bluuebunny/crossref_metadata_2025_split huggingface dataset:
Preliminary Investigations on the Effect of Cooling Water Boundary Conditions on PCS Performance
Authors: Fei Likai , Gao Bin , Fan Pucheng , Yuan Haowei , Wang Husheng , Yang Fuming , Du Lianyue , Zheng Lin 
Published: November 2022

Abstract
The coverage of cooling water film outside the containment shell is one of the most important boundary conditions for Passive Containment Cooling System (PCS) of AP/CAP reactor. The reduced coverage of the water film reduces the heat removal capacity of the containment vessel, which may result in the temperature and pressure within the containment can not be kept at a safe level. In this paper, the effect of cooling water flow rate and water film coverage on the heat removal capacity of PCS was studied experimentally under the loss-of-coolant accident (LOCA). Based on the CAP1400 passive containment water distribution test facility, the cooling water boundary conditions were set according to the combination of different cooling water flow rate and water film coverage rate, and the mass energy was released corresponding to the typical design basis accident condition. The response of parameters within the containment obtained from the test was compared and analyzed with the design condition. The experimental results show that the water film coverage rate has a greater impact on the parameters response within the containment than that of the cooling water flow rate.
Neutronics Design and Modelling of a Nuscale-Based Small Modular PWR
Authors: Qiu Yanfei 
Published: November 2022

Abstract
Small Modular Reactors (SMRs) are characterized by high safety, small size, and modular construction. In this paper, an integrated design of a small nuclear reactor, with thermal power of 200 MW and fuel enrichments below 5%, is designed and analyzed to study the burnup behaviour and reactivity coefficients. The core has been simulated by using the transmission neutron Monte-Carlo program Serpent. The results establish that the core is able to provide an 18-month operating cycle. Comparisons of different enrichments and loading patterns are carried out to explore potential future improvements. The relationship between the refuelling cycle length, geometric size of the core, fuel enrichment, arrangement of the burnable absorber and operating power is demonstrated. It is also confirmed that the SMR design has a sufficient negative feedback coefficient to ensure nuclear safety.
Numerical Investigation on Thermal-Hydraulic Characteristics of Supercritical CO2 in Multiple Compact Channels
Authors: Liang Guohu , Chen Deqi , Jin Feng , Hu Lian , Bu Shanshan , Liu Hanzhou 
Published: November 2022

Abstract
Printed circuit heat exchanger (PCHE) is promising in the application of energy conversion system with its high compactness and efficiency. In this paper, the flow and heat transfer performance of supercritical carbon dioxide (SCO2) PCHE are numerically investigated with multiple channels and different channel lengths. The results show that the heat transfer performance of the Zigzag channel PCHE is increased more than 11% comparing to the wavy channel, while it brings more flow resistance. The pressure drop in the Zigzag cold and hot channels was about two and four times of the wavy channel, respectively. It is recommended for the wavy channel PCHE on account of its better overall performance factor than zigzag channel PCHE. Through the analyses of flow characters among multiple channels, it shows good homogeneity of flow distribution for the flow distribution coefficient in each channel close to 1.0, which is able to effectively avoid uneven flow distribution resulting in high local temperature of the heat exchanger. Furthermore, the overall performance of PCHEs with different core lengths are compared, presenting nearly 10% difference on the thermal efficiency.
Coolability Evaluation of the Debris Bed on Core Catcher in a Sodium-Cooled Fast Reactor With a Whole Vessel Model
Authors: Yamano Hidemasa , Kubo Shigenobu , Sasa Kyohei , Shibata Akihiro , Hourcade Edouard , Dirat Jean François 
Published: November 2022

Abstract
This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The DHRS consists of two circuits of direct reactor auxiliary cooling system and one circuit of reactor vessel auxiliary cooling system. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most fuel accumulates on the core catcher in a short term.
Light Water-Cooled High-Performance Research Reactor Physics Design and Analysis
Authors: Guo Zihao , Wang Guang , Su Genghua , Shi Xiuan 
Published: November 2022

Abstract
Research Reactor can provide a platform for advanced reactors’ fuel and material irradiation test, medical isotopes production and neutron science research, which plays an important role in nuclear energy and technology domain. Nowadays, most of research reactor in worldwide are reaching the retirement and decontamination stage. For the increasing demand of the material irradiation and medical isotopes in the world, several new and advanced research reactors should be constructed to replace the void in the future. Differed from the design scheme of the traditional PWR, that of research reactor focus on increasing the power density as high as possible for the huge amount of neutron flux. According to the full reference of the existing advanced research reactor, this article starts a reactor physics design scheme and illustrate a unique reactor physics design. The neutron flux, length cycle and instrument flexibility of this design are better than the reference design.
Construction of the Failed Fuel Information Database Platform
Authors: Liu Xin , Li Changzheng , Wang Yang , Zhang Hongyang , Deng Yongjun 
Published: November 2022

