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Impact of fuel temperature on nuclear core design calculations Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-18 Dušan Čalič, Luka Snoj, Marjan Kromar
The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel
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Conceptual design for a 5 kWe space nuclear reactor power system Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-18 Huaping Mei, Dali Yu, Shengqin Ma, Jiansong Zhang, Yongju Sun, Chao Chen, Meisheng He, Haixia Wang, Yang Li, Liang Wang, Taosheng Li, Jie Yu
Enhancing the capabilities of unmanned space exploration, such as satellite monitoring and space science missions, requires efficient and reliable nuclear power systems. A viable solution is found in the 1–10 kWe power level of space nuclear reactor power systems, offering advantages such as a manageable research and development process, and relatively low investment requirements. This paper introduces
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Solution of OECD/NEA PWR MOX/UO2 benchmark with a high-performance pin-by-pin core calculation code Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-16 Hyunsik Hong, Jooil Yoon
Expanding upon the framework of the steady-state pin-by-pin 2D/1D decoupling method, a novel and high-performance pin-by-pin transient calculation method has been introduced. This transient method, consistent to the steady-state formulation, is designed for time-dependent calculations utilizing a 3D diffusion-based finite difference method (FDM). The inherent complexity of the large 3D problem is effectively
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Noticeable localized corrosion of solid boric acid on 304 stainless steel Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-16 Xinzhu Li, Wen Sun, Guiling Ning
With the aim to determine the potential corrosion effects of solid boric acid (BA) on light water reactors or other BA-involved equipment, the corrosion behaviors of solid BA on 304 stainless steel (SS) at different temperatures were investigated. Upon comparing the corrosion behaviors of solid BA at different temperatures, significant localized corrosion was observed on 304 SS surfaces at 150 °C following
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Uncertainty quantification based on similarity analysis of reactor physics benchmark experiments for SFR using TRU metallic fuel Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-13 YuGwon Jo, Jaewoon Yoo, Jong-Hyuk Won, Jae-Yong Lim
One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU
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Method of the known cross sections for calibration of the fast neutron spectrometer with a single-crystal stilbene based detector Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-13 I.V. Urupa, E.V. Ryabeva, R.F. Ibragimov, V.D. Sapozhnikov
The present work is devoted to implementation of the stilbene-based neutron spectrometer energy calibration method. The results of experiments with portable neutron generators and PuBe source and scattering materials with known cross sections are used for this method. It is shown that the submitted method makes it possible to carry out fast neutron spectrometry in the energy range from 1 to 15 MeV
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A new surrogate method for the neutron kinetics calculation of nuclear reactor core transients Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-13 Xiaoqi Li, Youqi Zheng, Xianan Du, Bowen Xiao
Reactor core transient calculation is very important for the reactor safety analysis, in which the kernel is neutron kinetics calculation by simulating the variation of neutron density or thermal power over time. Compared with the point kinetics method, the time-space neutron kinetics calculation can provide accurate variation of neutron density in both space and time domain. But it consumes a lot
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Seismic fragility analysis of shield building considering strength ratio of mainshock and aftershocks Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-09 Xue Zhang, Chunfeng Zhao, Lunhai Zhi, Rui Pang, Y.L. Mo
The shield building of the AP1000 nuclear power plant serves as a crucial protective barrier against radioactive substances. However, past research indicates that structures are susceptible to experiencing aftershocks, which may lead to unforeseeable damage and potential radioactive material leakage. To address this issue, a finite element model of the shield building was established with the damage
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Prediction of small-scale leak flow rate in LOCA situations using bidirectional GRU Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-09 Hye Seon Jo, Sang Hyun Lee, Man Gyun Na
It is difficult to detect a small-scale leakage in a nuclear power plant (NPP) quickly and take appropriate action. Delaying these procedures can have adverse effects on NPPs. In this paper, we propose leak flow rate prediction using the bidirectional gated recurrent unit (Bi-GRU) method to detect leakage quickly and accurately in small-scale leakage situations because large-scale leak rates are known
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Study on producing radioisotopes based on fission or radiative capture method in a high flux reactor Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-09 Wei Xu, Jian Li, Lei Shi
