bluuebunny/crossref_metadata_2025_split huggingface dataset:
Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China
Authors: Chen Haiying , Wang Shaowei , Tian Xinlu , Liu Fudong 
Published: October 2021

Abstract
The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.
Validation of Evaluation Method of Feedback Reactivity for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors
Authors: Yoshimura Kazuo , Doda Norihiro , Tanaka Masaaki , Yamano Hidemasa , Igawa Kennichi 
Published: October 2021

Abstract
Negative reactivity caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In this paper, the numerical results of the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-302R and BOP-301) are discussed in order to validate the evaluation method of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD. During the transient of the ULOHS tests as one of the representative issues, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. By comparing the numerical results and the experimental data, the profiles of the increase of the core inlet temperature and the decrease of the reactor power calculated by Super-COPD were comparable with those of the experimental data. Applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS event. The refinement of the plenum model of the cold pool by taking account of the thermal stratification was indicated for the future work in order to improve the profile of the core inlet temperature.
Corrosion Behavior of Iron-Chrome Alloys in Liquid Bismuth
Authors: Takai Toshihide , Furukawa Tomohiro , Watanabe Shigeki , Ishioka Noriko S. 
Published: October 2021

Abstract
For the mass production of astatine-211, a promising radiopharmaceutical for cancer treatment, the National Institute for Quantum and Radiological Science and Technology has proposed the innovative “Liquid Bismuth Target System.” The target window in this system must be made from a material that resists the highly corrosive liquid bismuth environment. To meet this requirement, a promising target window material was selected in corrosion experiments performed in stagnant liquid bismuth. Based on knowledge of corrosion in liquid lead–bismuth eutectic gained during the development of fast reactors and accelerator-driven subcritical systems, FeCrMo–alloy, FeCrAl–alloy, and austenitic stainless steel (as a reference) were selected as the specimen materials. Experiments were carried out under saturated dissolved oxygen and low oxygen conditions, and the corrosion behaviors of the specimens were evaluated, mainly by scanning electron microscopy. The FeCrAl–alloy exhibited the most excellent corrosion resistance, followed by FeCrMo–alloy. Both materials are suitable candidates for the target window. Although austenitic stainless steel was less corrosion resistant than the former two materials, it is a likely applicable for the target window under appropriately limited operation conditions (such as irradiation current and exposure time) of the liquid bismuth target system.
Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel
Authors: Huang Jin Feng 
Published: October 2021

Abstract
After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR.
The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR.
We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity.
The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor.
In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.
Research on Nuclear Turbine Control and Protection System Based on DCS Integrated Technical Solution
Authors: Shi Guilian , Wang Jikun , Gao Jingbin 
Published: October 2021

Abstract
Turbine Control System (DEH) and Turbine Protection System (ETS) are important auxiliary systems for turbines. Normally DEH and ETS are supplied by turbine manufacturers. This paper investigates the operation experience in Nuclear Power Plants (NPPs) in China, there are many problems including information security, maintenance inconvenience, long supply cycle guarantee, high cost and so on in DEH and ETS. This paper analyzes the problems and tries to get the reasons why DEH&ETS have these problems. The Distribute Control System (DCS) technology has been widely used in nuclear power plant in now days, so this paper puts forward a technical solution based on the safety DCS and non-safety DCS platform to realize the DEH and ETS, and according to the feasibility analysis of products, and the test results based on engineering prototype, the solution can solve the problems of DEH and ETS effectively, some performances of the DEH&ETS have been improved, and the solution has a good prospect in further.
A Sodium-Cooled Fast Reactor Simulation System and its Application In Teaching Research Based on VPOWER Platform
Authors: Ji Chengzhi , Xie Biheng , Guo Xiaoyu , Han Wenbin , Hao Yisheng , Chen Junyi , Huang Shanfang , Wang Kan , Wei Hongbin , Liang Yanming 
Published: October 2021

