bluuebunny/crossref_metadata_2025 huggingface dataset:
Seismic Isolation Scheme Design of 200MW Nuclear Heating Reactor
Authors: Hou Nannan , Dong Jianling
Published: November 2022

Air pollution and the energy problem have become concerned and urgently needed to be solved all over the world in recent years. Coal-fired boilers in industrial parks are an important source of air pollutants. Low temperature heating reactor built in industrial zone is an ideal substitute for coal-fired boiler because of its excellent inherent safety and cleanliness. The new type nuclear heating reactor named NHR200-II was proposed by Institute of Nuclear and New Energy Technology of Tsinghua University in 2015[1]. Compared with the previous generation NHR200, the thermal parameters of NHR200-II are improved which can not only be used for regional heating but also realize the multi-purpose like power supply, steam supply and seawater desalination. And the economic benefits are also greatly improved. The design of NHR200-II is based on the design, construction and operation experience of NHR5 which has maintained running safely up to now and NHR200. The main technical features include three circuit design, integrated layout of primary circuit, full power natural circulation, and automatic self-pressurization.

NHR200-II uses the traditional anti-seismic technology at present. However, the requirements of seismic performance of reactors have generally increased after Fukushima disaster. This paper tries to apply seismic isolation in NHR200-II to meet the improved requirements of seismic performance and realize the standardized design by which the magnitude of design ground motion is not restricted in the future. Base isolation, which means setting isolation devices like elastomeric bearings or spherical sliding bearings between the foundation and the superstructure, has been widely applied in normal buildings and bridges, but still needs particular design for nuclear reactors.

This paper describes NHR200-II seismic isolation system, which consists of the foundation, raft, isolation layer and the whole nuclear island as superstructure. In addition, the paper also tells the design process of the system, including designing of the raft, selecting of isolators and their layout design based on the environment and mechanical property, etc. And the particular calculation method of the performance parameters and acceptance criteria are also shown in this paper. Finally, the overall design method of seismic isolation is given as a conclusion.
Automation of High-Fidelity Modeling of PWR Fuel Rods and Verification
Authors: Chunyu Zhang , Yang Wang , Yongjun Deng , Zifan Song , Yong Lu
Published: November 2022

Definition of a high-fidelity finite element model of a fuel rod is complex and error-prone. In order to make the models affordable for the purpose of design and analysis, this study developed an automated procedure for generation of the finite element models of fuel rods within the general-purpose finite element modeling environment, i.e., ABAQUS/CAE, by using the parametric modeling technique with embedding the best modeling strategy that balances both efficiency and accuracy. The generated models are then verified against typical experimental measurements. Based on the results, several physics models, such as the thermal conductivity model and the fission gas release model, are evaluated. It was shown the automated modeling program can build a wide range of rods within several minutes. Important parameters such as the centerline temperature, the burnup distribution, the elongation of the fuel active zone and the diameter change of the cladding can be predicted with high accuracy under real power histories. The fission gas release can be qualitatively analyzed as well. A combination of the automated modeling program and the fuel performance analysis program therefore allows rapid high-fidelity modeling of real fuel rods and quantitative evaluation of their performance, which provides an important tool for assisting the design of advanced fuels.
Integral Simulation on the Thermo-Mechanical Behavior of the UMo/Al Dispersion Fuel Plate Using Finite Element Method
Authors: Yang Guangliang , Feng Wenpei , Ding Tao , Chen Hongli
Published: November 2022

