Publisher: Elsevier   (Total: 3161 journals)

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Showing 1 - 200 of 3161 Journals sorted alphabetically
Academic Pediatrics     Hybrid Journal   (Followers: 39, SJR: 1.655, CiteScore: 2)
Academic Radiology     Hybrid Journal   (Followers: 26, SJR: 1.015, CiteScore: 2)
Accident Analysis & Prevention     Partially Free   (Followers: 106, SJR: 1.462, CiteScore: 3)
Accounting Forum     Hybrid Journal   (Followers: 28, SJR: 0.932, CiteScore: 2)
Accounting, Organizations and Society     Hybrid Journal   (Followers: 43, SJR: 1.771, CiteScore: 3)
Achievements in the Life Sciences     Open Access   (Followers: 7)
Acta Anaesthesiologica Taiwanica     Open Access   (Followers: 6)
Acta Astronautica     Hybrid Journal   (Followers: 446, SJR: 0.758, CiteScore: 2)
Acta Automatica Sinica     Full-text available via subscription   (Followers: 2)
Acta Biomaterialia     Hybrid Journal   (Followers: 30, SJR: 1.967, CiteScore: 7)
Acta Colombiana de Cuidado Intensivo     Full-text available via subscription   (Followers: 3)
Acta de Investigación Psicológica     Open Access   (Followers: 2)
Acta Ecologica Sinica     Open Access   (Followers: 11, SJR: 0.18, CiteScore: 1)
Acta Histochemica     Hybrid Journal   (Followers: 5, SJR: 0.661, CiteScore: 2)
Acta Materialia     Hybrid Journal   (Followers: 322, SJR: 3.263, CiteScore: 6)
Acta Mathematica Scientia     Full-text available via subscription   (Followers: 5, SJR: 0.504, CiteScore: 1)
Acta Mechanica Solida Sinica     Full-text available via subscription   (Followers: 9, SJR: 0.542, CiteScore: 1)
Acta Oecologica     Hybrid Journal   (Followers: 12, SJR: 0.834, CiteScore: 2)
Acta Otorrinolaringologica (English Edition)     Full-text available via subscription  
Acta Otorrinolaringológica Española     Full-text available via subscription   (Followers: 2, SJR: 0.307, CiteScore: 0)
Acta Pharmaceutica Sinica B     Open Access   (Followers: 2, SJR: 1.793, CiteScore: 6)
Acta Psychologica     Hybrid Journal   (Followers: 26, SJR: 1.331, CiteScore: 2)
Acta Sociológica     Open Access   (Followers: 1)
Acta Tropica     Hybrid Journal   (Followers: 6, SJR: 1.052, CiteScore: 2)
Acta Urológica Portuguesa     Open Access   (Followers: 1)
Actas Dermo-Sifiliograficas     Full-text available via subscription   (Followers: 3, SJR: 0.374, CiteScore: 1)
Actas Dermo-Sifiliográficas (English Edition)     Full-text available via subscription   (Followers: 2)
Actas Urológicas Españolas     Full-text available via subscription   (Followers: 3, SJR: 0.344, CiteScore: 1)
Actas Urológicas Españolas (English Edition)     Full-text available via subscription   (Followers: 1)
Actualites Pharmaceutiques     Full-text available via subscription   (Followers: 7, SJR: 0.19, CiteScore: 0)
Actualites Pharmaceutiques Hospitalieres     Full-text available via subscription   (Followers: 3)
Acupuncture and Related Therapies     Hybrid Journal   (Followers: 8)
Acute Pain     Full-text available via subscription   (Followers: 15, SJR: 2.671, CiteScore: 5)
Ad Hoc Networks     Hybrid Journal   (Followers: 11, SJR: 0.53, CiteScore: 4)
Addictive Behaviors     Hybrid Journal   (Followers: 18, SJR: 1.29, CiteScore: 3)
Addictive Behaviors Reports     Open Access   (Followers: 9, SJR: 0.755, CiteScore: 2)
Additive Manufacturing     Hybrid Journal   (Followers: 13, SJR: 2.611, CiteScore: 8)
Additives for Polymers     Full-text available via subscription   (Followers: 22)
Advanced Drug Delivery Reviews     Hybrid Journal   (Followers: 188, SJR: 4.09, CiteScore: 13)
Advanced Engineering Informatics     Hybrid Journal   (Followers: 13, SJR: 1.167, CiteScore: 4)
Advanced Powder Technology     Hybrid Journal   (Followers: 17, SJR: 0.694, CiteScore: 3)
Advances in Accounting     Hybrid Journal   (Followers: 9, SJR: 0.277, CiteScore: 1)
Advances in Agronomy     Full-text available via subscription   (Followers: 17, SJR: 2.384, CiteScore: 5)
Advances in Anesthesia     Full-text available via subscription   (Followers: 30, SJR: 0.126, CiteScore: 0)
Advances in Antiviral Drug Design     Full-text available via subscription   (Followers: 2)
Advances in Applied Mathematics     Full-text available via subscription   (Followers: 12, SJR: 0.992, CiteScore: 1)
Advances in Applied Mechanics     Full-text available via subscription   (Followers: 12, SJR: 1.551, CiteScore: 4)
Advances in Applied Microbiology     Full-text available via subscription   (Followers: 24, SJR: 2.089, CiteScore: 5)
Advances In Atomic, Molecular, and Optical Physics     Full-text available via subscription   (Followers: 15, SJR: 0.572, CiteScore: 2)
Advances in Biological Regulation     Hybrid Journal   (Followers: 4, SJR: 2.61, CiteScore: 7)
Advances in Botanical Research     Full-text available via subscription   (Followers: 1, SJR: 0.686, CiteScore: 2)
Advances in Cancer Research     Full-text available via subscription   (Followers: 35, SJR: 3.043, CiteScore: 6)
Advances in Carbohydrate Chemistry and Biochemistry     Full-text available via subscription   (Followers: 9, SJR: 1.453, CiteScore: 2)
Advances in Catalysis     Full-text available via subscription   (Followers: 5, SJR: 1.992, CiteScore: 5)
Advances in Cell Aging and Gerontology     Full-text available via subscription   (Followers: 5)
Advances in Cellular and Molecular Biology of Membranes and Organelles     Full-text available via subscription   (Followers: 14)
Advances in Chemical Engineering     Full-text available via subscription   (Followers: 29, SJR: 0.156, CiteScore: 1)
Advances in Child Development and Behavior     Full-text available via subscription   (Followers: 11, SJR: 0.713, CiteScore: 1)
Advances in Chronic Kidney Disease     Full-text available via subscription   (Followers: 11, SJR: 1.316, CiteScore: 2)
Advances in Clinical Chemistry     Full-text available via subscription   (Followers: 26, SJR: 1.562, CiteScore: 3)
Advances in Colloid and Interface Science     Full-text available via subscription   (Followers: 21, SJR: 1.977, CiteScore: 8)
Advances in Computers     Full-text available via subscription   (Followers: 14, SJR: 0.205, CiteScore: 1)
Advances in Dermatology     Full-text available via subscription   (Followers: 16)
Advances in Developmental Biology     Full-text available via subscription   (Followers: 14)
Advances in Digestive Medicine     Open Access   (Followers: 13)
Advances in DNA Sequence-Specific Agents     Full-text available via subscription   (Followers: 7)
Advances in Drug Research     Full-text available via subscription   (Followers: 26)
