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Environmental fatigue correction factor model for domestic nuclear-grade low-alloy steel Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-22 Jun Gao; Chang Liu; Jibo Tan; Ziyu Zhang; Xinqiang Wu; En-Hou Han; Rui Shen; Bingxi Wang; Wei Ke
Low cycle fatigue behaviors of SA508-3 low-alloy steel were investigated in room-temperature air, high-temperature air and in light water reactor (LWR) water environments. The fatigue mean curve and design curve for the low-alloy steel are developed based on the fatigue data in room-temperature and high-temperature air. The environmental fatigue model for low-alloy steel is developed by the environmental
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Hydrogen isotope exchange behavior of protonated lithium metal compounds Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-22 Chan Woo Park; Sung-Wook Kim; Youngho Sihn; Hee-Man Yang; Ilgook Kim; Kwang Se Lee; Changhyun Roh; In-Ho Yoon
The exchange behaviors of hydrogen isotopes between protonated lithium metal compounds and deuterated water or tritiated water were investigated. The various protonated lithium metal compounds were prepared by acid treatment of lithium metal compounds with different crystal structures and metal compositions. The protonated lithium metal compounds could more effectively reduce the deuterium concentration
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Efficiency of various structural modelling schemes on evaluating seismic performance and fragility of APR1400 containment building Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-19 Duy-Duan Nguyen; Bidhek Thusa; Hyosang Park; Md.Samdani Azad; Tae-Hyung Lee
The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer
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An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-18 Ouadie KABACH; Abdelouahed CHETAINE; Abdelfettah BENCHRIF; Hamid AMSIL
Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured
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Bragg-Curve Simulation of Carbon-Ion Beams for Particle-therapy Applications: a study with the GEANT4 toolkit Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-18 M. Kh. Hamad
We used the GEANT4 Monte Carlo MC Toolkit to simulate carbon ion beams incident on water, tissue, and bone, taking into account nuclear fragmentation reactions. Upon increasing the energy of the primary beam, the position of the Bragg-Peak transfers to a location deeper inside the phantom. For different materials, the peak is located at a shallower depth along the beam direction and becomes sharper
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Earthquake Response of a Core Shroud for APR1400 Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-17 Myung Jo Jhung; Youngin Choi; Chang-Sik Oh
The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using
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Testing the pollution haven hypothesis on the pathway of sustainable development: Accounting the role of nuclear energy consumption Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-15 Danish; Salah Ud-Din Khan; Ashfaq Ahmad
The environmental effects of China’s nuclear energy consumption in a dynamic framework of the pollution haven hypothesis are examined. This study uses a dynamic autoregressive distributed lag simulation approach. Empirical evidence confirms that the pollution haven hypothesis does not exist for China; i.e., foreign direct investment plays a promising role in influencing environmental outcomes. Furthermore
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Mitigation of Seismic Responses of Actual Nuclear Piping by a Newly Developed Tuned Mass Damper Device Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-15 Shinyoung Kwag; Seunghyun Eem; Jinsung Kwak; Hwanho Lee; Jinho Oh; Gyeong-Hoi Koo
The purpose of this study is to reduce seismic responses of an actual nuclear piping system using a tuned mass damper (TMD) device. A numerical piping model was developed and validated based on shaking table test results with actual nuclear piping. A TMD for nuclear piping was newly devised in this work. A TMD shape design suitable for nuclear piping systems was conducted, and its operating performance
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Position error compensation of the multi-purpose overload robot in nuclear power plants Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-15 Guodong Qin; Aihong Ji; Yong Cheng; Wenlong Zhao; Hongtao Pan; Shanshuang Shi; Yuntao Song
The Multi-Purpose Overload Robot (CMOR) is a key subsystem of China Fusion Engineering Test Reactor (CFETR) remote handling system. Due to the long cantilever and large loads of the CMOR, it has a large rigid-flexible coupling deformation that results in a poor position accuracy of the end-effector. In this study, based on the Levenberg-Marquardt algorithm, the spatial grid, and the linearized variable
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UNCERTAINTY QUANTIFICATION AND PROPAGATION WITH PROBABILITY BOXES Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-14 L. Duran-Vinuesa; D. Cuervo