Fuel plate performance under irradiation is an important topic for the research reactor. The fuel plate is expected to deform after a long-time operation, which is a threat to the reactor safety. In this research, irradiation behaviors like interaction layer growth, swelling, creep and plasticity of the UMo/Al dispersion fuel plate are carefully reviewed from literatures. Taking these behaviors into consideration, the longtime three-dimensional stress update algorithms for the fuel meat and cladding are developed respectively based on the large-deformation incremental constitutive relation. Since local properties in the fuel meat change with burnup, equivalent material properties models are used to take account of this characteristic. These models and algorithms are implemented into Abaqus user subroutines to simulate the thermo-mechanical performance of the fuel plate under the typical operating condition in research reactors. In order to illustrate the macroscopic deformation of the fuel plate and reduce the computational cost, 1/4 symmetrical geometry of the fuel plate is employed in the simulation. The evaluation and distribution of temperature, stress and deformation of the fuel plate are obtained and analyzed. Results show that the growth of interaction layer affects the temperature profiles heavily. The stress concentration mainly locates at the contact interface between fuel meat and cladding, especially at the side of the interface.
Modeling and Analysis of Tritium Fuel Cycle for Magnetic Confinement Fusion Reactor
Authors: Xu Jinggang , Zeng Qin , Wang Sheng , Wang Huaihuai , Zhou Junjie , Xiong Yangbin , Yang Jingchen
Published: November 2022

The realization of tritium self-sufficiency has become an important step to the study of tritium cycle for fusion reactor. To realize tritium self-sufficiency of tritium cycle system of fusion reactor, a typical tritium cycle model of magnetically confined fusion reactor is established on the basis of CFETR in this paper. Tritium cycle model and tritium self-sufficiency are also studied. Based on the mean residence time method, the mass balance equations of the model are developed. Considering the tritium retention of the feeding system and plasma, the specific scheme of the model for calculating the minimum initial startup tritium inventory and the minimum tritium breeding ratio required for tritium self-sufficiency is given. The tritium inventory and the tritium treatment capacity requirements of each subsystem are obtained. In addition, the tritium inventory is also obtained through Simulink modeling and analysis. This is also a method to solve the problem, which can be used as a reference for the modeling and analysis of fusion reactor.
Evaluation of Effect of Depressurization Timing on the Thermal Load To Primary Containment Vessel During Severe Accident at BWR
Authors: Matsumoto Taisei , Yamaji Akifumi , Fujiwara Daisuke
Published: November 2022

According to accident of a Boiling Water Reactor (BWR), the operators generally do not depressurize the Reactor Pressure Vessel (RPV) until low-pressure core injection systems are recovered. However, when the core water level decreases below a certain limit, the RPV is depressurized to prevent the RPV failure at high pressure, which may lead to failure of the Primary Containment Vessel (PCV) by direct heating of the ejected melt. In such a case, if recovery of the core cooling cannot be expected, earlier depressurization may be preferable from the viewpoint of reducing thermal load on PCV, because it may effectively reduce the amount of fission product (FP) aerosol deposition on the RPV dome and reduce the risk of PCV top head heating via convective and radiative heat transfers from the RPV dome. However, significance of such risk is not evident as the steam dryer and separator may hinder direct aerosol deposition on the RPV dome and the RPV dome is thermally insulated. Moreover, the narrow space between the RPV dome and the PCV top head may hinder convective heat transfer. In this study, sensitivity analyses have been carried out using MELCOR-2.2 to support such understanding.
Experimental and Numerical Research on Droplet Drag Coefficient and Deformation Coefficient in Nuclear Power Plants
Authors: Li Ru , Tian Ruifeng , Chen Bowen , Wang Bo
Published: November 2022