Advances in Ecological Research     Full-text available via subscription   (Followers: 45, SJR: 2.524, CiteScore: 4)
Advances in Engineering Software     Hybrid Journal   (Followers: 30, SJR: 1.159, CiteScore: 4)
Advances in Experimental Biology     Full-text available via subscription   (Followers: 9)
Advances in Experimental Social Psychology     Full-text available via subscription   (Followers: 52, SJR: 5.39, CiteScore: 8)
Advances in Exploration Geophysics     Full-text available via subscription   (Followers: 2)
Advances in Fluorine Science     Full-text available via subscription   (Followers: 9)
Advances in Food and Nutrition Research     Full-text available via subscription   (Followers: 68, SJR: 0.591, CiteScore: 2)
Advances in Fuel Cells     Full-text available via subscription   (Followers: 17)
Advances in Genetics     Full-text available via subscription   (Followers: 21, SJR: 1.354, CiteScore: 4)
Advances in Genome Biology     Full-text available via subscription   (Followers: 12, SJR: 12.74, CiteScore: 13)
Advances in Geophysics     Full-text available via subscription   (Followers: 8, SJR: 1.193, CiteScore: 3)
Advances in Heat Transfer     Full-text available via subscription   (Followers: 26, SJR: 0.368, CiteScore: 1)
Advances in Heterocyclic Chemistry     Full-text available via subscription   (Followers: 11, SJR: 0.749, CiteScore: 3)
Advances in Human Factors/Ergonomics     Full-text available via subscription   (Followers: 26)
Advances in Imaging and Electron Physics     Full-text available via subscription   (Followers: 3, SJR: 0.193, CiteScore: 0)
Advances in Immunology     Full-text available via subscription   (Followers: 37, SJR: 4.433, CiteScore: 6)
Advances in Inorganic Chemistry     Full-text available via subscription   (Followers: 10, SJR: 1.163, CiteScore: 2)
Advances in Insect Physiology     Full-text available via subscription   (Followers: 2, SJR: 1.938, CiteScore: 3)
Advances in Integrative Medicine     Hybrid Journal   (Followers: 6, SJR: 0.176, CiteScore: 0)
Advances in Intl. Accounting     Full-text available via subscription   (Followers: 3)
Advances in Life Course Research     Hybrid Journal   (Followers: 9, SJR: 0.682, CiteScore: 2)
Advances in Lipobiology     Full-text available via subscription   (Followers: 1)
Advances in Magnetic and Optical Resonance     Full-text available via subscription   (Followers: 8)
Advances in Marine Biology     Full-text available via subscription   (Followers: 21, SJR: 0.88, CiteScore: 2)
Advances in Mathematics     Full-text available via subscription   (Followers: 17, SJR: 3.027, CiteScore: 2)
Advances in Medical Sciences     Hybrid Journal   (Followers: 9, SJR: 0.694, CiteScore: 2)
Advances in Medicinal Chemistry     Full-text available via subscription   (Followers: 6)
Advances in Microbial Physiology     Full-text available via subscription   (Followers: 5, SJR: 1.158, CiteScore: 3)
Advances in Molecular and Cell Biology     Full-text available via subscription   (Followers: 26)
Advances in Molecular and Cellular Endocrinology     Full-text available via subscription   (Followers: 8)
Advances in Molecular Toxicology     Full-text available via subscription   (Followers: 7, SJR: 0.182, CiteScore: 0)
Advances in Nanoporous Materials     Full-text available via subscription   (Followers: 5)
Advances in Oncobiology     Full-text available via subscription   (Followers: 2)
Advances in Organ Biology     Full-text available via subscription   (Followers: 2)
Advances in Organometallic Chemistry     Full-text available via subscription   (Followers: 18, SJR: 1.875, CiteScore: 4)
Advances in Parallel Computing     Full-text available via subscription   (Followers: 7, SJR: 0.174, CiteScore: 0)
Advances in Parasitology     Full-text available via subscription   (Followers: 6, SJR: 1.579, CiteScore: 4)
Advances in Pediatrics     Full-text available via subscription   (Followers: 27, SJR: 0.461, CiteScore: 1)
Advances in Pharmaceutical Sciences     Full-text available via subscription   (Followers: 19)
Advances in Pharmacology     Full-text available via subscription   (Followers: 17, SJR: 1.536, CiteScore: 3)
Advances in Physical Organic Chemistry     Full-text available via subscription   (Followers: 10, SJR: 0.574, CiteScore: 1)
Advances in Phytomedicine     Full-text available via subscription  
Advances in Planar Lipid Bilayers and Liposomes     Full-text available via subscription   (Followers: 3, SJR: 0.109, CiteScore: 1)
Advances in Plant Biochemistry and Molecular Biology     Full-text available via subscription   (Followers: 11)
Advances in Plant Pathology     Full-text available via subscription   (Followers: 6)
Advances in Porous Media     Full-text available via subscription   (Followers: 5)
Advances in Protein Chemistry     Full-text available via subscription   (Followers: 19)
Advances in Protein Chemistry and Structural Biology     Full-text available via subscription   (Followers: 20, SJR: 0.791, CiteScore: 2)
Advances in Psychology     Full-text available via subscription   (Followers: 69)
Advances in Quantum Chemistry     Full-text available via subscription   (Followers: 7, SJR: 0.371, CiteScore: 1)
Advances in Radiation Oncology     Open Access   (Followers: 3, SJR: 0.263, CiteScore: 1)
Advances in Small Animal Medicine and Surgery     Hybrid Journal   (Followers: 3, SJR: 0.101, CiteScore: 0)
Advances in Space Biology and Medicine     Full-text available via subscription   (Followers: 7)
Advances in Space Research     Full-text available via subscription   (Followers: 430, SJR: 0.569, CiteScore: 2)
Advances in Structural Biology     Full-text available via subscription   (Followers: 6)
Advances in Surgery     Full-text available via subscription   (Followers: 13, SJR: 0.555, CiteScore: 2)
Advances in the Study of Behavior     Full-text available via subscription   (Followers: 37, SJR: 2.208, CiteScore: 4)
Advances in Veterinary Medicine     Full-text available via subscription   (Followers: 20)
Advances in Veterinary Science and Comparative Medicine     Full-text available via subscription   (Followers: 15)
Advances in Virus Research     Full-text available via subscription   (Followers: 6, SJR: 2.262, CiteScore: 5)
Advances in Water Resources     Hybrid Journal   (Followers: 56, SJR: 1.551, CiteScore: 3)
Aeolian Research     Hybrid Journal   (Followers: 6, SJR: 1.117, CiteScore: 3)
Aerospace Science and Technology     Hybrid Journal   (Followers: 394, SJR: 0.796, CiteScore: 3)
