bluuebunny/crossref_metadata_2025_split huggingface dataset:
Retracted: “Power Control System Design for Heat Pipe Cooled Reactor Based on Model Predictive Control” [2022 29th International Conference on Nuclear Engineering, Volume 15] Shenzhen, China and Virtual, Online, August 8–12, 2022, Conference Sponsors: Pressure Vessels and Piping Division, ISBN: 978-0-7918-8650-2, Copyright © 2022 by ASME. Paper No. ICONE29-89495, V015T16A014; 8 pages; doi: https://doi.org/10.1115/ICONE29-89495
Authors: ASME  
Published: November 2022

Abstract
This paper was removed at the author’s request on December 1, 2022.
Copyright © 2022 by ASME
Preliminary Design and Neutronics Analysis for a Medium Temperature Heat Pipe Cooled Reactor
Authors: Li Sina , Yan Binhuo , Tang Guojun , Peng Donghong , Lu Hongzhi 
Published: November 2022

Abstract
Traditional heat pipe cooled reactors have the characteristics of high core temperature and high fuel enrichment, which makes its development limited by several unfavorable factors, such as the limitation of high temperature materials and low burnup, etc. In this paper, a new heat pipe cooled reactor with a rated thermal power of 1.5MWt is proposed to address the problems of high operating temperature and high fuel enrichment. Low-enriched uranium is used as fuel. Mercury heat pipe is used instead of the traditional high temperature heat pipe in order to maintain a low operating temperature of around 350 °C. A superheated steam/water cycle with an efficiency of 33% can ensure the high thermal efficiency of the reactor. The Monte Carlo software MCNP5 is used to analyze the core physics characteristics, including the control rod worth, neutron energy spectrum, neutron distribution, temperature coefficient and core lifetime. A radial power peaking factor of 1.46 with controls inserted can satisfy the heat transfer requirements. A high control rod worth and high transient negative temperature coefficient can also ensure a high inherent safety of this reactor. The core lifetime is around 8 years.
Dependence Analysis of Nuclear Power Plant Operators Based on HRA
Authors: Xu Jie , Zhou Jianzhang , Liu Deyi , Wu Jian , Wang Zilong , Zhao Ming , Luo Yang , Chen Honghao 
Published: November 2022

Abstract
As an important part of probabilistic safety assessment (PSA), human reliability analysis (HRA) provides human error probabilities for assessing the risk of nuclear power plants (NPPs). Human reliability analysis (HRA) uses a systematic method to qualitatively and quantitatively assess the mechanism and probability of human error in order to assess the human contribution to the risks of NPPs. Dependence anslysis is an important part of HRA. Without dependency analysis, the result of human error probabilities may be over optimistic, which will lead to the underestimation of the safety risk level of the NPPs. Based on the feedback of human factor dependence from the experts review of level 1 PSA in NPPs. This paper summarized the technical insights of HRA dependence analysis, and used two dependence analysis methods to re-analyze the human factor dependence in the PSA model of nuclear power plant.
In this paper, starting from the problem of human dependence raised by experts in the peer assessment of first-level PSA in power operation, some technical insights on human dependence were summarized in the analysis and discussion of the existing HRA dependence method. Meanwhile, the human dependence of power PSA was analyzed again by SPAR-H method and NUREG-1921 dependence method.
Analysis of Thermal Stratification Effects in the Ring Headers of the Passive Residual Heat Removal System of NHR-200-II
Authors: Geng Yiwa , Liu Xiongbin 
Published: November 2022