in the last decade, the best estimate plus uncertainty methodologies in nuclear technology and nuclear power plant design have become a trending topic in the nuclear field. Since BEPU was allowed for licensing purposes by the most important regulator bodies, different uncertainty assessment methods have become popular, overall non-parametric methods. While non-parametric tolerance regions can be well
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ASUSD nuclear data sensitivity and uncertainty program package: validation on fusion and fission benchmark experiments Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-12 Bor Kos; Aljaž Čufar; Ivan A. Kodeli
Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic
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Safety assessment of nuclear fuel reprocessing plant under the free drop impact of spent fuel cask and fuel assembly Part I: Large-scale model test and finite element model validation Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-12 Z.C. Li; Y.H. Yang; Z.F. Dong; T. Huang; H. Wu
This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is
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Effect of mechanical alloying on the microstructural evolution of a ferritic ODS steel with (Y-Ti-Al-Zr) addition processed by Spark Plasma Sintering (SPS) Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-10 E. Macía; A. García-Junceda; M. Serrano; S.J. Hong; M. Campos
The high-energy milling is one of the most extended techniques to produce Oxide dispersion strengthened (ODS) powder steels for nuclear applications. The consequences of the high energy mill process on the final powders can be measured by means of deformation level, size, morphology and alloying degree. In this work, an ODS ferritic steel, Fe-14Cr-5Al-3W-0.4Ti-0.25Y2O3-0.6Zr, was fabricated using two
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Evaluation of the true-strength characteristics for isotropic materials using ring tensile test Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-09 A.S. Frolov; I.V. Fedotov; B.A. Gurovich
The paper proposes a technique for reconstructing the true hardening curve of isotropic materials from ring tensile tests. Neutron irradiated 42XNM alloy tensile properties were investigated. The calculation of the true hardening curve for tensile and compression tests of standard cylindrical samples was performed at the first step. After that, the FEM-model was developed and validated using the ring
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A Study on Calculation of Friction Coefficient and Packing Stress Using Static Diagnosis Test For A Balanced Globe Valve In Nuclear Power Plants Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-08 Jaehyung Kim; Taemook Lim; Ho-Geun Ryu
Valve assembly used in nuclear power plants must be qualified and supervised. New technical standards such ASME QME-1 2007 particularly require detailed qualification using experiment and analysis. Particularly, diagnostic tests and engineering studies are required for qualification of ASME QME-1 2007. Among these studies, research on the measurement of friction coefficient and packing stress is important
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Leak flow prediction during loss of coolant accidents using deep fuzzy neural networks Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-08 Ji Hun Park; Ye Ji An; Kwae Hwan Yoo; Man Gyun Na
The frequency of reactor coolant leakage is expected to increase over the lifetime of a nuclear power plant owing to degradation mechanisms, such as flow-acceleration corrosion and stress corrosion cracking. When loss of coolant accidents (LOCAs) occur, several parameters change rapidly depending on the size and location of the cracks. In this study, leak flow during LOCAs is predicted using a deep
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Investigation on cavitating flow and parameter effects in a control valve with a perforated cage Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-06 Hong Wang; Zhimao Zhu; Miao Zhang; Jie Li; Weiqi Huo
Valve is widely used in the various industry areas to adjust and control the flow. Cavitation frequently takes place and sometimes is inevitable in various types of valve to cause the erosion damage. Therefore, how to control and minimize the effect of cavitation is still an important topic. This study numerically investigates the cavitating flow in a control valve with a perforated cage. The effects
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Effect of thermal aging on the mechanical, intergranular corrosion and corrosion fatigue properties of Z3CN20.09M cast duplex stainless steel Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-06 Wenxin Ti; Huanchun Wu; Fei Xue; Guodong Zhang; Qunjia Peng; Kewei Fang; Xitao Wang
The effect of thermal aging at 475°C and 750°C of Z3CN20.09M cast duplex stainless steel (CDSS) on microstructure, mechanical and intergranular corrosion properties were investigated by transmission electron microscope (TEM), nano indenter, scanning electron microscope (SEM) and corrosion fatigue test system. The result indicated that the spinodal decomposition and G precipitated were occurred after