AEU - Intl. J. of Electronics and Communications     Hybrid Journal   (Followers: 8, SJR: 0.42, CiteScore: 2)
African J. of Emergency Medicine     Open Access   (Followers: 6, SJR: 0.296, CiteScore: 0)
Ageing Research Reviews     Hybrid Journal   (Followers: 12, SJR: 3.671, CiteScore: 9)
Aggression and Violent Behavior     Hybrid Journal   (Followers: 487, SJR: 1.238, CiteScore: 3)
Agri Gene     Hybrid Journal   (Followers: 1, SJR: 0.13, CiteScore: 0)
Agricultural and Forest Meteorology     Hybrid Journal   (Followers: 18, SJR: 1.818, CiteScore: 5)
Agricultural Systems     Hybrid Journal   (Followers: 32, SJR: 1.156, CiteScore: 4)
Agricultural Water Management     Hybrid Journal   (Followers: 46, SJR: 1.272, CiteScore: 3)
Agriculture and Agricultural Science Procedia     Open Access   (Followers: 4)
Agriculture and Natural Resources     Open Access   (Followers: 3)
Agriculture, Ecosystems & Environment     Hybrid Journal   (Followers: 58, SJR: 1.747, CiteScore: 4)
Ain Shams Engineering J.     Open Access   (Followers: 5, SJR: 0.589, CiteScore: 3)
Air Medical J.     Hybrid Journal   (Followers: 8, SJR: 0.26, CiteScore: 0)
AKCE Intl. J. of Graphs and Combinatorics     Open Access   (SJR: 0.19, CiteScore: 0)
Alcohol     Hybrid Journal   (Followers: 12, SJR: 1.153, CiteScore: 3)
Alcoholism and Drug Addiction     Open Access   (Followers: 12)
Alergologia Polska : Polish J. of Allergology     Full-text available via subscription   (Followers: 1)
Alexandria Engineering J.     Open Access   (Followers: 2, SJR: 0.604, CiteScore: 3)
Alexandria J. of Medicine     Open Access   (Followers: 1, SJR: 0.191, CiteScore: 1)
Algal Research     Partially Free   (Followers: 11, SJR: 1.142, CiteScore: 4)
Alkaloids: Chemical and Biological Perspectives     Full-text available via subscription   (Followers: 2)
Allergologia et Immunopathologia     Full-text available via subscription   (Followers: 1, SJR: 0.504, CiteScore: 1)
Allergology Intl.     Open Access   (Followers: 5, SJR: 1.148, CiteScore: 2)
Alpha Omegan     Full-text available via subscription   (SJR: 3.521, CiteScore: 6)
ALTER - European J. of Disability Research / Revue Européenne de Recherche sur le Handicap     Full-text available via subscription   (Followers: 11, SJR: 0.201, CiteScore: 1)
Alzheimer's & Dementia     Hybrid Journal   (Followers: 55, SJR: 4.66, CiteScore: 10)
Alzheimer's & Dementia: Diagnosis, Assessment & Disease Monitoring     Open Access   (Followers: 6, SJR: 1.796, CiteScore: 4)
Alzheimer's & Dementia: Translational Research & Clinical Interventions     Open Access   (Followers: 6, SJR: 1.108, CiteScore: 3)
Ambulatory Pediatrics     Hybrid Journal   (Followers: 5)
American Heart J.     Hybrid Journal   (Followers: 58, SJR: 3.267, CiteScore: 4)
American J. of Cardiology     Hybrid Journal   (Followers: 67, SJR: 1.93, CiteScore: 3)
American J. of Emergency Medicine     Hybrid Journal   (Followers: 48, SJR: 0.604, CiteScore: 1)
American J. of Geriatric Pharmacotherapy     Full-text available via subscription   (Followers: 13)
American J. of Geriatric Psychiatry     Hybrid Journal   (Followers: 15, SJR: 1.524, CiteScore: 3)
American J. of Human Genetics     Hybrid Journal   (Followers: 39, SJR: 7.45, CiteScore: 8)
American J. of Infection Control     Hybrid Journal   (Followers: 29, SJR: 1.062, CiteScore: 2)
American J. of Kidney Diseases     Hybrid Journal   (Followers: 37, SJR: 2.973, CiteScore: 4)
American J. of Medicine     Hybrid Journal   (Followers: 50)
American J. of Medicine Supplements     Full-text available via subscription   (Followers: 3, SJR: 1.967, CiteScore: 2)
American J. of Obstetrics and Gynecology     Hybrid Journal   (Followers: 264, SJR: 2.7, CiteScore: 4)
American J. of Ophthalmology     Hybrid Journal   (Followers: 67, SJR: 3.184, CiteScore: 4)
American J. of Ophthalmology Case Reports     Open Access   (Followers: 5, SJR: 0.265, CiteScore: 0)
American J. of Orthodontics and Dentofacial Orthopedics     Full-text available via subscription   (Followers: 6, SJR: 1.289, CiteScore: 1)
American J. of Otolaryngology     Hybrid Journal   (Followers: 25, SJR: 0.59, CiteScore: 1)
American J. of Pathology     Hybrid Journal   (Followers: 32, SJR: 2.139, CiteScore: 4)
American J. of Preventive Medicine     Hybrid Journal   (Followers: 30, SJR: 2.164, CiteScore: 4)
American J. of Surgery     Hybrid Journal   (Followers: 39, SJR: 1.141, CiteScore: 2)
American J. of the Medical Sciences     Hybrid Journal   (Followers: 12, SJR: 0.767, CiteScore: 1)
Ampersand : An Intl. J. of General and Applied Linguistics     Open Access   (Followers: 7)
Anaerobe     Hybrid Journal   (Followers: 4, SJR: 1.144, CiteScore: 3)
Anaesthesia & Intensive Care Medicine     Full-text available via subscription   (Followers: 67, SJR: 0.138, CiteScore: 0)
Anaesthesia Critical Care & Pain Medicine     Full-text available via subscription   (Followers: 25, SJR: 0.411, CiteScore: 1)
Anales de Cirugia Vascular     Full-text available via subscription   (Followers: 1)
Anales de Pediatría     Full-text available via subscription   (Followers: 3, SJR: 0.277, CiteScore: 0)
Anales de Pediatría (English Edition)     Full-text available via subscription  
Anales de Pediatría Continuada     Full-text available via subscription  
Analytic Methods in Accident Research     Hybrid Journal   (Followers: 6, SJR: 4.849, CiteScore: 10)
Analytica Chimica Acta     Hybrid Journal   (Followers: 44, SJR: 1.512, CiteScore: 5)
Analytica Chimica Acta : X     Open Access  
Analytical Biochemistry     Hybrid Journal   (Followers: 214, SJR: 0.633, CiteScore: 2)
Analytical Chemistry Research     Open Access   (Followers: 13, SJR: 0.411, CiteScore: 2)
Analytical Spectroscopy Library     Full-text available via subscription   (Followers: 14)
Anesthésie & Réanimation     Full-text available via subscription   (Followers: 2)
Anesthesiology Clinics     Full-text available via subscription   (Followers: 25, SJR: 0.683, CiteScore: 2)
Angiología     Full-text available via subscription   (SJR: 0.121, CiteScore: 0)
Angiologia e Cirurgia Vascular     Open Access   (Followers: 1, SJR: 0.111, CiteScore: 0)
Animal Behaviour     Hybrid Journal   (Followers: 236, SJR: 1.58, CiteScore: 3)
Animal Feed Science and Technology     Hybrid Journal   (Followers: 7, SJR: 0.937, CiteScore: 2)
Animal Reproduction Science     Hybrid Journal   (Followers: 7, SJR: 0.704, CiteScore: 2)
Annales d'Endocrinologie     Full-text available via subscription   (Followers: 3, SJR: 0.451, CiteScore: 1)

