bluuebunny/crossref_metadata_2025_split huggingface dataset:
On Multiaxial Strain Characteristics of Piping Systems Subjected to Earthquake Loads
Authors: Morishita Masaki , Otani Akihito 
Published: November 2022

Abstract
The JSME Code Case N-CC-008 provides design rules for piping seismic design by inelastic response time history analysis and strain-based fatigue criteria. Since the strain time histories obtained from the analysis are multi-axial and random, the Code Case uses the Rainflow method for cycle counting in its procedure of evaluating the ranges of equivalent strain and the fatigue usage factor. The Rainflow method is used for identifying pairs of peaks/valleys of the representative strain and their occurrence times. Then, with the other components of strain and stress at these times, the range of equivalent strain is calculated sequentially. On the other hand, ASME Boiler and Pressure Vessel Code Section VIII, Division 2 reads “The Rainflow cycle counting method is recommended but not applicable for non-proportional loading.” The authors, then, analyzed the behavior of multi-axial strain from the inelastic response analyses of some dynamic tests of piping systems to test if the piping strains due to seismic excitations are proportional or non-proportional, or somewhere in between. The IS method, which was proposed by Itoh et.al for quantifying the degree of non-proportionality of multi-axial fatigue cycles, was used for the analysis. The analysis results showed that the multi-axial strain in piping systems induced by seismic excitation is almost proportional. Based on this result, the authors concluded that the Rainflow cycle counting method is well applicable for the piping seismic fatigue evaluation. In addition to these numerical analyses with the IS method, some discussions are also made in this paper from the viewpoint of the relation between vibration mode and strain components (compatibility condition between deformation and strain) to support the conclusion that the piping strain by seismic loads is proportional.
TCR Central Shutdown Rod Fine Motion Control
Authors: Varma Venugopal K. , Fountain Eliott , Burkhardt Earle , Cetiner Sacit , Ottinger Dustin 
Published: November 2022

Abstract
The Transformational Challenge Reactor (TCR) is a helium-cooled 3 MWt test reactor with a design that leverages advances in materials and manufacturing, computing, and artificial intelligence. Gravity, pneumatics, or springs are typically used to quickly release the central shutdown rod — which contains the neutron absorber — into the reactor core, thus stopping the nuclear reaction. The rod’s motor/actuator typically resides outside the reactor, and the motion is transmitted through a penetration into the pressure vessel. Some control rod designs incorporate magnetic latches with a coil outside the pressure boundary to ensure precise position control.
The proposed design comprises a magnetic coupling to position the shutdown rod, with no penetration into the pressure vessel for the entire drive length of TCR shutdown rod. Electromagnets are used in nonpower nuclear reactors for shutdown rod control, but they reside inside the reactor pressure vessel. This paper describes the use of an electromagnet outside the pressure vessel to position and release the shutdown rod. A prototype was developed at Oak Ridge National Laboratory to demonstrate the concept, and the ferritic core and material design was optimized to maximize the electromagnet’s lift force. An elevated temperature test was also conducted to ensure that the system will perform effectively inside an operating reactor.
Study on Preparation and Radioactive Methyl-Iodine Adsorption Performance of TEDA Ionic Liquids
Authors: Li Xin , Li Yongguo , Yu Jie , Chen Xin , Hou Jianrong , Li Jian , Qiu Dangui , Shi Xu 
Published: November 2022

