-
Design and simulation of neutron radiography system for an aqueous homogeneous solution reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-22 B. Jandaghian, J. Mokhtari, M.H. Choopan Dastjerdi
AHR reactors are known for the production of radiopharmaceuticals. The ARGUS reactor is one of the most famous of these reactors, which has been designed originally for radioisotopes production. The ARGUS reactor produces a suitable neutron flux that can be used for other applications such as neutron radiography (NR). Here, a NR system is designed step by step using the Monte Carlo method. The length
-
High-energy radiation shielding characteristics of SeTeSnZn chalcogenide glasses (STSZ ChGs) Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-22 Vishnu Saraswat, A. Dahshan, Z. Khattari, Neeraj Mehta
The focus of the current research is to see the effectiveness of some quaternary ChGs as shielding materials for incoming photon energy in the range of 0.015–15.0 MeV. The ChGs addressed in this investigation belong to a quaternary glass system of SeTeSnZn (x = 0, 2, 4, 6). The online Phy-X/PSD program has been used to deduce several safety factors. The MAC values of the studied glass have the following
-
Developed concrete reinforced by waste of lead fishing weights and steel nails as gamma rays and neutrons shields for nuclear power reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-22 Ahmed Faisal Oan, Aly Saeed, R.M. El Shazly
Wastes of steel nails WSN and wastes of lead fishing weights WLFW were successfully recycled in concrete to enhance its ability to attenuate neutrons and gamma rays. A series of studies on the effect of partially replacing coarse aggregates of concrete with different percentages (5, 10, and 20%) of WSN or WLFW on the density, slump, compressive strength, and attenuation of neutrons and gamma rays were
-
Numerical simulation study of boiling Critical Heat Flux characteristics of graphene nanofluids Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-22 Yandong Hou, Jianwei Huang, Rui Cai, Wenyu Liu, Chao Zhang, Weichao Li, Chuntian Gao, Yan Xiang
IVR-ERVC (In-Vessel Retention – External Reactor Vessel Cooling) is a critical accident management method for ensuring the integrity of the reactor pressure vessel (RPV) lower head. One of the most crucial aspects within this method is to enhance the CHF (Critical Heat Flux) on the outer surface of the reactor pressure vessel (RPV) lower head. This paper explores the application of graphene nanofluids
-
Changes in chemical composition of TixAl1−xN coatings immersed in oxygen-saturated Lead–Bismuth Eutectic at low and moderate temperatures (250 °C ≤ T ≤ 410 °C) Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-21 Essam Serag, Emile Haye, Ben Caers, Paul Schuurmans, Stéphane Lucas
Lead–Bismuth Eutectic (LBE) will serve as a liquid metal coolant in accelerator-driven systems, posing a corrosive threat to exposed materials. To address this challenge, TiAlN coatings (0.38 ≤ x ≤ 0.58) were applied onto AISI 316 L austenitic stainless steel using the reactive bipolar magnetron sputtering technique. These coated samples underwent immersion in static oxygen-saturated LBE at temperatures
-
The potential of bi2–xZrxO3+x/2@ZrO2 (BZO) as a heavyweight radiation shielding material: Fabrication and characterization using a hydrothermal technique directly from zircon mineral Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-20 K.A. Mahmoud, Mohammad W. Marashdeh, Mohannad Al-Hmoud, Mamduh J. Aljaafreh, Sitah Alanazi, Islam G. Alhindawy
The hydrothermal technique was utilized to produce a new zirconia doped Bi ions to be suitable for gamma ray shielding applications. The monoclinic phase of zirconia in the manufactured samples is revealed by X-ray diffraction, and each sample's variation in the intensity and position of the diffraction peaks prominently represented the different percentage of BiZrO. Furthermore, the scanning electron
-
Thermodynamic performance of an innovative space nuclear Brayton cycle with S–N2O Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-20 Xinyu Miao, Haochun Zhang, Fangwei Ma, Tong Lu, Ersheng You
The supercritical Brayton cycle system is renowned for its characteristics of elevated high-power density, compact structure, lightweight design, and extended operational lifespan, rendering it as a preeminent energy conversion approach for space nuclear power systems. Using nitrous oxide (NO) as the working fluid in a supercritical Brayton cycle system enables heightened efficiencies at comparatively
-