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Similar Journals
Journal Cover
Nuclear Engineering and Design
Journal Prestige (SJR): 1.061
Citation Impact (citeScore): 1
Number of Followers: 13  
 
  Hybrid Journal Hybrid journal (It can contain Open Access articles)
ISSN (Print) 0029-5493
Published by Elsevier Homepage  [3161 journals]
  • Computational and experimental analysis of gas/liquid two-phase flow in
           rod bundles with mixing-vane spacer grids
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Brian M. Waite, Horst-Michael Prasser, Michael Z. PodowskiAbstractThe existing evidence clearly shows that the physical phenomena governing gas/liquid two-phase flows are quite complicated even in the case of smooth conduits and simple geometries. Needless to say, the development of experimental, analytical and computational methods for predicting such flows in complex geometries is an even more complicated and challenging task. A configuration of interest to a broad range of industrial applications, including nuclear reactors, deals with the flow of two-phase mixture along the channels formed between narrow arrays of multiple parallel cylindrical elements (tubes or rods). The alignment of such elements is normally accomplished by installing spacer grids placed at regular distances along the flow. The presence of spacers actually affect flow conditions, including the velocity field, pressure drop, heat transfer and, in the case of two-phase flows, phase distribution.The objective of this paper is to present a comparative analysis of the results of a combined experimental, theoretical, and computational study of phase distribution around and downstream from complex-geometry spacer grids with split-vane type mixing devices. The main emphasis has been given to the analysis of the effect of proper interpretation of the experimental data on the modeling consistency. The importance of the understanding of uncertainties and limitations associated with the results of multidimensional computer simulations performed using mechanistic modeling principles based on an average bubble size is also discussed.
       
