bluuebunny/crossref_metadata_2025_split huggingface dataset:
Logic Equation of Operating Situations and Man-Machine Interface Design for NPP
Authors: Si Tianqi , Wang Yuqi , Du Yu 
Published: November 2022

Abstract
In a nuclear power plant (NPP), the operation process is often very complicated. According to the requirements of operation technical specifications (TS) for event handling, operators should pay more attention to the current state of the unit in order to control NPP, so the plant has set different kinds of operating situations. Based on the definition of operating situations in existing NPPs, if the operator wants to know the current situation of the unit, it needs to check the relevant parameters and then have a judgment, or through other indirect means, which is not conducive to improving work efficiency. In view of the problems mentioned above, for the NPP based on the digital control system (DCS) and advanced main control room (MCR), this paper presents a design strategy of the operating situation logic equation, which include the design principle, calculation method and friendly man-machine interface display. These logic equations are used to automatically calculate current situation by instrumentation and control (I&C) system via selecting and combining typical parameters with logic signals, make the operator be able to know the unit state in any case, improve the work efficiency and reduce the operation burden. The method in this paper also provides a reference for the design of new nuclear power units.
Hydrogenation of Zircaloy-4 in a Defective Fuel Pellet for Pressurized Water Reactors
Authors: Enivweru Doctor , Wang Qingyu , Ayodeji Abiodun , Njoroge John 
Published: November 2022

Abstract
The integrity of structural components is a major concern in reactor safety. Cladding materials can fail when hydrogen uptake is in excess of the terminal solid solubility (TSS). Previous studies on hydrogenation or secondary hydriding (SH) focused mainly on loss of coolant accident (LOCA) and fuel ballooning. Most of these studies are carried out in autoclave conditions and they preclude the contribution of ionizing radiations (neutrons and gamma) to hydrogen generation during normal reactor condition. This study is devoted to the hydrogenation of Zircaloy-4 (Zry-4) cladding material during normal reactor operation and during the early phase of core degradation in a LOCA event. Radiolysis and corrosion are considered as hydrogen sources during normal reactor operation while corrosion is considered in the early phase of core degradation. Using appropriate initial and boundary conditions peculiar to pressurized water reactors (PWRs), diffusion equation is solved in a defective Zry-4 fuel system using COMSOL Multiphysics 5.2. Results show that the average corrosive source term (1.279E−4 mol m−3 s−1) is higher than the radiolytic source term (3.6594E−7 mol m−3 s−1) by a factor of 350 during normal reactor operation. By integrating in two kinetics regimes and setting 6 years as the maximum lifetime of Zry-4, the amount of H dissolved in the cladding material at a temperature of 633 K are 6.54E−03 and 2.02 wt. ppm due to radiolysis and corrosion respectively. These values are safe compared to the TSS of dissolution (CTSSD) of Zry-4 which is 117 wt. ppm at the same referenced temperature. However, in the early phase of core degradation where the source term is 9.405E−01 mol m−3 s−1, SH is observed in 13 days. A comparison is made between hydrogenation by integration in two regimes and hydrogenation using Sievert’s law. The results obtained in this study are useful for the prediction of the onset of hydrogen embrittlement, and to determine the service time of Zry-4 cladding material.
Research and Development of Wet Winding Motor Reactor Coolant Pump for Generation 3 Nuclear Power Plant
Authors: Tianbin Li , Lin Ma 
Published: November 2022