Radioisotopes tend to play important roles in many fields, such as industry, healthcare, agriculture, aerospace, etc. Radioisotope production is mainly through accelerators or research reactors, and high flux research reactor is one of the most effective approaches for radioisotope production. The physical basis of preparing radioisotope relies on nuclear reactions occurring in the reactor core, which
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Development and application of the helically coiled once-through steam generator module for dynamic simulation of nuclear hybrid energy system Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-05 Keon Yeop Kim, Young Suk Bang
Small Modular Reactors (SMRs) adopt the Helically Coiled Once-Through Steam Generators (OTSG) extensively for its compactness and higher heat transfer efficiency. As a heat exchanger between the primary side (reactor coolant system) and the secondary side (feedwater and steam system) of nuclear steam supply system, the inlet/outlet conditions both of shell side and tube side of OTSGs have significant
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Extensive analysis of several Indian and Yemeni soils' gamma-ray shielding characteristics: An experimental and simulation approach Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-04 Shamsan S. Obaid, M.I. Sayyed, A.S. Alameen, D.K. Gaikwad, K.A. Mahmoud
The linear attenuation coefficients (LAC) of four soils (Black cotton (S1), Sandy (S2), Clay (S3), and Sandy (S4)) samples were measured at photon energies released from radioisotopes Co (122 keV), Ba (356 keV), Na (511 and 1275 keV), Cs (662 keV), Mn (840 keV), and Co (1330 keV) using a gamma spectrometer includes a NaI (Tl) scintillation detector. The experimental measurements were confirmed utilizing
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Corrigendum to “Experiment of proof-of-principle on prompt gamma-positron emission tomography (PG-PET) system for in-vivo dose distribution verification in proton therapy” [Nucl. Eng. Technol. 55 (2023) 2018–2025] Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-04 Bo-Wi Cheon, Hyun Cheol Lee, Sei Hwan You, Hee Seo, Chul Hee Min, Hyun Joon Choi
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Hybrid vibro-acoustic model reduction for model updating in nuclear power plant pipeline with undetermined boundary conditions Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-04 Hyeonah Shin, Seungin Oh, Yongbeom Cho, Jinyoung Kil, Byunyoung Chung, Jinwon Shin, Jin-Gyun Kim
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Generation of security system defense strategies based on evolutionary game theory Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-04 Bowen Zou, Yongdong Wang, Chunqiang Liu, Mingguang Dai, Qianwen Du, Xiang Zhu
The physical protection systems of Nuclear Power Plant are utilized to safeguard targets against intrude by attacker. As the methods employed by attackers to intrude Nuclear Power Plant become increasingly complex and diverse, there is an urgent need to identify optimal defense strategies to interrupt adversary intrusions. This paper focuses on studying the defense of security personnel against adversary
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An efficient numerical modeling approach for coupled electrical cabinets in nuclear power plants Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-04 Sudeep Das Turja, Md. Rajibul Islam, Dong Van Nguyen, Dookie Kim
Seismic quantification of nonstructural components like electrical cabinets is essential to ensure the uninterrupted operation of nuclear facilities during earthquake events. This process requires experimental tests, which can be expensive, time-consuming, and limited by safety concerns and precision. As an alternative to that, numerical simulations should be done in such a way that they are capable
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Fission counter array for pulse-mode measurements of high-flux and high-energy neutrons Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-03 Pilsoo Lee
This manuscript describes a neutron counting system based on cylindrical fission counters that can monitor neutron activity for high-energy neutron flux above 10 MeV under electrically noisy environments with intense gamma rays. Miniature fission counters with depleted uranium as sensitive material and modular electronics were built for digital signal processing and high-countrate operation. The counters
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Effects of microalloying element addition on mechanical properties of SA508 Gr.1A low-alloy steels Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-02 Se-mi Hyun, Min-Chul Kim, Seokmin Hong, Jongmin Kim, Seok Su Sohn
SA508 Gr.1A low-alloy steel is being considered as a candidate material for main steam line piping in nuclear power plants. Therefore, improving its strength and toughness is essential for enhancing the leak-before-break (LBB) margin. In this study, six types of model alloys were fabricated by varying the contents of microalloying elements (C, Cu, B, Ti, and Nb) to enhance the mechanical properties