Abstract
As the most important impregnation component of iodine adsorption materials, Triethylenediamine (TEDA) has a disadvantage of being volatile. This disadvantage has a negative impact on the adsorption performance of iodine adsorber, resulting in an increase in the operation cost of nuclear facilities and the amount of solid waste. Based on this problem, TEDA ionic liquids (TEDAILs) was prepared by quaternization of TEDA, which is a method to effectively solve the volatile problem of TEDA. The changes of the yield of TEDAILs and the proportion of single-cationic ionic liquids (SCILs) were investigated under different reactant ratio, reaction temperature and reaction time. According to the law of change, the optimal preparation condition of TEDAILs was revealed. In addition, the composition of the TEDAILs was analyzed by infrared spectrum and nuclear magnetic spectrum, and the thermal stability was also verified by thermogravimetric analysis. On this basis, TEDAILs loaded activated carbon fibers (TEDAILs-ACF) was designed and prepared, and the iodine adsorption performance was tested. Verification test results show that the complete decomposition temperature of TEDAILs is about 362 °C, which is about 187% higher than that of TEDA. Moreover, the adsorption efficiency of radioactive methyl iodine by TEDAILs-ACF under normal temperature and humidity is 99.94%. In summary, TEDAILs-ACF as radioactive methyl-iodine adsorption material has potential application prospect.
DFMEA Application in Rotor Design of Million-Kilowatt Nuclear Power Stream Turbine
Authors: Binghan Xue 
Published: November 2022

Abstract
DFMEA is a systematic and scientific method for new product design and development, which is applicable throughout the entire process of new product design and development. In this paper, DFMEA method is adopted in the systematic analyzing of the million-kilowatt nuclear power steam turbine rotor new product design and development process. On the basis of the introduction of product characteristics and current development requirement, DFMEA is introduced to the design and development process, applying the block diagram and structure-function matrix to locate the possible failure factors. Subsequently, the evaluation criteria of Severity (S), Occurrence (O), and Detection (D) are introduced and defined. According to the criteria, the Risk Priority Numbers (RPN) and Action Priority (AP) of key risks are sorted, therefore indicating the terms critical to product quality improvement.
Through the application of DFMEA, an effective risk identification and analysis of the million-kilowatt nuclear power steam turbine rotor is carried out, achieving the results that operational failures are advanced to the design stage to implement countermeasures, therefore realizing the improvement of product and system operation reliability, optimization of project cycle and manufacturing cost reduction. DFMEA promotes the optimization of overall new product development process. Meanwhile it provides ideas for the subsequent design optimization methods of similar kinds.
Volume Change Analysis of Primary Loop in a Small Modular Thorium-Based Molten Salt Reactor
Authors: Zhu Guifeng , Zou Yang , Yan Rui , Guo Wei , Kang Xuzhong , Xu Tiangui , Yang Pu , Zhou Bo 
Published: November 2022

Abstract
In a molten salt reactor, the volume change in the primary loop may lead to an overflow of fuel salt, which will challenge the radioactive containment. Several factors, such as the thermal expansion, graphite dimensional change caused by fast neutron irradiation, fuel feed online and fuel salt reconstitution that affect volume change are analyzed based on a designed small modular thorium-based molten salt reactor (SM-TMSR) in this paper. The results show that the thermal expansion of nickel alloy is comparable to that of fuel salt, the net volume change due to isothermal expansion is not obvious, about 0.04 m3 per 100 °C in SM-TMSR, which can be easily contained in a pump bowl without an additional overflow tank. Furthermore, the volume decrement due to graphite shrink by irradiation (based on the IG-110 data) is more obvious than the volume increment by fuel feed online, which will lead to a decline of liquid level during the long-term operation. 0.13 m3 excess volume of fuel salt should be provided in the pump of SM-TMSR. Finally, as we analyzed, fuel salt reconstitution will add to the difficulties in volume control, reactivity control and the control of heavy metal mole ratio in fuel salt. It is recommended that the used fuel salt can be recycled into a new power module without any treatment, and the increased volume of fuel salt caused by feed online in last power module can be accommodated in a larger downcomer.
Study on Application of Hollow Springs in Seismic Isolation in Nuclear Power Plants
Authors: Hou Nannan , Dong Jianling 
Published: November 2022