Abstract
Within an eight-year period, starting in September 2011 and ending in July 2019, Shanghai Electric-KSB Nuclear Pumps & Valves Co., Ltd (hereinafter called SEC-KSB) has managed to develop a new pump type that can be used as reactor coolant pump (hereinafter called RCP) in generation 3 pressurized water reactor design. Wet winding motor reactor coolant pump (pump type RUV 650–870) is a customized and especially designed to meet the high safety criteria of generation 3 nuclear power plant, which is a hermetically sealed (without shaft seal system) type pump, and it is first time used as reactor coolant pump in pressurized water reactor in the worldwide. Started with the pump commissioning test in December 2014, total eight times high temperature and high-pressure full load testing had to be carried out in order to eliminate design or manufacturing flaws. In early 2019 within three months of intensive testing, under all the specified design transient scenarios, the RUV 650-870 pump had successfully passed all the qualification test items. On July 31. 2019 an expert qualification meeting was organized by China Machinery Industry Federation and China General Machinery Industry Association, the qualification committee states that the development of first set wet winding motor reactor coolant pump is successful in the worldwide, and the pump technical performance meet the requirements of the equipment specification, test specification and related industry standards, the RUV 650-870 pump can be used for nuclear power plant projects. On January 2022, the first production pump was delivered to nuclear power plant.
Research on Cracked Fuel Relocation of Dual-Cooled Annular Fuel Element Under Normal and LOCA Conditions
Authors: Deng Yangbin , Yin Yuan 
Published: November 2022

Abstract
The application of the dual-cooled annular fuel element can significantly improve the safety and economy of current pressurized water reactors. Due to the geometric differences, the fuel relocation caused by the fuel cracking in annular rods was much more complex and influential on thermal-hydraulic performance than that in solid rods. In this research, the Discrete Element Method (DEM) was applied to perform the simulation of fuel fragment relocation in the annular fuel rod under both long-term normal operation and LOCA conditions, in order to get deep insights into the fuel relocation mechanisms and develop the mathematical relocation model. Under normal operations, it was found that the radial fuel relocation was bidirectional of both inward and outward, resulting in the radial size reduction of both internal and external gas gaps. The maximum reduction of total gaps was about 58% of the as-fabricated value, and the maximum allowable recovery fraction of fuel relocation was about 55%. Under LOCA conditions, the ballooning of the cladding was full considered in the fuel relocation. As a consequence, the substantial axial fuel relocation was found, which resulting a size reduction of active fuel length and a local fuel accumulation at the position of ballooning.
Study on Probabilistic Safety Analysis Method of Spent Fuel Reprocessing Facilities
Authors: Zhang Jinghan , Zhao Jun , Hou Jie 
Published: November 2022

Abstract
Probabilistic safety analysis (PSA) method is a systematic analysis method based on probability theory. Based on the PSA method, the relationship and interaction between various factors in a large and complex system can be recognized clearly, and the accident scenarios can be modeled correctly. After years of development, PSA has gradually become an important tool for evaluating the safety of nuclear power plants, and has been widely used. For the safety analysis of spent fuel reprocessing facilities, deterministic safety analysis methods are still mostly used. It is beneficial to introduce quantitative probabilistic safety analysis methods into spent fuel reprocessing facilities. For spent fuel reprocessing plants, there are many features which are totally different from nuclear power plants. Therefore, we are supposed to establish a new framework for PSA of spent fuel reprocessing plants. In this study, the spent fuel reprocessing facilities safety evaluation methods and the PSA method were investigated, based on the risk characteristics of spent fuel reprocessing facilities, every technical element of PSA method was adjusted. We studied the possible problems existing in the application of PSA method and proposed corresponding solutions, and then established a PSA method system suitable for spent fuel reprocessing facilities. Finally, we used PSA method to evaluate the safety of the selected typical processes of spent fuel reprocessing facilities and got the analysis result which is useful for practical application.
Discussion on Functional Safety and Cyber Security of I&C System in Nuclear Facilities
Authors: Yang Anyi 
Published: November 2022

Abstract
The safety of I&C system of nuclear facilities is one of the key points of nuclear safety supervision. Due to the use of proprietary technology and generally not connected with IT systems, traditional I&C system of nuclear facilities mainly focuses on functional safety in design. Relevant standards such as IEC 61508, IEC 61511 and IEC 61513 all focus on ensuring functional safety. The extensive use of information technology in nuclear energy field has brought about cyber security threats, and cyber-attacks against nuclear facilities have been reported occasionally. The International Atomic Energy Agency (IAEA) and nuclear safety regulators around the world have issued laws, regulations and technical standards to strengthen the supervision of cyber security of nuclear facilities. However, how to properly deal with the relationship between functional safety and cyber security has become a hot issue when strengthening the supervision of cyber security of nuclear facilities. Based on the relevant international standards and national practices, this paper further clarifies the relevant concepts, review current construction practices, analyzes and discusses the relationship between functional safety and cyber security. The research lays a foundation for the subsequent establishment of applicable nuclear power cyber security standard system.
Experiment Investigation on Onset of Flow Instability in Upward Tube Under Low Pressure Condition
Authors: Yang Bing , Zhao Minfu , Wei Junhan 
Published: November 2022

