bluuebunny/crossref_metadata_2025_split huggingface dataset:
Improvement of Transverse Leakage Term Based on Fourier Series Expansion in the 2-D/1-D Method
Authors: Zhu Kaijie , Kong Boran , Hou Jie , Zhang Han , Guo Jiong , Li Fu 
Published: October 2021

Abstract
The 2-D/1-D concept is widely used in various neutron transport codes to obtain sub-pin level flux distribution. The 3D problem is decomposed into 2-D plane and 1-D problems in order to solve the time-consuming problem of direct transport calculation. The method of characteristics (MOC) is usually used in radial 2-D transport calculation in order to capture radial complex geometry. The discrete ordinate (SN) method is selected as the axial 1-D transport sweeper in this work. The two solvers are coupled by transverse leakage terms. Therefore, the accuracy of 2-D/1-D method highly depends on the approximation of the transverse leakage term. For example, in KUCA benchmark test, the isotropic approximation will introduce more than 700 pcm errors to the eigenvalue since the angular correlation of the leakage term is ignored. However, the memory consumption of storing strictly angle related leakage terms is too large. In this work, an alternative method is proposed. The Fourier series expansion is used to approximate the leakage terms, while only a few moments need to be stored. Several benchmark problems have been tested to verify the accuracy of the Fourier-series-based transverse leakage terms. The results show that the improved leakage term can significantly improve the accuracy of eigenvalue with relatively low memory consumption.
Neutronic/Thermal-Hydraulic Coupling Analysis of Xi'an Pulsed Reactor Based on RMC and COBRA-TF
Authors: Li Ruihan , Cao Jianzhu , Wang Lipeng , Zhang Xinyi , Guo Xiaoyu , Liang Jingang 
Published: October 2021

Abstract
This article carries out neutron-hydraulic coupling research of Xi’an Pulsed Reactor (XAPR). The coupling is based on Monte Carlo neutron transport code RMC and sub-channel thermal-hydraulic code COBRA-TF (CTF) because of the high accuracy of Monte Carlo method and the high efficiency of subchannel method. The interface code is developed in python for data transfer between RMC and CTF. The coupling simulation is firstly performed under 1MW, 1.5MW and 2MW steady state. The temperature and power distribution are obtained. The results are compared with the experimental data. Temperature of the fuel rod has a difference in the range of 0.6 °C to 29 °C with the experimental values. Then we perform burnup simulation of XAPR running for 3600 days continuously under the power of 2MW and obtain the neutron and thermal-hydraulic parameters with burnup. Two coupling schemes are applied in the burnup simulation. One scheme is to make neutron-hydraulic coupling using both RMC and CTF in each burn step and the other is only coupling in the first and last burn step while calculation in other steps is only carried out by RMC. The result show great difference between the two schemes. Though currently there are no experimental data for this model to verify, the result obtained using the latter scheme is obviously unreasonable, implying the necessity of making neutron-hydraulic coupling in each burn step.
The Vertical Leaching Migration Research on 137Cs in Soil Around Shidaowan Plant of CAP1400
Authors: Zhang Qiong 
Published: October 2021

Abstract
The risk of radioactive material leakage caused by nuclear accidents increased with the increasing of number of Nuclear Power Plants in China. One of the serious threat of radiation harm to the human body is believed to be 137Cs. It can exist in the environment for a long time, due to the half-life of which is as long as about 30 years. In this research, the brown soil and aeolian sandy soil were selected as two kinds of classic soil around the site of Shidaowan demonstration plant of CAP1400 to carry out the 137Cs migration model experiments. The source of radioactivity was put on the 16 of experiment columns that were filled by undisturbed soil, the simulated raining weather was according to the local rainfall amount with two sprinklers, and the sunlight was replaced by fluorescent lamp. The result showed that the vertical migration model of radionuclides in soil is one-dimensional convective dispersion equation. three-year experiments showed that most of 137Cs in soil were distributed in the range of 0–4cm. The simulation results using the Hydrus 1D with 10 years that most of 137Cs in soil were distributed in the range of 0–10cm, so the nuclear accident caused by the serious nuclides have little impact on groundwater.
Verification of PWR-Core Analysis Code CORAL Using VERA Core Physics Benchmark
Authors: Yang Wen , Chao Fei , Qiu Jinrong , Li Xing , Liu Baolin 
Published: October 2021