Abstract
To ensure the safe operation of nuclear power plant, employees must complete nuclear power plant-related safety training. College students who are majored in nuclear engineering in China cannot deeply participate in nuclear power plant work during the college period. So, they do not have a deep understanding of the theoretical concepts of nuclear power systems in class. For research and teaching work on Sodium-cooled fast reactor (SFR), the Department of Engineering Physics at Tsinghua University has developed the sodium-cooled fast reactor nuclear power plant simulation software (ECFR) with Neoswise Science & Technology Co. Ltd. ECFR is the dynamic process model software of the SFR nuclear power plant which is based on VPOWER platform. This paper presents the application of ECFR in teaching. Experimental courses have been conducted for Tsinghua students. Students can observe the main system structure of the power plant and perform accident analysis through ECFR. Those experiments can help students grasp the latest nuclear power technology better and motivate them to study effectively. Furthermore, the future improvement direction of ECFR is also discussed, which includes the improvement measures of human-computer interaction and system design.
Application of High Accuracy Numerical Methods for the Natural Circulation Problem
Authors: Chao Fei , Li Longze , Yang Wen , Qiu Jinrong , Tai Yun , Shan Jianqiang 
Published: October 2021

Abstract
Natural circulation refers to the phenomenon of fluid circulation driven by the driving force formed by the density difference and height difference. Natural circulation is very common in nuclear power system, which has an important impact on reactor safety. Therefore, accurate simulation of natural circulation phenomenon is one of the most important research contents in current reactor thermal hydraulic analysis. The low order difference method is widely employed to discretize basic conservation equations for current main thermal-hydraulics analysis code such as RELAP5, CATHARE, TRACE, etc. However, the low order difference scheme has the characteristics of large numerical truncation error and low calculation accuracy when simulating natural circulation problem. In this paper, the high-order accuracy numerical algorithm for two-fluid two-pressure model is used to simulate Welander natural circulation problem. Numerical results imply that the higher order difference scheme can accurately predict the unstable boundary and chaotic behavior with fewer grids, while the lower order difference scheme may get wrong results when the mesh is not fine enough, that is, the unstable mode of natural circulation can be predicted as stable mode. Therefore, high-order accuracy numerical algorithm could prevent excessive numerical diffusion effectively and improve the prediction accuracy, which demonstrates the advantage of using high-order schemes.
Towards Malicious Action Detection for Nuclear Security via Integrated Deep Learning Based Image Recognition and Natural Language Processing
Authors: Demachi Kazuyuki , Sudo Masaki , Chen Shi 
Published: October 2021

Abstract
In recent years, deep learning has been very successful in various fields. In particular, since image data contains a lot of information, it is expected that deep learning will be applied to nuclear security technology for detecting the malicious action. However, most human action recognition using deep learning only sets OK / NO in advance for the combination of object identification result, so it is difficult to apply them to complicated nuclear security situation. On the other hand, in most nuclear facility, the nuclear security rules are determined as a detailed text. So, it is desirable to detect malicious actions in captured images according to these rule text. If it is possible to flexibly determine whether the captured image scene violates or matches the nuclear security rules, it is necessary to greatly improve its practicality. However, this requires different deep learning interfaces for image identification and natural language processing.
Most of the deep learning research currently being conducted is closed only within each field, and there is no mutual access, that is, an interface. The human brain, on the other hand, usually links and processes all kinds of cognitive information, including language, sight, speech, and touch. Like the brain, deep learning abilities should develop anew when different types of deep learning are interconnected. One of the reasons this is difficult is that there is no common data format between different types of deep learning.
In this study, it was determined that the logical expression is the most suitable as a common data format. The logical expression is a method of expressing the correlation between an object and a person. In this research, a deep learning interface method between image recognition and natural language processing was developed via the logical expression. This allows flexible judgment of wrong actions according to the rule text.
Hybrid Dynamic Response Test Focusing on the Support Structure of Piping Systems
Authors: Okuda Yukihiko , Nishida Akemi , Sakai Michiya , Shiogama Yuzo , Li Yinsheng 
Published: October 2021