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Comparisons of Performance and Operation Characteristics for Closed- and Open-loop Passive Containment Cooling System Design Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-06 Jungjin Bang; Hangon Kim; Dong-Wook Jerng
Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance
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Numerical study of oxygen transport characteristics in lead-bismuth eutectic for gas-phase oxygen control Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Chenglong Wang; Yan Zhang; Dalin Zhang; Zhike Lan; Wenxi Tian; Guanghui Su; Suizheng Qiu
One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and
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Region-Wise Evaluation of Gamma-ray Exposure Dose in Decontamination Operation after a Nuclear Accident Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Hae Sun Jeong; Won Tae Hwang; Moon Hee Han; Eun Han Kim; Jo Eun Lee; Cheol Woo Lee
The gamma-ray exposure doses in decontamination operation after a nuclear accident were evaluated with a consideration of various geometrical conditions and specific gamma-ray energies. The calculation domain is organized with three residence types and each form is divided into two kinds of geometrical arrangements. The position-wise air KERMA values were calculated with an assumption of evenly distributed
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Fuzzy-PID controller for Motion Control of CFETR Multi-Functional Maintenance Platform Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Dongyi Li; Kun Lu; Yong Cheng; Wenlong Zhao; Songzhu Yang; Yu Zhang; Junwei Li; Huapeng Wu
The motion control of the divertor maintenance system of the China Fusion Engineering Test Reactor (CFETR) was studied in this paper, in which CFETR Multi-Functional Maintenance Platform (MFMP) was simplified as a parallel robot for the convenience of theoretical analysis. In order to design the motion controller of parallel robot, the kinematics analysis of parallel robot was carried out. After that
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Performance evaluation of an adjustable gantry PET (AGPET) for small animal PET imaging Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Hankyeol Song; In Soo Kang; Kyu Bom Kim; Chanwoo Park; Min Kyu Baek; Seongyeon Lee; Yong Hyun Chung
A rectangular-shaped PET system with an adjustable gantry(AGPET) has been developed for imaging small animals. The AGPET system employs a new depth of interaction(DOI) method using a depth dependent reflector patterns and a new digital time pickoff method based on the pulse reconstruction method. To evaluate the performance of the AGPET, timing resolution, intrinsic spatial resolution and point source
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Development and Performance Evaluation of Large-Area Hybrid Gamma Imag er (LAHGI) Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Hyun Su Lee; Jae Hyeon Kim; Junyoung Lee; Chan Hyeong Kim
We report the development of a gamma-ray imaging device, named Large-Area Hybrid Gamma Imager (LAHGI), featuring high imaging sensitivity and good imaging resolution over a broad energy range. A hybrid collimation method, which combines mechanical and electronic collimation, is employed for a stable imaging performance based on large-area scintillation detectors for high imaging sensitivity. The system
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A Simple Data Assimilation Method to Improve Atmospheric Dispersion Based on Lagrangian Puff Model Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-03 Ke Li; Weihua Chen; Manchun Liang; Jianqiu Zhou; Yunfu Wang; Shuijun He; Jie Yang; Dandan Yang; Hongmin Shen; Xiangwei Wang
To model the atmospheric dispersion of radionuclides released from nuclear accident is very important for nuclear emergency. But the uncertainty of model parameters, such as source term and meteorology data, may significantly affect the prediction accuracy. Data assimilation (DA) is usually used to improve the model prediction with the measurements. The paper proposed a parameter bias transformation
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Efficiency calculation of the nMCP with 10B doping based on mathematical models Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-02 Jianqing Yang; Jianrong Zhou; Lianjun Zhang; Jinhao Tan; Xingfen Jiang; Jianjin Zhou; Xiaojuan Zhou; Linjun Hou; Yushou Song; XinLi Sun; Quanhu Zhang; Zhijia Sun; Yuanbo Chen
The nMCP (Neutron sensitive microchannel plate) combined with advanced readout electronics is widely used in energy selective neutron imaging because of its good spatial and timing resolution. Neutron detection efficiency is a crucial parameter for the nMCP. In this paper, a mathematical model based on the oblique cylindrical channel and elliptical pore was established to calculate the neutron absorption
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State-of-the-Art Progress of Gaseous Radiochemical Method for Detecting of Ionizing Radiation Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-02 S.G. Lebedev; V.E. Yants