Abstract
Earthquake is always an important consideration when constructing nuclear power plants. There are many studies on how to resist or avoid earthquake damage. And recently more and more seismic isolation technology has been applied to nuclear structure in nuclear power plants using isolators mainly made of rubber which can not only bear the vertical load, but also reduce the response of seismic on the upper structure due to their low lateral stiffness so as to improve the safety of the nuclear structure in the earthquake. Compared with traditional anti-seismic technology, seismic isolation can make the design of the main structure of nuclear power plants achieve standardization unrestricted by the magnitude of design ground motion.
The two main categories of seismic isolators are elastomeric bearings and spherical sliding bearings. And low damping bearings (LDR), lead rubber bearings (LR) and spherical sliding bearings (FP) are currently considered sufficiently well characterized for use in nuclear facilities. However, the rubber which is the main material of elastomeric bearings cannot withstand high temperature, so the properties of the bearings might be affected and even lost under the complex and severe environment of nuclear power plants, which could compromise the safety of nuclear reactor. In addition, both rubber bearings and spherical sliding bearings can only apply to horizontal seismic isolation.
The metal springs have more stable performance than rubber under high temperature condition and also has better isolation effect. Springs are usually made of solid spring wire by metal forming. In production practice, springs are used for vertical isolation because of their low vertical stiffness, and calculations show that they also work in the lateral direction. Springs have been widely used for vibration isolation in many light load and high requirements situations like precision instruments at this stage, but need more research to be applied in nuclear structure of nuclear facilities.
Compared with air spring, under the same load-bearing capacity, the stiffness of solid spring is higher, which leads to the higher natural frequency of the isolation system, so it is not suitable for the situation that requires low frequency and heavy load. This paper tries to use springs made of hollow wire to solve the problem. In this paper, some research results of hollow springs are given, including frequency and damping property when filled with damping materials. In addition, a calculation formula of hollow springs’ mechanical property, including the vertical and lateral stiffness and carrying capacity, is derived, and the finite element calculation are used to validate the formula. And the paper also gives a finite element calculation case on how the hollow spring reduce the natural frequency of an isolation system.
Application of Nuclear Insurance Risk Survey Results in Nuclear Power Plant Safety Improvement Program
Authors: Ping Jiang , Jiangtao An , Changliang Liu 
Published: November 2022

Abstract
From the prospective of loss prevention, economic risk reduction and catastrophe risk reduction, nuclear insurance risk survey provides professional and foresighted safety improvement recommendations based on the international nuclear and insurance experience. As a common international risk management practice, nuclear insurers provide loss prevention evaluation services, which is considered as peer review by nuclear operators. With the development of Chinese insurance industry and increasing loss prevention demand from Chinese nuclear power plant, Chinese insurance industry gradually integrates international practice, and establishes a systematic, Chinese characterized nuclear insurance survey methodology. This thesis introduces insurance survey background and importance, analyses the areas in which insurance survey could support nuclear power risk management. This paper focuses on risk survey methodology and implementation plan based on insurance service practices, and analyses survey result trends. The methodology has been implemented in nuclear power risk management regime. Recent cases have verified the insurance survey could support nuclear power plant safety improvement. Besides, based on the current practice, this paper suggest the potential ways to improve the interface between insurance risk survey and nuclear power risk management, improve common risk and enhance survey effectiveness.
Effect of Water Jet Behavior on the Operating Range in Supersonic Steam Injector
Authors: Minamizono Yukiya , Kaneko Akiko 
Published: November 2022

