bluuebunny/crossref_metadata_2025 huggingface dataset:
Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit
Authors: Zhang Shengtao , Yi Ke
Published: November 2022

Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient.

This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in certain PWR nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.
Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Basic Device Specifications
Authors: Morita Koji , Liu Wei , Arima Tatsumi , Arita Yuji , Sato Isamu , Matsuura Haruaki , Sekio Yoshihiro , Sagara Hiroshi , Kawashima Masatoshi
Published: November 2022

A new subassembly-type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device can passively provide a large negative reactivity to the core. In this study, the nuclear and thermal properties of the device subassembly with metallic fuel were evaluated to determine the basic device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWe-class MOX-fueled SFR core showed that a conventional homogeneous core can maintain stable cooling of the core prior to coolant boiling in the driver fuel subassemblies. On the other hand, the negative reactivity required to terminate the event by device operation was found to be slightly larger in the low sodium void reactivity core than in the conventional homogeneous core.
Application of Vacuum Hot Helium Leak Examination Technology for ITER Blanket Shield Blocks
Authors: Li Weiping , Jiang Guoyan , Luo Wei , Fang Renjun
Published: November 2022

The Blanket Shield Blocks is an important part of the International Thermonuclear Experimental Reactor (ITER). In order to ensure its sealing performance, each shield blocks needs to be taken a vacuum hot Helium leak testing. A method for high-sensitivity examination using a vacuum thermal helium leak test system is proposed in this paper. The shield blocks can remove its own impurity gas and increase the activity of Helium atoms by increasing the environment temperature. The examination sensitivity is improved by increasing the pressure difference between the inside and outside of the shield block. The sensitivity of this method can realize the helium leak test with the system sensitivity of 5 × 10−11 Pa·m3/s under the conditions of examination environment temperature of 250°C, ambient pressure of 1 × 10−5Pa, and products internal helium pressure of 4MPa.

The results indicate that the technology has a high degree of automation and data traceability, which can avoid the influence of the product’s own outgassing on the examination, and has a high sensitivity to the examination of subtle air leakage.
Review on Electromagnetic Compatibility Research and Practice in Nuclear Power Plants
Authors: Yao Wenqing , Shang Xuelian , Chen Rigang
Published: November 2022

This paper studies the electromagnetic compatibility (EMC) for the instrumentation & control system, communication devices and electric equipments in commercial nuclear power plants (NPPs). It reviews the EMC research activities and practices for NPPs in United States, France, Russia and China. The paper first familiars the readers with electromagnetic interference (EMI) knowledge and examples that are most related to NPP EMC issues, including EMI source, coupling path and sensitive equipment. It then summarizes the historical research activities, focusing on the development of EMC standards for NPPs in those four countries. It is concluded that the current EMC standard systems are not uniform and lack of specific test procedures and quantification methods. They are insufficient for the nuclear industry to execute. At last, recent popular research topics for NPPs are reviewed including systematic EMC design, EMC design and control measures, EMC prediction and electromagnetic environment monitoring. Similar research work done in the general EMC community or by other industries has been introduced. It indicates that for NPPs EMC design especially at the system level is still immature, electromagnetic simulation can be a good tool to assist the EMC design and that EMC control regulations and environment monitoring become more and more necessary as wireless components and devices are being used.
Reliability Analysis of Off-Plant Power for Multi-Units
Authors: Honghao Chen , Cheng Ren , Jian Wu , Deyi Liu , Yang Luo , Shengjia Zou , Zilong Wang
Published: November 2022

Globally, multi-units are located at the same nuclear power plant site. After several serious nuclear accidents, the overall risk assessment of a nuclear power plant with multiple reactors has attracted much attention. Combined with the concept of PSA (probabilistic safety analysis) for multi-units, this paper analyzes the off-plant power supply configuration of 9 units in Qinshan Nuclear Power Base. In view of the 6KV emergency busbar directly affected by the instability of the off-site power grid, the correlation among several units is fully considered, and the exploratory research and analysis of off-site power grid reliability is carried out in Qinshan Nuclear Power Base. Taking the reliability of external power grid of Qinshan Nuclear Power Base as an example, the reliability of emergency bus and the influence range and contribution of various analysis factors to the evaluation results are quantitatively analyzed through fault tree modeling. The results of model analysis are in line with the actual situation. The analysis concept based on the multi-unit PSA methodology should be used in a wider range. For example, thematic analysis should be carried out for other highly correlated originating events (common cold source, equipment common cause failure, etc.), and the development and progress of multi-unit PSA should be promoted through practice with a point-to-surface approach. In the follow-up research, we should improve the model system, take the actual situation into full consideration, and develop a set of overall risk assessment system that can reflect the real-time risk level of multiple unit site.
Numerical Investigations of the LBE Flow and Heat Transfer Characteristics in the Helical-Coiled Tube Bundle
Authors: Shen Cong , Liu Limin , Liu Maolong , Gu Hanyang
Published: November 2022