Accident simulation and control strategy of feedwater system of sodium-cooled fast reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-17 Aodi Sun, Muping Li, Shuhan Yang, Peiwei Sun, Xinyu Wei
Improper operation leading to an emergency shutdown in a nuclear power plant can result in significant energy loss and have a substantial impact on the economy. This study focuses on a sodium-cooled fast reactor (SFR) nuclear power plant, analyzing system characteristics under different accidents through feedwater system simulation and improving control system performance. The results show that the
-
Supercharging performance of anti-cavitation schemes in pool-type reactor: Experimental verification and numerical investigation Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-16 Fubing Ma, Xiaobo Zeng, Zhiting Yue, Changqi Yan, Guangming Fan, Shuai Hao
-
Dynamic mode decomposition coupled with multilevel octree grid algorithm for large-scale discrete ordinates neutron transport calculations Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-16 Jiaxing Wang, Bin Zhang, Shouhai Yang, Yixue Chen
Large-scale discrete ordinates neutron transport calculations are known to be excessively slow due to the large number of grids that need to be solved and the limitations of the source iteration method employed. This paper introduces a novel algorithm called DMD-MLTG as a practical means for solving this problem. The DMD-MLTG algorithm utilizes the dynamic mode decomposition (DMD) method to generate
-
Sensitivity and uncertainty analyses for advanced nuclear systems (ALFRED, ASTRID, ESFR and MYRRHA) Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-13 S. Panizo, C. Alfonso, A. Jiménez-Carrascosa, N. García-Herranz, V. Bécares, P. Romojaro, F. Álvarez-Velarde, O. Cabellos, A. Cuesta-Matesanz, L. Fiorito, A. Stankovskiy, G. Van den Eynde
Within the frame of the EU H2020 program SANDA project, sensitivity and uncertainty analyses have been performed for the ESFR, ASTRID and ALFRED reactor concepts and the multi-purpose flexible irradiation facility MYRRHA. Relevant reactor parameters, namely the effective multiplication factor, the effective delayed neutron fraction, the Doppler reactivity coefficient, the void worth and the worth of
-
Towards establishment of an efficient approach for validation of PWR full core Monte Carlo simulations at hot zero power conditions Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-12 L. Berry, A. Vasiliev, M. Hursin, D. Rochman, M. Frankl, H. Ferroukhi
-
Development and adaptation of meta-heuristic optimization methods in nuclear fuel management of soluble boron-free system-integrated modular advanced reactor to effectively increase the operation cycle length Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-12 Amir Karimi Jafari, Farrokh Khoshahval
The optimal arrangement of fuel assemblies inside a nuclear reactor core is one of the most challenging topics in nuclear engineering. Three different powerful optimization methods (particle swarm optimization, genetic algorithm method, and dragonfly algorithm) are developed and applied on the System-integrated Modular Advanced ReacTor (SMART). A new program named “DYNAMIC-FMO” is written in MATLAB
-
Numerical approximation of a two-dimensional fractional neutron diffusion model describing dynamics of neutron flux in a nuclear reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-11 P. Roul, Vikas Rohil, Gilberto Espinosa-Paredes, K. Obaidurrahman
The goal of this work is to develop a high-order numerical technique to solve a two-dimensional fractional neutron diffusion (FND) model describing dynamical behavior of a lead-cooled fast reactor. This method is based on technique for temporal discretization and a high-order compact Alternating Direction Implicit (ADI) finite difference scheme for the spatial discretization. Numerical simulations
-
Accident scenario analysis and control scheme design for a micro Molten Salt Reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-11 D.Y. Cui, X.X. Li, Y. Dai, Y. Zou, J.G. Chen, X.Z. Cai
-
Analysis of the correlation for energy deposition in PWR coolant by coupled neutron-photon transport calculations Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-11 Christian Castagna, Daniele Tomatis, Shai Kinast, Erez Gilad
-
Simultaneous use of bismuth trioxide and mill scale for ternary blended geopolymer composite in radiation shielding applications Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-10 Rahul Sharma, Shaik Hussain, Naidu Seetala, John Matthews, Reed Edwards, Sudhir Amritphale, Elizabeth Matthews
