bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.