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Loading pattern optimization of VVER-1000 reactor core based on the discrete golden eagle optimization algorithm Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-02 Sajjad Abbasi Fashami, Mahdi Zangian, Abdolhamid Minuchehr, Ahmadreza Zolfaghari
The main features of the loading pattern optimization (LPO) problem, such as high-dimensionality, multi-modality, and non-linearity, make it difficult to achieve a truly optimal configuration. In recent years, metaheuristic methods have been successfully used to solve this problem. In this research, a discrete golden eagle optimization (DGEO) algorithm has been developed to solve the LPO problem in
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JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-02 Anže Pungerčič, Alireza Haghighat, Luka Snoj
Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally
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Real-time measurements and modeling of sodium combustion aerosol dynamics in test chamber to improve the evaluation of SFR containment aerosol behaviour Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-01 Usha Pujala, Amit Kumar, Subramanian Venkatesan, Sujatha Pavan Narayanam, Venkatraman Balasubramanian
The initial size distribution and morphological parameters of sodium aerosols are critical in evaluating the accidental suspended aerosol behaviour in Sodium-cooled Fast Reactor (SFR) containment. Mass-based measurements were more familiar in characterizing the sodium aerosols. Real-time number size distribution measurements are carried out in this study. The sensitivity analysis of sodium aerosol
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Titanium alloys: A closer-look at mechanical, gamma-ray, neutron, and transmission properties of different grade alloys through MCNPcode application Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-01 Ghada ALMisned, Omer Guler, Duygu Sen Baykal, G. Kilic, H.O. Tekin
Titanium alloys play a vital role in optimizing the effectiveness and security of nuclear reactors, strengthening structural durability, and facilitating the effective handling of nuclear waste. The aim of this study is to investigate the gamma-ray, neutron, and transmission properties of four common titanium alloys through the examination of the deposited energy amount in the liquid sodium coolant
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Extended cognitive reliability and error analysis method for advanced control rooms of nuclear power plants Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-04-01 Xiaodan Zhang, Shengyuan Yan, Xin Liu
This study proposes a modified extended cognitive reliability and error analysis method (CREAM) for achieving a more accurate human error probability (HEP) in advanced control rooms. The traditional approach lacks failure data and does not consider the common performance condition (CPC) weights in different cognitive functions. The modified extended CREAM decomposes tasks using a method that combines
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Comparative analysis of modeling approaches for sulfide-induced corrosion of copper disposal canisters in a 3-dimensional domain Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-30 Heejae Ju, Nakkyu Chae, Jung-Woo Kim, Hong Jang, Sungyeol Choi
Copper canisters are commonly employed to contain HLW for the long-term, making it crucial to understand how corrosion affects the canister. This study conducted a comparative analysis of two widely used calculation methods for modeling canister corrosion within a 3-D DGR domain. The first method, termed , assumes an immediate sulfide-copper reaction and has been traditionally used due to its conservative
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Experimental examination on physical and radiation shielding features of boro-silicate glasses doped with varying amounts of BaO Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-28 M.I. Sayyed, Abdelmoneim Saleh, Anjan Kumar, Fatma Elzahraa Mansour
Investigations were conducted on the addition of barium's impact on the radiation shielding and physical attributes of five different glasses, designated S1–S5, with varying BaO contents. Using two point sources namely Co and Cs along with a scintillation detector [NaI(TL)], experimental measurements were made of the shielding parameters of γ-rays, namely the effective atomic number (Z), electron density
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Research on void drift between rod bundle subchannels Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-27 Shasha Liu, Zaiyong Ma, Bo Pang, Rui Zhang, Luteng Zhang, Quanyao Ren, Liangming Pan
Void drift between subchannels in a rod bundle is a crucial phenomenon affecting the calculation accuracy of thermal-hydraulic parameters in SMRs. It holds significant importance in enhancing the precision of safety analysis for SMRs. Existing research on experiment and model of void drift between rod bundle subchannels is relatively rare, and the accuracy of model calculations requires improvement
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Electricity mix scenarios simulation for Korean carbon neutrality in 2050 Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-26 Pilhyeon Ju, Sungyeol Choi, Jongho Lee