The present study evaluates Mill scale which is a steel industry waste and bismuth trioxide simultaneously as a potential radiation shielding material in geopolymer composite. An innovative and first of its kind lead-free design has been developed for making radiation shielding materials using mill scale and bismuth trioxide as shielding aggregates and industrial wastes such as fly ash and blast furnace
-
Quantitative analysis of contributing factors to the resilience of emergency response organizations in nuclear power plants Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-10 Jaehyun Kim, Sungheon Lee, Awwal Mohammed Arigi, Jonghyun Kim
In the nuclear field, emergency response organizations are established upon the occurrence of events that have the potential to release radiation. These emergency response organizations consist of many sub-organizations such as the regulatory body, local government, hospitals, fire stations, and organizations of the nuclear utility such as an emergency operations facility and technical support center
-
AI-driven non-intrusive uncertainty quantification of advanced nuclear fuels for digital twin-enabling technology Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-10 Kazuma Kobayashi, Dinesh Kumar, Syed Bahauddin Alam
In response to the urgent need to establish AI/ML-integrated Digital Twin (DT) technology within next-generation nuclear systems, advancements in modeling methods and simulation codes are necessary. The increased complexity of models demands significant computational resources to quantify their uncertainties. To address this challenge, a data-driven non-intrusive uncertainty quantification method via
-
Experiment study on the flow mechanism of the molten pool relocation using simulant materials at room temperature Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-09 Xinhai Zhao, Yicong Lan, Peng Chen, Yapei Zhang, Haoli Wang, Chao Guo, Simin Luo, Wenxi Tian, G.H. Su, Suizheng Qiu
-
A simulation study of the ability to detect power distribution perturbations in the texas A&M TRIGA reactor with self-powered neutron detectors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-06 Anthony Birri, Jonathan T. Gates, Daniel C. Sweeney, Kathleen C. Goetz, N. Dianne Bull Ezell
Given the variety of ways that nuclear reactor core power may be perturbed, reactor operators and developers are keen on understanding the accuracy and convergence time during which perturbations in reactor power distribution may be synthesized (i.e., inferred) from an array of in-core radiation detectors. A simulation study was conducted as described herein using a highly detailed model of the Texas
-
Experimental study of IVR-ERVC CHF limits for CAP1400 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-06 Jinquan Yan, Mingguang Zheng, Kemei Cao, Ning Guo, Kun Zhang, Jiayun Wang, Wei Lu
In-Vessel Retention (IVR) is designed as one of the key mitigation strategies for Chinese Advanced Passive pressurized water reactor (CAP1400), to keep melt corium retaining in reactor pressure vessel (RPV) under severe accidents situations. External Reactor Vessel Cooling (ERVC) is selected as the main measure to achieve the purpose of IVR. The key safety evaluation criteria for the effectiveness
-
Theoretical study on pressure oscillation dominant frequency of bubbles submerged jet under rolling condition Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-05 Pengbo Wei, Zhenghang Luo, Shilin Song, Weixiong Chen, Quanbin Zhao, Daotong Chong
Steam direct contact condensation is a highly efficient heat and mass transfer phenomenon, usually accompanied by strong mass, momentum and energy exchanges at the water-steam interface, which can rapidly realize cooling and pressure relief, but also accompanied by pressure oscillations. When the relevant equipment is used in offshore nuclear power plants and naval non-energetic systems, the additional
-
Ultrasound-assisted removal of contaminants on stainless steel surfaces using nitrogen ultrafine bubble water Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-05 Masaumi Nakahara, Sou Watanabe, Shuya Kimura, Misa Sasaki, Hiromitsu Inagaki, Tetsuji Moriguchi
In this study, a novel ultrasound-assisted removal/decontamination process using ultrafine bubbles (UFBs) of 1 μm or less diameter gas bubbles is proposed to reduce secondary radioactive waste for decommissioning of nuclear installations. The removal/decontamination technique does not require abrasive media for abrasive-blasting decontamination or chemical reagents for chemical decontamination. The