Abstract
Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis software and methods, the US CASL project proposed the VERA reactor physics benchmark problem. This benchmark problem is based on the initial loading of Watts Bar Nuclear unit 1 as a model ranging from a simple 2D pin problem cell to the full cycle depletion and refueling of problem a 3D reactor core configuration. Aiming at the VERA benchmark problems, physics calculations are performed using the CORAL code, and the results of effective multiplication factor, assembly power distribution, control rod worth value and reactivity coefficient are obtained. By comparing with the KENO results provided in the benchmark problems, the calculation results of CORAL code are in good agreement. This shows that the CORAL code has the ability to calculate from 2D lattice to the 3D reactor core, and its computational accuracy is basically equivalent to the KENO code.
Evaluation of the Applicability of Plutonium Transmuted From Minor Actinides by Fusion Reactor as Fertile Fuel in Boiling Water Reactor
Authors: Shimizu Masaki , Shishido Hiroki , Hashizume Hidetoshi 
Published: October 2021

Abstract
This study proposes to make effective use of plutonium transmuted from minor actinides (MA) by fusion reactors as fertile fuel in a light water reactor (LWR). The plutonium transmuted from MA, particularly Pu-238, has a large neutron capture cross section and becomes fissile Pu-239 by the reaction. If the plutonium transmuted from MA is loaded into LWR appropriately, there is a possibility to maintain a constant effective multiplication factor. This study evaluated the effects of the plutonium transmuted from MA on a full MOX boiled water reactor (BWR) core by Monte Carlo-based neutron transport and burnup calculation. We revealed that some fuel assemblies achieved the effective multiplication factor is greater than 1.0 and its decrease was significantly small (−0.013) during 400 days operation. However, under this condition, the power peaking factor was 1.4, which should be unacceptable from the viewpoint of thermal design. Inventories of heavy nuclides in fuel cycles was evaluated using a simple diagram and its result indicated that the accumulation of MA in the fuel cycle was reduced by introducing the MA transmutation and the Pu recycle system. Furthermore, the amount of MA production and MA transmutation are balanced by introducing two fusion reactors with 3 GW thermal output into the fuel cycle, and the MA inventory was equilibrated.
Chinese Public and Nuclear Experts May Have Different Perceptions Of Severe Nuclear Accidents
Authors: Wu Hsingtzu , Huang Leyao 
Published: October 2021

Abstract
Nuclear power has been a controversial social issue, and societal acceptance is critical to its development and future. In addition, risk informed rules and regulations rely on the public’s understanding. However, there seems a communication gap about nuclear safety between nuclear experts and the public in China, and three questionnaire surveys were conducted to better understand Chinese public’s perceptions of a severe nuclear accident. The sample sizes were 117, 280 and 1071. Most of the respondents were students or white-collar workers born after 1990. In these three surveys, we found that more than 85% of respondents consider a less severe accident as a severe nuclear accident, and most respondents considered an incident to constitute a severe nuclear accident. The results demonstrate that nuclear experts and Chinese public may have different definitions of a severe nuclear accident. Therefore, we suggest that the definition of severe accidents should be better explained to the public to benefit the communication about risk informed rules and regulations. In addition, our three different surveys yielded a similar result, and we anticipate that a questionnaire survey with a larger sample size would do the same.
Conceptual Study of Neutron Physics of Nuclear Fuel Cycle for Ceramic Fast Reactor
Authors: Yan Xuesong , Zhang Yaling , Gao Yucui , Yang Lei 
Published: October 2021