The article provides a review of the research results obtained during of more than 20 years concerning using the gaseous radiochemical method (GRCM) for detecting of ionizing radiation. This method based on threshold nuclear reactions with production of radioactive noble gas which does not interact with the materials of gaseous tract. The applications of GRCM in the diagnostics of neutrinos, neutrons
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Economic Evaluation of Thorium Oxide Production from Monazite using Alkaline Fusion Method Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-02 Sanjith Udayakumar; Norlia Baharun; Sheikh Abdul Rezan; Aznan Fazli Ismail; Khaironie Mohamed Takip
Monazite is a phosphate mineral that contains thorium (Th) and rare earth elements. The Th concentration in monazite can be as high as 500 ppm, and it has the potential to be used as fuel in the nuclear power system. Therefore, this study aimed to conduct the techno-economic analysis (TEA) of Th extraction in the form of thorium oxide (ThO2) from monazite. Th can be extracted from monazite through
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High Resolution Size Characterization of Particulate Contaminants for Radioactive Metal Waste Treatment Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-02 Min-Ho Lee; Wonseok Yang; Nakkyu Chae; Sungyeol Choi
To regulate the safety protocols in nuclear facilities, radioactive aerosols have been extensively researched to understand their health impacts. However, most measured particle-size distributions remain at low resolutions, with the particle sizes ranging from nanometer to micrometer. This study combines the high-resolution detection of 500 size classes, ranging from 6 nm–10 μm, for aerodynamic diameter
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SrAl2Si2O8 Ceramic Matrices for 90Sr Immobilization Obtained via Spark Plasma Sintering-Reactive Synthesis Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-02 E.K. Papynov; A.A. Belov; O.O. Shichalin; I.Yu. Buravlev; S.A. Azon; A.V. Golub; A.V. Gerasimenko; A. Parotkina Yu; A.P. Zavjalov; I.G. Tananaev; V.I. Sergienko
In the present study, an original spark plasma sintering-reactive synthesis (SPS-RS) method for mineral-like ceramic materials based on SrAl2Si2O8 feldspar-like skeleton structure was used for the first time, promising solid-state matrices for reliable immobilization of high-energy 90Sr. The method is based on the “in-situ” reaction of a mixture of SrO, Al2O3 and SiO2 oxides when heated by a unipolar
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Development of a structural integrity evaluation program for elevated temperature service according to ASME code Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-01 Nak Hyun Kim; Jong Bum Kim; Sung Kyun Kim
A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities
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High performance γ-ray imager using dual anti-mask method for the investigation of high-energy nuclear materials Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-01 Taewoong Lee; Wonho Lee
As the γ-ray energy increases, a reconstructed image becomes noisy and blurred due to the penetration of the γ-ray through the coded mask. Therefore, the thickness of the coded mask was increased for high energy regions, resulting in severely decreased the performance of the detection efficiency due to self-collimation by the mask. In order to overcome the limitation, a modified uniformly redundant
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Diagnostic Methods Applied to Esfahan Light Water Subcritical Reactor (ELWSCR) Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-02-01 Mohammad Arkani
In this work, Esfahan light water subcritical reactor (ELWSCR) is analysed using experimental and theoretical diagnostic methods. Important neutronic parameters of the system such as prompt neutron lifetime, delayed neutron fraction, prompt neutron decay constant, negative reactivity of the core, fuel and moderator temperature coefficient of reactivity, and overall and local void coefficient of reactivity
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Design and Heat Transfer Optimization of a 1 kW Free-piston Stirling Engine for Space Reactor Power System Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-30 Zhiwen Dai; Chenglong Wang; Dalin Zhang; Wenxi Tian; Suizheng Qiu; G.H. Su
The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that
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Design and Neutronic Analysis of the Intermediate Heat Exchanger of a Fast-Spectrum Molten Salt Reactor Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-29 Jamiyansuren Terbish; W.F.G. van Rooijen
Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal
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Topology optimization of design of tie-down structure for transportation of metal cask containing spent nuclear fuel Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-28 Gil-Eon Jeong; Woo-Seok Choi; Sang Soon Cho