-
Corrosion behaviors of Inconel 617 and Incoloy 800H in impure helium with different CO contents at high temperatures Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-05 Wei Zheng, Haoxiang Li, Bin Du, Huang Zhang, Huaqiang Yin, Xuedong He, Tao Ma
The helium coolant of the very high-temperature gas-cooled reactor (VHTR) is expected to contain trace impurities, which will cause corrosion to superalloys at high temperatures. The high-temperature corrosion behaviors of Inconel 617 and Incoloy 800H were investigated in the impure helium with various CO levels. For Inconel 617, the microclimate reaction occurred at high temperatures, releasing large
-
Development of Modelica-based one-dimensional thermodynamic cycle library and its application in simulation and multi-objective optimization of a He–Xe closed-Brayton-cycle system Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-04 Ao Zhang, Xiang Wang
-
An adaptive navigation model for path finding in radioactive environment Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-03 Mengkun Li, Jiamei Tang, Li Liu, Chao Dong, Yan Li, Ting Wang
-
An integrated assessment method of real-time source term for high temperature gas-cooled reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-01 Ao Liu, Tao Liu, Jingang Liang, Liguo Zhang, Jiejuan Tong
To address high temperature gas-cooled reactor (HTR) nuclear power plant (NPP) safety analysis and emergency response, this paper proposes an integrated source term assessment (ISTA) method, guided by the principles of reality, rapidity and rationality. The fundamental concept of ISTA is to establish a connection between the theoretical model of source term (ST) calculation and actual conditions of
-
Performance analysis of indirect contact heat pipe radiator for space nuclear power system Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-01 Zengen Li, Haochun Zhang, Shuting Zhao, Cheng Zhang, Yan Xia
The heat pipe radiator, with excellent thermal conductivity and anti-collision performance, is expected very suitable for heat emission in nuclear-power spacecraft. Indirect contact heat pipe radiator could operate longer life with lower capital cost. In this paper, the fluid-solid coupling heat transfer method is integrated to solve the energy model and S–S radiation model based on the k-epsilon model
-
Modeling gamma detectors in OpenMC: Validation of a newly implemented pulse-height tally Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-01 Christopher Fichtlscherer, Milon Miah, Friederike Frieß, Malte Göttsche, Moritz Kütt
Gamma spectroscopy measurements can be simulated using a pulse-height tally functionality of Monte Carlo particle transport codes. Such a functionality must account for the complete simulation history of a particle’s energy deposited in a particular volume. OpenMC, an open-source application for neutron and photon particle transport which is widely used in the nuclear engineering community, previously
-
Numerical modelling of catalytic hydrogen combustion in passive autocatalytic recombiners: A review Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-04-01 Alexander A. Malakhov, Maria H. du Toit, Alexander V. Avdeenkov, Dmitri G. Bessarabov
-
Thermal and hydraulic characteristics analysis of horizontal lead-bismuth reactor assembly based on sub-channel analysis code Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-30 Yixin Zhang, Chenglong Wang, Wenxi Tian, Suizheng Qiu, G.H. Su, Lin Chao
In this paper, the self-developed subchannel analysis program SACOS-PB was used to analyze the thermal hydraulic characteristics of 127 rods in horizontal lead-bismuth reactor under steady state and typical ocean transient conditions. The results of the steady-state analysis showed that the coolant flow and temperature of the vertical assembly were symmetrically distributed, the coolant flow of the
-
CHF and heat transfer enhancement by SiO2 nanofluids on a inclined downward facing heating surface Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-30 Saikun Wang, Huai-En Hsieh, Zhibo Zhang, Shiqi Wang, Xintian Cai
In this experiment, the factors affecting critical heat flux (CHF) in the pool boiling research with the heating face down were investigated. Including the type of cooling liquid and the angle of the heating surface. The types of coolants include reverse osmosis (R·O) water and SiO nanofluids (0.02 g/L~0.12 g/L). The morphology and types of nanoparticles were characterized by microscope (SEM) and scanning
-
