bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Tritium Behavior Issues in High-Temperature Gas-Cooled Reactors
Authors: Zhou Ziling , Li Chuan , Gui Nan , Xie Feng , Wen Yanwei , Shan Bin , Fu Jia , Fan Qunchao 
Published: October 2021

Abstract
The very-high-temperature reactor (VHTR) has been included in the conception of the Generation IV nuclear energy system proposed by the United States Department of Energy Nuclear Energy Research Advisory Committee and the Generation IV International Forum. After the Fukushima accident, concern has grown regarding the safety of nuclear power plants. For its inherent safety, superior proliferation resistance, and high thermal efficiency, the VHTR and its prototype, the high-temperature gas-cooled reactor (HTGR), attracted worldwide interest. Tritium exists in the fuel element and graphite reflectors in the core, primary coolant, some liquid and solid wastes, and gaseous effluent of HTGRs. Due to its permeation behavior at high temperatures, tritium is the most important nuclide in the secondary loop of HTGRs. Considering the future application of HTGRs in hydrogen production, tritium will be non-negligible and could be difficult to separate from hydrogen. In addition, the interaction between tritium and graphite is expected to be very strong. The absorption, desorption, and diffusion of tritium at the interface between the primary coolant and graphite are of great interest. Experimental measurements of the diffusion coefficients of tritium and its isotopes do not agree well with theoretical predictions based on the consideration of isotope effects. In this paper, research progress on the permeation, absorption, desorption, and diffusion behaviors of tritium in HTGRs is summarized and discussed. The chemical form of tritium in the primary coolant, and its spectroscopic characteristics, internal structures, and transport behaviors are organized and analyzed. Basic properties related to tritium, such as isotope effect, nuclear spin, radioactivity, etc. are also discussed.
Thermal Hydraulics Analysis of a High-Performance Once-Through Steam Generator With Annular Narrow Slot Tube
Authors: Han Jinyu , He Wen , Guan Shunran , Zhao Chenru , Bo Hanliang 
Published: October 2021

Abstract
The integrated small modular reactors (SMR) have been attracting more and more attentions in the field of nuclear energy. For PWR-type SMR with full-power natural circulation, a built-in steam generator with high heat transfer performance is required to enhance the natural circulation capability of the core. The heat exchanger with annular narrow slot tube takes the advantages of heat transfer enhancement of narrow slot and double-sides heat exchange in both internal and external sides, and results in higher heat transfer efficiency compared with the conventional heat exchanger with straight tube. Therefore, it is regarded as a promising design scheme of built-in steam generator with great potential. However, there have been relatively few researches on the design and calculation of steam generator with narrow slot tube. On the other hand, the flow boiling heat transfer in micro/mini channels (generally with hydraulic diameter less than 3 mm) is more complicated than that in a regular flow channel. Heat transfer correlations developed especially for micro/mini-channels with high accuracy are required for accurate predictions.
In the present paper, thermal-hydraulic design calculations and analysis for the once-through steam generator (OTSG) with annular narrow slot tube. The thermal-hydraulic performance was quantitatively evaluated and compared with that of the conventional OTSG with straight tube, such as the OTSG designed by B&W, based on the heat transfer coefficient, heat exchange area, area density, power density, steam generator volume power and other parameters.
Finally, the working performance of the steam generator and the coupled heat transfer characteristics with the primary loop of natural circulation under steady-state operating conditions were studied, and the feasibility and applicability of the OTSG with annular narrow slot tube used in a full-power natural circulation integrated SMR were discussed.
Research on the Diagnosis Model of Break Diameter During the Blowdown Process of SBLOCA
Authors: Ke Bingzheng , Gao Puzhen , Cheng Kun , Wu Lingyan , Wang Bo , Chen Bowen , Tian Ruifeng , Wen Jiming 
Published: October 2021