Abstract
The fuel failures occur in the process of design, manufacture and operation the whole time, which will affect the reliability of the fuel. In order to study the method of fuel reliability analysis, it is necessary to establish a database platform of failed fuel information. Failed Fuel Information Database Platform (FAFID) platform established for the first in CGN with demand-based peer group. The platform adopts Browser/Server architecture, Tomcat for web application server software, MySQL for database server software and IE browser for client. After the completion of the platform, it realizes the function of saving the key parameters of fuel assembly/fuel rod in the process of design, manufacturing and operation, and the information can also be classified. The results of poolside inspection and hot cell examination are also arranged in the plate model block. In addition, the platform can also compare the differences in key fuel data, intuitively display the current situation of failed fuel in the form of various statistical charts and tables, so as to help technicians analyze the causes of failed fuel fast and efficient. At the same time, it can also help business personnel better predict and analyze the possible sources of failure. The platform has become a tool for business personnel to comprehensively analyze the causes of failed fuel.
Validation of TULIP via ICSBEP Critical Benchmark
Authors: Chen Wenjie , Du Xianan , Zheng Youqi , Wang Yongping , Wang Rong , Wu Hongchun 
Published: November 2022

Abstract
NECP-SARAX is a neutronics code system for fast spectrum reactor developed by Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In previous work, NECP-SARAX has shown high performance on fast spectrum reactor analysis. Recently, neutron-moderating materials are employed in advance reactors design where the pure fast spectrum is softened to intermediate and thermal energy spectrum. Due to the larger fission cross-section below the fast energy range, the volume of reactors reaching criticality can be reduced. Compared that in fast energy range, the temperature reactivity negative feedback resulting from the Doppler effect in thermal spectrum range is more significant, which is conducive to the safety and miniaturization of the reactors. To meet the design requirement of this kind of reactor, the assembly-wise neutron spectrum calculation module TULIP of NECP-SARAX has recently been extended to generate the cross sections for both the thermal and fast spectrum reactor system. Therefore, in this paper, the validation works of TULIP code have been performed. In order to systematically validate the accuracy of TULIP code, a series of benchmarks with neutron-moderating material are selected from the ICSBEP, such as HEU-MET-FAST-001-002,HEU-MET-FAST-027-001, U233-SOL-THERM-015-001. The numerical results showed that the TULIP code had accurate neutron spectrum calculation capability for the advanced nuclear reactor design.
Feasibility Study on Power Ramp Test Under Atmospheric Pressure and Ordinary Temperature
Authors: Wei Xiangyu , Zhang Wenhua , Zhao Yingchun 
Published: November 2022

Abstract
Pellet-to-cladding mechanical interaction is an important physical phenomenon during reactor power change as well as a multi-phenomenal fuel rod failure mechanism involving stress, strain and material irradiation properties. In order to avoid the failure caused by PCMI, a large number of power ramp tests have been carried out by international organizations over the past decades. The typical way of power ramp test in a high temperature and pressure loop requires stringent test capabilities and high test costs. The key parameters of the PCMI phenomena are stress and strain of cladding, which are generally chosen as the evaluation indicators of PCMI. If it is possible to simulate the pellet-to-cladding contact state, i.e. the gap between pellet and cladding under basic irradiation power level in atmospheric pressure and ordinary temperature environment, then the high stress and strain state of the cladding during the subsequent power ramp test could also be simulated in the same environment, which means lower test costs and test loop requirements. Therefore, a sensitivity analysis using the fuel rod performance analysis code RoPE, was carried out on factors such as initial pellet-to-cladding gap and fuel densification in the power ramp test rod design. By adapting the manufacturing parameters of the test rod and coolant conditions, the high stress and strain state of the cladding could be simulated in the test environment at normal temperature and pressure. The sensitivity analysis provides a theoretical basis for conducting power ramp tests in an atmospheric pressure and ordinary temperature loop.
Scaling and Designing Activities of Integral Test Facility for HPR1000 Reactor
Authors: Lu Donghua , Li Liangguo , Su Qianhua , Xing Jun 
Published: November 2022

