nuclear Power Plant
Authors: Yang Jia , Cao Ranran , Bai Wei 
Published: November 2022

Abstract
At present, most intercepting facilities in water intake open channel of nuclear power plant are passive. Moreover, the correlation between the layout of intercepting facilities and hydraulic conditions of water intake open channel is not considered. Hence, the layout of intercepting facilities is unreasonable. The cooling water intake system of nuclear power plants faces risks. In this paper, The flow field characteristics of water intake open channel under typical sea hydrologic conditions were studied. Then, the migration path and aggregation rule of floating or suspended objects are simulated along with the flow. The blockages-removal effect of floats and marine organisms are studied after different intercepting engineering measures are carried out in water intake open channel. The technical route combining mathematical model and physical model test is adopted. Finally, the principle and scheme of the interception engineering layout of active intercepting and guiding are proposed. The optimized interception engineering layout can effectively guide suspended or floating blockages to the outside of water intake, or guide blockages in water intake open channel to the area without disaster risks, where the blockages are also easy to be salvaged. It can significantly reduce the large quantity of blockages into the cooling water intake system, which will also reduce the blockage loads for all intercepting facilities. Thereby, the defense capability of nuclear power plant cold source system against potential water intake blockages is greatly improved.
A Method to Evaluate Defect Size of Failed Fuel Rods in CGN 1000MWe PWRs
Authors: Fu Pengtao , Li Zhijun 
Published: November 2022

Abstract
Fuel cladding is the first barrier to confine the release of fission products to the environment during normal operation. Once the fuel rods defect, the fission product accumulated in the pellet-cladding gap of failed fuel rods will release into the primary loop and lead to high activity levels in a pressurized water reactor. It is necessary to determine the status of fuel failure to avoid its degradation in the operation.
This paper introduces the mechanism of the generation and release of fission products in fuel rods and the primary loops in a pressurized water reactor, and provides the theoretical method of the regression between release-to-birth R/B and decay constant λ of radio-iodine. It has been verified by one typical case of fuel failure in CGN 1000 MWe PWR, but the post-irradiation examinations in hot cell reveal that six defects exist in one failed rod and the actinides have an obvious release from the failed rods due to secondary hydriding. The actual status of defect size of failed rods is much more complicated than expected in most analysis codes and modelings and it is suggested to introduce some parameters, e.g. the equivalent defect size, to describe the status of fuel failure in future research. The related operation experience in CGN has also been presented to help to better understand the status of fuel failure in PWR units.
Partially Averaged Navier Stokes Simulations of Turbulent Thermal Mixing in T-Junction
Authors: Bilal Ashhar , Gao Puzhen 
Published: November 2022

Abstract
High Cycle Thermal Fatigue is a major degradation mechanism associated with the nuclear power plant piping where hot and cold streams of fluids mix together with a phenomenon called turbulent thermal mixing. It induces significant thermal fluctuations on the piping wall called thermal stripping. T-junction has been known as a typical component susceptible to high cycle thermal fatigue. Accurate prediction of turbulent flow and thermal fields at the wall is necessary to predict the thermal fatigue degradation in such piping systems. Computational Fluid Dynamics (CFD) with advanced turbulent modelling is required to predict the flow and temperature fluctuations at the wall. Wall resolved LES is typically known to give good results for velocity and temperature fluctuations for such a flow phenomenon. However, LES is computationally expensive, and the cost of computation increases with the higher Reynolds number flows. Partially Averaged Navier Stokes (PANS) is a Hybrid turbulence Model. It is a bridging method from Reynolds Averaged Navier-Stokes (RANS) to Direct Numerical Simulation (DNS). PANS is usually controlled by two resolution control parameters, fk (unresolved to total kinetic energy) and fε (unresolved to total dissipation), fk = 0 being the DNS and resolving all scales of turbulence, and fk = 1 is RANS simulation. In this study PANS simulations for the Vattenfall T-junction Benchmark case are performed. fk is considered constant throughout the domain. Average and Root mean square temperature and Velocity profiles are compared with experimental data. It is concluded that PANS model can simulate turbulent thermal mixing phenomenon in a T-junction.
Summary of the History of Improvement and Optimization of Fuel Management Strategies for Two 1000 MWe PWR Units
Authors: Xiaoling Zhou , Zhixian Xu , Bingquan Zhao , Wen Li , Shengdong Zhong , Jinghui Gao 
Published: November 2022