Abstract
Two-phase flow instability plays an very important role in operation safety and reliability of heat exchange equipment. The mechanism of two-phase flow instability was very complicated. Two-phase flow instability may occur when the fluid volume in the circulation loop changes greatly due to heating or cooling. In this paper, the onset of flow instability (OFI) was investigated under the condition of: P = 2.2MPa, G = 2000∼4000kg/(m2·s), ΔTsat = 100∼15 °C, q = 1000∼1500kW/m2 in upward tube (2400 × 8.2 mm) with deionized water. When the fully developed subcooling nucleate boiling (FDB) happens in the test section, the fluctuating flow appears between the circulation loop and the pressurizer, also the test section. The inlet and outlet water temperature, the pressure drop of the test section and the exit valve, the liquid level of the pressurizer and the other system related thermal parameters become unstable after the FDB condition, which is caused by the generation and departure of large amount of bubbles in the test section. The point of FDB can be considered to the onset of flow instability in the circulation loop. An empirical correlation was developed by comparing the heat flux leading to OFI with that leading to saturation: qOFI = 2.11qsat0.73, and the deviation is within ±15% for all the conditions.
Summary of Methods for Studying the Chemical States of Nuclides In Nuclear Energy Systems
Authors: Guo Jingni , Wang Yu , Zhou Ziling , Xie Feng , Tong Jiejuan , Wang Kerong , Li Peng , Jiang Jing 
Published: November 2022

Abstract
The chemical states of nuclides affect their physical and chemical properties including thermal conductivity, melting point, adsorption and desorption behavior, diffusion process, and chemical reactivity. It is an important issue in nuclear energy systems and spent fuel reprocessing. In this review, we summarize the theoretical calculations and experimental measurements to determine the chemical states of nuclides in nuclear energy systems. Software such as FactSage, HSC Chemistry and The Geochemist’s Workbench are generally used to determine the macroscopic-scale thermodynamic parameters such as thermodynamically stable phases. Quantum chemistry calculation software such as Gaussian 03 is employed for microscopic first-principles calculations to elucidate chemical reaction channels, microstructures, bonding characteristics, and rate constants. The experimental methods to determine chemical states of nuclides include X-ray, spectroscopy and chemical extraction.
Pre-Conceptual Core Design of a Micro-LBE-Cooled Fast Reactor
Authors: Yeoh Eing Yee , Liu Menglu , Luo Zhen , Chen Xiaosong , Li Linsen , Zheng Gang 
Published: November 2022

Abstract
Generation IV nuclear reactors feature better performance in safety, economy, spent fuel burning and a higher breeding ratio, thus are becoming an important direction for the development of future nuclear industry. In the world, the special supply-demand relationship and flexibility requested by users of the energy market make small and micro nuclear reactors a key development trend, while small and micro lead-cooled fast nuclear reactors show great competitiveness. In the present paper, a micro LBE (lead-bismuth eutectic) - cooled Fast Reactor is proposed, on which a preliminary conceptual core design has been conducted. This micro LBE-cooled Fast Reactor has thermal power of 12 MW, while the reactivity controlled by 12 control drums. Neutronics analysis was performed by a Monte-Carlo code cosRMC jointly developed by State Power Investment Corporation Research Institute of China and Tsinghua University. The calculation results showed that the reactivity control performance meets the safety criteria, giving a coolant reactivity of −0.0976 pcm/K at BOL, −0.1955 pcm/K at MOL, and −0.1195 pcm/K at EOL, coolant voiding coefficient −31.24 pcm/% at BOL, −24.16 pcm/% at MOL, and −17.72 pcm/% at EOL, and fuel Doppler coefficient −0.6981 pcm/K at BOL, −0.7239 pcm/K at MOL, and −0.6902 pcm/K at EOL. The power distribution was flattened enough with a radical power peak factor 1.185 at BOL, 1.136 at MOL, and 1.155 at EOL.
Applying the PIRT Approach to the Meteorological Hazards Wind and Rainfall in Relation to Industrial Needs: Benefits and Lessons Learned
Authors: Le Morvan Tiphaine , Défossez Arièle , Kreitz Mickaël 
Published: November 2022