  • Nonlinear vibrations of a nuclear fuel rod supported by spacer grids
    • Abstract: Publication date: Available online 14 January 2020Source: Nuclear Engineering and DesignAuthor(s): Giovanni Ferrari, Giulio Franchini, Prabakaran Balasubramanian, Francesco Giovanniello, Stanislas Le Guisquet, Kostas Karazis, Marco AmabiliAbstractThe internal components of Pressurized Water Reactors (PWRs), particularly nuclear fuel assemblies, must be able to withstand Flow Induced Vibrations (FIVs) during operating conditions and during extreme accident conditions, such as earthquakes. Nuclear fuel assemblies are composed of long slender tubes, filled with uranium pellets that are bundled together by periodic support provided by spacer grids. Spacer grids are square structures used to increase thermal mixing in the core and provide support to the fuel rods and guide tubes allowing for the installation and removal of nuclear fuel rods. Nevertheless, spacer grids constitute a nonlinear flexible boundary condition experiencing friction forces and impacts complicating the dynamics of the fuel rod-spacer grid system. In order to improve safety margins in the design of nuclear fuel assemblies, it is of great interest to understand the nonlinear behavior at the interface of the spacer grids with the fuel rods, investigating the complexity due to the nonlinear evolution of the system stiffness and damping properties. In particular, the effect of the constant evolution of the vibration amplitude as a function of the change of the excitation forces on the dynamics of the fuel rod response is still undetermined. Experiments were carried out in quiescent water and in air to understand the nonlinear vibration response of a single zirconium fuel rod supported by spacer grids. The vibration response under a step-sine harmonic excitation at different force amplitude levels in the frequency neighborhood of the fundamental mode was measured. The response of the rod displayed nonlinear phenomena such as the shift of the resonant frequencies, multiple solutions with some instabilities (jumps) and hysteresis, and a weak one-to-one internal resonance. Tests were performed on an empty rod and on a rod filled with tungsten pellets representative of nuclear fuel. The pellets were let free to move and were subsequently blocked axially to reproduce the effect of the beginning-of-life constraint in operational nuclear plants. The experimental data were processed by means of a simplified identification procedure to extract the damping parameters of the vibrating system. The equivalent viscous damping is found to increase and to be a function of the level of excitation and of the peak vibration amplitude.
       
