bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Applicability Analysis of Fusion Reactors and China Nuclear Safety Code HAF102
Authors: Qian Yijie , Li Chong 
Published: November 2022

Abstract
With the development of ITER project, China is preparing to build China Fusion Engineering Experimental Reactor (CFETR), at present; there is no domestic nuclear fusion regulation framework. The nuclear safety regulation framework based on the fission technology is relatively solid and referable, which can provide reference for the establishment of fusion regulation framework. HAF102 “Safety Code on Nuclear Power Plant Design” is one of the most important regulations in Chinese nuclear safety framework. HAF102 specifies the nuclear safety requirements and provides guidance for the design of fission reactors in terms of defense in depth, safety functions and safety classification, these design principles are also crucial for fusion reactors. Therefore, this paper analyzes the general applicability of fusion reactors and HAF102-2016, identifies the non-conformance terms and categories the terms into generally applicable and partially applicable. The application analysis aims to identify the important safety issues that should be considered in the early stage of the development of fusion reactors, in addition, the inapplicable provisions of HAF102 are identified based on the characteristics of fusion reactors. The analysis lays a preliminary foundation for the establishment of a fusion regulation framework on safety issues in China, and on the other hand, it provides recommendations for the design principles of fusion reactors.
Study for the Construction of Gas Leak Detection System in Nuclear Power Plants Using Ultrasound
Authors: Kai Seigo , Moriya Takeshi , Shoji Naruki , Endo Gen , Takahashi Hideharu , Kikura Hiroshige 
Published: November 2022

Abstract
On March 11, 2011, a serious accident occurred at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. due to the Great East Japan Earthquake and tsunami. Currently, nitrogen gas is used to fill the reactor containment vessel, pressure vessel, and plant piping to prevent a hydrogen explosion. Gas leak detection technology is needed to maintain and control the integrity of nitrogen gas containment and to ensure plant safety, as well as to identify the leak location in the event of a leak and understand the nature of the leak before plugging the leak location. However, because of the high radiation environment in the plant, it is not practical for a person to enter the plant to detect gas leakage. Therefore, there is a need for a leak detection method for gases that can be used in a radiation environment, and that is compact, measurable, and can be loaded by robots. In this study, we focused on acoustic gas leak detection as a leak detection technique. The characteristics of ultrasound generated by gas leakage were investigated for the development of gas leak detection technology and source localization system in nuclear reactors.
Comparative Study on Creep-Fatigue Damage Assessment Between R5 Procedures and a Full Inelastic Analysis
Authors: Hurst Antony M. , Chang Louis C. W. , Cho Nak-Kyun 
Published: November 2022

Abstract
This research aims to investigate the differences in structural integrity assessment results between R5 Volume 2/3 procedures and a detailed inelastic analysis. The membrane sidewall/roof bimetallic stub tube subjected to creep-fatigue loading is selected as an example. Based on a linear elastic analysis, the maximum equivalent stress is found to be at the weldment located at the top of the stub tube. The R5 Volume 2/3 procedures enhance the start of a dwell stress and inelastic strain ranges by applying a weld strain enhancement factor and a stress concentration factor, and the assessment predicts a total creep-fatigue damage of 54% for a 45 year life. As a comparison, a detailed inelastic analysis has been conducted for the same bi-metallic stub tube using an incremental, step-by-step analysis. Temperature dependent material properties are used to simulate stress-strain response and a user subroutine is employed for creep deformation. Assessment results for the detailed inelastic analysis confirm the weldment is similarly affected as indicated by the elastic analysis results but the comparable total predicted creep-fatigue damage is less than 35%. Cold eye review of both assessment results concluded that the significant deviation is attributed to the additional factors applied for the assessment of weldments, including weldment geometry, flaw effects and weld micro-cracking, as provided in the R5 Volume 2/3 procedures.
Manufacturing Technology of Mn-Ni-Mo Alloy Steel Heavy Forgings for HPR1000 Nuclear Power Plant
Authors: Zhifeng Zhang , Xiang Li , Xiaofang Wang 
Published: November 2022

Abstract
HPR1000 nuclear power technology is the mainstream technology of commercial pressurized water nuclear power in China, which is adopted by most nuclear power plants under construction or planned. Mn-Ni-Mo alloy steel heavy forgings are widely used in the equipment of pressurized water nuclear reactor. In order to achieve the aim of safe operation for at least 60 years, the technical specifications of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 are high and strict. Due to the high technology and difficult manufacturing, only a few factories have the manufacturing capacity of forgings with stable quality worldwide, which can’t meet the continuous growth demand of Chinese market. Therefore, whether Mn-Ni-Mo alloy steel heavy forgings can be delivered on time has become an important factor restricting the construction cycle of HPR1000 nuclear power plants.
In order to meet the market demand, Shanghai Electric SHMP Casting & Forging Co., Ltd (SCF) has developed the manufacturing technology of heavy forgings for HPR1000. According to the technical specifications of HPR1000, SCF has carried out systematic research in smelting, forging and heat treatment technologies. By 2020, SCF has realized the mass production of heavy forgings for HPR1000 nuclear power plant.
This paper reports new technologies obtained by SCF in the development of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 nuclear power plant.
Sensitivity Analysis on Key Parameters of Severe Accident Source Term of PWR
Authors: Ou Pingwen , Chen Yongzheng , He Dongyu , Chen Peng 
Published: November 2022