Spent nuclear fuel, which can degrade during long-term storage, must be transported intact in normal transport conditions. In this regard, many studies, including those involving Multi-Modal Transportation Test (MMTT) campaigns, have been conducted. In order to transport the spent fuel safely, a tie-down structure for supporting and transporting a cask containing the spent fuel is essential. To ensure
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Study on Analog-based Ex-Core Neutron Flux Monitoring Systems of Korean Nuclear Power Plants for Digitization Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-28 Young Baik Kim; Felipe P. Vista; Kil To Chong
The analog-based Ex-core Neutron Flux Monitoring System (ENFMS) in Korean Nuclear Power Plants (NPPs) has been performing its intended functions successfully for a long time. On the other hand, the primary concern with the extended use of analog systems is the aging effect, such as mechanical failures, environmental degradation, and obsolescence. The transition to a digital-based Man-Machine Interface
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Comparative Analysis of Internal Flow Characteristics of LBE-cooled Fast Reactor Main Coolant Pump with Different Structures under Reverse Rotation Accident Conditions Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-25 Yonggang Lu; Xiuli Wang; Qiang Fu; Yuanyuan Zhao; Rongsheng Zhu
Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution
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Effect of packing structure on anisotropic effective thermal conductivity of thin ceramic pebble bed Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-23 Shuang Wang; Shuai Wang; Bowen Wu; Yuelin Lu; Kefan Zhang; Hongli Chen
Helium cooled solid breeder blanket as an important blanket candidate of the Tokamak fusion reactor uses ceramic pebble bed for tritium breeding. Considering the poor effective thermal conductivity of the ceramic breeder pebble bed, thin structure of tritium breeder pebble bed is usually adopted in the blanket design. The container wall has a great influence on the thin pebble bed packing structure
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Improvement of signal and noise performance using single image super-resolution based on deep learning in single photon-emission computed tomography imaging system Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-23 Kyuseok Kim; Youngjin Lee
Because single-photon emission computed tomography (SPECT) is one of the widely used nuclear medicine imaging systems, it is extremely important to acquire high-quality images for diagnosis. In this study, we designed a super-resolution (SR) technique using dense block-based deep convolutional neural network (CNN) and evaluated the algorithm on real SPECT phantom images. To acquire the phantom images
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A novel reconstruction algorithm based on density clustering for cosmic-ray muon scattering inspection Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-23 Linjun Hou; Quanhu Zhang; Jianqing Yang; Xingfu Cai; Qingxu Yao; Yonggang Huo; Qifan Chen
As a relatively new radiation imaging method, the cosmic-ray muon scattering imaging technology can be used to prevent nuclear smuggling and is of considerable significance to nuclear safety. Proposed in this paper is a new reconstruction algorithm based on density clustering, aiming to improve inspection quality with better performance. Firstly, this new algorithm is introduced in detail. Then in
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Reproduction Strategy of Radiation Data with Compensation of Data Loss Using a Deep Learning Technique Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-23 Woosung Cho; Hyeonmin Kim; Duckhyun Kim; SongHyun Kim; Inyong Kwon
In nuclear-related facilities, such as nuclear power plants, research reactors, accelerators, and nuclear waste storage sites, radiation detection, and mapping are required to prevent radiation overexposure. Sensor network systems consisting of radiation sensor interfaces and wxireless communication units have become promising tools that can be used for data collection of radiation detection that can
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Performance-Based Earthquake Engineering Methodology for Seismic Analysis of Nuclear Cable Tray System Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-22 Baofeng Huang
The Pacific Earthquake Engineering Research (PEER) Centre has been developing a performance-based earthquake engineering (PBEE) methodology, which is based on explicit determination of performance, e.g., monetary losses, in a probabilistic manner where uncertainties in earthquake ground motion, structural response, damage estimation, and losses are explicitly considered. To carry out the PEER PBEE
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Effects of electronic energy deposition on pre-existing defects in 6H-SiC Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-22 Wenlong Liao; Huan He; Yang Li; Wenbo Liu; Hang Zang; Jianan Wei; Chaohui He
Silicon carbide is widely used in radiation environments due to its excellent properties. However, when exposed to the strong radiation environment constantly, plenty of defects are generated, thus causing the material performance downgrades or failures. In this paper, the two-temperature model (2T-MD) is used to explore the defect recovery process by applying the electronic energy loss (Se) on the