  • Characterisation of RBMK-1500 graphite: A method to identify the neutron
           activation and surface contamination terms
    • Abstract: Publication date: Available online 14 January 2020Source: Nuclear Engineering and DesignAuthor(s): V. Remeikis, R. Plukienė, A. Plukis, V. Barkauskas, A. Gudelis, R. Druteikienė, R. Gvozdaitė, L. Juodis, G. Duškesas, E. Lagzdina, D. Germanas, D. Ridikas, S. KrutovcovAbstractIn this study, we provide the results of radiological characterization of the RBMK-1500 graphite using modeling and nuclear spectrometry; we also introduce a method to identify neutron activation and surface contamination terms for the spent graphite waste. The simplified model of the reactor core created by programme package SCALE 6.2 for 4 × 4 fragment of the core was used for simulation of neutron activation of graphite impurities in the RBMK-1500 reactor. Calculations were supplemented by the non-destructive measurements of gamma-ray emitting nuclides and destructive analysis of selected samples of the graphite stack of Ignalina NPP Unit 1 reactor. Our analysis demonstrates that the partial contribution of different contamination sources can be identified by combining modeling and measurements. The findings on radionuclides such as 14C, 36Cl, 60Co, 134Cs, 137Cs, 154Eu as well as actinides in Ignalina NPP Unit 1 graphite having a significant impact on graphite activity and radiological characterization are discussed. The proposed method is also applicable for identification of contamination source in the other activated components of the reactor core.
       
  • Development of fault displacement PRA methodology and its application to a
           hypothetical NPP
    • Abstract: Publication date: Available online 14 January 2020Source: Nuclear Engineering and DesignAuthor(s): Ryusuke Haraguchi, Futoshi Tanaka, Katsumi Ebisawa, Toshiaki Sakai, Hideaki Tsutsumi, Ayumi Yuyama, Kunihiko Sato, Yoshinori Mihara, Yuji Nikaido, Shinichi YoshidaAbstractThe authors have developed fault displacement (FD) probabilistic risk assessment (PRA) methodology, through application to a hypothetical nuclear power plant (NPP).Core damage frequency (CDF) and its uncertainty of a hypothetical NPP for dip-slip faults and strike-slip faults were evaluated using a probabilistic FD hazard, FD fragilities of building and components, and fault tree and event tree models. Important initiating events, accident sequences, components and structure failures, as well as the range of FD that characterize the core damage risk profile were identified. Through the study, the authors confirmed the feasibility of the methodology, and also identified the important source of uncertainties within the methodology that require further development. Risk insights obtained from the FD PRA can be used to investigate countermeasures to reduce the FD risk.
       
  • Simulation and application of Bi-directional corrugated honeycomb aluminum
           as filling material for impact limiter of nuclear spent fuel transport
           cask
    • Abstract: Publication date: Available online 10 January 2020Source: Nuclear Engineering and DesignAuthor(s): Youdong Xing, Siyi Yang, Zhongfang Li, Shiqing Lu, Pengfei Zhang, Yukun An, Ertuan Zhao, John ZhaiAbstractThe filler material of impact limiter plays a vital role in the nuclear spent fuel transport cask. In this paper, combined with the size of the NAC-STC (a type of nuclear spent fuel transport cask), the energy absorbed by the impact limiter during the 9-meter drop process is derived, and it is analyzed that the plateau stress of the filling material is not less than 12.35 MPa, so it can ensure the safety of the transport cask. A small size bi-directional corrugated honeycomb aluminum was designed as a filling material for the impact limiter of the transport cask. By studying the characteristics of the bi-directional corrugated honeycomb aluminum and the mechanical properties and energy absorption characteristics, it is concluded that the material has a longer energy absorption range and average plateau stress can reach 15 MPa, which indicates the material has a good energy absorption characteristic, and can theoretically be used as an energy absorbing material for impact limiter. Further, combined with the application of the similarity theory, it shows that the energy absorption ratio of bi-directional corrugated honeycomb aluminum is 103:1 when the size is 10:1. Finally, through the simulation of drop test, it is further verified that the bi-directional corrugated honeycomb aluminum material is used as the filling material of the impact limiter can meet the safety.
       
  • Plant-level dynamic modeling of a commercial-scale molten salt reactor
           system
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Vikram Singh, Alexander M. Wheeler, Belle R. Upadhyaya, Ondřej Chvála, M. Scott GreenwoodAbstractOngoing research by the authors has led to a dynamic modeling approach that was verified against experimental data from the Molten-Salt Reactor Experiment (MSRE). These lumped-parameter models, characterizing the reactor dynamics, are nonlinear and represent changes in mass, energy, and temperature in all parts of the reactor plant. The reactivity feedback effects due to changes in temperature and Xe-135 concentration are taken into account. These models can be used to study both the time and frequency response to perturbations caused during normal operating conditions and during anomalies resulting from failure of certain sub-systems. A plant-level dynamic model of a representative molten salt reactor system developed based on the same methodology is presented here. A once-through steam generator providing superheated steam and reheated-regenerative Rankine cycle balance-of-plant system are coupled to the plant model. Results from simulation are presented for various transients and the resulting response of the plant is analyzed. Frequency response of the plant is also presented for various boundary conditions on the secondary side. Modeling results suggest that the inherent temperature-related feedbacks result in load-following characteristics that can be leveraged to engineer control systems offering a great deal of autonomy.
       