Abstract
The integral test facility is very useful to study the behavior of the pressurized water reactor (PWR) at accidents. As more and more passive safety techniques were adopted in the reactor system, the integral effect test facilities acted very important role to verify these techniques and the prediction of software. An integral effect test facility for the HPR1000 reactor was designed and constructed based on the scaling analysis. The scaling criteria were derived on the model of natural circulation and blowdown of the constant bulk volume in the primary system. The phenomenon were identified and ranked to ensure the scaling can reproduce them in the test facility as the same as the prototype does. The height ratio is 1:4 and the diameter ratio is 1:6 for the test facility. Totally 177 simulators were used to simulate the thermal hydraulics of the fuel assemblies in the practical reactor core. This makes the core keep the same array as the prototype. Each simulator has one electrical heater which power is controlled by the computer. The power of the core has axial cosine profile and three radial zones to reproduce the physical non-uniform distribution in the reactor core.
Probabilistic Safety Assessment on Water Intake Blockage of Nuclear Power Plant
Authors: Wu Yan , Feng Bingchen , Meng Gan , Zhang Xiaoming 
Published: November 2022

Abstract
The marine organism outbreak can lead to water intake blockage, system degradation, equipment failure, and even worse, trip or lose of heat-sink event. According to a survey conducted by the American Electric Power Research Institute (EPRI) in 2009, among the 77 plants surveyed, more than 55% of plants suffered unplanned shutdown or unplanned power reduction events caused by water intake blockage. In recent years, several water intakes blockage events caused by marine organism were reported by domestic nuclear power plants. which leaded to power reduction or shutdown event and heat sink SSCs (systems, structures, and components) degrade or failure. These events greatly threaten operation management, equipment maintenance and emergency response of nuclear power plants. Hence, it is necessary to carry out detailed risk assessment on the water intake blockage event caused by marine organism outbreak.
Aiming at a large-sized pressurized water reactor (PWR) nuclear power plant, this paper builds a realistic model to analyze the risk caused by marine organism using PSA (probabilistic safety assessment) method, and obtains quantitative results of the unit. This paper also identifies the problems and deficiencies in plant’s response to the water intake blockage events, and provides us with risk insight to improve the safety level of nuclear power plants combining the analysis results.
The Research and Application of Configuration Risk Management in Daya Bay NPP
Authors: Zhang Kuan , Cheng Changsheng , Guo Dongyuan , Li Qiongzhe , Zhang Heng 
Published: November 2022

Abstract
National Nuclear Safety Administration issuance “The Technical Policy of Nuclear Power Plant Configuration Risk Management” in December 2019 to support the risk management of multiple system/equipment failures and the flexible operation of nuclear power plant (NPP). Subsequently, Suzhou Nuclear Power Research Institute PSA teams and Daya Bay NPP nuclear safety advisory/Operation operator/planning Engineer established a joint project team to study the implementation of configuration risk management. The configuration risk management system framework and operation process have been determined through the in-depth discussion of multiple disciplines and departments, and the trial operation of risk management system has been started in January 2021 to optimize iterative software and implementation processes. This paper introduces the composition of Daya Bay configuration risk management system, the responsibilities of each discipline in the implementation of configuration risk management, the establishment method and results of risk threshold and risk management matrix, the functions of operation and maintenance risk assessment tools, the characteristics of risk calculation engine, the implementation process of operation configuration risk management and maintenance configuration risk management, the problems found during the trial operation, the management measures for subsequent upgrades, so as to provide reference for other NPPs to carry out the configuration risk management.
Overview of AFCEN Non-Linear Benchmark in China – Improving Rules for Vessel Ratcheting Evaluation
Authors: Tan Yu , Shao Xuejiao , Liang Enming , Zuo Shuchun , Lv Feng , Metais Thomas , Liu Han 
Published: November 2022

Abstract
The French nuclear standard, RCC-M 2017 edition, has seen the inclusion of an Appendix on non-linear calculation methods: the Appendix ZC. This non-mandatory Appendix encompasses a set of rules and guidelines which can help the designer perform non-linear analyses. It includes the definition of models adequate to use, fatigue Ke factor optimization methods, etc...
In fatigue analyses per the RCC-M B-3200 paragraph, rules on thermal ratcheting limits are exposed in paragraph B-3234.8. These rules aim at limiting any thermal ratcheting of the structure and the criteria exposed are sometimes difficult to meet, even though there is no threat to safe operation of the concerned equipment.
Non-linear analyses in this case could be used to help meet the criteria but would need extra rules to be defined: these are considered to be added to the Appendix ZC.
In early 2019, an international benchmark effort was launched on this topic. The case is a typical nozzle geometry subjected to both thermal and pressure transients.
In China, some main nuclear entities have taken part in this benchmark effort: NPIC, SNPI, CNPDC and CNPE, along with the coordination of EDF and Yuansuan.
• let participants get familiar with the benchmark and performing linear elastic calculations along with simplified RCC-M code evaluation methods. The outcome is a participant comparison of the respect or not of the code criteria.
• perform an analysis based first on one’s own parameter identification and an Armstrong-Frederick model (with 1 and 2 parameters) and then perform the comparison with a common set of parameters.
• use a defined cyclic stress-strain curve including full hysteresis loops to derive a common identification methodology and compare results.
This paper aims at providing the public with a progress of the benchmark case (using ANSYS and code_aster software), the Chinese participants findings, and recommendations for the structural integrity community on the first two bullet points above.
A Location Algorithm for Orphan Radioactive Source
Authors: Xiao Qianxi , Cai Jiejin 
Published: November 2022