Due to the advantages of compactness and enhanced heat transfer efficiency, the Helical-Coiled Once-Through Steam Generator (HCOTSG) is one of the best choices to be applied in the Lead-bismuth Fast Reactor (LFR). In order to investigate the flow and heat transfer characteristics in the shell side of lead-bismuth eutectic (LBE) HCOTSG for the purpose of safety analysis and optimization design, a CFD method for calculating the heat transfer and flow resistance in helical-coiled tube bundles is proposed. Based on available relevant experimental studies, the numerical model is validated. Validation results indicate that the liquid LBE cross flow through helical-coiled tube bundles can be well simulated using SST k-ω model and turbulent Prandtl number model. The maximum deviation for both friction factor and Nusselt number is less than 15%. After validating the reliability of the numerical model, geometry models with different helix angles of the shell side are established. Effects of helix angle on the flow resistance and heat transfer performances are presented. The flow and heat transfer characteristics in the helical-coiled tube bundle are numerically simulated and analyzed. This work provides a new method for studying the flow and heat transfer characteristics and optimization design of liquid metal H-OTSG.
Research on Public Communication of Small Reactor
Authors: Zhu Rongxu , Zhao Feng , Zhang Xiaofeng , Guo Jiandi , Zhang Meng
Published: November 2022

Based on the conclusions of small reactor environmental impact assessment and social stability risk assessment of small reactors, this paper conducts a special study on the public communication of small reactors. This article discusses the work content, scope and form of public communication of small reactor, and give suggestions on the scope and form of public publicity, public participation and information disclosure. It is matched with small reactor public communication theoretical system and operation manual.

The research on the public communication work of small reactor can provide necessary theoretical basis and technical support for the specific public communication work of small reactor. Provide decision support for local policies and construction departments to implement public communication. It is hoped that this study can provide some reference for the public communication of small reactors in the exploratory stage.
Applicability Study on Human Reliability Analysis for Severe Accidents in Nuclear Power Plants
Authors: Zhao Ming , Chen Shuai , Xu Jie , Zhou Jianzhang , Qing Tao
Published: November 2022

Three severe nuclear power accidents (Fukushima accident, Chernobyl accident and Three Mile Island accident) have proved that human errors in the mitigation of severe accidents will exacerbate the deterioration of accidents and cause immeasurable consequences. Therefore, there is an urgent need for finding human reliability analysis methods to effectively analyze and predict possible human errors in severe accident mitigation.In order to compare the applicability of the Human Reliability Analysis (HRA) method under the severe accident in the nuclear power plant, this paper put forward four criteria, qualitative analysis, quantitative analysis, traceability and availability, based on the characteristics of the personnel in the process of severe accident mitigation. Several commonly used HRA methods were compared and analyzed. The results show that the common HRA methods which applied to the analysis have shortcomings, so it is necessary to establish a new HRA method, or to improve the original methods to solve the issues under severe accident conditions and the IDAC model and Phoenix HRA methodology are considered better choice. Combined with computer programs,the IDAC model can simulate the process of accident and the dynamic response of personnel, which helps to explain the dynamic interaction characteristics of human behavior and environment, but the model still has some limitations in the aspects of uncertainty analysis and simulation data availability. While Phoenix method provides some Performance influencing factors and crew failure modes that can reflect the response characteristics of emergency personnel. This applicability study can provide guidance for the establishment of HRA method in severe accidents.
Study on Radial Flow Characteristics of U-Pattern 4-Wire and 7-Rod Bundles
Authors: Zhang Longzai
Published: November 2022

Wound wire bundles are often used in reactor structural components. The winding wire can not only prevent the vibration of the pipes, but also strengthen the heat exchange between the pipes. Because of its advantages in heat transfer, the wound wire bundle structure has been used in both core and OTSG.In this paper, the u-shaped 4-wound wire bundle structure is regarded as the research object, and four commonly turbulence models are used for calculation based on CFD program. Radial flow characteristics under different turbulence models are obtained through the given velocity inlet. What’s more, it is found that the SST k-ω model are shown better performance, when applied in the 4-wound wire bundle structure with the consideration of computer performance and calculation accuracy. The radial velocity of the three axial paths and the flow pressure at the center of the B-B slit show periodic distribution with a period of one quarter pitch. Due to the interaction between wire winding, there appears obvious flow pulsation in the radial direction of the central sub-channel in the structure of 4-wound wire bundle.
Study on a High Reliability Power Supply Scheme for Distributed Control System of Nuclear Power Plant
Authors: Hongpeng Tian , Lin Zhang
Published: November 2022