Abstract
To make the nuclear fuel cycle more economical and convenient, as well as prevent nuclear proliferation, the conceptual study of a simple high-temperature dry reprocessing of spent nuclear fuel (SNF) for a ceramic fast reactor is proposed in this paper. This simple high-temperature dry (HT-dry) reprocessing includes the Atomics International Reduction Oxidation (AIROX) process and purification method for rare-earth elements. After removing the part of fission products from SNF by a HT-dry reprocessing without fine separation, the remaining nuclides and some uranium are fabricated into fresh fuel which can be used back to the ceramic fast reactor. Based on the ceramic coolant fast reactor, we studied neutron physics of nuclear fuel cycle which consists operation of ceramic reactor, removing part of fission products from SNF and preparation of fresh fuels for many time. The parameters of the study include effective multiplication factor (Keff), beam density, and nuclide mass for different ways to remove the fission products from SNF. With the increase in burnup time, the trend of increasing 239Pu gradually slows down, and the trend of 235U gradually decreases and become balanced. For multiple removal of part of fission products in the nuclear fuel cycle, the higher the removal, the larger the initial Keff.
Machine Learned Metamodeling of a Computationally Intensive Accident Simulation Code
Authors: Conner Landon , Worrell Clarence L. , Spring James P. , Liao Jun 
Published: October 2021

Abstract
The nuclear power industry is increasingly identifying applications of machine learning to reduce design, engineering, manufacturing, and operational costs. In some cases, applications have been deployed and are realizing value, in particular in the higher volume and data rich manufacturing areas of the nuclear industry. In this paper, we use machine learning to develop metamodel approximations of a computationally intense safety analysis code used to simulate a postulated loss-of-coolant accident (LOCA). The benefit of an accurate metamodel is that it runs at a fraction of the computational cost (milliseconds) compared to the LOCA analysis code. Metamodels can therefore support applications requiring a high volume of runs, for example optimization, uncertainty analysis, and probabilistic decision analysis, which would otherwise not be possible using the computationally intense code. We first generate training data by running the safety analysis code over a design of experiment. We then perform exploratory data analysis and an initial fitting of several model forms, including neighbor-based models, tree-based models, support vector machines, and artificial neural networks. We select neural network as the most promising candidate and perform hyperparameter optimization using a genetic algorithm. We discuss the resulting model, its potential applications, and areas for further research.
Evaluation of Normalization Method With Crack Mouth Opening Displacement for Developing J-R Curves
Authors: Zhu Xian-Kui 
Published: October 2021

Abstract
Fracture toughness is an important material property used to characterize material resistance against crack tearing. In ASTM E1820 standard fracture toughness test method developed for metallic materials, the elastic unloading compliance method is recommended for measuring incremental crack growth and evaluating fracture resistance curve from a single specimen test in the conventional test conditions. For dynamic loading or other harsh conditions, on the other hand, ASTM E1820 suggests the normalization method as an alternative approach for estimating incremental crack length during stable crack tearing and for evaluating the J-integral resistance (J-R) curve from a single specimen test.
The normalization method is a semi-analytical methodology that was developed in the 1990s based on the load separation approach with use of load-line displacement (LLD) for a single edge notched bend (SENB) or compact tension (CT) specimen. This method has been extensively utilized to develop J-R curves using SENB and CT specimens for different steels in various loading and environmental conditions. Recently, the LLD-based normalization method was simply extended to use crack mouth opening displacement (CMOD) in determination of J-R curves for SENB and single edge notched tension (SENT) specimens. Based on the historical work for the LLD-based normalization method, this paper develops a fundamental basis for use of the CMOD-based normalization method. Experimental data measured from a single SENB specimen test are then used to evaluate the normalization method in determination of J-R curves using both LLD and CMOD. Finally, how to use the LLD-based or CMOD-based normalization method is suggested for developing a J-R curve.
Comparison of Vehicle-Mounted High-Pressure Hydrogen Storage Cylinders in United Nations Standards and Chinese Standards
Authors: Ding Ju , Yuan Yiwen , Yu Yixi 
Published: October 2021