Effect of silver chloride deposition technique on modified bentonite operating properties for radioactive iodide-ions localization in geological disposal facility for radioactive waste Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-29 Ekaterina A. Tyupina, Artem V. Pryadko, Olga M. Klimenko
This work discusses the influence of the technique of silver chloride deposition (via one-stage or two-stage method) onto performance attributes of modified bentonite for iodine ions localization in radioactive waste repositories. It has been established that bentonite modified via both techniques in wide range of AgCl content has high sorption properties (more than 95%) towards iodide anions, the
-
Preliminary analysis of in-orbit operation accidents for an ultra-small Lithium-cooled space reactor power system Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-27 Li Ge, Huaqi Li, Jianqiang Shan
To study the safety characteristics of an ultra-small (5.2 MWt) lithium-cooled space reactor power system (ULCR) under postulated accidents, a transient analysis code for the ultra-small space reactor coupled with a closed Brayton cycle dynamic conversion and heat pipe radiator was established. A preliminary analysis of four typical postulated accidents occurring at 100 percent power in-orbit operation
-
A concurrent fault diagnosis method for electric isolation valves in nuclear power plants based on rule-based reasoning and data-driven methods Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-27 Xin Ai, Yongkuo Liu, Longfei Shan, Chunli Xie, Hongkuan Zhou
In order to improve the fault diagnosis ability of electric isolation valves in nuclear power plants, especially for the diagnosis of concurrent faults, a novel concurrent fault diagnosis method based on rule reasoning and data-driven methods is proposed in this study. The concurrent fault diagnosis method is divided into two layers: the rule-based reasoning layer and data-driven layer. In order to
-
Equilibrium core modeling of a pebble bed reactor similar to the Xe-100 with SCALE Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-27 Annie Berens, Friederike Bostelmann, Nicholas R. Brown
As the nuclear industry moves towards licensing and constructing advanced reactors, new attention has been focused on the advanced reactor designs that have past operational experience, such as pebble-bed high-temperature gas-cooled reactors (PB-HTGRs). Pebble-bed reactor designs have many advantages, such as their higher operating temperatures and online refueling capabilities. However, high-fidelity
-
A review of CFD studies on thermal hydraulic analysis of coolant flow through fuel rod bundles in nuclear reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-27 Mingjun Wang, Haoran Ju, Jian Wu, Hanrui Qiu, Kai Liu, Wenxi Tian, G.H. Su
Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD)
-
Impact of fission product (Ce, Sn, Sr, Se) oxides on UO2 oxidation to U3O8 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-26 Wei Han, Jie Gao, Mei Li, Meng Zhang, Rugeng Liu
In order to investigate the impact of fission products on the oxidation of spent fuel in the cladding separation process, the oxidation behavior and kinetic mechanism of UO and impact of M (M = Ce, Sn, Sr, Se) oxides on the oxidation of UO to UO were investigated by TG/DSC, XRD, FT-IR and Raman. It was found that the oxidation of UO to UO went through two stages: UO→UO; UO→UO. Doping M oxide in UO
-
Enhancing cladding mechanical modelling during DBA/LOCA accidents with FRAPTRAN: The TUmech one-dimensional model Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-26 Pau Aragón, Francisco Feria, Luis E. Herranz, Arndt Schubert, Paul Van Uffelen
The analysis of the cladding mechanical performance is essential for predicting rod failure during the heat-up phase of a loss-of-coolant accident (LOCA). Previous modelling and simulation exercises have shown that there is room for improvement in the accuracy of fuel performance simulations under such conditions. In this work, TUmech, a simplified standalone version of the mechanical model in TRANSURANUS
-
An intelligent fault detection and diagnosis monitoring system for reactor operational resilience: Unknown fault detection Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-26 Mario Mendoza, Pavel V. Tsvetkov
-
Source term inversion coupling Kernel Principal Component Analysis, Whale Optimization Algorithm, and Backpropagation Neural Networks (KPCA-WOA-BPNN) for complex dispersion scenarios Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-22 Xinpeng Li, Jiayue Song, Li Yang, Huanting Li, Sheng Fang