Distributed Control System (DCS) is regarded as the “brain” of nuclear power plant due to its function required in guarding nuclear safety. Power supply system is vital to supply electric power for DCS to maintain normal operation. Distribution design requirements of DCS power supply system are discussed firstly on analyzing the requirements of IEEE, IEC, etc. Combining with the comparative analysis of power supply schemes of DCS system in typical nuclear power plants based on CPR1000, AP1000 and EPR, key factors restricting the reliability of DCS power supply system are analyzed based on relevant design rules and standards in this paper. The method of improving the reliability of DCS power supply system is explored by analyzing the type of power supply, system design, equipment selection and design of electrical raceway. Thus a new high reliability power supply scheme for DCS system is presented based on mature equipment manufacturing technology. The scheme enhances the reliability, stability and flexibility of the distribution system by reducing the converting process on distribution path, increasing the design of cross-tie between divisions, and making full use of the advantages of DC distribution system. Through system operation analysis, equipment analysis and fault response analysis, this paper discussed the high reliability and feasibility of the new scheme, providing a new research direction for the design of DC system in nuclear power plant.
The Effect of Specimen Size and Surface Roughness on Tensile Strength of Nuclear Graphite
Authors: Guan Yuxin , Zhan Yingjie , Wang Shuiyong , Wang Qingwu , Ti Wenxin , Gong Bing
Published: November 2022

The tensile strength of nuclear graphite is an important parameter in evaluating and calculating the structural strength of High Temperature Reactor-Pebbled Modules (HTR-PM). The tensile strength of nuclear graphite with varied sizes and surface roughness was tested, and the size effect tensile strength formula was created. The results reveal that when the size of nuclear graphite increases, its tensile strength falls. However, in the 4 mm to 8 mm range, the size impact is less noticeable. The Weibull modulus m0.95 value is calculated based on the experimental data, and the size effect formula is established. The difference between the estimated value and the experimental value computed using the size effect formula is less than 10%. Tensile strength may be calculated using the size effect formula for nuclear graphite specimens of various sizes. The impact of surface roughness on tensile strength is linked to the grinding direction and specimen size. After a certain level of surface roughness, the tensile strength diminishes when the surface scratches are perpendicular to the tensile force. The stronger the effect, the smaller the specimen size. When surface scratches are parallel to the tensile tension, the tensile strength does not vary much as the surface roughness increases.
Experimental Study on Structural Vibration Characteristics of Single Rod in Different Flow States of Water
Authors: Yin Yibo , Wen Jiming , Yu Xinyang , Li Xin , Gao Puzhen , Tian Ruifeng , Sun Junshuai
Published: November 2022

In the nuclear reactor, flowing state of surrounding fluid will change structural vibration characteristics of fuel rods, which may lead to severe resonance and damage of fuel rods. Therefore, it is necessary to study structural vibration characteristics of fuel rods in different fluid flow states. In this paper, based on experiment, structural vibration characteristics of a single rod in different water flow states and different restraint modes are obtained by hammering method. The results show that different restraint modes have influence on dry mode, wet mode and structural vibration characteristics of the rod. Under condition of no excitation, when flow rate is lower than 1m/s, the response frequency and flow velocity change trend is the same under different restraint modes, response frequency of test rod with increases of flow rate. When flow rate is more than 1m/s, the response frequency varies with flow rate in different restraint modes. Under the fixed support, the excitation ways and excitation positions have no effect on natural frequency of rod. For simply supported, the excitation positions have great influence on the natural frequency of rod. It has reference significance for the safety design of fuel rods.
Evaluation System of Human-Factor Training Effect Based on Hierarchical Model
Authors: Zhong Xu , Xu Xia , Liao Xianglin , Tu Yichun , Song Yeshu , Liu Jiajin , Zhao Qian , Chen Xuefang
Published: November 2022

Nuclear power enterprises attach great importance to staff training and have invested a lot of resources to establish training and authorization systems based on Systematic Approach Training (SAT), including demand analysis, outline design, material preparation, training implementation and evaluation feedback five parts. However, the lack of effective training effect evaluation means in practice makes it difficult to continuously improve the training system. This paper takes human-factor training in nuclear power plants as an example, analyzes the characteristics of human factors training, such as complex training object, rich content and diverse methods, compares the Kirkpatrick model, Context, Input, Process, Product (CIPP) model and Context, Input, Reaction, Output (CIRO) model. The kirkpatrick four-layer model is selected as the theoretical basis to establish the evaluation system for the effect of nuclear power human factor training, including a closed-loop training evaluation process, five evaluation methods and corresponding tools. After a period of practice in a nuclear power operation team, this system has been proved to be able to achieve effective evaluation and continuous improvement of the human-factor training effect in nuclear power plants, and effectively improved the long-term performance of the operation team, the non human-factor mistake record has been more than 900 consecutive days. In view of the problems found in the system practice, the paper also puts forward suggestions for further improvement.
Probabilistic Analysis of Containment Overpressure Failure in Level 2 PSA
Authors: Liu Yu , Li Wenjing , Lu Wenkui , Wang Gaopeng
Published: November 2022