Abstract
The hydrogen fuel cell vehicle industry has formed two major international standardization organizations, ISO, and International Electrotechnical Commission. ISO has formed a relatively mature evaluation system for electric vehicle performance indicators and test methods. However, in China, there are still deficiencies in the safety requirements and test evaluation methods in the field of hydrogen fuel vehicle components. The vehicle-mounted high-pressure hydrogen storage cylinder is a special equipment in Chinese current special equipment supervision and regulation system An important energy storage component of hydrogen fuel cells, it has a development history of more than 50 years in the world. However, Chinese national standard committee has not yet established a standard for type IV hydrogen storage cylinders, which had been used in Japan, U.S, and E.U. In this paper, the United Nations’ automotive high-pressure hydrogen storage cylinder standard ISO 19881:2018 and my Chinese standard GB/T 35544-2017, had been compared and analyzed, especially the specific differences in the manufacturing and inspection. China has been preparing to manufacture the type IV hydrogen cylinder standard nowadays. So it is very important to learn from foreign advanced experience, The purpose of this article is provide relevant instructs for manufacture, inspection and test, to formulate China’s vehicle-mounted hydrogen storage cylinders, and improves the conversion and application of standards.
Inspection Method of Finned Tube and Finned Heat Exchanger
Authors: Ding Ju , Zhang Min , Liu Shuhong , Wang Shenghui , Wang Jielu 
Published: October 2021

Abstract
The finned tube heat exchanger is one of the earliest and most successful discoveries in the process of improving tube heat exchange. This method is still the most widely used of all kinds of tube heat transfer surface enhancement heat transfer methods. It is not only suitable for single-fin tube heat exchangers, which are widely used in power, chemical, petrochemical, air-conditioning engineering and refrigeration engineering.
Conventional heat exchanger with smooth tubes can be inspected through the pressure test during the manufacturing process. Finned tubes and finned heat exchangers with inner thread structure have some difficult to pass the water pressure test. The same situation exists in regular inspections. Due to structural reasons, it is difficult to carry out regular surface inspections[1]. For these two situations, two different testing methods are required to ensure quality. This article introduces in detail the methods of inspecting finned tubes and finned heat exchangers. Hierarchical comparison of alternatives in hydrostatic testing project, and the eddy current detection technology of the finned tube under the condition of in-service air cooling. The far-field eddy current method is chosen for inspection. And by comparing the standard sample tube, it is mainly used to adjust the sensitivity of the eddy current detector and ensure the accuracy of the test results[2]. The results show that the eddy current detection technology can be more accurate and reliable. The corrosion of the finned tube under service air cooling is detected, and a reliable basis is provided for judging the use of the finned tube and finned heat exchanger[3].
Comparison of Tensile Test Results in High Pressure Gaseous Hydrogen Using Conventional and Tubular Specimens
Authors: Michler Thorsten , Wackermann Ken , Ebling Fabien , Oesterlin Heiner 
Published: October 2021

Abstract
Hydrogen environment embrittlement (HEE) of X60M pipeline steel was assessed by tensile testing with tubular specimen containing high pressure hydrogen in the axial hole. Residual oxygen content as well as strain rate were varied in order to define appropriate test parameters. Maximum HEE effects were found for oxygen contents below 2 vppm and strain rates below 5*10−4 s−1, which is in the same range compared to tensile testing in an autoclave using conventional specimens. Results at such tests parameters revealed that the tensile reduction of area in hydrogen is about 10% higher compared to a similar steel tested in an autoclave at identical test conditions using conventional specimens. Further investigations are required to verify, if this difference is caused by the specimen geometry.
Study on Ultrasonic Phased Array Detection and Imaging Techniques for Oil Storage Tank Floor Covered With Sludge
Authors: Cheng Jingwei , Chen Xuedong , Fan Zhichao , Chen Wei , Wang Zhe , Cheng Wei , Xu Wanbao 
Published: October 2021