As the realization of carbon neutrality has been a main assignment for coping with the global climate change, it became necessary to analyze upcoming changes in electricity mix with economic and technical viewpoints. This paper presents a newly-developed simulation model that reflects the daily intermittency of renewable energy by applying daily average power supply-demand patterns for each season
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Asymmetric nexus between nuclear energy technology budgets and carbon emissions in European economies: Evidence from quantile-on-quantile estimation Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-25 Shuifa Shen, Muhammad Zahir Faridi, Raima Nazar, Sajid Ali
Our research seeks to assess the influence of nuclear energy technology on carbon emissions in the top 10 European economies comprising the topmost nuclear energy R&D budget (France, Germany, Russia, the Netherlands, the UK, Finland, Spain, Sweden, Italy, and Switzerland). Unlike prior investigations predominantly relying on panel data methodologies without considering the distinctive characteristics
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A robust nano-indentation modeling method for ion-irradiated FCC single crystals using strain-gradient crystal plasticity theory and particle swarm optimization algorithm Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-25 Van-Thanh Pham, Jong-Sung Kim
Addressing the challenge of identifying an appropriate set of material and irradiation parameters for accurate simulation models using crystal plasticity finite element method (CPFEM), this study proposes a novel two-stage method for nano-indentation modeling of ion-irradiated face-centered cubic (FCC) materials. It includes implementing the strain-gradient crystal plasticity (SGCP) theory with irradiation
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Realistic estimation framework of radioactive release distributions into the environment during nuclear power plant accidents Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-22 Wasin Vechgama, Jaehyun Cho
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Research on design requirements for passive residual heat removal system of lead cooled fast reactor via model-based system engineering Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-21 Mao Tang, Junqian Yang, Pengcheng Zhao, Kai Wang
Traditional text-based system engineering, which has been used in the design and application of passive residual heat removal system (PRHRS) for lead-cooled fast reactors, is prone to several problems such as low development efficiency, long iteration cycles, and model ambiguity. This study aims to effectively address the above-mentioned problems by adopting a model-based system engineering (MBSE)
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Study on the shielding performance of bismuth oxide as a spent fuel dry storage container based on Monte Carlo simulation Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-21 Guo-Qiang Zeng, Shuang Qi, Peng Cheng, Sheng Lv, Fei Li, Xiao-Bo Wang, Bing-Hai Li, Qing-Ao Qin
For traditional spent fuel shielding materials, due to physical and chemical defects and cost constraints, they have been unable to meet the needs. Therefore, this paper carries out the first discussion on the application and performance of bismuth in neutron shielding by establishing Monte Carlo simulation on the neutron flux model of shielded spent fuel. Firstly, functional fillers such as bismuth
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Development of uncertainty quantification module for VVER analysis in STREAM/RAST-V two-step method Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-21 Jaerim Jang, Yunki Jo, Deokjung Lee
This paper introduces the creation of a module for Uncertainty Quantification (UQ) specifically designed for VVER analysis through the implementation of the STREAM/RAST-V two-step approach. The aim was to expand the range of use by developing a UQ module tailored for analyzing VVER. This research presents two innovative computational functionalities: (1) development of a library for the pin-based pointwise
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Simulation-guided design of a target-cooling system for cyclotron-based isotope production Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-21 Sang Chul Mun, Gyeol Chan Kang, Choong Mo Kang, Jung Young Kim, Kyo Chul Lee, Seyoung Oh
Isotopes are an important aspect of modern medical and scientific research and cyclotron-based isotope production is of particular interest. Cooling devices are required to manage the heat generated by high-energy particle beams to ensure that they can be delivered securely. However, there is considerable scope for further advancements in the design of cooling systems. Therefore, this study uses simulations
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Neutronics analysis of the ion cyclotron resonance heating antenna of the China Fusion Engineering Test Reactor Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-21 Gaoxiang Wang, Chengming Qin, Shanliang Zheng, Yongsheng Wang, Kun Xu, Huiqiang Ma
Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted
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Whole-core analysis of Watts bar benchmark with three-dimensional MOC code STREAM3D Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-19 Murat Serdar Aygul, Wonkyeong Kim, Deokjung Lee