There are many multiphase flow phenomena in steam generators (SG) of nuclear power plants (NPPs) and the movement of droplets affects the separation efficiency of dryers in SG. And the most important factor affecting droplet movement is drag force on droplet. Drag force coefficient (DRC) and deformation coefficient (DEC) as characterization coefficient of drag force have important research significance. In this paper, water and silicone oil are used as continuous phase and dispersed phase, respectively. Study on variation of the DRC when Re is in the range of 30–1200 is carried out. Firstly, the motion of droplets (MDs) of different sizes is visually studied by using the high-speed camera and images of MD is acquired. The droplet contour is recognized based on principle of boundary differentiation and the centroid coordinates of the droplet are determined, thus DRC and DEC are obtained. Besides, the relationship between DRC and DEC are researched. Additionally, Lattice Boltzmann Method (LBM) is used to simulate the droplet motion. The simulation results are compared with the experimental results, thus verifying the feasibility of LBM method to simulate the MD. The results show that when velocity of droplet (VD) is low, DRC is inversely proportional to VD. While VD is high, DRC is constant. When the droplet diameter (DD) is small, the final VD is proportional to DD, and when DD is large, the final VD is proportional to the square of DD. The DEC is linearly related to We. The larger the We, the smaller the DEC. Shan-Chen model of LBM is feasible for droplets with low Reynolds number (Re), while the simulation of droplets with high Re is the future prospect.
High Efficient Mineralization of Cesium in Waste Liquid by Hydrothermal Method
Authors: Gao Kai , Feng Wendong , Yan Xiaojun , Guo Xiaobin , Cui Xiaoyan , Guo Xiliang
Published: November 2022

During the nuclear fuel cycle, a large amount of radioactive cesium is produced, and at the same time, it is difficult to safely dispose of because of its extremely soluble in water and low boiling point. The safe disposal of radioactive cesium is a problem that needs to be solved urgently. The currently used methods to treat cesium-containing wastewater such as cement fixation, glass solidification and ion exchange, have certain drawbacks, and urgently need to be further optimized. Pollucite is an analcite molecular sieve. Many studies have shown that it is one of the potential final choices for radioactive cesium. Pollucite has an appropriate pore size, without destroying the structure of pollucite, radioactive cesium cannot diffuse out of pore of pollucite, pollucite also has a good stability and high cesium loading rate. In previous studies, pollucite needs to be synthesized under high temperature conditions above 1000°C. This method will cause the volatilization of cesium and increase the difficulty of exhaust gas treatment. How to quickly synthesize pollucite at a lower temperature has become an important research direction for the safe disposal of caesium. Different from the traditional calcination method to synthesize pollucite, this paper uses hydrothermal method to simulate the generation environment of zeolite in nature, and realizes the synthesis of pollucite at a lower temperature. Using cesium contained in simulated wastewater as synthetic raw materials, combined with commonly used silicon sources and aluminum sources as synthetic raw material, through exploring different reaction conditions such as reaction temperature, reaction time, and alkali addition, the hydrothermal synthesis of pollucite was finally successfully realized. Then the influence of different reaction conditions on synthesis and cesium removal efficiency was investigated, and its growth mechanism was also analyzed. Finally, it was realized that pollucite was successfully synthesized under 150°C hydrothermal conditions using common chemical reagents as raw materials, which further promoted pollucite application in the safe disposal of radioactive cesium.
Logic Equation of Operating Situations and Man-Machine Interface Design for NPP
Authors: Si Tianqi , Wang Yuqi , Du Yu
Published: November 2022

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Deuterization Process and Result Analysis of Resin in Heavy Water Purification System of a Reactor
Authors: Zhang Wen Juan , Xu Fengxia , Hua Xiao
Published: November 2022

This paper introduces in detail the whole process of resin replacement of heavy water purification system in a reactor reconstruction project in Algeria, including old resin discharge, pipeline cleaning, new resin filling, new resin deuterization, water quality detection, heavy water storage and continuous operation of the system. By sorting out deuterization data and tracking and analyzing the water quality detection results, it is considered that the resin deuterization process of the heavy water purification system of a reactor is smooth. After deuterization, the effluent concentrations of the three resin columns of the heavy water purification system meet the operation requirements. The whole deuterization process tries to save precious high concentration heavy water for Algeria. After deuterization, the resin column is connected to the heavy water purification system. After continuous operation and sampling detection, it is determined that the purification effect is ideal and the pH value is not affected. The resin of the heavy water purification system is successfully replaced and can be put into use. At the same time, aiming at some problems in the deuterization process, this paper analyzes the causes in detail, summarizes the experience and lessons, and gives optimization suggestions and precautions in the deuterization process.
Research on the Construction of the Advanced Nuclear Power Standards System Platform
Authors: Zhang Shaolin , Yang Wen , Dong Ruilin , Liang Yun , Liu Shangyuan
Published: November 2022

Advanced nuclear power standards system is a new type of nuclear power standards system composed of all kinds of standards in the nuclear energy field.