Abstract
In order to reduce the harm to public safety caused by the loss of radioactive sources. A source searching method is proposed to locate orphan gamma source in the environment with obstacles and radiation shielding. The method is divided into two parts: predicting the probability of a radioactive source by Gaussian process regression (GPR) and path planning for the mobile robot using the probabilistic roadmap (PRM) method. The mean function and covariance function of the Gaussian process need to be selected in advance and the hyperparameters are calculated in the training process. Predicted values of GPR are given in the form of Gaussian distribution, which make us able to calculate the probability that there is a radioactive source in the grid and construct a probability map to guide the robot. The path connecting the start grid to the target grid is planned by the PRM method in a very short time, the cost function of A* algorithm in the query phase controls the pattern of the robot. At the beginning of the search, the robot prefers to move to areas with more information, after data above the threshold is detected, the robot switches to the tracking mode, accesses to the radioactive source rapidly. The effectiveness of the proposed method is verified in a simulation experiment in which a 10m × 10m radiation field with five walls is simulated. The searches of gamma point source are all successfully realized in 5 different cases and compared with the total number of discrete grids, the method uses only a small part of the grid data to realize the location. Overall, the results show that the proposed method is efficient for searching and locating orphan radioactive source.
Study on the Influence of Power Reduction Process on the Leakage Rate of Steam Generator
Authors: Zhang Zhao , Yan Jun-Jie , Wang Wen-Bo , Chong Dao-Tong , Su Min-Hua , Wang Zi-Chun , Shen Rong-Fa 
Published: November 2022

Abstract
The steam generator tubes in PWR unit are prone to mechanical damage and corrosion hazard, tube cracks and steam generator leakage often occurs on alloy 600 tubes. Due to the better corrosion resistance of alloy 690, the tube leakage has never occurred in CPR1000 unit. However, based on the systematic study of the alloy 690 behavior in the PWR operation condition, cracks will also occur at the end of burnup life (EOL); in addition, whether the limit leakage value for steam generator operation originating from EDF units under full power operation is valid for CPR1000 unit should be investigated.
For the same critical crack size, in the process of unit power reduction, the change of the pressure difference between the primary and secondary sides of steam generator may cause the leakage rate to exceed the operating limit. An overall model for CPR1000 unit was developed, the changes were investigated of the pressure difference and leakage rate on the primary and secondary sides of the steam generator under different unit load reduction rates, the results showed that for the critical size crack at the bottom of the hot leg, during the load reduction transient process, the calculated leakage rate will exceed the limit value of 44 L/h in the Operating Technical Specifications (OTS) of CPR1000 unit. Finally the corresponding operation strategy was proposed.
Comparative Analysis and Research on Nuclear Power Regulation & Standard System and Standardization Works Among China, the U.S. and Japan
Authors: Liu Shangyuan , Wei Ran , Zhang Shaolin , Wu Feifei , Du Jian 
Published: November 2022

Abstract
As the world’s leading countries in nuclear power fields, China, the United States, and Japan have relatively sound nuclear power regulation and standard systems. The nuclear power standardization works have also been gradually improved and optimized with the iteration of good practices and the feedback of best experience. Nevertheless, it couldn’t be denied that the paths followed by these countries are not exactly same, due to diversities of state institutions and mechanisms, as well as inherently different national politics, history, technology, and culture. Making periodical revisions on development approach, each country establishes unique, well-applicable, self-consistent system framework and operating model.
With a systematic review, this paper will summarize the current R&D of the countries’ nuclear power regulation & standard system, as well as the latest states on the development of standardization works on both domestic and abroad scales, following by some fruitful comparative analyses. Using the literature research method and Differentiation Competitive Strategy methodology, this paper will also critically analysis advantages and shortcomings respectively, which would be helpful to understand the regulation and system operations comprehensively and objectively, for each country. Additionally, inspirations would be brought to seek an opportunity deepening extensive collaborations for the forthcoming nuclear power international standardization works. This paper will make comments and put forward suggestions focusing on strengthening R&D cooperation among multilateral countries. Ultimately, to improve the diversification and interoperability of nuclear emerging technologies and industrial chain, also accelerate global promotion and common prosperity, via the nuclear field standardization approach.