Abstract
In the case of a severe accident, such as a loss of power supply, it is necessary to keep the nuclear power plant in a safe condition. Therefore, steam-water injectors (SI) have been attracting attention as a device that can cool the plant without power supply in the event of a severe accident. In a SI, the steam condenses by direct contact between steam and water interface. As the steam condenses, negative pressure is generated, and a force that attracts water is generated. The SI can process a large amount of cooling water with this force. It can also function as a reactor condenser for nuclear reactors due to its high condensation performance. There are some characteristics of SI that the discharge pressure is expected to exceed the inlet pressure after being accelerated at the throat, and it has a very simple structure with converging-diverging part, which is expected to reduce the cost of installation and maintenance. Previous studies have indicated that GROLMES et al. (1968) inferred the flow in the reduced test section by measuring the void fraction and MIWA et al. (2018) suggested a relationship between internal flow and the operating range of SI, referring to the results inferred by GROLMES et al. (1968). Although these existing studies have suggested some relationship between the internal flow and the operating range, the relationship between the stability of the water jet and the behavior of the SI has not been clarified.
Therefore, the aim of this study is to clarify the effect of water jet behavior on the operating range of SI. The taper angle of the water jet outlet was varied and the relationship between the water jet behavior and the operating range was investigated. In this study, experiments were conducted using two types of taper angles at the water jet outlet: 20 and 7 degrees. As a result of detailed observation inside the SI, it was confirmed that the water jet with the taper angle of 7 degrees was less spreading than that with the taper angle of 20 degrees. In the operating range of this experiment, no significant difference was observed when the taper angle of the water jet was changed. However, it is suggested that the water jet with a certain degree of dispersion might be better in terms of heat exchange due to large surface area. These results suggest that the stability of the water jet is a necessary but not a sufficient condition for operation.
Effect of Brownian Diffusion on Film Condensation With Non-Condensable Gas
Authors: Peng Cheng , Song Youya , Meng Xianpin 
Published: November 2022

Abstract
The innovative design of passive containment cooling system (PCCS) is one of the most significant safety improvements for the third generation of advanced pressurized water reactors (APWRs), however, its impact on the hydrogen risk in severe accidents for nuclear power plants (NPPs) has not been determined yet. On the one hand, natural convection through liquid film condensation on the inner wall of the containment may mitigate the hydrogen risk, but on the other hand, rapid steam condensation may also ruin the existing inerting environment within the containment. Brownian diffusion is one of the essential mechanisms which drives the steam condensation near the wall. In this paper, the distributions of the velocity field, temperature field, and concentration field in the gas mixture boundary layer are firstly obtained through the build-up of a two-dimensional mathematical model allowing for the Brownian diffusion behavior of the mixing gas under the presence of non-condensable gas. Compared with the COPAIN test data and simulation results from the GASFLOW code obtained from previous literature, the applicability and prediction error of the present model are evaluated. Besides, the comparison with the prediction results based on Nusselt’s classical liquid film condensation equation demonstrates that the Brownian diffusion of the mixing gas, including steam and air, will reduce the heat transfer rate through the liquid film by 1–2 orders of magnitude. The theoretical model from the present work can support the development of new CFD codes and help get new understandings of complex thermal-hydraulic phenomena in hydrogen risk under various containment cooling conditions.
Fuel Management Strategy of LBE-Cooled Fast Reactor BLESS-D
Authors: Luo Zhen , Yeoh Eing Yee , Chen Xiaosong , Li Linsen , Xing Mian 
Published: November 2022

Abstract
As one of the six nuclear reactor candidates selected by the Generation IV International Forum (GIF), lead-cooled fast reactor (LFR) has become one of the most promising concepts and attracted more attention from the institutes. A lead-bismuth eutectic (LBE) cooled fast reactor called BLESS-D has been proposed by State Power Investment Corporation Research Institute of China. The fuel management is critical to reactor design because it affects reactor economics. Although the research on fuel management of PWR has matured, there are few studies on fuel management for LFR. Fuel management design includes the entire process from initial cycle to transition cycle, and to equilibrium cycle. To obtain a better refueling scheme and optimize the key parameters of equilibrium cycle, a new refueling scheme is first proposed, and then traditionally, the simulation covering the entire refueling operation is required cycle by cycle, resulting in consuming a massive computing resource. If the scheme fails to meet expectations, expenditure will multiply until the modified scheme meets the safety design criteria. Fuel management design using “pseudo-equilibrium cycle” instead of traditional method is carried out in this work. The “pseudo-equilibrium cycle” method can directly solve the core key parameters of equilibrium cycle by replacing fuel model with approximate nuclides densities estimated from initial core arrangement and refueling scheme. In this paper, a two-batch refueling scheme is proposed with “pseudo-equilibrium cycle” method and then transition cycle are designed to ensure the feasibility that the fresh core successfully transition to the “pseudo-equilibrium cycle” state. Afterwards, neutronics parameters are solved for each cycle from the fresh core and results show that when the burnup calculation comes to the 5th cycle, the reactor key parameters, including assembly peaking factor, linear power density, delayed neutron fraction, and prompt neutron lifetime are in good agreement with “pseudo-equilibrium cycle”, which proves that the “pseudo-equilibrium cycle” method can be used accurately and efficiently to design the refueling scheme.
Study on the Influencing Factors of Oxidation Degradation for High Concentration Polyvinyl Alcohol Wastewater in Heterogeneous Fenton System
Authors: He Ye , Zhou Dongsheng , Liu Feng , Li Xuefei , Su Xingdong , Li Jiawen , Zhan Jie 
Published: November 2022