Containment is the final barrier to prevent any leakage and release of radioactive materials. The containment internal pressure may exceed the design value during severe accident, which could threaten the containment integrity. Therefore, the study on fragility and failure probability of containment under accident conditions has become a crucial issue in nuclear safety. In this paper, firstly, the PWR containment failure modes are identified based on relevant researches and engineering practices. Secondly, the containment failure caused by overpressure is divided into the structural failure and functional failure, which are discussed in detail. Finally, combined with both deterministic and probabilistic analysis methodology, fragility curve of PWR containment was evaluated as a step on the evaluation of containment failure probability due to the overpressure in Level 2 PSA.
Research on Accident Diagnosis Method of Reactor System Based on XGBoost Using Bayesian Optimization
Authors: Liu Yongchao , Li Xiangyu , Liang Biao , Wang Bo , Tan Sichao , Gao Puzhen
Published: November 2022

Traditional machine learning algorithms have problems such as overfitting, low accuracy, and difficulty in hyperparameter optimization when performing fault diagnosis.In order to improve the accident diagnosis ability of nuclear power plant reactor system, this paper combines Bayesian optimization (BO) algorithm with eXtreme Gradient Boosting (XGBoost) algorithm to develop a reactor accident diagnosis model.First, data preprocessing and feature quantity analysis are performed on accident data samples.Then, the BO algorithm is used to optimize the hyperparameters of the XGBoost model. Finally, the BO-XGBoost model is used to diagnose the operating conditions of seven nuclear power plants, and the diagnostic effects of various traditional machine learning classification algorithms are compared and analyzed.The results show that the BO-XGBoost model can achieve more efficient and accurate identification of reactor accident types, and the model has better generalization ability.This research can help nuclear power plant operators to accurately identify the types of reactor accidents, assist decision-making, and ensure the safe operation of nuclear power plants.
The Research of the Customer Needs of a Type of Innovative Lead-Bismuth Fast Reactor Base on the TRIZ Method
Authors: Yuan Weiquan , Chen Zhao , Shi Kangli , Su Chang , Tang Jinghua , Cui Dawei , Duan Chengjie
Published: November 2022

Lead-bismuth fast reactor is one of the six generation IV reactors proposed by GIF. It is a kind of innovative reactor which has higher safety performance and economy. This kind of reactor can be used for producing electricity, producing heat for industry and other applications. Extracting the oil under the ground is one of them because the high quality steam produced by the reactor facilitates the extraction of oil. Considering that its situation of application is strongly different from the normal reactors which produce electricity and which are built for a long time, the customer needs of this kind of reactor should be well researched before its design.

The TRIZ is a kind of systematic analysis method for innovation. It is a series of tools directs the inventor to invent in an optimal way. In TRIZ tools, there are some methods for recognizing customer needs. Main Parameter of Value (MPV) is one of them. Main Parameter of Value is a description of the parameter(s) that contribute most to Customer Value. MPVs are the most important parameters that should be considered in the design.

This article formulates MPV analysis method and applies the MPV analysis on finding the main parameters for the design of the lead-bismuth fast reactor for oil. With this method, the most importance parameters of reactor are determined and ranked. This work can provide the experience of MPV analysis on the reactor design.
Development and Verification of Neutron and Photon Ultrafine Group Library for Fast Reactor Physical Calculation
Authors: Zhang Teng , Ma Xubo , Jia Guanqun , Ma Xuan , Zhang Bin , Hu Kui
Published: November 2022

In order to improve the accuracy of fast reactor physical analysis, two libraries with 1968-group neutron and 21-group photon were generated based on ENDF/B-VIII.0 and ENDF/B-VII.1 data by using NJOY2016. A code, named TXMAT2.0, was developed to process the two libraries to generate ultrafine group neutron and photon cross sections and Kinetic Energy Release in Material (KERMA) factors. To perform the verification of the two libraries, ICSBEP benchmarks for critical verification, and the sample 1D benchmark were selected. Several results were in good agreement with reference data. For the RBEC-M benchmark, the power distribution based on the ultrafine group library was good.
Analysis and Mitigation Strategy on the Decreasing Opening of Thermal Pressurizer Pilot Operated Safety Relief Valves
Authors: Hu Wensheng , Guo Chaofan , Zhao Haisheng , Ruan Zhaoji , Min Jidong , Du Pencheng , Wang Xinxin
Published: November 2022