  • Bayesian inference and non-linear extensions of the CIRCE method for
           quantifying the uncertainty of closure relationships integrated into
           thermal-hydraulic system codes
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Guillaume Damblin, Pierre GaillardAbstractUncertainty Quantification of closure relationships integrated into thermal-hydraulic system codes is a critical prerequisite in applying the Best-Estimate Plus Uncertainty (BEPU) methodology for nuclear safety and licensing processes. This issue has been subject to several international initiatives, such as BEMUSE and PREMIUM projects, as well as some statistical developments of which the “CIRCE” method. This method has been designed at the end of the twentieth century, then extensively used for quantifying the uncertainty of closure relationships integrated into the CATHARE thermal-hydraulic system code.The purpose of the CIRCE method is to estimate the (log)-Gaussian probability distribution of a multiplicative factor applied to a reference closure relationship in order to assess its uncertainty. Even though this method has been implemented with success in numerous physical scenarios, it can still suffer from substantial limitations such as the linearity assumption and the difficulty of properly taken into account the inherent statistical uncertainty. In the paper, we will extend the CIRCE method in two aspects. On the one hand, we adopt the Bayesian setting putting prior probability distributions on the parameters of the (log)-Gaussian distribution. The posterior distribution of the parameters is then computed with respect to an experimental database by means of Markov Chain Monte Carlo (MCMC) algorithms. On the other hand, we tackle the more general setting where the simulations do not move linearly against the multiplicative factor(s). MCMC algorithms then become time-prohibitive when the thermal-hydraulic simulations exceed a few minutes. This handicap is overcome by using Gaussian process (GP) emulators which can yield both reliable and fast predictions of the simulations.The GP-based MCMC algorithms will be applied to quantify the uncertainty of two condensation closure relationships at a safety injection with respect to a database of experimental tests. The thermal-hydraulic simulations will be run with the CATHARE 2 computer code.
       
  • Low resolution modelling of mixing phenomena in PWR fuel assemblies
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): B. Mikuž, F. RoelofsAbstractSimulation of single-phase turbulent flow in the entire reactor core is a challenging task due to a complex geometry and a huge computational domain, which requires large number of computational cells. In order to reduce the size of a computational mesh, simplifications in the geometry are practically inevitable. Among the most challenging parts with many small details that affect the flow through a fuel assembly are mixing grids. There are several different designs of the mixing grids. One of the most frequently used designs consists of a strap, springs and dimples, which are used as a structural support for the fuel rods, and mixing vanes, which deflect the fluid and increase the mixing. Neglecting the effects of the mixing grid would result in a less accurate reproduction of the flow and temperature field, which may lead to too small inter sub-channel mixing in the fuel assembly. Hence, application of a porosity model is proposed for the mixing grid, which largely reduces the computational mesh. Additional momentum sources mimic the effect of the mixing vanes. This approach aims to fill the gap between three-dimensional well-resolved Computational Fluid Dynamics (CFD) and one-dimensional sub-channel or system codes. As such, they are able to provide valuable insights to the flow behaviour in the reactor core.The present study examines the possibility to model the effect of the mixing grid by applying momentum sources in the governing equations. A split-type mixing grid is considered, which is one of the most frequently used designs in a Pressurized Water Reactor (PWR) fuel assembly. The applied split-type mixing grid generates a particular pattern of swirl motion within each sub-channel as well as it increases the cross flow between neighbour sub-channels of the fuel bundle. Non-homogeneous momentum sources have been applied, which mimic the deflection of the flow and blockage by the mixing vanes. The obtained pattern of the secondary flow closely resembles the secondary flow, which has been obtained with a reference CFD simulation that included all detailed geometrical aspects of the grid spacer with mixing vanes. The magnitudes of the momentum sources have been tuned to generate similar magnitudes of the secondary flow and stream-wise vorticity as observed in the results of the reference CFD simulation. In the next step, this approach has been applied on successively coarser meshes to investigate its effectivity on low resolution (very coarse) meshes. It turned out that the proposed model is able to mimic the mixing effect between the sub-channels. However, discrepancies are observed in the development of the secondary flow in the stream-wise direction. In the last step, the coolant flow in a flow domain consisting of 9 fuel assemblies is reproduced for normal conditions inside a PWR core as well as for two off-normal scenarios, which consider two different blockages at the entrance of the central fuel assembly. This approach has a promising potential to perform a CFD simulation of a realistic fluid mixing inside and between adjacent fuel assemblies in an entire PWR reactor core.
       
  • Research and development for safety and licensing of HTGR cogeneration
           system
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Hiroyuki Sato, Takeshi Aoki, Hirofumi Ohashi, Xing L. YanAbstractHigh temperature gas-cooled Reactor (HTGR) is expected to extend the use of nuclear heat to a wider spectrum of industrial applications such as hydrogen production, high efficiency power generation, etc., due largely to high temperature heat supply capability as well as inherent safe characteristics. Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first HTGR in Japan, towards the realization of industrial use of nuclear heat. On the basis of licensing experience through the HTTR construction, JAEA initiated an activity to establish an international safety standard for licensing of commercial HTGR cogeneration systems fully taking into account safety features of HTGRs.This paper explains a roadmap towards licensing of commercial HTGR cogeneration systems. A test plan using the HTTR to support the establishment of safety standards and safety analysis methods is also presented.
       