Abstract
This paper takes a certain nuclear power plant (NPP) which consists of two 1000 MWe PWR units as the research object, and summarizes various fuel management strategies experienced since it started commercial operation in 2002 and 2003. These fuel management strategies include:
1) high-leakage annual refueling, refueling fuel enrichment is 3.2% and cycle length reaches about 270 EFPD;
2) fuel type mixed refueling (AFA2G and AFA3G mixed), refueling fuel enrichment increases from 3.2% to 3.7% and cycle length reaches about 320 EFPD;
3) low-leakage 1/4 refueling, refueling fuel enrichment increases from 3.7% to 4.2% and cycle length reaches about 316 EFPD that fuel economy has been greatly improved;
4) low-leakage 18-months refueling, refueling fuel enrichment increases from 4.2% to 4.45% and the refueling cycle has been extended from annual refueling to an average of 16 months;
5) two-enrichment refueling, refueling fuel enrichment is extended from single 4.45% to two enrichment 4.45% and 4.00%, which improves the flexibility of fuel management to well meet the load shedding needs.
This paper reviews each fuel management strategy’s general refueling pattern, main design parameters and actual operating parameters of the core, fuel economy and the feedback of related problems, finally further suggests a few optimizations of future fuel management strategies.
Study on Different Ways for Selecting NPP Design Extension Conditions
Authors: Ma Yuan , Feng Churan , Ma Chao 
Published: November 2022

Abstract
With the further improvement of the safety requirements, the Design Extension Conditions need to be fully considered as an important design requirement in the design of advanced nuclear power plants. For Chinese new nuclear power plants and future new reactor types, in principle, a set of Design Extension Conditions must be obtained on the basis of engineering judgment, determinism and probability theory evaluation to further improve the safety of nuclear power plants. In current practice, PSA methods and models are often used to identify and determine extremely unlikely events and multiple failure events in the selection of the Design Extension Condition sequence list, while considering the design extended conditions given by determinism and engineering judgment. However, even with the same frequency screening values, the way in which the PSA model is used can be different, so it is necessary to study the effect of different ways on the results of DEC selection. In this paper, combined with the determination process of the DEC-A list of a third-generation reactor type, two different ways of identifying the DEC-A event sequences based on the PSA event sequence and based on the cutset-importance, are described, and the results are compared and discussed.
Implementation and Verification of the DAGMC Module in Monte Carlo Code RMC
Authors: Shen Pengfei , Liang Jingang , Liu Shichang , Wang Kan 
Published: November 2022

Abstract
The ability to perform shielding simulations is one of the advantages of Monte Carlo codes. However, the shielding structure of an advanced nuclear reactor or other facilities is usually designed in CAD geometry and cannot be easily transformed into CSG geometry. Thus, the research for performing shielding simulations in CAD models in Monte Carlo codes is of significant importance. DAGMC is an advanced toolkit used to simulate the Monte Carlo particle transport process in CAD models. Based on the RMC (V3.5) code developed by the REAL lab of the Department of Engineering Physics of Tsinghua University, this paper implemented the DAGMC module to enhance the ability to handle CAD models. The implementation scheme supports weight window parameters and cell importance parameters’ definition to carry out shielding simulations in CAD models. We have done an extensive investigation on various problems to exemplify its validity. In the CAD-based deep penetration calculations, the adaptive variance reduction (AVR) method is adopted and improves the FOM by 601 and 207 times for single and multiple material problems.
Study on Nuclear Safety Management Based on Multiple Nuclear Power Plants Experience Feedback Management
Authors: Zhang Shuang , Xiaoyan Sun 
Published: November 2022