The pressurizerpilot operated safety relief valves (abbr. POSRV) protect the reactor coolant system against overpressure situation under accident conditions. It can be used us safety valves, relief valves or isolation valves. It’s one of the most important valves in PWR. The POSRV of a nuclear power unit adopts the thermal pilot structure design which is the first application of the third generation nuclear power unit. The POSRV composed of main valve and metallic pilot detector with special sealing design which can withstand the high temperature and high pressure medium in the primary circuit. During the first cycle operation of the nuclear unit, the opening of the isolation valve of the POSRV decreases, the discharge line temperature of metallic pilot detector continuous rise and the leakage rate of the primary coolant circuit increases abnormally, which bring great retreat risk to the unit. Combined with the structural principle of the POSRV and the change trend of key parameters, the analysis shows that the reason for the decrease of the valve opening is the leakage of the three-way sealing surface of the metallic pilot detector, which leads to the introduction of pressurized medium into the valve head of the main valve, then reduces the valve opening and affects the leakage rate of the primary coolant circuit. By formulating mitigation strategies such as energizing the solenoid valve to change the pilot valve state, draining the leakage medium in the primary coolant circuit and the unit operation decision-making scheme, the valve availability and unit safety are ensured, which provides a reference for similar equipment problems.
Preliminary Study on the Migration of Polonium-210 in Lead-Based Fast Reactor
Authors: Qiu Yali , Zeng Qin , Chen Hongli
Published: November 2022

210Po has become a non-negligible radioactive safety problem in the fourth generation Lead-based Fast Reactor (LFR) due to its high toxicity and volatility. It will be produced by neutron irradiation of coolant and emits α ray. This may cause potential internal exposure hazards when workers are working in the plant. Polonium produced in the primary coolant circuit of the Reactor will be released to cover gas due to its volatility. This process is slow but not negligible. Therefore, the migration of polonium in the whole reactor and the distribution in the containment room for the M2LFR-1000 was preliminary analyzed. According to the radionuclide migration theory, the migration of polonium was analyzed theoretically, and the distribution of polonium was simulated by FLUENT. Two situations are simulated, one is normal operating condition and the other is leakage accident of covering gas. Preliminary calculation shows that polonium is distributed homogeneously throughout the space with the accumulation of radionuclides only at the certain special place, and the concentrations below safe levels under normal operating conditions. The method of analysis procedure can also be applied to analyze the distribution of other radionuclides.
The Experimental Study on Critical Heat Flux in Vertical Eccentric Internally Heated Annuli
Authors: Guo Junliang , Kong Huanjun , Gui Miao , Peng Yujiao , Shan Jianqiang
Published: November 2022

An experiment was carried out to obtain data on critical heat flux (CHF) in vertical internally heated annuli cooled with Refrigerant (R-134a) at high-pressure subcooled and low-quality conditions (outlet pressure:1.78–2.72 MPa, mass flux: 587–2135kg/m2s, critical quality: −0.64–0.12). The test section has two forms, concentric, eccentric (ε = 0.435, 0.783). At inlet conditions, parametric trends of the CHF data in concentric annuli have been shown consistent with those obtained in experiments in tubes. At local conditions, the reversed trend of CHF with increasing mass flux occurs a lower critical quality. The effect of eccentricity on CHF depends mainly on local flow conditions. At high subcooled conditions, eccentricity causes punishment to CHF and CHF decreases with the increase of eccentricity. The eccentricity effect appears to be diminished at low subcooled conditions. An interesting phenomenon occurs in high pressure and high flow (HPHF) conditions, where CHF with eccentricity of 0.783 is higher than CHF with eccentricity of 0.435. This is because the larger cross flow causes bubbles to be more easily leave away from the narrow gap in annuli with high void migration. The effect of eccentricity on CHF was studied from the mechanism of two different types of CHF (DNB and Dryout) in vertical eccentric internally heated annuli.
High-Temperature Deformation Constitutive Relationship of Zr-Sn-Nb Alloy for Nuclear Power
Authors: Gao Yingqi , Wang Ben , Cao Jianguo , Luo Qianqian , Cao Yuan , Zhang Zhihao , Wang Lian , Gao Bo
Published: November 2022