Abstract
The small break loss of coolant accident (SBLOCA) is a typical accident in the nuclear power plant, and the diagnosis of accident parameters in a small break loss of coolant accident (SBLOCA) is a very important research direction in the field of nuclear safety. The timely and accurate judgment of the break parameters can help the operator understand the serious situation of the SBLOCA and avoid more serious consequences. In this study, a physical model of the blowdown process during SBLOCA was established by using the theory of two-phase critical flow, flash evaporation, and energy conversion. Then a diagnostic model for the break diameter of SBLOCA was established by combining the NSGA-II and the physical model, and the experimental results obtained by the blowdown experiment of simulated pressurizer under a low pressure condition were used to verify the accuracy of results obtained by the diagnostic model. The results show that the diagnosis model can be used to obtain the break diameter during the blowdown process of SBLOCA. In different conditions of the blowdown process, the errors between the diagnosis results and the experiment results do not exceed 20%. The calculation time of the diagnostic model is less than 400s.
Numerical Analysis on the Thermal-Hydraulic Characteristics for the Reactor Main Vessel Cooling System of Chinese Sodium Cooled Fast Reactor
Authors: Song Ping , Feng Tangtao , Zhang Dalin , Chen Lie , Li Shaodan , Lin Yuansheng , Qiu Suizheng 
Published: October 2021

Abstract
As a critical component for pool-type Sodium cooled Fast Reactor (SFR), the main vessel plays a major role in holding the entire sodium in the pool. While the Reactor Vessel Cooling System (RVCS) would contribute to the cooling of the main vessel. This paper develops a one-dimensional code and the numerical analysis has been conducted to evaluate the thermal-hydraulic characteristics for RVCS for China Experimental sodium-cooled Fast Reactor (CEFR). Firstly, the code is applied to the analysis on the steady characteristics of RVCS and the results are compared with the design value. The simulation results show that the main vessel is cooled effectively and the comparison results are in good agreement with the design value. Then, in order to study the transient characteristics of RVCS, the code is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The simulation finds that a reverse flow in RVCS shows up and RVCS could take the decay heat out effectively, which contributes a lot to the core cooling. The code evaluated the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The results are very promising but further experimental data is still needed to validate the code.
Corrosion Behavior of Superalloys in High Temperature Gas Cooled Reactor in Impure Helium With Corrosion Time
Authors: Li Haoxiang , Du Bin , Zheng Wei , Wang Qiuhao , Yin Huaqiang , He Xuedong , Fan Hua , Ma Tao 
Published: October 2021

Abstract
There are trace impurities such as CO, CH4, H2, H2O in helium of main cooling loop of high temperature gas cooled reactor, which will cause corrosion of superalloy materials used in steam generator and intermediate heat exchanger at high temperature. Previous experiments and literatures show that the formation of chromium rich protective oxide layer on the surface of superalloy is an important factor to prevent severe corrosion of superalloy. In this paper, the corrosion tests of four candidate materials Inconel 617, Incoloy 800H, Hastelloy X and T-22 for high temperature reactor steam generator were carried out. The four superalloys were exposed at 950 °C for 50, 100, 150 and 200 hours in helium with a certain impurity content. The corrosion results were analyzed by weighing, scanning electron microscopy (SEM) and electron probe microanalysis (EPMA). The corrosion of these four alloys occurred in different degrees. In this paper, we mainly analyse the conclusion of 50h and 200h corrosion of these four kinds of superalloys. On the whole, Hastelloy X and T-22 have good corrosion resistance and low mass gain; Inconel 617 and Incoloy 800H have poor corrosion resistance, obvious internal oxidation and high mass gain.
Research on Electric and I&C Equipment Safety Function Classification of Nuclear Power Plant
Authors: Yuqi Wang , Qian Sun 
Published: October 2021