Abstract
As an integrated nuclear engineer and operator, EDF has a strong background in studying extreme natural hazards for its Nuclear Power Plants (NPPs) fleet safety demonstration. The state-of-the art evolution and the development of natural hazards Probabilistic Safety Analysis provide an opportunity to go more in depth in our studies. To reinforce or, if need be, to adjust our hypotheses regarding the detailed characterization of natural hazards, we decided to investigate the detailed phenomenology associated with strong winds and rainfall in France.
The PIRT process (Phenomena Identification and Ranking Table) allows a physics-oriented approach to complex phenomena by systematically analyzing relevant parameters and their influence on a scenario of interest. This approach appears like a legitimate way to go more in depth in the phenomenology associated with strong winds and rainfall.
This approach, used in the nuclear energy field, for example, as support for complex transients numerical modelling, is here tested on complex natural phenomena.
Strong winds, rainfall and other associated phenomena are very complex if one wishes to consider a high detail level. It was decided during the exercise to limit the technical complexity to a level relevant to industrial concerns.
Even with this simplification, the lessons learned will allow EDF to be more precise in future studies, such as those regarding strong winds directions and some phenomena specific to certain geographical locations.
This article describes the PIRT approach and modifications that were made to the classical PIRT steps both for strong winds and rainfall. We also outline the main lessons we learned, from the PIRT application and regarding the analysis we were able to gain regarding specific needs for our studies.
We found that the PIRT approach can be used to study meteorological phenomenology with the following adaptations:
- An adapted detail level — some very complex aspects, coming under Research & Development (R&D) or advanced meteorological expertise have not been taken into account in the present exercise
- Use the PIRT as a way to present and organize advanced knowledge on a subject. The aim is to shed more light on an end user issue, like to better comprehend the underlying physical processes for strong winds safety demonstration studies: speed, direction, duration ...
The exercise was possible thanks to the strong collaboration with Météo France, the national French Meteorological Agency, who was the referent expert bringing the technical and scientific knowledge on meteorology.
Modelling the Brittle-Ductile Transition Temperature of Irradiated A508-3 Steel With CPFEM
Authors: Lin Pandong , Nie Junfeng , Liu Meidan 
Published: November 2022

Abstract
Crystal Plasticity Finite Element Model (CPFEM) is the powerful tool to bridge the meso-scale and marco-scale and study the mechanical properties of metals systematically. In this paper, crystal plasticity theory coupling with irradiation effect is proposed. Note that it is based on density of dislocation and irradiation-induced full-absorption and partial-absorption dislocation loop and their interaction between each other. Then the model is numerically implemented in UMAT on ABAQUS platform. Secondly, A508-3 steel, the typical BCC crystal material, has specific brittle-ductile transition. The cleavage failure probability theory model (CFPTM) which assumes that the failure probability of specimen depends on its Cauthy stress filed is presented. The tensile tests for A508-3 steel are simulated by CPFEM in a combination of CFPTM to obtain the brittle-ductile transition temperature (BDTT). Results show that failure probability for specimen increases with increasing strain and decreasing temperature. In terms of BDTT, the numerical and experimental result are in close agreement whether there is under irradiation condition or not. Furthermore, the obvious irradiation embrittlement phenomenon is observed.
Study and Development of Irradiation Test Device for Fuel Rods In Research Reactor
Authors: Zhao Wenbin , Si Junping , Zhang Liang , Sun Sheng , Yang Wenhua , Jin Shuai , Lei Jin , Liu Hao 
Published: November 2022