Accurate and rapid source term estimation is critical for consequence assessment and emergency decision-making in nuclear accidents. Neural network methods provide a promising approach to achieving this goal, but they are mainly demonstrated in ideal scenarios with flat terrain. In this study, a source term inversion model combining the Kernel Principal Component Analysis, the Whale Optimization Algorithm
-
Conceptual design and feasibility analysis of a heat pipe cooled cermet fuel space nuclear reactor for mars surface application Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-22 Yuhan Fan, Rui Yan, Shihe Yu, Liang Chen, Yang Zou
The compact, safe, light-weight and highly reliable Space Nuclear Reactor (SNR) with heat pipe cooling is an excellent candidate for the potential needs of Mars surface applications. This work proposed a heat pipe reactor concept design which can provide 1 MWth power for more than 8 years. Reactivity control is accomplished by 6 rotating control drums. Cermet fuel with dispersion granules characteristics
-
Comprehensive study on structural, elastic and radiation shielding abilities of novel quaternary Bi2O3–TeO2–Li2O–Al2O3 glasses Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-21 Aishah Zarzali Shah, Mohd Hafiz Mohd Zaid, Khamirul Amin Matori, Yazid Yaakob, Abdul Rahman Sarmani, Rosdiyana Hisam
Novel quaternary glasses with varying compositions of BiO-(75–)TeO–20LiO–5AlO were synthesized using a melt-quenching method to analyze their physical, elastic properties and to evaluate their potential as radiation shielding materials. The X-Ray Diffraction (XRD), Fourier Transform Infrared (FTIR) and ultrasonic measurement were utilized to study those properties. The radiation shielding properties
-
Causal Model Framework for Nuclear Power Plant Licensing Process Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-21 Lauren Kiser PhD, Luis Daniel Otero PhD
Interests in clean energy revived the nuclear power industry. For the first time in decades, innovative technologies and plant designs are being considered by regulatory agencies. This paper explores a Bayesian Network and AHP approach to causal modeling of the Combined License review process for new nuclear power plants (NPP). Historically lengthy and expensive, NPP licensing is critical to ensuring
-
Column percolation leaching test for rare-earth processed residue: Standard laboratory column elution experiments towards development permanent disposal facility (PDF) Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-20 Eli Syafiqah Aziman, Aznan Fazli Ismail, Muhammad Abdullah Rahmat, Nursyamimi Diyana Rodzi, Muhammad Ariff Baihaqi Jamaludin
-
Hierarchical FFT-LSTM-GCN based model for nuclear power plant fault diagnosis considering spatio-temporal features fusion Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-19 Yushun Wang, Jingquan Liu, Gensheng Qian
As safety-critical infrastructure, nuclear power plants (NPPs) require enhanced safety measures and risk minimization. To achieve this goal and to aid operator decision-making while reducing human error, various fault diagnosis (FD) methods have been proposed. Among these, deep learning-based approaches have demonstrated significant success in FD because of their ability to effectively extract information
-
A review of multiphysics tools and methods to evaluate high temperature pebble bed reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-19 Edward M. Duchnowski, Nicholas R. Brown
The Generation IV high temperature pebble bed reactor has garnered interest over the years as a viable solution for reducing overall carbon emissions. The potential for enhanced safety, increased efficiency, and reduced overall cost makes the high temperature pebble bed reactor design a top candidate for commercialization. However, detailed and accurate computational simulation of the concept is required
-
A new solution of the fractional neutron point kinetics equations using symmetry and the Heaviside's expansion formula Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-18 Carlos-Antonio Cruz-López, Gilberto Espinosa-Paredes
A new analytical solution of the Fractional Neutron Point Kinetics Equations is developed in the present work, considering multi-groups of precursors of delayed neutrons as well as a constant reactivity. Such solution is obtained in a short way, as a direct generalization of the integer case, being only necessary to use an elementary change of variable as well as the Heaviside's Expansion Formula.