Abstract
Classification of System, Component and Structure (SSC) is the base as well as high level demand of nuclear power plant. Equipment classification including electric and Instrument and Control (I&C) equipment is the precondition of correct design regulation and standard. Safety function classification is key pass of electric and I&C equipment classification. This paper researches the method of nuclear power plant electric and I&C equipment safety function classification. Firstly from view of function, it explains the importance of function classification. Then function analysis and classification of equipment is implemented by design order. Lastly from view of accident analysis, function classification is validated, and a complete approach of function classification is formed. The purpose of this paper is the NPP electric and I&C equipment safety function classification as an example, to study and summarize the method of the electric and I&C equipment safety function classification, and to provide the basis for specific items design work according to design requirements. At the same time, a practical method is provided for other similar NPP electric and I&C equipment classification work. The electric and I&C equipment function classification of nuclear power plant satisfy the basic principles requirement of relative nuclear power rules and codes. It provides an important basis of equipment classification for next nuclear power plants.
Flux Rate Calculation and Analysis of the Integrated Small Pressurized Water Reactor Based on Monte Carlo Method
Authors: Yang Wen , Chao Fei , Tai Yun , Li Longze 
Published: October 2021

Abstract
The neutron and photon flux rates are important parameters for safe reactor operation, refueling and decommissioning, scientific applications and radiation protection. For the Integrated small pressurized water reactor, the advanced reactor core analysis code CORAL, the source calculation code ORIGEN-II and the Monte Carlo code SuperMC are used to establish the reactor flux rate calculation model under normal operation and shutdown refueling condition. The results show that (1) In the normal operation of the reactor, the neutron flux rate is attenuated by 10 orders of magnitude from the outermost component to the inner surface of the pressure vessel, and the shielding effect of the coolant on neutrons is more significant. The neutron flux of the inner surface of the pressure vessel in 40 years is 3.723 × 1014 ncm2; the neutron flux in 60 years is 5.585 × 1014 ncm2. The photon flux rate is reduced by 10 orders of magnitude from the periphery of the core to the outer surface of the pressure vessel. High-quality density materials have better photon shielding effects. (2) In the case of reactor shutdown and refueling, the neutron flux rate is much smaller than the photon flux rate. On the outer surface of the pressure vessel, the maximum neutron and photon dose rates are 7.74 × 10−10 mSv · h−1 and 6.97 × 10−5 mSv · h−1, respectively, which belong to the supervision area. When the cover is opened, the radiation dose value of the workplace at the top of the reactor is less than 0.0025 mSv · h−1, which can ensure the radiation safety of the operation.
Linear Active Disturbance Rejection Control of Steam Bypass System for a Pressurized Water Reactor
Authors: Zhang Xianshan , Sun Peiwei , Wei Xinyu , Gou Xiaolong , Tan Guocheng , Tian Yajie 
Published: October 2021

Abstract
China has recently promised to reach carbon neutral before 2060. To achieve this ambitious goal, it is necessary to substantially increase the electricity production from clean sources, such as renewable power and nuclear power, to replace the previous share of coal power generation. Most of the nuclear reactors currently in operation in China are pressurized water reactors (PWR). The nuclear steam supply system (NSSS) in PWR is a complex nonlinear system. The steam bypass system ensures the steam generator’s cooling capacity in accident conditions and is essential for safe and efficient plant operation. Because of the broad transport delay and non-linearity in the steam bypass control system, it is difficult for the currently widely used proportional-integral-derivative (PID) control to obtain satisfactory control effects. Therefore, it is necessary and meaningful to design a suitable steam bypass control system for NSSS.
The investigation of the steam bypass system is performed by MATLAB & Simulink for a large-scale PWR. The nonlinear model is developed by applying the mass and energy conservation equations of fluids and the critical flow correction. The state-space multivariable models are derived by linearizing the nonlinear model at different operation points. Open-loop simulation experiments have been carried out to study the steam bypass system’s dynamic, and the state-space model simulation results are validated by comparing with the results from the nonlinear model. The open-loop results show that the overall system has a considerable transport delay. Thus, a linear active disturbance rejection controller (LADRC) is designed for the steam bypass system. This research assumes that the model and the amount of delay are nearly unknown; however, the proposed controller still performs better than the existing Ziegler-Nichols (Z-N) tuned PID controller. The LADRC has a shorter settling time than the PID controller; the control system can satisfy the design purposes.
Study on the Dynamic Modeling of the Micro-High Temperature Gas Cooled Reactor for Control System Design
Authors: Qiu Leilei , Wei Xinyu , Sun Peiwei , Liao Shengyong 
Published: October 2021