Abstract
Nuclear power plant experience feedback management includes event reporting, screening, analysis, corrective action management and assessment. In the early stage of nuclear power development in China, experience feedback was conducted based on single nuclear power plants. After years of development, the number of nuclear power plants increases year by year, institutes serving multiple nuclear power plants (multi-plants) develop multi-plants experience feedback management system, to improve event management quality, reduce events recurring probability, promote good practice, carry out experience feedback work effectively, reduce the impact of repeated events, continuously improve nuclear safety quality management level, take the implementation of experience feedback and the assessment of its effectiveness as a priority of nuclear safety management. According to the experience feedback assessment of nuclear safety field - operations personnel management, the vulnerabilities of in nuclear safety management have been identified, including insufficient knowledge and substandard operation by operations personnel, and inadequate monitoring of nuclear safety related equipment by operators. The following measures are recommended: improve the on-the-job training outline of operations personnel, and regularly carry out professional knowledge training of machinery, electrical, instrumentation, etc; formulate a special inspection plan for work standardization, and incorporate the special inspection into the task supervision system for tracking management; establish a tracking process for key equipment defects in nuclear safety issues to ensure the timeliness and effectiveness of tracking nuclear safety issues. This article mainly focuses on multi-plants experience feedback organization, information system, program management and nuclear safety indicators, in order to optimize nuclear safety management and prevent the degradation of nuclear safety level.
Study on Maintenance Strategy Optimization of Nuclear Power Plants After the Implementation of Maintenance Rule
Authors: Xiliang Ma , Yongkui Xie 
Published: November 2022

Abstract
After NNSA issued “Technical Policy for Improving Maintenance Effectiveness of Nuclear Power Plants (Trial)” in 2017, relevant units in China actively carried out pilot maintenance rule and gradually carried out optimization research on related maintenance activities. Suzhou Nuclear Power Research Institute Co., Ltd. has participated in most maintenance rule related projects in China. This paper introduces the development of maintenance rule in China and the current implementation of maintenance rule for nuclear power plants in China. Combined with the current situation of technical specification and the construction of configuration risk management in China’s nuclear power plants, the optimization of maintenance strategy based on maintenance rule is studied, which lays a foundation for more risk-informed applications, in order to strengthen the safety of nuclear power plants more effectively and reduce the unnecessary operation burden without affecting safety.
Structure Control of a HLW Immobilized Zirconolite Glass-Ceramic Matrix
Authors: Li Haiqing , Pan Yutong , Gao Chao , Wang Shuming 
Published: November 2022

Abstract
Resources and environment are two crucial (key) elements for the sustainable development of human society. This paper aims to explore a new glass-ceramic system for the development of high-level radioactive solid-waste vitrification technology, clarify the function of multielemental radionuclides (heterogeneous ion) on the structural and chemical durability characteristics.
By changing the crystallization temperature and crystallization time, the heat treatment system of the zirconolite-based glass ceramic waste form was optimized; the K value method was used to obtain the quantitative analysis of the crystal phase; CeO2 was used as a simulated nuclide to explore the effect of CeO2 doping on the phase. The influence of composition, and the valence distribution of cerium in the glass ceramics was analyzed by XPS. The results show that the nucleation temperature is 810 °C for 2 h, and then the crystallization temperature is 950 or 1000 °C for 2 h, the glass ceramics containing only the 2M zirconolite and residual glass. CeO2 doping will lead to the increase of the lattice constant. The ratio of Ce4+ to Ce3+ in the waste form with different CeO2 content is maintained at 6:4. The normalized leaching rate of the sample with the highest doping content is 5.949 × 10−6 g·m−2·d −1.
Current Situation and Development Trend of Decommissioning of Nuclear Power Plants
Authors: Guochang Cao , Yunding Kang , Zhihua Li 
Published: November 2022

Abstract
The complete life of a nuclear power plant consists of three stages: construction, operation and decommissioning. In the last stage of the operation life of nuclear power plants, they have to face the problem of decommissioning. The safe decommissioning of nuclear power plants is the guarantee for the sustainable development of nuclear power. After the termination of the operation of nuclear power plants, in order to ensure the safety of the public and the environment, the plant site should be purified, demolished and cleaned to reach the restricted utilization level or unrestricted opening. This paper studies the operation and decommissioning status of nuclear power plants from both international and domestic levels, and analyzes the real time cycle and development trend of decommissioning projects. The research shows that in the international nuclear power plant decommissioning market, the nuclear power pioneer countries have successively established national laboratories or research centers to carry out research on the decommissioning technology of nuclear power plants, developed a large number of advanced technologies and tooling equipment, and formed their own technical characteristics and technical advantages; the nuclear power plant decommissioning is increasingly researched in China, but the decommissioning practice has not been started yet in this country. The industry and relevant parties should carry out systematic construction for the decommissioning of nuclear power plants from the aspects of capital sources, regulations, standards, technical reserves, organizational structure and talent reserves.