-
Gamma-ray shielding effectiveness, thermal, and dielectric properties of filler-reinforced high-density polyethylene/boron carbide composites Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-16 A.M. Reda, M.A. Alsawah, M. Hosni, R.M. Ahmed
The present research prepared samples of pure high-density polyethylene (HDPE)/boron carbide reinforced with iron dioxide, aluminum dioxide, iron, and aluminum. The effectiveness of the prepared samples as a shielding material against gamma-rays and their thermal and dielectric properties have been evaluated. The shielding characteristics of the prepared composites were investigated theoretically using
-
An experimental study on critical heat flux for low flow and low pressure of water in a hexagonal 7-rod bundle channel Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-15 Sen Chen, Leitai Shi, Huaqi Li, Da Li, Lixin Chen, Xiaofei Luo, Lei Zhu, Weitong Li, Xiaoyan Tian
This research explored novel experimental research on critical heat flux (CHF) in a hexagonal 7-rod bundle channel at low flow and low-pressure conditions. The heater rod in the experiment was composed of Inconel-625 alloy and the heated length was 390 mm, and the diameter was 37.2 mm. The experimental parameters of inlet temperature, outlet pressure as well as mass flux ranged from 39.7–71.7 °C, 130–197 kPa
-
Performance analysis of integrated Nuclear-Solar Energy system sharing same molten salt thermal energy storage Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-14 G.J. Soto, M.J. Wagner, T.W. Neises, B.A. Lindley
Advanced nuclear reactors may be deployed with integrated thermal energy storage to improve flexibility and maximize revenue. This presents opportunities for thermal integration with concentrating solar power (CSP) to generate component synergies and/or improve performance. In this study, a computational model is developed for an integrated nuclear and CSP system that both share the same molten salt
-
Numerical simulation study on the filtration performance of metal fiber filters Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Song Ma, Yanmin Zhou, Zhongning Sun, Haifeng Gu, Yin Wang, Xiang Yu
-
Lithium stabilization of amorphous ZrO2 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Gareth F. Stephens, Jack A. Wilson, Simon F. Curling, Guanze He, P. John Thomas, David W. Williams, Susan Ortner, Chris Grovenor, Michael J.D. Rushton, Aidan Cole- Baker, Simon C. Middleburgh
Small modular reactors (SMRs) are a key option to aid the worldwide net zero targets for carbon emissions. Some pressurised water reactors aim to operate with a boron-free coolant chemistry for simplification in plant design. In the absence of boron, Li has been found to accelerate the corrosion of the zirconium-based alloy fuel cladding under certain conditions and concentrations within pressurised
-
A study with PUREX aqueous-organic pair in Taylor-Couette mixing field Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Shekhar Kumar
-
Assessment of thermal shock resistance of refractory magnesia lining under simulated core melt impingement for application to SFR core catcher Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Prabhat Kumar Shukla, Hemanth Rao E, Anish Kumar, Sanjay Kumar Das, Ponraju D, Venkatraman B
Despite a hypothetical event, whole core melt accident in a sodium cooled fast reactor is investigated as a part of plant safety analysis. To prevent breach of primary boundary in such accidents, a Core Catcher (CC) is provided at the bottom of the main vessel for retention of corium in subcritical and coolable state. For future Indian FBRs, refractory magnesia-lined CC is planned for protection against
-
Review of thorium-containing fuels in LWRs Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Maria Hendrina Du Toit, Frederik Van Niekerk, Shervineh Amirkhosravi
Stemming from a renewed interest in the suitability of thorium as fertile isotope, this study presents an extensive survey of some of the recent advances on the use of thorium in light water reactors, with the view of informing researchers and policy makers. The isotopic properties of fertile and fissile material are discussed, with an emphasis on the suitability and unique characteristics of thorium
-
Utilizing even Plutonium Isotopes as burnable absorbers for controlling the reactivity and power distribution in Pressurized Water Reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Sayed Saeed Mustafa, Esmat A. Amin
In this paper, a new approach is proposed for controlling the reactivity and power distribution in pressurized water reactor cores without the implementation of burnable absorbers. This approach depended on the homogeneous mixing of uranium oxide with even plutonium isotopes; Pu-238, Pu-240, and Pu-242. Three Plutonium fuels; (99% UO + 1% Pu-238), (99% UO + 1% Pu-240), and (98% UO + 0.5% Pu-238 + 1%
-
Applying U.S. metal fuel experience to new fuel designs for fast reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-11 Douglas C. Crawford, Douglas L. Porter
-
Experimental study on fluid-structure interaction of multiple support barrels in liquid metal fast reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-10 Yu Liu, Dexuan Duan, Chao Deng, Yuxuan Zhu, Yuchao Wang, Yang Zhang, Daogang Lu
In an earthquake, the main equipment support barrels in the pool-type liquid metal fast reactor (LMFR) may experience strong vibrations due to the fluid-structure interaction (FSI) phenomenon. Accurate parameters for FSI characteristics of multiple cylinders are crucial for the seismic design and analysis of these support barrels. Limited experimental data is available for pressure characteristics