Abstract
NHR-200-II is a small integral pressurized water reactor with a core thermal power of 200MW. Two parallel columns of passive residual heat removal (PRHR) system are connected to the two intermediate circuits of NHR-200-II to remove core decay heat in postulated accidents. It was confirmed by scaled experiments that the decay heat removal capacity of the PRHR system was severely deteriorated once the isolation valves for steam generator on the intermediate circuits failed to close. Thermal stratification effects in the ring headers of PRHR system played a significant role in the deterioration of PRHR capacity. A RELAP5 system model and a CFD model were developed for the PRHR system of NHR-200-II, in both models, we focused on the prediction of thermal stratification phenomena in the ring headers of PRHR. Both the case of normal PRHR scenarios (i.e. the SG branches were successfully isolated) and the case of isolation failure of the SG branches during PRHR startup were numerically simulated with the developed CFD and RELAP5 models. The numerical results from both CFD and RELAP5 models showed that the thermal stratification effects in normal PRHR scenarios can be neglected, however, both models gave similar conclusions that the thermal stratification effect was strong in the ring headers in case of isolation failure of the SG branches. Possible physical mechanism of thermal stratification in PRHR of NHR-200-II was explained depending on results of numerical simulations.
Research of Supercritical Carbon Dioxide Thermal Power System Based on Lead-Cooled Fast Reactor
Authors: Li Zhihui , Liu Chunyuan , Xing Mian , Zheng Gang , Wang Yu 
Published: November 2022

Abstract
The supercritical carbon dioxide (S-CO2) Brayton cycle with re-compression has the advantages of compact layout, simple structure, thermal high efficiency, clean working quality, its application in lead-cooled fast reactor power conversion system helps the miniaturization and modularization of the whole system. In addition, it is considered to be an effective solution to the problem of the recuperator point for the lead-cooled fast reactor power system. The Aspen plus software was applied in this paper to establish the mathematical model for supercritical carbon dioxide Brayton cycle with re-compression and the 1MWe lead-cooled fast reactor power system was calculated. The effects of several key factors such as the turbine inlet temperature, the turbine inlet pressure, the efficiency of the turbine, compressor and recuperator were analyzed. The results show that the turbine inlet pressure, the turbine inlet temperature and the turbine efficiency have remarkable effects on thermal efficiency of the system. Thermal efficiency of the system increases with the inlet pressure, inlet temperature and efficiency of the turbine increasing. The research results of this paper could provide important theoretical reference both for thermal cycle parameters for 1MWe lead-cooled fast reactor and system design of power cycle based on the lead-cooled fast reactor.
Effect of Surface Machining on Stress Corrosion Cracking Sensitivity of 304L and 316L Austenite Stainless Steel
Authors: Wu Huanchun , Meng Xinming , Han Yaolei , Zhang Yanwei , Peng Qunjia 
Published: November 2022

Abstract
304L/316L austenitic stainless steels (SSs) often used for main pipe of PWR nuclear power plant. Surface machining has a significant effect on microstructure and residual stress of SS, which further affects its stress corrosion cracking (SCC) sensitivity. The present study results show that, the SCC sensitivity of machined specimens increases significantly tested in 300 °C high temperature and high pressure water environment by the slow strain rate tensile test (SSRT). It increased from less than 10% of polished specimens to 20–30% of machined specimen for SCC sensitivity. It was found that the machined specimen showed as cleavage cracking, while the polished specimen was shown as ductile cracking with obviously fracture necking phenomenon. The mechanism of SCC of two kinds of SSs materials focused on the deformation layer of machined specimen was analysed, and the model of SCC initiation caused by nanocrystalline layer was established.
Optimization of Core Parameters Based on Artificial Neural Network Surrogate Model
Authors: Chen Shu , Ding Peng , Hu Shuowen , Xia Wenqing , Liu Min , Yu Fengwan , Li Wenhuai 
Published: November 2022

Abstract
The methodology of artificial intelligence (AI), particularly artificial neural network (ANN), would be in favor of nuclear energy system development. These ANN simulators may provide more efficient means than the traditional nuclear design codes, especially for the design of the key parameters, such as core geometry and layout, material composition.
In this paper, a neutronics calculation code SARAX and the corresponding multilayer perceptron (MLP) surrogate model were used as simulators for core parameters optimization of a reference lead based fast reactor. The pellet radius, enrichments and active height are the interested core parameters, and core burnup and power distribution are the target characteristics in the study. The training of structure and weight parameters in MLP network are based on about 5000 calculations of SARAX code. Test results of neural network show a good agreement between MLP surrogate model and SARAX code. The feasibility of the MLP surrogate model to be used in core parameters optimization was also discussed. Results showed that, the core surrogate model based on MLP could be quickly constructed and regulated, and be a more efficient simulators in a innovate reactor optimization.
The above work is completed in Sinan Platform, a multidisciplinary intelligent design platform developed by China Nuclear Power Technology Research Institute Co. Ltd.
Experimental Study on Aerosol Penetration Through Concrete Cracks Under Inertial Collision and Steam Condensation
Authors: Wang Shanpu , Li Yuxiang , He Liwen , Tong Lili , Cao Xuewu 
Published: November 2022