Zr-Sn-Nb zirconium alloys are widely used in the field of nuclear power instead of traditional zirconium alloys and can be used as structural materials for nuclear power fuel assemblies. In this paper, the constitutive relation of high temperature plastic deformation of Zr-Sn-Nb alloy for nuclear power is described. Gleeble-3800 thermal simulation tester was used to conduct compression tests at different temperatures (450∼700°C) and deformation rates (0.01∼10s−1), stress-strain curves under different deformation conditions were drawn based on stress-strain data. According to the analysis of flow stress curve types, Zr-Sn-Nb alloy shows obvious temperature sensitivity and strain rate sensitivity. Under the condition of low temperature and high strain rate, the power dissipation efficiency is calculated according to Murty instability discriminant. Zirconium alloy material shows flow instability due to the formation of adiabatic shear band. In order to control the successful rolling of microstructure, careful process design must be carried out to avoid the flow instability zone. Based on the types of stress-strain curves, the curves are divided into work hardening stage and dynamic recovery (DRV)/ dynamic recrystallization (DRX) stage. A piecewise constitutive model is established, which can accurately predict the machining properties of Zr-Sn-Nb alloy for nuclear power.
Development and Experiment on High Temperature and Radiation Resistance Concrete Material for Spent Fuel Dry Storage
Authors: Xueling Zhang , Cheng Cheng , Yongjie Pan , Yuanyuan Wei , Qionghui Tang , Zengfen Li , Shouhai Yang
Published: November 2022

To solve the problem of low thermal safety margin, bad heat conductivity and lack of experiment data for performance against accumulative radiation in current concrete developed for spent fuel dry storage all over the world, a specific high performance concrete material with three mixed components and high tension steel fibre has be proposed and developed. It is proved by the experiment data that 1) the concrete material can withstand 180 °C for long term, 300 °C for 168 hours and 800 °C (as an accidental case) for 20 minutes, after which there is no observable cracks and flakes; 2) the heat conductivity of the material can be enhanced 27%, with the help of steel fibre inserted, which can significantly enhance the heat load of spent fuel assemblies to be loaded in dry storage container; 3) the material can withstand a total neutron dose of 1 × 1017n/cm2 and gamma dose of 2.4 × 107Gy, which is expected during 60 years of spent fuel storage. The performance of the material is better than that of current concrete material used for spent fuel dry storage all over the world. The material enhances the thermal safety, structural safety and radiation protection safety during spent fuel dry storage, which can be applied at the same time in other technical area with high temperature and high radiation conditions.
Risk Monitoring of Spurious Safety Injection in Pressurized Water Reactor Nuclear Power Plants
Authors: Huang Yu , Jiao Zhenying , Wang Zhenying , Wang Taike , Kong Kaihe , Li Min , Zhang Jianwen
Published: November 2022

The typical third generation pressurized water reactor nuclear power plants trigger safety injection signal to actuate the emergency core cooling system based on the saturation margin of coolant in hot legs of primary circuit, i.e. ΔPsat in hot legs, which represents the difference between saturation pressure and actual pressure in hot legs. In order to identify the risk of spurious safety injection during falling the reactor back to safety shutdown state after steam generator isolation following an accident on secondary side for which safety injection is unnecessary, this paper analyzes thermo hydraulic phenomenon in the hot legs where the steam generators are isolated. By selecting reasonable monitoring parameters, the risk monitoring curve of spurious safety injection, which clearly marks off the risk zone of spurious safety injection, is plotted based on the actuation logic of safety injection. The curve has been integrated into the post-accident operating state monitoring and tracing system for primary circuit, and has been verified and validated thanks to the full scope simulator. It is concluded that the risk of spurious safety injection can be dynamically and effectively monitored following the accidents on secondary side, so that operators can timely take appropriate measures to avoiding spurious safety injection and fall the reactor back to safety shutdown state smoothly.
Research on Key Assembling Technology of Tube Bundle Assembly of Hualong-1 Type Steam Generator
Authors: Wanbing Ye , Yuanbin Liu , Yaomao Chen , Sanjun Zhang , Aili Wang
Published: November 2022

Fangchenggang Phase II nuclear power project adopts the third generation advanced nuclear power technology of Hualong-1. The steam generator tube bundle assembly of the project adopts HL1000 type, which was manufactured for the first time in China. The tube bundle assembly is composed of three parts: tubes, support assembly of straight section of tube bundle and anti-vibration bar assembly of U tube bend. The structure of tube bundle assembly is novel, involving several kinds of parts and components, high assembly accuracy requirements and strict assembly sequence. During manufacturing of the tube bundle assembly, key manufacturing technologies including the installation and alignment of 9 pieces of Lattice Grid Assembly (diamond shaped holes) and tubesheet holes, anti-sagging control of tube bend section, integrated connection of anti-vibration components and anti-offset of anti-vibration components were successfully in command. The development of HL1000 tube bundle components was successfully completed, the independent research and manufacturing of products was realized, and the blanks of this new reactor manufacturing technology in China was filled in. This paper summarizes and analyzes the key assembly technology of HL1000 tube bundle assembly, so as to provide experience for the design and manufacture of steam generators of other reactor.
Analysis of Severe Accident Induced by Loss of Coolant Accident for a Small Pressurized Water Reactor
Authors: Bi Xiaolong , Sun Peiwei , Wei Xinyu
Published: November 2022