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Circumferential Steady-State Creep Test and Analysis of Zircaloy-4 Fuel Cladding Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-18 Gyeong-Ha Choi; Chang-Hwan Shin; Jae Yong Kim; Byoung Jae Kim
In recent studies, the creep rate of Zircaloy-4, one of the basic property parameters of the nuclear fuel code, has been commonly used with the axial creep model proposed by Rosinger et al. However, in order to calculate the circumferential deformation of the fuel cladding, there is a limitation that a difference occurs depending on the anisotropic coefficients used in deriving the circumferential
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Axial strength of Zircaloy-4 samples with reduced thickness after a simulated Loss of Coolant Accident Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-18 Jean Desquines; Tatiana Taurines
To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 μm. After high temperature oxidation under steam at 1200°C with measured ECR ranging from 10 to 18 % and water quenching, machined samples were axially loaded until fracture. The tests
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Analysis of the Performances of the CFD Schemes Used for Coupling Computation Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-16 Guangliang Chen; Hongwei Jiang; Huilun Kang; Rui Ma; Lei Li; Yang Yu
In this paper, the coupling of fine-mesh computational fluid dynamics (CFD) thermal-hydraulics (TH) code and neutronics code is achieved using the Ansys Fluent User Defined Function (UDF) for code development, including parallel meshing mapping, data computation, and data transfer. Also, some CFD schemes are designed for mesh mapping and data transfer to guarantee physical conservation in the coupling
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Verification and Validation of Isotope Inventory Prediction for Back-End Cycle Management Using Two-Step Method Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-15 Jaerim Jang; Bamidele Ebiwonjumi; Wonkyeong Kim; Alexey Cherezov; Jinsu Park; Deokjung Lee
This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history
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Development of a Radiological Emergency Evacuation Model Using Agent-Based Modeling Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-12 Yujeong Hwang; Gyunyoung Heo
In order to mitigate the damage caused by accidents in nuclear power plants (NPPs), evacuation strategies are usually managed on the basis of off-site effects such as the diffusion of radioactive materials and evacuee traffic simulations. However, the interactive behavior between evacuees and the accident environment has a significant effect on the consequential gap. Agent-based modeling (ABM) is a
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Magnesium potassium phosphate cements to immobilize radioactive concrete wastes generated by decommissioning of nuclear power plants Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-12 Jae-Young Pyo; Wooyong Um; Jong Heo
This paper evaluates the efficacy of magnesium potassium phosphate cements (MKPCs) as waste forms for the solidification of radioactive concrete powder wastes produced by the decommissioning of nuclear power plants. MKPC specimens that contained up to 50 wt.% of simulated concrete powder wastes (SCPWs) were evaluated. We measured the porosity and compressive strength of the MKPC specimens, observing
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Analysis of Dislocation Density in Strain-Hardened Alloy 690 Using Scanning Transmission Electron Microscopy and Its Effect on the PWSCC Growth Behavior Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-11 Sung-Woo Kim; Tae-Young Ahn; Dong-Jin Kim
The dislocation density in strain-hardened Alloy 690 was analyzed using scanning transmission electron microscopy (STEM) to study the relationship between the local plastic strain and susceptibility to primary water stress corrosion cracking (PWSCC) in nuclear power plants. The test material was cold-rolled at various thickness reduction ratios from 10% to 40% to simulate the strain-hardening condition
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Neutronics study on small power ADS loaded with recycled inert matrix fuel for transuranic elements transmutation using Serpent code Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-09 Thanh Mai Vu; Donny Hartanto; Pham Nhu Viet Ha
A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO2 spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that
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Dynamic Characteristics of Single Door Electrical Cabinet under Rocking: Source Reconciliation of Experimental and Numerical Findings Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-09 Bub-Gyu Jeon; Ho-Young Son; Seung-Hyun Eem; In-Kil Choi; Bu-Seog Ju