This paper presents a high-fidelity simulation of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) 3D whole-core Watts Bar benchmark using the UNIST in-house STREAM3D (Steady State and Transient Reactor Analysis code with Method of Characteristics) neutronic code. The benchmark encompasses various whole-core exercises, including single physics problems,
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Multiscale modeling of smectite illitization in bentonite buffer of engineered barrier system Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-19 Xinwei Xiong, Jiahui You, Kyung Jae Lee, Jin-Seop Kim
With the increasing usage of nuclear energy, how to properly dispose nuclear waste becomes a critical issue. In this study, a multiscale modeling approach combining the experimental findings is presented to address the illitization process, its impact on transport properties, and system behavior of bentonite buffer in engineered barrier systems (EBS). Through the pore-scale modeling, reactive transport
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Analysis of values-beliefs-norms of decommissioned nuclear power plant reestablishment acceptance in developing countries: a perspective from the Philippines Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-18 Leo Miguel V. Tolentino, Ardvin Kester S. Ong, Josephine D. German
Amid the ongoing discourse on clean energy solutions, the reopening of decommissioned plants, such as the Bataan Nuclear Power Plant (BNPP) in the Philippines has become a focal point in the country. This study delved into the complex web of factors influencing public acceptance of BNPP, employing the values-beliefs-norms theory. By utilizing partial-least square structural equation modeling, the research
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In-situ TEM investigation of zirconium alloy under Kr+ single-beam and Kr+-He+ dual-beam synergetic irradiation Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-18 Zhen Wang, Qing-Xue Yan, Zhong-Qiang Fang, Chen-Yang Lu
The in-situ TEM irradiation experiments of zirconium alloy were conducted at 573 K, 673 K, and 773 K utilizing a 400 keV Kr + single beam and a 400 keV Kr+ and 30 keV He + dual beam. The results show that a large number of dislocation loops have been characterized in the matrix of the zirconium alloy under irradiation. With increasing the irradiation damage dose, some dislocation loops have reacted
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An explicit approximation of the central angle for the curved interface in double-circle model for horizontal two-phase stratified flow Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-16 Taehwan Ahn, Dongwon Jeong, Jin-Yeong Bak, Jae Jun Jeong, Byongjo Yun
Stratified flow in horizontal tubes is frequently observed in gas-liquid two-phase flow system. In the two-fluid modeling, it is important to define the interface shape in solving the balance equations to determine the key parameters such as the interfacial transfer terms, void fraction, and pressure drop. A double-circle model is usually introduced to depict the concave-down interface in a horizontal
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Crevice chemistry and corrosion in high temperature water: A review Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-16 Young-Jin Kim, Chi Bum Bahn, Seung Heon Baek, Wonjun Choi, Geun Dong Song
Crevice corrosion is a localized attack of metal that occurs in occluded areas of materials as a result of a degradation of the oxide passivity on the metal surface in contact with stagnant environments. Materials suffer crevice corrosion when generally the crevice opening gap is so narrow that the migration or diffusion of ionic species into the crevice can be restricted and consequently results in
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Parametric study on stress distribution of thin disk specimen of rupture disk corrosion test influencing SCC initiation using finite element analysis Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-15 Tae Young Kim, Sung Woo Kim, Dong Jim Kim, Sang Tae Kim
Rupture disk corrosion test (RDCT) method has been recently developed for real-time measurement of initiation of stress corrosion cracking (SCC) in a high-temperature water. This work presents a parametric study on the stress distribution of a thin disk specimen of RDCT to consider the fixture shape and friction using finite element analysis (FEA). The FEA results showed a dome-shaped deformation of
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Corrigendum to “Li4SiO4 slurry conditions and sintering temperature for fabricating Li4SiO4 pebbles as tritium breeder for nuclear-fusion reactors” [Nucl. Eng. Technol. 55 (2023) 2966] Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-15 Young ah Park, Ji Won Yoo, Yi-Hyun Park, Young soo Yoon
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Improvement of doses rate prediction using the Kalman Filter-based algorithm and effective decay constant correction Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-15 Cheol-Woo Lee, Hyo Jun Jeong, Sol Jeong, Moon Hee Han
This study proposes an algorithm that combines a Kalman Filter method with effective decay constant correction to improve the accuracy of predicting radiation dose rate distribution during emergencies. The algorithm addresses the limitations of relying solely on measurement data by incorporating calculation data and refining the estimations.