Abstract
Ultrasonic testing is one of the most common non-destructive testing (NDT) methods for periodic inspection of crude oil storage tanks. Ultrasonic phased array system can effectively and efficiently image different types of defects in the structures through direct contact method. However, the sludge remained on the tank floor gives rise to difficulties on detecting the defects in the floor plate using existing techniques. Therefore, this paper focuses on the development of novel phased array testing methods without requiring complete removal of the sludge. The appropriate experimental apparatus and its parameters used for through-sludge testing are selected based on the acoustic characteristics of the oil sludge composition. Furthermore, a novel phased array full matrix capture method is explored based on firings of multi-element aperture. The total focusing method (TFM) is studied by considering the existence of sludge. The experimental results from the sludge thickness study demonstrate that the TFM imaging technique can more accurately measure the change in thickness of the metal sample at the submillimeter scale. Additionally, high resolution imaging of side drilled holes in size of 1 millimetre can be realized on the samples covered with sludge. Last, the effect of aperture size on the detectability is investigated. Consequently, the results obtained from different sludge experiments show good reliability and measurement consistency of our proposed methods. It delivers great potential to accurately locate and quantitatively evaluate the corrosion defects frequently found in the tank floor.
A Convolution Neural Network Based Fault Diagnosis Model for Slug Flow Condition of Oil-Gas Separator
Authors: Liu Jiaquan , Hou Lei , Wang Xin , Xu Lei , Zhang Rui , Yang Kai 
Published: October 2021

Abstract
The process of oil-gas separation is an important part of oilfield surface engineering. The oil-gas separator is the key equipment in this process. The composition of produced fluid from oil well is complex and the flow state of it changes frequently, which leads to the abnormal operating condition of the separator and affects the continuity of oil production. Therefore, the operating condition diagnosis of the oil-gas separator is extremely important for the system of oil production. Compared with the traditional threshold-based alarm method, the data-driven diagnosis method is more accurate and intelligent. According to the multivariate time series properties of the data of supervisory control and data acquisition (SCADA) system, the convolution neural network (CNN) is proposed to extract the multi-dimensional features of the oil-gas separator. Then based on the extracted features, the slug flow conditions of oil-gas separator are diagnosed, whose process is visually analyzed by t-distributed stochastic neighbour embedding (t-SNE). Finally, the data-driven diagnosis method of the oil-gas separator is realized, and the simulation data show that the average fault diagnosis rate of this method reaches 97.78%, which is 26.95% higher than that of the traditional method. The method proposed can diagnose the operating conditions of the separator accurately and quickly.
A Simple Method to Estimate Airborne Release Fractions Associated With the Pressurized Release of Radioactive Materials
Authors: Hensel Steve J. , Sprankle Ray A. 
Published: October 2021

Abstract
Pressurized release of radioactive materials, such as plutonium bearing oxides, could occur during a fire accident event where a robust container holding the radioactive materials ruptures catastrophically due to excessive pressurization. The common source of pressurization is the vaporization of water adsorbed onto the plutonium bearing oxides. A simplified pressure based approach derived from NUREG/CR-6410 is presented which allows calculation of airborne release fractions for pressures up to and above 500 psig. A parametric study is used to determine a bias (conservative or non-conservative) when the approach is applied to DOE-STD-3013 welded stainless steel containers which are used to store plutonium bearing oxides at the US Department of Energy Savannah River Site. The event considered is a 3013 container subjected to a hypothesized facility fire where the container fails catastrophically and ruptures resulting in a vented high pressure release of the plutonium bearing oxide materials. The parametric study shows that the simple correlation for calculating airborne release fractions is biased high by about 40% compared with the detailed NUREG/CR-6410 method. Care should be taken in using the simplified approach where the ratio of vessel net free volume to maximum allowed oxide mass is greater than 0.66 cc/g.