At present, the research results of the two stages of the subject and the achievements of nuclear power standard construction in recent years are scattered. It is necessary to establish a single nuclear power standard database to process and develop the standards in the field of nuclear power and various research results, reports, process documents, data and other databases, so as to provide a good data environment for the majority of nuclear engineering scientists and technicians.

In order to solidify the research results of advanced nuclear power standard system and nuclear power standard data, the application support platform of nuclear power standard system adopts innovative SaaS cloud service mode to build an Internet-based cloud service platform with process engine and knowledge map technology.

Through the integration of informatization and standardization, standardize the standardization process, we can provide convenient and fast information services, vigorously expand the scope of participants in nuclear power standardization and the depth of participation of all parties, and realize the standard system management mode of autonomy, industry participation and dynamic updating.
Analysis of Different Requirements of Dissimilar Metal Joint Welding Procedure Qualification Between RCC-M and ASME BPVC
Authors: Pei Wang , Dongxu Chen , Shumin Yang
Published: November 2022

In order to enhance the comprehension of the welding procedure qualification standards, RCC-M (Design and Construction Rules for Mechanical Components of PWR Nuclear Islands) and ASME BPVC (American Society of Mechanical Engineers Boiler and Pressure Vessel Code), used by the components fabrication for nuclear power industry in the People’s Republic of China, this paper takes the qualification of dissimilar metal joint between inlet-nozzle and safe end of steam generator in Generation III nuclear power unit as an example, taking a comparative approach, compares the requirements of different welding variables one by one between two standards and then lists the differences and analyzes the differences. It shows that RCC-M has more strict requirements than ASME BPVC on the qualification of dissimilar metal joint with buttering, the type of test couple qualified by RCC-M is more similar to product than that by ASME BPVC, and qualification by RCC-M may need more test couples than that by ASME BPVC, if choosing same value of variables which have overlap between these two standards, test coupon qualified by RCC-M can used for that qualified by ASME BPVC, but not vice versa. At last, the paper gives some implementation advises and guidance of subsequent qualification works during design and fabrication.
Analysis on the Nuclear Safety Supervision Mode of the World’s Major Nuclear Power Countries and Its Enlightenment to the Improvement of China’s Nuclear Safety Supervision Technical Support Ability
Authors: Wu Han , Yu Guoxin , Teng Keyan , Zheng Xiangyang
Published: November 2022

In the view of the International Atomic Energy Agency (IAEA), the decision-making of safety supervision of civil nuclear facilities needs independence. In order to deeply evaluate and review a series of issues related to the safety of civil nuclear facilities, national regulatory authorities usually need to consult independent institutions or external committees with professional capabilities. Such organizations providing technical services are called Technical Support Organizations (TSO), which can be national laboratories, research institutions or consulting units. In some cases, support from professional organizations in other countries may also be required.

Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. All along, China has continuously strengthened the construction of nuclear safety supervision capacity, and the safety supervision capacity has been gradually improved.

China has established a nuclear and radiation safety supervision system, including National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional nuclear and radiation safety supervision stations, Nuclear and Radiation Safety Center (NSC) and other units. China has established a nuclear safety expert committee, and mainly relies on the National Nuclear and Radiation Safety Supervision Technology R & D Base (R & D Base) as the central level test, verification and R & D platform to provide consultation and technical support.

The R & D base has completed the construction of some test, verification and R & D functions. China has built a series of important laboratories in the R & D base, it still has not formed R & D system. In the future, China need to improve the technical support capacity of nuclear safety supervision.