Small pressurized water reactors (PWRs) have become a new trend in the current nuclear energy development due to their many advantages, such as compact equipment layout, high thermal efficiency, and strong cycle capability. Compared with large PWRs, small PWRs are designed to reduce the coolant inventory and increase the core power density, which is not good for nuclear safety. Severe accident studies on large PWRs cannot be directly applied to small PWRs. Loss of coolant accident is one of the main inducements of reactor core melting, which needs to be focused on prevention and treatment. It is of great significance for the safe operation of small PWRs to analyze and study severe accident induced by loss of coolant accident.

In this paper, MELCOR is used to establish the severe accident analysis model of the primary loop system of a small PWR, and the loss of coolant accident is introduced to obtain the accident sequence from the shutdown of the reactor until the core degradation. At the same time, the core pressure, core liquid level and other key parameters are analyzed.

The results show that in the case of a severe accident, compared with the large PWR, the small PWR takes a faster time for the pressure of the primary circuit and the containment pressure to reach equilibrium after the break accident occurs. The unbalanced radial power distribution causes the cladding of the 3rd ring to fail first. In the later stage of the severe accident, the melt plays a major role in heating the coolant. During the entire core degradation process, the upper fuel assemblies start to melt first, and the core does not completely collapse. The research results can provide reference for the formulation of severe accident management guidelines for small PWRs.
Development of Intelligent Straightening Equipment Control System for Pipe Components
Authors: Lin Wang
Published: November 2022

In the field of nuclear industry, pipe components are often used as common parts. Because the processing process will bend the pipe components, it needs to be straightened before leaving the factory to meet the practical application. In order to improve the detection accuracy and processing efficiency of pipe components, the control system of intelligent straightening equipment is designed according to the manual straightening method. According to the characteristics of end jump detection, an algorithm based on secondary data filter is designed to achieve high precision measurement. According to the material and structure characteristics of the components, an adaptive algorithm of downdraft is designed to realize difference recognition and automatic calculation. By studying the manual process and combining with the theory of automatic control, an efficient and safe automatic testing and straightening process is designed, which realizes the standardization and intelligence of straightening. Using STM32F407 self-developed main control unit, data communication, algorithm calculation, logic control, data access and fault detection are realized. Combined with touch screen, the whole process of detection and alignment is visualized. The intelligent straightening equipment with this control system has been used in practical production. The detecting precision and straightening efficiency are greatly improved. It has friendly human-machine interaction mechanism and reliable safety protection mechanism to meet the actual production needs.
Research on Operational Intervention Levels for the “Ground” Scenario in High Temperature Gas-Cooled Reactor
Authors: Liu Ao , Liu Tao , Zhang Liguo
Published: November 2022

Nuclear emergency is the last barrier of defense in the defense-in-depth system of nuclear safety. Operational Intervention Level (OIL) is an important principle in nuclear accident emergency response. When a nuclear accident occurs, together with other operating standards, OIL guides emergency personnel to take appropriate protective actions rapidly, so as to reduce the harm to the personnel and environment to the lowest. In 2017, International Atomic Energy Agency (IAEA) released a new publication on OIL——ERP-NPP-OILs (2017), which is based on light water reactors (LWRs). Due to the differences between high temperature gas-cooled reactor (HTR) and LWR in core type and accident process, the determination OILs will also be different. This paper conducts the application of method in ERP-NPP-OILs (2017) to HTR especially for OIL1 and OIL2, which is mainly focuses on the “ground” scenario. Taking high temperature gas-cooled reactor – pebble-bed module (HTR-PM) as an example, the calculation process and result of OIL1 are given. It has foundational significance for further completing the OIL system matched with HTR characteristics.
Design and Evaluation of a Prototype Portable Panel PET Imaging System
Authors: Wen Jing , Pei Changxu , Niu Meng , Li Yingguo , Li Yurong , Zhang Qinghua , Huang Chuan , Yin Yongzhi , He Lixia
Published: November 2022