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Development of an AI-based remaining trip time prediction system for nuclear power plants Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-14 Sang Won Oh, Ji Hun Park, Hye Seon Jo, Man Gyun Na
In abnormal states of nuclear power plants (NPPs), operators undertake mitigation actions to restore a normal state and prevent reactor trips. However, in abnormal states, the NPP condition fluctuates rapidly, which can lead to human error. If human error occurs, the condition of an NPP can deteriorate, leading to reactor trips. Sudden shutdowns, such as reactor trips, can result in the failure of
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Reactor core design with practical gadolinia burnable absorbers for soluble boron-free operation in the innovative SMR Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-14 Jin Sun Kim, Tae Sik Jung, Jooil Yoon
The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive
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Evaluation of MUF uncertainty based on GUM method for benchmark bulk handling facility Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-14 Hyun Cheol Lee, Jung Youn Choi, Hana Seo, Hyun Ju Kim, Yewon Kim, Haneol Lee
The Republic of Korea is performing independent national inspections under the IAEA's State System of Accounting for and Control (SSAC), and developing an evaluation methodology for the material unaccounted for (MUF) to reinforce capabilities with the purpose of assessment for the accounting system of the facility handling bulk nuclear materials. In relation to this, a new approach for MUF evaluation
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Assessing the nuclear weapons proliferation risks in nuclear energy newcomer countries: The case of small modular reactors Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-13 Philseo Kim, Sunil S. Chirayath
While several nuclear energy newcomer (NEN) countries have shown interest in small modular reactors (SMRs) as a potential energy source, this interest can generate new uncertainties regarding future nuclear weapons proliferation risks. Therefore, this research seeks to determine whether future SMR deployment in NEN countries will contribute to nuclear weapons proliferation, and how the risks can be
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Physical and γ-ray shielding properties of Vietnam's natural stones: An extensive experimental and theoretical study Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-11 Ta Van Thuong, O.L. Tashlykov, A.M. Shironina, I.P. Voronin, E.V. Kuvshinova, D.O. Pyltsova, E.I. Nazarov, K.A. Mahmoud
The current work deals with investigation of the gamma ray shielding properties for various natural stones from Vietnam to be applied in the radiation shielding applications. The physical and chemical properties affecting the γ-ray shielding performance were examined. The MH-300A density meter was utilized to measure the density (ρ, g/cm) of stone samples, as well as the chemical composition of Vietnamese
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Research on mechanism of gas leakage in microchannels of steel containment vessels for nuclear power plants Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-09 Min He, Yueyao Chen, Zhen Wu, Gangling Hou, Jialong Wang, Zhuangfei Li, Yuzhu Wang, Hanze Li
Steel containment vessels for nuclear power plants can experience gas leakage due to minute defects such as cracks, corrosion, and aging, leading to gas leakage. A gas leakage model for microchannels is established to elucidate the mechanism underlying gas leakage within microchannels caused by these defects, specifically addressing the issue of unidirectional gas flow. Computational Fluid Dynamics
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Invulnerability analysis of nuclear accidents emergency response organization network based on complex network Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-09 Wen Chen, Shuliang Zou, Changjun Qiu, Jianyong Dai, Meirong Zhang
Modern risk management philosophy emphasizes the invulnerability of human beings to cope with all kinds of emergencies. The Nuclear Accidents Emergency Response Organization (NAERO) of Nuclear Power Plant (NPP) is the primary body responsible for nuclear accidents emergency response. The invulnerability of the organization to disturbance or attack from internal and external sources is crucial in the
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Radiation tolerant capacitor-SRAM without area overhead Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-08 Eunju Jo, Hosang Yoon, Hongjoon Park, Woo-young Choi, Inyong Kwon