Abstract
To develop a more realistic seismic response evaluation method of nuclear power plants, it is necessary to evaluate the seismic behavior considering the joints of each component that are treated as independent models during design evaluation, such as buildings, equipment, and piping systems. Particularly, the piping support structure, which is the joint between the building and piping, is important in the seismic evaluation of the piping system. While the current seismic evaluation of piping support structures is performed within the elastic range, it is important to consider the realistic elastic-plastic response of piping support structures for fragility assessment in seismic probabilistic risk assessment. However, the seismic evaluation method that considers the elastic-plastic response of piping support structures has not yet been established, and there is a need to improve the seismic response evaluation method.
In this study, a hybrid dynamic response test for simulating the seismic behavior of the piping support structure, including the elastic-plastic response, has been conducted. Specifically, static cyclic loading tests and hybrid dynamic response tests were conducted using four types of piping support structures to understand the basic mechanical behavior. This report presents the details of the tests and test results.
Research on Quality Assurance of Raw Materials for Nuclear Power Equipment
Authors: Pu Chenghao , Jin Gang , Yi Zilong , Huang Jiaqi , Han Dongao , Li Maolin , Wu Qi 
Published: October 2021

Abstract
In terms of quality management of raw materials for nuclear power equipment, equipment manufacturers implement quality management for raw materials according to different standards, to ensure the quality and reliability of nuclear power equipment. However, through some typical cases, it is found that many manufacturers do not effectively operate the quality assurance system in the equipment raw material quality management, and there are not enough anti-counterfeiting measures, resulting in unqualified raw material quality affecting the quality of nuclear power equipment. For example, when a manufacturer purchased ASME code grade bars for manufacturing nuclear valves, the raw material supplier forged part of the test reports from the third-party laboratory to make the results meet the requirements of the standard. However, the manufacturer did not find the problem in the procurement process. By comparing the quality assurance requirements of nuclear power units under different standards, this paper analyzes the problems in the process of raw material purchasing and quality management, combined with some measures and requirements to prevention CFSI, puts forward some suggestions for optimizing the quality assurance of raw materials for nuclear power equipment.
Characteristics of Debris From Simulated Molten Fuel Coolant Interaction Experiments
Authors: Hemanth Rao E. , Shukla Prabhat Kumar , Varghese Paulson , Polaki S. R. , Vetrivendan E. , Das Sanjay Kumar , Ponraju D. , Athmalingam S. , Venkatraman B. 
Published: October 2021

Abstract
Studies on Molten Fuel Coolant Interaction (MFCI) resulting from a hypothetical core melt accident in Sodium cooled Fast Reactor is important as the MFCI phenomena is associated with generation of dynamic pressure which has a potential for mechanical damage to the reactor vessel and structures. During MFCI, molten core at high temperature interacts with relatively cooler liquid sodium and undergoes fragmentation. The size, shape and physical state of debris depend on the associated fragmentation mechanism, super heat of the melt and sodium temperature. The fragmented particles settle on core catcher and form as debris bed, which continue to generate decay heat. Characteristics of debris govern the debris bed coolability and needs thorough investigation. Experiments have been conducted by releasing a high temperature melt at 2400 °C into liquid sodium at 400 °C to study the fragmentation phenomena. Molten mixture of alumina and Iron generated by aluminothermy process was used for simulating the corium. After the experiment, sodium was drained and debris was retrieved by vacuum evaporation method. Size distribution and mass median diameters of the resultant debris was determined by sieve analysis. The debris was segregated into metallic/ nonmetallic particles using magnetic separation method and subjected to microscopic investigation for obtaining morphological characteristics. This paper outlines about the characteristics of debris generated by interaction of simulated corium with sodium towards assessment of fragmentation mechanism and associated debris coolability.
Development of Effective Momentum Model for Steam Injection Through Multi-Hole Spargers: Unit Cell Model
Authors: Wang Xicheng , Grishchenko Dmitry , Kudinov Pavel 
Published: October 2021