Abstract
The complicated severe accident phenomena in typical Pressurized Water Reactor (PWR) Generation III may have a strong influence on source term release into environment and radiological consequence. The study on sensitivity analysis is beneficial to the identification of important factors in severe accident source term analysis and the quantification of their impact.
ASTEC, the integral code of severe accident analysis developed by IRSN, is used to analyze the sensitivity of key parameters of severe accident source term for typical PWR Generation III, with the simulation of safety systems and source term phenomena, in the representative sequence with fast accident progression, Large Break Loss of Coolant Accident (LBLOCA). With the consideration of the design features of typical PWR Generation III and research status of severe accident source term, the key parameters for sensitivity analysis are identified and selected based on the whole process of radionuclides release, including gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction, dose rate and pH value in sump, washing effect, etc.
The sensitivity is quantified by iodine release mass to containment, which is one of the most dangerous radionuclides due to its threat to environment and human thyroid after inhalation and ingestion. The gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction and washing effect are presented in results as the major contributors to the variation of severe accident source term evaluation.
Preliminary Study on Three-Dimensional Characteristics of Modular High Temperature Gas-Cooled Reactor
Authors: Zhang Haojie , Zheng Yanhua , She Ding , Shi Lei 
Published: November 2022

Abstract
The two-dimensional (2D) system analysis codes are usually used to design and analyze the modular pebble-bed High Temperature Gas-cooled Reactors (HTR). However, in some cases, three-dimensional (3D) characteristics, for example the 3D power distribution and temperature distribution due to a part of control rods movement, also need to be concerned. In this paper, based on the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM) design, a 3D system analysis code was used to analyze some three-dimensional characteristics in steady state, as well as the transient process caused by the asymmetric movement of control rods in the side reflectors.
Under normal operations of HTR-PM, the control rods are inserted at certain heights and the primary coolant enters the Reactor Pressure Vessel (RPV) from the outer pipe of the coaxial hot gas duct. As a result, the fuel temperature has a relative circumferential deviation of about 1.64% on the outside of the pebble bed, while a very small deviation at the center of the bed. The temperature distribution of RPV is significantly affected by the position of the helium inlet as well. The analysis results showed that, during the transient process caused by the withdrawal of two control rods, the relative circumferential deviation of the fuel particle temperature on the outside of the pebble bed would increase to around 8.47%. However, because of heat conduction in relatively long distance, the movement of the control rods has very little effect on the circumferential temperature deviation of the fuel particles at the center of the bed. These 3D characteristics of HTR under different conditions demonstrate the good applicability of 2D and 3D system codes and will provide support for more accurate safety analyses in the future.
Transient Thermal-Hydraulic Analysis for Thermal Load Fluctuation Test Using HTTR
Authors: Aoki Takeshi , Sato Hiroyuki 
Published: November 2022

Abstract
High temperature gas-cooled reactor (HTGR) has a potential to produce competitive and large amount of carbon-free hydrogen because of its capability to supply high temperature heat of nuclear energy. It is required to establish the control method and system for the HTGR hydrogen production system to maintain its normal operation against the abnormality in the hydrogen production facility. Performances of the control method and system such as readiness and stability have to be demonstrated by transient thermal-hydraulic analysis with a code developed based on RELAP5/MOD3 code.
The reactor response against the disturbance in the reactor inlet coolant temperature is revealed in the HTGR hydrogen production system using the control systems. The analytical results showed that the reactor outlet coolant control system enabled to control the variation of the reactor outlet coolant temperature within 4°C against 30°C of large disturbance in the reactor inlet coolant temperature. A variation of the reactor outlet coolant temperature has to be less than 5°C in the simulation to maintain its normal operation in the HTGR hydrogen production system. Thus, the effectiveness of the control method was confirmed. The simulations code and results will be validated through a HTTR thermal load fluctuation test.
Reliability Analysis of Phased Mission Systems Using GO-FLOW Methodology
Authors: He Zhanyu , Yang Jun , Matsuoka Takeshi , Yang Ming 
Published: November 2022