Abstract
For the purpose of performing high-temperature irradiation of the fuel rod in the low-temperature and low-pressure water-cooled research reactor and obtaining the post-irradiation performance parameters of the fuel rod, a fuel rod irradiation test device was designed based on the High Flux Engineering Test Reactor (HFETR). The key parameters such as thickness of the lead-bismuth eutectic (LBE) alloy layer, structure of the cooling water flow channel, the structure of throttle plug, and flow rate of the cooling water were determined. The flow field distribution in the test section was obtained by numerical simulation method. The stress values of the device under design and accident conditions were obtained by finite element analysis. The test of hydraulic characteristic out-reactor were performed by using the simulation device. The results show that, the average velocity of the cooling water outside the test assembly was 1.9 m/s under the design flow rate, which met the cooling requirement when the linear power of the fuel rod was 280 W/cm. The structure of the test device met the strength requirement with sufficient safety margin. In the experiment out-reactor, the test device was able to cover the design value of 4.03 t/h on the differential pressure of 200 kPa and 300 kPa with coolant flow rate adjustment ranges of 0.73 t/h ∼ 4.28 t/h and 1.05 t/h ∼ 5.25 t/h, which met the test requirements. The device designed in this paper can be used to perform the irradiation test in research reactor, which provides guarantee for the research and development of new type of the fuel.
Early Mitigation Measures for High Flux Reactor Loss-of-Coolant Accidents
Authors: Wang Zhuang , Zhu Gan , Xie Heng 
Published: November 2022

Abstract
The thermal hydraulic characteristics of high flux research reactor (HFRR) are different from those of pressurized water reactor (PWR). Therefore, the emergency core cooling system of HFRR must be reconsidered and redesigned to ensure the core safety of HFRR in the event of loss-of-coolant accident (LOCA). This paper designs and analyses a set of mitigation measures for LOCA in the high flux research reactor. Relap5 is used for thermal hydraulic system analysis and breaks at three different sites are simulated. The primary coolant system includes four loops, each of which consists of a main pump, a heat exchanger and the corresponding pipes and valves. The three break sites are the main pump outlet, the main pump inlet and the heat exchanger inlet respectively. The core pressure, mass flow, coolant temperature, etc. in the transient process are studied and analyzed. And the responses of the mitigation measures during LOCA are organized and summarized. In this research, critical heat flux is calculated by Sudo correlation and the onset of nucleate boiling is calculated by Bergles-Rohsenow correlation. Through analysis, it can be found that the current mitigation measures can ensure the safety of core in the early stage of LOCA.
Analytical and Experimental Research on Reactor Coolant Pump Mechanical Seals With Hydrodynamic Deep-Grooves
Authors: Tang Liming , Pan Yongchun , Tan Xin , Jin Le , Wang Yan 
Published: November 2022

Abstract
Hydrodynamic mechanical seal with deep grooves is used on reactor coolant pumps (RCP) of KSB SE. It has several good features, such as simple structure, excellent wear resistance, low pressure leakage rate, etc. In this paper, an analysis method across different calculation platforms is used to analyze the RCP mechanical seals. The finite element analysis based on ANSYS is to simulate the thermal deformation of mechanical seal rings, while an algorithm based on Reynolds equation with JFO cavitation boundary is used to evaluate the seal film, which is less than 1 μm thick. A specific test rig was built up to carry out the operational test for RCP mechanical seals. In this test rig, full pressure, normal temperature and rotary speed can be achieved, meanwhile some special conditions can be simulated, e.g. thermal impact, pressure drop, and loss power. The internal pressure, flow rate, temperature, and rotary speed can be detected in-situ. The low-pressure leakage rate is the flow through the gap between seal faces, which is the most important operational parameter to represent the tribological features of seal faces. In this test, the low-pressure leakage rate can be detected by a weight-based measurement system. After 200 hours test, all detecting parameters kept constant. The detected low-pressure leakage rate was around 80 mL/h, which has good consistency with the analysis results.