Abstract
Estimation of aerosol penetration through the cracked concrete wall with a submillimeter height has attracted more attention for radioactivity optimization evaluation of containment, which emphasizes the aerosol deposition inside the cracks without catastrophic destruction of containment. While, no deposition has been considered in current radiological assessments and the penetration of aerosol is simply assumed to be released as the leakage of gas. Therefore, the results of assessment are too conservation for the leakage of radioactive aerosol without considering any deposition.
In this work, an experimental facility is set up to study the penetration of particles through cracked concrete walls with different geometric dimensions. Titanium dioxide powder is chosen due to its density and nontoxicity. A sampling system is designed for the measurement of aerosol concentration under high temperature and containing condensable steam. One regular concrete crack in the shape of a straight rectangular slot and one irregular concrete crack with different streamwise tortuosities, τ, which is defined as the radio of the actual length of a crack pathway and the thickness of concrete structure, have been used in the tests of gas leakage and aerosol exposure. The height of cracks can be calculated from the consequences of gas leakage characteristic, based on the Suzuki’s flow model. Research results of aerosol exposure tests indicate that increasing the flow rate, particle diameter and tortuosity all can enhance the aerosol penetration, and micron particles cannot penetrate concrete cracks, compared with submicron particles. The aerosol penetration through the straight concrete crack is in good agreement with the prediction by the correlation proposed by Van De Vate, but no agreement is reached for the irregular crack with some tortuosity. Meantime, the effect of steam condensation on aerosol penetration is also been discussed preliminarily.
Study on Probability Safety Assessment Method of Small Aircraft Crash Event in a Nuclear Power Plant
Authors: Wang Handing , Liang Xiaoyu 
Published: November 2022

Abstract
In nuclear power plants, one of the external human-caused events is aircraft crash. The method used to evaluate the risk of aircraft crash must comply with the requirements of national nuclear safety guidelines. When estimating the distance of aircraft from facility, the events for which the aircraft distance satisfying the screening guidelines may be left out of account firstly. While for those unmet, it is necessary to apply probability safety assessment to further study whether the risk of aircraft crash into nuclear power plants is within the probabilistic screening level. If the aircraft impact can not be eliminated through distance or probability screening, the next step is to identify the corresponding aircraft type, and determine the design basis of aircraft crash event as an input. For the purpose, this paper analyzes the impacts of small military airports near nuclear power plants, and researches into the probability safety assessment method for military aircraft impacting nuclear power plants, i.e., airport and non-airport operation risks. According to the study, the risk of airport operation can be ignored as the conditional probability f (x, y) representing the military aircraft falling to the site shows very low. Conservatively considering the military aircraft impact “four-factor formula parameters”, the aircraft crash impact frequency of non-airport operation is 6.23E−08/year, also the total impact frequency of military aircraft crash event, which is less than the screening level in terms of probability and meets the requirements of nuclear safety guidelines. This study aims at providing certain reference value for the establishment of design basis to protect nuclear power plants from aircraft crash.
Fuel Route Risk Assessment Framework and Pilot Plant Application
Authors: Wang Ming , Yang Chunju , Xiao Linmei , Hou Wenyu , Xu Zhixin , Zhang Bing , Wang Jinkai , Cui Lei 
Published: November 2022