Abstract
A Phased Mission System (PMS) is defined as a system subject to multiple, consecutive and non-overlapping operation phases of tasks during its mission, which is commonly found in complex technological and engineering practices such as aerospace, nuclear and military operations, etc. Over the past few decades, extensive efforts have been devoted to advance the theories, methods, techniques and tools used for reliability analysis of PMS. The general methods include fault tree based combinatorial approaches and Markov chains. In this paper, a success-oriented GO-FLOW method with a new exact algorithm is presented for reliability analysis of multi-phase mission systems. The feasibility and correctness of the extended GO-FLOW method (GFA) for PMS analysis are proved by comparing the results with fault tree analysis based on Sum of Disjoint Products (SDP) generation algorithm for two example case studies. The comparison results show that the GO-FLOW method can be effectively applied for reliability analysis of PMS in a very compact way. Consistent results can be obtained using extended GO-FLOW algorithm when compared to the fault tree analysis.
A Knowledge Retrieval Method for Layout Design of Nuclear Power Reactor Coolant System Based on Hypernetwork and Deep Structured Semantic Model
Authors: Wang Bin , Hu Bingtao , Zhang Zhifeng , Feng Yixiong , Tan Jianrong 
Published: November 2022

Abstract
The layout design of nuclear power reactor coolant system requires a large amount of knowledge that satisfies many disciplines, which will waste designers a lot of time retrieving relevant knowledge in the design process. In order to obtain the knowledge efficiently and accurately, a knowledge retrieval method based on hypernetwork and deep structured semantic model (DSSM) was proposed. The knowledge hypernetwork model consisted of a designer sub-network, a design task subnetwork, and a design knowledge resource sub-network. Nodes in each sub-network and between different sub-networks were connected through special edges, which represented correlation degree information. Then an improved DSSM model was used to evaluate relevance at the semantic level between user query information and knowledge elements in hypernetwork. Correlation scores will be obtained based on relevance at the semantic level, and knowledge elements with lower scores will be removed during the process. Finally, the Bayesian method was used to calculate the knowledge recommendation probability to obtain the most relevant knowledge retrieval results. The knowledge retrieval results were sorted from high to low according to the calculated probability. A case study conducted in this work showed that the proposed approach was effective in capturing relevance at the semantic level and supporting efficient and accurate knowledge retrieval services.
Construction and Application of Mitigation System Performance Index (MSPI) Platform for a China Nuclear Power Plant
Authors: Zhang Jie , Wang Jing , Zhang Peiying , Tan Kun , Yang Minxue 
Published: November 2022

Abstract
The mitigation system performance index is used to evaluate the ability of the safety-related system to be put into operation to mitigate the accident when an accident occurs in the nuclear power plant (NPP). This index can effectively monitor the status of the safety system and reasonably reflect the impact of equipment performance on the safety level of the nuclear power plant. Typical pressurized water reactor (PWR) nuclear power plant mitigation systems include emergency alternating current (AC) power system, high pressure safety injection system, auxiliary feedwater system, residual heat removal system and cooling water support system. Based on the probabilistic safety assessment (PSA) model, this paper develops a nuclear power plant MSPI computing platform to monitor and judge the risk level of the mitigation system, and guide the nuclear power plant to carry out production activities such as risk-informed preventive maintenance, corrective maintenance and production planning. This paper takes the Essential Service Water System (SEC) of a nuclear power plant in China as an example to verify the proposed method, and conducts the trend analysis of MSPI in the 8 quarters of 2020–2021. The results show that MSPI can timely reflect the influence of the unavailability and unreliability of system equipment on the overall safety status of the power plant from the perspective of risk.
Research and Development of Rock Cavity Disposal Technology in China
Authors: Yin Wencheng , Pan Yuelong , Zhang Xueling , Liu Yu , Ma Jian , Li Tongtong 
Published: November 2022

Abstract
Rock cavity disposal is regarded as a preferred option of radioactive waste final disposal solutions with less disturbance to surface and better isolation from people and surface environment which facilitates the long term surveillance and control of facility both during operation and after closure. This paper provides the research progress of key aspects of rock cavity disposal technology, such as site selection criteria, engineering design scheme of disposal vaults, technique process design of the handling of waste packages, research & development of intelligent control system and equipment of waste transportation, documentation of safety case and supporting safety assessment, numerical simulation of radionuclides migration process, and etc. The first practice of rock cavity disposal in China is also discussed. It shows that developing the rock cavity technology is reasonable, feasible and has significant potential in China. Finally, this paper looks forward to the research focus in the next stage. It is suggested to study the compatibility between the complexity level of engineering barrier measures and the waste hazards and to improve the reliability of future evolution and scenario assumption. The nuclides migration behavior study and migration parameter determination involving more key nuclides in waste from other origins are also of importance.
Research on Severe Accident Management Guidelines for HPR1000 NPP
Authors: Pingting Jiang , Peng Chen 
Published: November 2022