Seismic qualifications of electrical equipment, such as cabinet systems, have been emerging as the key area of nuclear power plants in Korea since the 2016 Gyeongju earthquake, including the high-frequency domain. In addition, electrical equipment was sensitive to the high-frequency ground motions during the past earthquake. Therefore, this paper presents the rocking behavior of the electrical cabinet
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A novel radioactive particle tracking algorithm based on Deep Rectifier Neural Network Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-08 Roos Sophia de Freitas Dam; Marcelo Carvalho dos Santos; Filipe Santana Moreira do Desterro; William Luna Salgado; Roberto Schirru; César Marques Salgado
Radioactive particle tracking (RPT) is a minimally invasive nuclear technique that tracks a radioactive particle inside a volume of interest by means of a mathematical location algorithm. During the past decades, many algorithms have been developed including ones based on artificial intelligence techniques. In this study, RPT technique is applied in a simulated test section that employs a simplified
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Numerical Estimation of Errors in Drop Angle during Drop Tests of IP-Type Metallic Transport Containers for Radioactive Materials Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-04 Jongmin Lim; Yun Young Yang; Ju-chan Lee
For industrial package (IP)-type transport containers for radioactive materials, a free drop test should be conducted under regulatory conditions. Owing to various uncertainties observed during the drop test, errors in drop angles inevitably occur. In IP-type metal transport containers in which the container directly impacts onto a rigid target without any shock absorbing materials, the error in the
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Delta-form-based method of solving high order spatial discretization schemes for neutron transport Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-04 Xiafeng Zhou; Changming Zhong; Fu Li
Delta-form-based methods for solving high order spatial discretization schemes are introduced into the reactor SN transport equation. Due to the nature of the delta-form, the final numerical accuracy only depends on the residuals on the right side of the discrete equations and have nothing to do with the parts on the left side. Therefore, various high order spatial discretization methods can be easily
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Study on sputtering yield of tungsten with different particle sizes: Surface roughness dependence Nucl. Eng. Technol. (IF 1.846) Pub Date : 2021-01-03 Tae Hyun Kwon; Sangjune Park; Jeong Min Ha; Young-Sang Youn
The sputtering yield of tungsten pellets composed of different particle sizes of <1, 12, 44−74, and 149−297 μm was systematically investigated by bombardment with Ar+ ions accelerated at 2.0 keV in an ultra-high vacuum chamber. We found that the tungsten sample fabricated from larger particles had a higher surface roughness, based on the results obtained from a surface profiler. Using the data of the
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An investigation of the nuclear shielding effectiveness of some transparent glasses manufactured from natural quartz doped lead cations Nucl. Eng. Technol. (IF 1.846) Pub Date : 2020-12-31 Said M. Kassem; G.S.M. Ahmed; A.M. Rashad; S.M. Salem; S. Ebraheem; A.G. Mostafa
The influence of lead cations on natural quartz (QZ) from Egypt as a glass shielding material for the composition with nominal formula (10Na2O - (90 - x) QZ – xPbO (where x = 30, 35, 40, 45 and 50 mol %)) was examined. The studied samples are synthesized via the melt quenching method at 1050 0C. The X-ray diffraction XRD patterns were confirmed the glass nature for studied samples. Moreover, the optical
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Effect of oxygen containing compounds in uranium tetrafluoride on its non-adiabatic calciothermic reduction characteristics Nucl. Eng. Technol. (IF 1.846) Pub Date : 2020-12-31 Sonal Gupta; Raj Kumar; Santosh K. Satpati; Manharan L. Sahu
Uranium ingot is produced by metallothermic reduction of uranium tetrafluoride using magnesium or calcium as reductant. Presence of oxygen containing compounds viz. uranyl fluoride and uranium oxide in the starting uranium fluoride has a significant effect on the firing time, final temperature of the charge, slag-metal separation and hence the metal recovery. As reported in the literature, the maximum
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Numerical Simulation of Three-dimensional Flow and Heat Transfer Characteristics of Liquid Lead-Bismuth Nucl. Eng. Technol. (IF 1.846) Pub Date : 2020-12-31 Shaopeng He; Mingjun Wang; Jing Zhang; Wenxi Tian; Suizheng Qiu; G.H. Su
Liquid lead-bismuth cooled fast reactor is one of the most promising reactor types among the fourth-generation nuclear energy systems. The flow and heat transfer characteristics of lead-bismuth eutectic (LBE) are completely different from ordinary fluids due to its special thermal properties, causing that the traditional Reynolds analogy is no longer recommended and appropriate. More accurate turbulence
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