Abstract
Damage to fuel assemblies and the release of radioactive substance may occur during fuel lifting and storage in nuclear power plant. For the purpose of quantitatively evaluate internal events, internal and external hazards induced risk of spent fuel pool and related fuel lifting facilities, to identify weaknesses of the fuel route design and provide corresponding risk insights, the methodology of fuel route probabilistic safety assessment (PSA) is systematically summarized in this study, and based on the design characteristics of a specific pressurized water reactor nuclear power plant, fuel damage frequencies and radioactive substance release frequencies in the fuel route are assessed, some possible further safety improvements for the fuel route are identified. The results of analysis case show that the fuel building ventilation system failure has the largest contribution to the overall fuel thermal damage frequency, while the spent fuel crane failure has the largest contribution to the overall fuel mechanical damage frequency. The drop of the spent fuel cask full of spent fuels in loading pit has a prominent contribution to the risk of radioactive substances release due to the mechanical damage. The safety risk in the fuel route can be effectively reduced by the diversified design of local cooling unit fans in fuel building ventilation system and type modification of the spent fuel pool crane.
Comparison of the Powderization Effect of Non-Equilibrium Plasma Oxidation and Thermochemical Oxidation Powders of Uranium Dioxide Solids for Actinide Analysis
Authors: Ma Zhuoran , Tatsuno Takaharu , Homma Yoshiya , Konashi Kenji , Suzuki Katsuya , Suzuki Tatsuya 
Published: November 2022

Abstract
In order to facilitate the dissolution of these insoluble nuclear debris from the Fukushima accident, it is necessary to investigate methods of pulverizing them to increase their surface area. Although solid uranium dioxide is known to become powder through volume changes caused by oxidation, thermal oxidation can lead to the volatilization of quasi-volatile radioactive materials, so it is desirable to cause reaction under the milder conditions. We therefore developed non-equilibrium plasma oxidation device to verify the powderization of uranium dioxide solids and to compare the results with thermochemical oxidation. For the results of the plasma oxidation experiment, Uranium dioxide solid (42mg) can be partly converted to powder by plasma oxidation (6.5h, O2:0.4 L/min) with a low temperature (less than 200 °C). And Partial oxidation of the powder, uranium dioxide to triuranium oxtoxide was confirmed by XRD. Small uneven bumps of 1μm or less appears on the surface of powders produced by oxidation using non-equilibrium plasma, thus increasing the surface area required for dissolution or chemical conversion to soluble compounds can be expected.
Deep-Learning Based Three-Dimensional Analysis of Bubble Flows From Light Field Images
Authors: Zhang Heng , Lyu Xue , Hang Qin , Du Yulu , Liu Bo , Liu Qun , Wang Guoyin , Liu Jinchao 
Published: November 2022

Abstract
An accurate measurement for the modality and motion parameters of bubbles is of great significance. In the existing bubble measurement methods, there are some problems that are desirable to be solved, such as system complexity, poor real-time performance, and the deviation due to two-dimensional measurement. To overcome these challenges, we developed a novel three-dimensional analysis method based on light field imaging diagnosis and deep learning algorithm. Different from traditional two-dimensional reconstruction, the bubble depth can be computed from light field images directly through digital refocusing technology. After calibrating, the conversion between the camera coordinate system and the real-world coordinate system is achieved through the sharpness evaluation algorithm. According to the corresponding relationship, the refocused image could be calibrated to the actual position in real world. Combined with the Multi-input Residual Convolution Neural Network, (MRCNN), the bubble depth could be computed fully automated from given images at high accuracy. Based on the above works, the three-dimensional reconstruction model for bubble flow can be established by coupling the depth and the parameters extracted through bounding boxes. The proposed method solves the problem that conventional imaging can only perform the two-dimensional measurement, which contributes to the error during the measurement process. Results show a promising performance on the three-dimensional reconstruction of bubble flow, validating the feasibility of the three-dimensional measurement method for bubbles in gas-liquid two-phase flow based on light field imaging diagnosis and deep learning algorithm.
Principal Component Analysis on Tangential Oscillation of Thermal Stratification in Pipe-Flow
Authors: Wei Dongwei , Zhou Mi , Zhang Naiqiang 
Published: November 2022

Abstract
Tangential oscillation of thermal stratification is a new finding in previous investigations on thermal-mixing pipe-flow at mixing tee. Due to this phenomenon, regular temperature changes can be created on the pipeline inner surface and lead to material damage caused by the degradation mechanism of thermal fatigue, which has been frequently reported as a reason for incidents on piping systems of nuclear power plants. Currently, numerical simulations have been performed to reconstruct this phenomenon by using LES (Large-Eddy Simulation) methods. The tangential oscillation can be recognized from the animation of spatial temperature distributions but is still difficult to present in writing form.
In this work, principal component analysis (PCA) has been applied to identify and analyze the tangential oscillation of thermal stratification. Time-dependent temperature data in the vertical cross-sections of stratified pipe-flow has been extracted from the previous LES calculation. The characteristics of the spatial temperature distributions has been estimated with the PCA method. The results show the angular shift of the temperature distribution and its periodical change with the physical time in the simulation. It proves that the PCA method is effective in presenting the tangential oscillation of thermal stratification and can be applied in the further investigation of this phenomenon.
Three-Dimensional Reconstruction of Single Bubble Shape and Trajectory Using Digital Image Processing and 3D Modeling Method
Authors: Zhang Heng , Liu Bo , Hang Qin , Lyu Xue , Liu Jinchao 
Published: November 2022