Abstract
Polyvinyl alcohol (PVA), a high molecular polymer, has good barrier property, acid or alkali resistance and degradability, which could be used as personal radiation protective equipment and pollution control articles in the nuclear industry, resulting in degradable radioactive technical wastes after use. The degradation treatment process of PVA degradable technology waste can realize validity separation of radionuclides and reduce the output of radioactive solid waste. In this study, MnxOy/γ-Al2O3, CuxOy/γ-Al2O3, FexOy/γ-Al2O3 and FexCuyMnzOw/γ-Al2O3 nanoparticles were prepared by the impregnation-calcination method, which were used as catalysts for the oxidative degradation of ultra-high concentration PVA organic waste solution, and the effects of initial pH of polyvinyl alcohol (PVA) solution, hydrogen peroxide (H2O2) dosing, catalyst dosing and initial concentration of PVA solution on the degradation of PVA were analyzed. The results showed that the best overall degradability was achieved by using FexCuyMnzOw/γ-Al2O3 nanoparticles as catalysts with the initial reaction concentration of PVA at 1%, the initial pH of the solution at 3, the reaction temperature at 60 °C, the catalyst dosage at 0.5 g/L and the H2O2 dosage at 60 mL/L. At this time, the concentration decrease rate of PVA could reach more than 95 %, the viscosity average molecular weight of degradation products reached 1395, the molecular weight decrease rate reached 99 %, and the degradation products mainly included butanone, benzaldehyde, hexanal, acetic acid and other unsaturated small molecules. In addition, the mechanism of hydroxyl radical generation by nanoparticles catalyzing hydrogen peroxide was discussed, and the reaction pathway of oxidative degradation of PVA macromolecules by hydroxyl radicals was proposed, which provided a reference for volume reduction treatment of PVA degradable waste in nuclear industry.
Analysis of Radiation-Induced Demagnetization Influence on the Performance of Permanent Magnet Damper
Authors: Wu Hongyu , Yan He , Diao Xingzhong 
Published: November 2022

Abstract
The magnetic vortex damper, a key unit of the control rod drive mechanism of high-temperature gas-cooled reactor (HTR), is a permanent magnet damper that produces eddy current resistance by rotating the conductor disc in magnetic field, and acts as a speed limit during the rod dropping process. To analyze the influence of neutron irradiation leaking from the core on the damping performance of the permanent magnet damper, the mechanism and the influencing factors of radiation-induced demagnetization of the Nd-Fe-B magnet were summarized through literature investigation and a magnetic vortex damper simulation model was established based on ANSYS Maxwell software and verified by experimental data. Current research shows that the magnet with higher intrinsic coercivity and length diameter ratio results in less demagnetization. The magnet permeance coefficient was simulated by the static magnetic field simulation to calculate its equivalent length diameter ratio. According to the literature experimental results, the equivalent length diameter ratio of the magnet, and its intrinsic coercivity, it was conservatively estimated that the remanence attenuation amplitude of the permanent magnet in this study should not exceed 1%. Based on this simulation model, the damping torque before and after the certain amplitude of magnet remanence attenuation was simulated and the corresponding maximum rod dropping speed was calculated. The simulation results show that the damping torque decrease is within 5%, which can meet the service requirements of high-temperature gas-cooled reactor over its service life.