Abstract
The Severe Accident Management Guideline (SAMG) is an integrated document for a nuclear power plant (NPP) to deal with severe accidents effectively, with the basic objective to maintain the integrity of the containment in both short term and long term as far as possible. After Fukushima Daiichi nuclear accident, insights on severe accident prevention and mitigation are put forward in the process of a newly designed nuclear power plant. Hua-long Pressurised Reactor (HPR1000) is a generation-III NPP in China, with necessary dedicated system and equipment that designed at the beginning during severe accidents, and is also planned to have associated procedures after accidents. The paper focuses on development process of SAMG for HPR1000 nuclear power plant. Firstly, the characteristics of the plant are sorted, with both dedicated equipment required for severe accident mitigation and other equipment. Then the main objective and principles in developing SAMG are defined. After that, the structure of SAMG is discussed and finally determined referred to the latest insights all over the world and also considering the good practice in China. Additionally, diagnosis process and severe accident strategy are described. In relation to this, flexible strategy in HPR1000 NPP is discussed and included in the integrated SAMG.
Design of Liquid Level Control System of Steam Generator Based on Neural Network PID Controller
Authors: Xiao Longhao , Sun Peiwei , Wei Xinyu 
Published: November 2022

Abstract
Steam generator (SG) is an important equipment of the nuclear power plant, and the stability of its liquid level affects the safe operation of the nuclear power plant. SG is a complex system with nonlinear, time-varying, nonminimum-phase, small stability margin and large time delay. In actual operation, it is difficult for classical PID control to ensure a satisfactory control performance.
In this paper, the neural network methods are used to optimize the parameters of the PID controller, and a neural network controller is designed. The controller of the system consists of two components: a classical PID controller, which realizes control through a closed loop; a single-hidden-layer neural network based on the BP (back propagation) model. The neural network calculates the coefficients of the classic PID controller through matrix operations. Two weighting matrices are adjusted according to the gradient descent method to reduce the loss function and realize the training process. The control system is deployed to a SG simulation model through Simulink. The typical working conditions are simulated and investigated. The control performance is compared with that of the classical PID controller.
Through analysis, it is confirmed that the neural network PID control system can meet the control requirements with fast response speed, short settling time, stable control effect under various working conditions, and strong anti-interference ability. The results prove that the neural network control has greater advantages and better application value than the classical PID controller.
Extension of the IVR Analysis Module in Modeling the Ex-Vessel Water Cooling
Authors: Yu Peng , Zhao Yuer , Wang Chen , Yang Xiaoming , Ma Rubing , Yuan Yidan 
Published: November 2022

Abstract
In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is an important severe accident mitigation strategy widely adopted in some advanced light water reactors, e.g. AP1000 and HPR1000. The reactor pressure vessel is supposed to be flooded externally by water, such that the decay heat could be efficiently removed and the radioactive corium materials retained inside the vessel. Lumped parameter models are used in engineering design to calculate the melt pool heat transfer and thus assess the effectiveness of IVR. In this methodology, a simple constant temperature is normally assumed for the ex-vessel surface accounting for the external water cooling. Since the RPV and cooling water are coupled in thermal behavior, modelling of the related heat transfer between them may be of interest. Under the PISAA framework, we developed the IVR module for the IVR analysis. In this paper, we further extend the code capability by considering the ex-vessel heat transfer. Possible conditions of the external cooling water e.g. pure liquid phase convection and boiling as well as the corresponding heat transfer correlations would be considered. The newly implemented model is then verified with acceptable results achieved. With this new feature, the coupled heat transfer analysis of the RPV and external cooling water is investigated. Compared with the original treatment with constant temperature boundary conditions, the new treatment with convective heat transfer affects the predictions of outer vessel temperature and residual wall thickness while the influence on the inner vessel temperature is negligible.
Heat Transfer for Forced Flow of Helium Gas in a Small Diameter Tube With Different Heated Length
Authors: Honjo Yushi , Xu Feng , Liu Qiusheng , Shibahara Makoto 
Published: November 2022

Abstract
As one of the cooling methods for the blanket in a fusion reactor, forced convection using helium gas has been considered as one of the promising candidates. In this study, in order to clarify the effect of tube length on heat transfer for forced flow of helium gas under various velocities and pressures, experiments on forced convection heat transfer for helium gas in a small diameter tube were conducted. A circular tube made of platinum with an inner diameter of 1.8 mm and a heated length of 50 mm was used in this experiment. The tube was heated with exponentially increasing heat inputs. As a result of the experiment, the heat transfer process can be considered in the quasi-steady state when the e-folding time is larger than about 1.5 s. In addition, the heat transfer coefficient increased with the increases in velocity and pressure. The heat transfer was also higher than that of conventional turbulent heat transfer correlation. By comparing with the experimental results of tube with different heated lengths, it was found that the heat transfer coefficients for the heated length of 50 mm were higher than those of the one with a length of 90 mm.