Abstract
The Steam injection through multi-hole spargers into the pressure suppression pool (PSP) is used in light water reactors to prevent containment over-pressure. The development of thermal stratification in the PSP can reduce its cooling capacity and results in higher containment pressures compared to completely mixed pool conditions. Explicit modelling of direct contact condensation (DCC) of steam at the steam-water interface is a challenge for contemporary codes. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to enable the prediction of thermal stratification and mixing transients induced by steam condensation in a large pool. The general idea of the EHS/EMS is to resolve the effect of the DCC phenomena on a large pool, instead of explicit modelling of the small-scale phenomena at steam-water interface.
The EHS/EMS models can be implemented using (i) respective boundary conditions at the boundary of the Steam Condensation Region (SCR) or (ii) using source terms in the heat and momentum transport equations. In previous work, EHS/EMS models were implemented using the second approach and validated against data from PPOOLEX and PANDA tests. It was found that results are sensitive to the spatial distribution of the source terms. Since the current data are not sufficient to provide a reasonable distribution, a preliminary study of the first method was done in this paper.
The goal of this work is to develop a ‘Unit Cell’ model by using respective boundary conditions for steam injection through multi-hole sparger. The condensed turbulent jet is resolved by introducing the liquid jet with the same effective momentum and heat as the injected steam. A uniform velocity profile solved by EMS model and the temperature boundary solved by EHS model is provided on each injection hole of the sparger wall. Validation is conducted against sparger test in PANDA facilities.
Study on Influence of Rolling and Heaving Motions on Differential Pressure and Flow Rate Measurements
Authors: Zhang Biao , Liu Jingyu , Li Xin , Qiao Shouxu , Li Dongyang , Tan Sichao 
Published: October 2021

Abstract
Ocean conditions will cause ships to be affected by rolling, heaving and other motions, which introduce additional inertial forces to the marine nuclear power system, introduce additional pressure drop to the fluids in each loop, bring errors to the measurement of thermal hydraulic parameters such as pressure difference and flow rate. Therefore, the measurement of pressure difference and flow rate under ocean conditions is studied and corrected in this paper. We built a rolling platform and a heaving platform and measured the differential pressure and flow rate under various working conditions. Through analysis, the relationship between differential pressure, flow fluctuation and swaying, undulating and other movements under moving conditions was obtained. Besides, the calculation method of attached pressure drop was analyzed, and the following conclusions were obtained: The influence of motion on differential pressure and flow rate is mainly composed of two parts: The additional pressure drop caused by the swinging motion of working medium in the pipeline and the pressure lead-in pipe, and the periodic fluctuation of friction pressure drop caused by the periodic fluctuation of flow velocity caused by swinging motion. The attached pressure drop caused by motion can be calculated through theoretical analysis. Through the analysis of experimental data and the comparison of theoretical calculation methods, it is found that the pressure drop measured by the experiment is in good agreement with that fitted by the theoretical calculation, and the modification scheme of pressure difference under ocean conditions is feasible to some extent.
Review of Determining Transformed Temperature Over 325°C Based on Monte Carlo Simulation for Environmental Fatigue Evaluation
Authors: Lee Bonghee , Nam Ilkwun 
Published: October 2021

Abstract
The NUREG/CR-6909, which had provided the technical background for the Environmental Fatigue Evaluation (EFE), was revised in May 2018 to incorporate the Japan Nuclear Energy Safety organization (JNES)’s extensive fatigue ε-N data and to address concerns from interested stakeholders over the original NUREG/CR-6909. But the revised NUREG/CR-6909 sets limit of the temperature ranges to 325°C for the application of the environmental fatigue correction factor (Fen). Therefore, the revised NUREG/CR-6909-based EFE is not applicable to the materials whose temperatures are higher than 325°C.
In this paper, Monte Carlo simulations have been performed to determine the transformed temperatures over 325°C for the materials which are not covered in the revised NUREG/CR-6909. To determine the distributions of fatigue lives in LWR environments, Monte Carlo simulations were performed using the fatigue life models in the LWR environments and transformed parameters provided in the revised NUREG/CR-6909 and JNES-SS-1005. The transformed parameters which affect the fatigue lives in the LWR environments were selected randomly using the uniform distribution. And a total of 100,000 simulations were performed for each material. The distributions of fatigue lives in LWR environments obtained from Monte Carlo simulations were fit to lognormal curves to estimate the cumulative distributions.
The estimated cumulative distributions were individually determined in the temperature ranges of 100 ∼ 325°C and 325 ∼ 350°C to investigate the temperature effects (FT,effect) over 325°C. The temperature effect is defined as ratio of fatigue lives in the LWR environments in the temperature range of 100 ∼ 325°C to that in the temperature range of 325 ∼ 350°C in this paper.
As a result of reviewing the temperature effects between the revised NUREG/CR-6909 and JNES-SS-1005, the transformed temperatures over 325°C for each material have been suggested for the application of the revised NUREG/CR-6909.