Abstract
The micro-high temperature gas cooled reactor (Mi-HTR) uses Tri-structural iso-tropic (TRISO) particle as fuel, helium gas as coolant, and graphite as moderator and structural material. The reactor adopts complete passive residual heat removal technology and is inherent security. The usage of the Mi-HTR is megawatt-level micro nuclear power plant. Because of the using of hexagonal prism graphite core, the dynamic characteristics of the Mi-HTR are different from large scale high temperature gas cooled reactors and pressurized water reactors. Therefore, the accurate and reliable reactor dynamic characteristics analysis are of significance for the design of the Mi-HTR control system. The neutron kinetics model of the Mi-HTR is built based on the point reactor kinetics with the reactivity feedback mechanism. The core thermal-hydraulic model of the Mi-HTR is built based on the fundamental conservation laws. By coupling the two models, a lumped-parameter model of the Mi-HTR is established. In order to obtain an accurate transfer function model for subsequent control system design, the mechanism model is linearized based on perturbation theory, and then the transfer function model with different inputs and outputs is obtained through Laplace transform.
The developed nonlinear model is implemented on MATLAB/Simulink. By introducing a variety of disturbances, the simulation results of transient responses are presented, which show that the static precision of the dynamic model is satisfactory, and the trend of the transient responses is consistent with the physical laws. The obtained transfer function model and the nonlinear model are introduced with the same disturbances, and the simulation results show that the responses of transfer functions are consistent with those of the nonlinear model. Therefore, the transfer function model can be utilized to control system design for the Mi-HTR.
Preliminary Evaluation on the Relocation Phase of Ex-Vessel Debris of Fukushima Daiichi Nuclear Power Plant Unit-3
Authors: Li Xin , Yamaji Akifumi , Furuya Masahiro , Sato Ikken , Madokoro Hiroshi , Ohishi Yuji 
Published: October 2021

Abstract
To deepen our understanding for the current debris status and investigate the debris-structure interactions in the pedestal region of Fukushima Unit-3, the Moving Particle Semi-implicit (MPS) method is further developed for simulation of multi-component liquid/solid relocation with solid-liquid phase changes aiming for plant-scale practices. The improvement of the existing MPS method mainly consists of two parts, namely 1) the improvement of numerical stability and accuracy, including a) applying second-order corrective matrix to the particle interaction models and b) particle shifting that can optimize particle configurations; and 2) improvement of calculation efficiency, including a) hybrid OpenMP and MPI parallelization and b) particle type-dependent speed-up algorithms to reduce calculation costs for particles with extremely high viscosity and low velocity.
In the current study, the improved MPS method is validated against the experiments carried out at Waseda University, in which molten salt droplets were released to interact with aluminum pillars and solidify on them. Good agreement of the total height of the solidified salt and its distribution on the pillars has been achieved. The successful validation has shown the capability of the current MPS method for simulations of Unit-3 pedestal region.
Configuration Risk Management Support for the Maintenance Rules at Qinshan NPP1
Authors: Wang Li , Wang Zilong , Liu Deyi , Xu Jie , Zhou Jianzhang , Zou Shengjia 
Published: October 2021

Abstract
This article first briefly introduces the “Technical Policy for Improving the Maintenance Effectiveness of Nuclear Power Plants (Trial)” issued by the National Nuclear Safety Administration (NNSA) in August 2017 and the implementation background of the Maintenance Regulations (MR) of the First Plant of Qinshan. Based on the general introduction of the implementation process of MR, the important roles played by the configuration risk management during the implementation of the MR project of Qinshan NPP1 are detailed: 1) The background of configuration risk management and December 30, 2019 “Technical Policy for Nuclear Power Plant Configuration Risk Management (Trial)” issued by the National Nuclear Safety Administration (NNSA); 2) describes the relationship between configuration risk management and maintenance rules and how to apply configuration risk management to implement maintenance rules The scope of activities to carry out configuration risk management and the requirements for the risk threshold of maintenance activities; 3) the method of risk assessment and the establishment of a risk management action matrix on the basis of the EPRI document; 4) the introduction of probabilistic safety assessment (PSA) in the role of configuration risk management; 5) The actual case of configuration risk management is used to show how to configure risk management to solve the requirements of risk assessment in maintenance rules before maintenance activities. As the first operating nuclear power plant in China to formally carry out MR work, the application of Qinshan NPP1 in the implementation of MR configuration risk management can provide a good experience for other domestic nuclear power plants.
Heat Transfer and Fluid Flow Characteristic of U-Shaped Flow Channel for Applications of VHTR
Authors: Takayama Yasuaki , Takeda Tetsuaki 
Published: October 2021