Positron Emission Tomography (PET) is a nuclear imaging modality used in clinical diagnostics and biomedical research to show metabolic processes in an organism. Movable PET has great significance for the practical application of tumor diagnosis. In this paper, we propose a panel PET imaging system with high portability, and further evaluate a virtual-pinhole PET (VP-PET) system with higher spatial resolution. To evaluate the resolution of the PET prototype, we build a rotation stage coupled with two PET detector modules. PET detectors are composed of 12 × 12 LYSO crystals and Position Sensitive Photomultiplier Tubes (PSPMT). The central position of the two coincidence detectors is consistent with the rotation center, where a 22Na point source with a 3 mm diameter and an activity of 1 μCi is placed. The stage rotates several times on the horizontal plane, which enables signal acquisition within the range of 360°. A TCA DAQ system from Natuq is used for signal waveform acquisition, and FPGA algorithm programming is performed. After removing the coincidence detection data of repeated angles, the filtered back-projection (FBP) algorithm is used to reconstruct the 22Na point source image. Since the pixel size of the LYSO crystal directly affects the spatial resolution of the PET system, we set experiments of five types of LYSO crystals with pixels from 0.9mm to 2.0mm on VP-PET geometry. Meanwhile, the relationship between the spatial resolution of VP-PET with the pixel size of LYSO crystal is compared and analyzed in the experiments. In addition, in the Geant4 simulation, the two-dimensional image reconstruction of MRI glioma phantom is carried out by the panel PET system, which further proves that the VP-PET system has accurate spatial resolution and high sensitivity.
Application of ALARP Concept in Nuclear Power Plant Design — A Case Study in HPR1000
Authors: Li Dongyuan , Qing Shi , Luo Haiying , Wang Jinkai , Xu Jiaoshen
Published: November 2022

As Low As Reasonably Practicable (ALARP) is one of the most important nuclear regulatory concepts in the UK Generic Design Assessment (GDA). ALARP requires that nuclear power plants have the duty to achieve the lowest level of residual risk provided this is reasonably practicable. Unlike As Low As Reasonably Achievable (ALARA), ALARP concept should be considered not only in radioprotection, but also in nuclear safety areas such as fault studies, structural integrity, etc. Different from the traditional licensing requirements in other countries, ALARP concept is not prescriptive but emphasizes self-evaluation, risk-orientation and balance of all aspects. It is therefore difficult to demonstrate that the design has reduced nuclear risks to an ALARP level without a systematic approach during the licensing. In this study, the UK context regulatory requirements of ALARP (including the Tolerability of Risk, Relevant Good Practice, Gross Disproportion, and Numerical Targets) are investigated based on which an ALARP approach for Hua-long Pressurized Reactor (HPR1000) is developed. In this approach, potential improvements can be identified through the review against Relevant Good Practice and risk insights. Risks can be reduced through the implementation of preferred options determined by optioneering process. The ALARP position of the whole plant design can be evaluated through holistic risk review. The ALARP approach has been successfully applied to HPR1000 and recognized by the UK authority during the GDA process. The ALARP approach facilitates the balance of safety, economy and environmental friendliness during nuclear power plant design, and can be guidance to the robust licensing and regulation.
Development and Verification of New Drum Filter Equipment for Nuclear Power Plants
Authors: Xiaojie Huang , Yishan Lin , Honghu Xie , Yanqing Li , Jinchun Yang
Published: November 2022

Drum filter is a large-scale filtration equipment in nuclear power plants, providing clean cooling water for nuclear power plants. At present, there are problems of easy blockage, short response time of the unit after blockage, and shutdown of the unit in severe cases during the operation of the drum filter in the in-service nuclear power plants. The analysis shows that the drum filter adopts the structure of “internal water intake”, and the garbage such as marine organisms remain on “inside” of the drum filter during operation, which makes it difficult to clean up and operate the inner garbage. When the marine organisms suddenly break out, if the amount of garbage cleaned by the drum filter is not enough to offset the amount of garbage entering, the garbage will accumulate continuously inside, which may lead the pressure difference between the inside and outside of the drum filter to suddenly increase, and then cause the circulating pump to trip or even the reactor to shut down. The inherent safety of the drum filter of this water intake method is low. To solve this problem, this paper puts forward a new concept of water intake of drum filter, that is, the concept of “external water intake”, which changes the water intake method of the drum filter, intercepts the garbage on the “outside” of the drum filter. Under the action of water flow, the garbage outside is not easy to remain on the net and block the drum filter, thereby fundamentally improving the inherent safety of the drum filter. A new type of “external water intake” drum filter was developed by using parametric and modular design method. The reliability of the drum filter was analyzed by using the FMEA failure mode and impact analysis method, and the weak links of the drum filter were found out, and optimization and improvement measures have been taken for the weak links. The ANSYS finite element analysis software is used to conduct a seismic analysis of the drum filter, which proves that the drum filter meets the requirements of the RCC-M specification and the seismic requirements of the third-generation nuclear power plants, which demonstrates the safety of the drum filter. Finally, through the physical prototype test, the design concept of “external water intake” was verified, and garbage cleaning capacity of the new drum filter meets the requirements of nuclear power plants. This type of drum filter has been popularized and applied to many HPR1000 projects under construction and newly built ones.