In short, the independence of nuclear safety regulatory decision-making requires to enhance from institutional independence to capability independence. The independence of China’s nuclear safety regulatory authority has been basically established. In the future, it is more important to improve the regulatory capacity. Only by establishing a perfect technology R & D system can nuclear safety decision-making be based on sufficient scientific basis.
Fire Probability Safety Analysis for a Compact Small Modular Reactor
Authors: Chen Yanzhu
Published: November 2022

Traditional fire protection design applied in the second generation nuclear power plant,was based on determinism and reinforced by Fire Probability Safety Analysis,in which Problems such as difficult selection of fire protection-related technical solutions, low efficiency of fire protection layout, and large rework workload for later technical solutions update.For highly compact SMRs with tight design schedule, it’s impossible to acceptable such shortcomings.

According to the requirements of SMRs fire protection design,this paper based on NUREG/CR-6850 fire probability safety analysis methodology, combined the design depth (the concept design phase)of SMRs, develop the SMRs fire PSA model, calculated quantitatively Contribution of core damage frequency (CDF) of fire PSA model.At the same time, based on the results of internal fire PSA analysis, the design weakness is found, and the design/improvement proposal based on fire PSA analysis is given.
Review on Reactor Coolant Pumps Trip Criteria After Small Break LOCA in PWR
Authors: Jiao Zhenying , Wang Zhenying , Huang Yu , Liu Liu , Wang Taike
Published: November 2022

The post-accident operation status of reactor coolant pumps (RCP) play an important rule in accident mitigation, especially for the loss of coolant accident (LOCA), the prolonged RCP operation after medium and small LOCA will aggravate the deterioration process of RCS water inventory. This paper first analyzes the adverse influence of delayed trip of RCPs after medium and small break on RCS, then the principle of determining the trip criteria of RCPs is suggested, the trip criteria of RCPs from the worldwide nuclear steam supply system venders are investigated and summarized. From the perspective of prolonging the operator ‘s non-intervention time, reasonable measures to alleviate the consequences of medium and small LOCA, such as the RCP automatic trip logic and RCS fast depressurization by cooling, etc., are of great significance to improving the safety level of the plant.
Modeling and Simulation of a Pressurized Water Reactor Nuclear Power Plant and Power System
Authors: Wang Haomin , Wei Xinyu
Published: November 2022

As the proportion of nuclear power in the power grid continues to increase, for the safe, stable, and economical operation of the power system, nuclear power plants must participate in the peak shaving operation of the power system. Therefore, it is very important to study the impact of the power system on nuclear power plants. In this paper, a nonlinear dynamic model of a pressurized water reactor (PWR) is built on the MATLAB/Simulink simulation platform. The model includes the point kinetics reactor core model, the thermal-hydraulics model, the piping & plenum model, the steam generator module, the reactivity module, the sensor model and the turbine module, in addition to the governor module and the simulation model of the power system. Based on the established dynamic model of nuclear power plant and an island power system module, the response of the model was tested under load variations. There are often some short circuit faults in the power system. To verify the stability of the model, the impact of the single-phase-to-ground fault in the power system on the nuclear power plant is analyzed. In this paper, the nuclear power plant can maintain stable operation in the event of a transient failure in the power system.
Development of an MPS Code for Corium Behavior Analysis
Authors: Jian Lijun , Lu Hongfan , Yu Peng , Yuan Yidan
Published: November 2022

After fukushima nuclear accident in Japan, phenomenology of serious accident has been paid more and more attention. And particle methods have a natural advantage in computational simulations involving deformation and free surfaces. The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: a heat transfer model for thermal conductivity and a phase change model for Solid-liquid conversion, an enthalpy-based viscosity model to realize a smooth transition of viscosity at the solid-liquid interface and a surface tension model for better simulation of surface shape. Validation of the developed simulation approach is carried out with a classical water column collapse example. The heat transfer model is validated by the example of a one-dimensional plate. The surface tension model is validated by a case of droplet oscillation. A comprehensive example of the melting of “Wood’s alloy” is carried out to verify the capacity of MPS method in the simulation of melting and expansion. The simulation results are in good agreement with the experimental results, which indicates that MPS method promises well in the field of severe accidents.