In memory semiconductors such as a static random access memory (SRAM), a common problem is soft errors under radiation environment. These soft errors cause bit flips, which are referred to as single event upsets (SEUs). Some radiation-hardened SRAM cells such as a Quatro SRAM, we-Quatro SRAM, and DICE SRAM cells have been reported for years. However, these designs have the disadvantage of taking up
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Experimental study of the influence of borehole parameters on prompt fission neutron uranium logging and its corrections Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-07 Pengfei Zhou, Bin Tang
In prompt fission neutron uranium logging, borehole environmental parameters affect the measured results and must be corrected. In order to explore the influence of borehole parameters on the interpretation of logging results, this paper builds a sandstone type uranium ore block model to simulate the field production drilling device based on the “Epithermal/Thermal neutron counting rate ratio” (E/T)
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Analysis of heat-loss mechanisms with various gases associated with the surface emissivity of a metal containment vessel in a water-cooled small modular reactor Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-07 Geon Hyeong Lee, Jae Hyung Park, Beomjin Jeong, Sung Joong Kim
In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major
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New polyester composites synthesized with additions of different sized ZnO to study their shielding efficiency Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-07 M. Elsafi, M.I. Sayyed, Aljawhara H. Almuqrin
This investigation developed a novel polyester composite based on the addition of zinc oxide (ZnO) of different sizes. We prepared nine samples Containing different percentages and sizes of ZnO as well as the control sample (Pol-ZnO0). The attenuation factors of Pol-micro ZnO were estimated using Phy-x software, while the HPGe detector and various gamma sources were used to experimentally measure the
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A comparative study of different radial basis function interpolation algorithms in the reconstruction and path planning of γ radiation fields Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-07 Yulong Zhang, Jinjia Cao, Biao Zhang, Xiaochang Zheng, Wei Chen
Accurate reconstruction of radiation field and path planning are very important for the safety of operators in the process of dismantling nuclear facilities. Based on radial basis function (RBF) interpolation algorithm, this paper discussed the application of inverse multiquadric radial basis Function (IMRBF) interpolation method to the reconstruction of gamma radiation field, and proved the feasibility
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Predicting the core thermal hydraulic parameters with a gated recurrent unit model based on the soft attention mechanism Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-06 Anni Zhang, Siqi Chun, Zhoukai Cheng, Pengcheng Zhao
Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters
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Characterization of small single photon avalanche diode fabricated using standard 180 nm CMOS process for digital SiPM Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-05 Jinseok Oh, Hakcheon Jeong, Min Sun Lee, Inyong Kwon
In this work, single photon avalanche diodes (SPADs) were fabricated using the standard 180 nm complementary metal-oxide semiconductor process. Their small size of 15–16 m and low operating voltage made it possible to easily integrate them with readout circuits for compact on-chip sensors, particularly those used in the radiation sensor network of a nuclear plant. Four architectures were proposed for
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Local mechanical properties of corrosion layers formed on T91 and SS316L steels after exposure to static liquid LBE at 500 °C for 1000 h obtained by nano-indentation Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-05 Zhikun Zhou, Juan Du, Chenwen Tian, Xuhao Peng, Yabo Wu, Xi Lv, Yixiong Zhang, Ziguang Chen
Static corrosion tests in LBE under oxygen-saturated and –depleted conditions at 500 °C for 1000 h were conducted to investigate the corrosion-induced diffusion layers on T91 and SS316L steels subject to dissolution and oxidation corrosion. Following nano-indentation experiments examined the variations of local mechanical properties within the corrosion layers. Under the present conditions, T91 steel
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Failure simulation of nuclear pressure vessel under LBLOCA scenarios Nucl. Eng. Technol. (IF 2.7) Pub Date : 2024-03-05 Eui-Kyun Park, Jun-Won Park, Yun-Jae Kim, Kukhee Lim, Eung-Soo Kim
This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as