Abstract
Plate type fuel element is widely used in integral and test reactor cores with the advantages of higher power density, deeper fuel burnup and more compact structure. In the semi-closed narrow gap rectangular channel, physics-structure coupling effect under low pressure condition will lead to very complex coolant flow and heat transfer phenomena. Knowledges of the bubble movement and deformation regime are important for studying the mechanism of gas-liquid two-phase flow. In this work, a visual experimental investigation was carried out for the bubble rising characteristics and bubble three-dimensional (3D) reconstruction in a rectangular channel. The process of the bubbles motion was observed by two vertically placed high-speed cameras. By controlling the power of the oxygen pump and the aperture of the air outlet respectively, the rate of bubble formation can be controlled, which eliminated the effects of bubble-to-bubble interactions by increasing the distance between the bubbles. Based on the extracted parameters obtained by digital image processing and assumption that each cross section along the height of bubble to be an ellipse, a new method for reconstructing bubble shape was proposed, using 3D modeling method which can more closely simulates objects to connect bubble elliptic cross sections smoothly under the condition that the cross sections of bubble were not changed as much as possible. Then the bubble shape and trajectory were reconstructed. The results showed that the 3D trajectory of bubble was oscillating and the reconstructed bubble shape not only conformed to the projected contour but also satisfied the effect of surface tension which lead to smooth shape of the bubbles in any time.
Novel Syntheses Method of Grafted Chelating Resin and Its Application Study for Depleted Zinc Production
Authors: Liu Feng , He Ye , Li Jiawen , Zhan Jie , Su Xingdong , Li Xuefei , Liu Xiajie 
Published: November 2022

Abstract
Depleted zinc injection into the reactor coolant system (RCS) of pressurized water reactors (PWRs) has demonstrated benefits for significantly reducing radiation fields. Liquid chromatographic separation of zinc isotopes to produce depleted zinc was attempted. A novel chelating resin PS-g-GMA@IDA containing iminodiacetic acid (IDA resins) for separation of Zn2+ and zinc isotope was synthesized by γ-ray induced co-irradiation grafting of glycidyl methacrylate (GMA) onto polystyrene (PS) spherical resin, followed by ring-opening processes with iminodiacetic acid. The obtained chelating resin PS-g-GMA@IDA before and after adsorption of Zn2+ were characterized by FT-IR spectra. This showed that the adsorption of Zn2+ by chelating resin involved two processes. Firstly, zinc ions displaced the hydrogen of the carboxyl group on chelating resin, then Zn2+ chelated with oxygen atom to form a ring structure through the coordination bonds. Adsorption abilities of PS-g-GMA@IDA were evaluated by batch adsorption experiments. The effect of pH and initial concentration on adsorption capacity of PS-g-GMA@IDA were investigated. The highest adsorption capacity 130.1 mg/g for (CH3COO)2Zn was obtained at pH = 6 within 24h when C0 were 200 ppm. EDTA-(NH4)2 was selected as eluent because of its high desorption rate of zinc. A dynamic column adsorption test was carried out at 25 °C using a 1 m column packed with PS-g-GMA@IDA, and the breakthrough occurred at about 500 ml efflunent volume. Column capacity calculated from the breakthrough curve was 56.83 mg/g. Na type PS-g-GMA@IDA was utilized to carry out column chromatographic experiment for zinc isotope separation. Under the conditions of 278±1 K, migration distance 1+9 m and EDTA-(NH4)2 as eluent, the the maximum enrichment value of 66Zn/64Zn and 68Zn/64Zn in the effluent is 1.0268 and 1.0451 respectively.