Abstract
A very high-temperature reactor (VHTR) is a next-generation nuclear reactor systems. A gas cooling system with natural circulation is considered as a candidate for the pressure vessel cooling system (VCS) of the VHTR. The Japan Atomic Energy Agency is pursuing the design and development of commercial systems such as the 300 MWe gas turbine high-temperature reactor 300 for cogeneration (GTHTR300C). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV). A cooling panel utilizing the natural convection of air has been proposed. However, the amount of removed heat is inferior to that of cooling by forced convection. In this study, we use an experimental apparatus to simulate the cooling panel of a VCS. The experimental apparatus is a U-shaped flow channel, and the heating surface side is a vertical rectangular flow channel. Air is used as the working fluid. A fine copper wire is used as the porous material. The porosity is varied from 0.996 to 0.999. We perform an experiment to investigate the heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Additionally, we perform an experiment on a smooth surface for comparison.
Probabilistic Safety Assessment on Unavailability of Auxiliary External Power Supply in Fangjiashan Nuclear Power Plant
Authors: Zou Shengjia , Zhao Ming , Liu Deyi , Zhang Naiyuan , Luo Yang , Wang Li , Zhang Jianguo , Chen Honghao 
Published: October 2021

Abstract
The auxiliary external power Qinfang 2P70 of Fangjiashan NPP need to be shutdown, which makes one way the external power unavailable, thus affecting the safety level of the plant. In this paper, Through the internal event level PSA model, internal event LERF model, internal fire model and internal water flooded model, during Fangjiashan at power, 2P70 shut down caused one way the external power unavailable, gives the safety assessment results, And put forward the close attention of the important equipment such as the main power grid and the diesel engine. The test and maintenance of the diesel engine and the SEC/RRI heat exchanger are prohibited. The risk of the unit is managed and controlled by a series of other measures. The risk monitor is used to supervise the unit in real time. Finally, the application risk monitor of the power plant is discussed. The necessity and the PSA method to evaluate the possibility of auxiliary transformer shutdown to generalize to generating units.
SWIFT-RIMPUFF Modeling of Air Dispersion at a Nuclear Powerplant Site With Heterogeneous Upwind Topography
Authors: Dong Xinwen , Fang Sheng , Zhuang Shuhan 
Published: October 2021

Abstract
The SWIFT-RIMPUFF can provide refined atmospheric dispersion modeling for nuclear emergency response, but its performance for the mesoscale range in a nuclear power plant (NPP) site with highly complex topographies hasn’t been fully investigated. In this study, a validation of SWIFT-RIMPUFF was performed based on a wind tunnel experiment simulating a real China’s multi-reactor NPP site with heterogeneous upwind topography and dense buildings to understand the potential discrepancies or limits. The results demonstrate that the SWIFT-RIMPUFF can reproduce the sharp changes of wind flows for both speed and directions near the buildings, but usually overestimate the wind speed in the complex topography. For vertical wind profiles, the accuracies show high dependencies on the local topography and building layout, and the deviation of those near the building is more obvious. The simulated ground concentrations match the topographic changes of high-altitude mountains. The concentration predictions in the downwind building area are acceptable which displays that the influence of building effects can be well introduced, but the simulations in the building area still show noticeable discrepancies when compared with those in the sea area. However, such deviations do not propagate to the downwind mountainous and sea areas, which the accuracies are quite satisfactory.