  • Development of strength evaluation method of ceramic reactor for
           iodine-sulfur process and hydrogen production test in Japan Atomic Energy
           Agency
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Hiroaki Takegami, Hiroki Noguchi, Nobuyuki Tanaka, Jin Iwatsuki, Yu Kamiji, Seiji Kasahara, Yoshiyuki Imai, Atsuhiko Terada, Shinji KuboAbstractJapan Atomic Energy Agency (JAEA) has been conducting R&D on the thermochemical iodine–sulfur (IS) process for nuclear-powered hydrogen production. The IS process is one of the promising candidates of heat application of the high-temperature gas-cooled reactors. JAEA achieved continuous hydrogen production for one week with a hydrogen production rate of approximately 20 NL/h by using a test apparatus made of glass and fluororesin material. Subsequently, as a next step, JAEA fabricated main chemical reactors made of industrial structural materials and confirmed their integrity in practical corrosive environments in the IS process. Based on the results of these confirmation tests, JAEA has constructed a 100 NL/h-H2-scale test facility made of industrial structural materials. This paper presents an outline and results of hydrogen production tests by using the test facility and reliability improvements by developing a strength estimation method for heat-resistant and corrosion-resistant ceramics components made of silicon carbide.
       
  • Experimental study on the circulating-cavity flow and an innovative
           central baffle design in a steam generator
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Yu Wang, Dao-gang Lu, Cong Wang, Qiong Cao, Xiang-bin Li, Shi-liang ZhouAbstractSteam generator (SG) is the key equipment in pressurized water reactors (PWR), which transfers heat from primary circuit to secondary circuit and has the feed water vaporized into the steam. It is related to the safe, reliable and economical operation of the nuclear power plant. Many researches have been done on SG, including numerical simulation and experimental research. Since it is involved with complicated steam-water (two phase) flow in high temperature and high pressure, it is not easy to measure the key parameters such as pressure, temperature and void fraction, especially to carry out the visual observation. So the detailed working information of SG such as void fraction and flow pattern is still unknown, while this information is very important for the improvement of SG performance. In order to obtain the working information of SG and study the dynamic flow process in secondary side, a visualization scaled-down mock-up experimental bench was set up. Although its operation parameters (temperature and pressure) are much lower than the actual one in PWR, the internal dynamic flow process of SG in this facility is kept similar to the actual one by the scaling analysis and design. Appling the high speed camera, particle image velocimetry (PIV) and self-made optical fiber probes, two phase flow behavior including flow pattern, velocity and void fraction were collected and analyzed in the experiment. From the experiment, a circulation phenomenon is discovered; which is defined as circulating-cavity flow (CCF). At U-bend area, CCF flows from hot side to cold side, which may result in flow-induced vibration (FIV) and jeopardize the U-tubes. In order to limit CCF, a baffle installed in the middle of U-bend area is proposed to suppress CCF. The experimental results show that this baffle can effectively suppress the CCF. This paper may contribute to the design and safety of SG.
       
  • RBF-based adaptive sliding mode controller with extended state observer
           for load following of nuclear power plant
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Jiuwu Hui, Jingqi YuanAbstractLoad following is one of the basic control task for nuclear power plants. This paper proposes a RBF-based adaptive sliding mode control strategy with extended state observer (ESO) for load following of nuclear power plants in the presence of disturbances. The nonlinear mathematical model of the nuclear reactor system is firstly described. In order to recover unmeasured states including average fuel temperature and relative delayed neutron precursor density, an extended state observer is designed to reconstruct both unmeasured states and disturbances at the same time, and its asymptotically stability condition is analyzed. Based on the nuclear reactor model and estimation information from the ESO, a RBF-based adaptive sliding mode control strategy is proposed, which consists of a dual closed-loop integral sliding mode control scheme while taking into account disturbances simultaneously. In addition, the chattering phenomenon of sliding mode control is alleviated by replacing the sign function with a saturation function. Simulation results are provided to verify the effectiveness of the overall control scheme compared with conventional PID control and sliding mode control approaches in accordance to load following performance, states reconstruction accuracy, and disturbances rejection ability.
       
  • Discussion on the accident behavior and accident management of the HTGR
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Chen Zhipeng, Wang Yan, Zheng YanhuaAbstractHigh Temperature Gas-cooled Reactor (HTGR), which plays an important role in the worldwide development of Generation-IV nuclear energy technology, is well-known for the high safety, high efficiency and process heat application. A commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) has been designed and is now under construction in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in the end of 2020.In this paper, the reactor behaviors of the HTR-PM during the typical design basis accidents (DBAs) and beyond design basis accidents (BDBAs) have been introduced. In the DBAs, the maximum fuel temperature will never exceed its design limitation of 1620 °C, below which under the design burn-up the “Tristructural-isotropic” (TRISO) coated particles can effectively retain their fission-product inventory. Even in the typical BDBAs with extremely low probability, there is enough time, e.g. several days, to adopt appropriate measures to mitigate the consequence, so that the large release of the radioactive materials would not happen.Accident management is important for the nuclear power plant. Besides, After the Fukushima Dai-ichi nuclear accident, in the worldwide, people are more concerned about the severe accident management of the nuclear power plant, and some standards and guidelines are issued. Based on the accident analyses of the HTR-PM, its accident management is studied and discussed. The designers believe that the accident management procedures can be simplified and no offsite emergency measures are needed for the HTR-PM due to the inherent safety design. Compared to the other types of the nuclear power plant, e.g. the Pressurized-Water Reactor (PWR) power plant, the Severe Accident Management Guideline (SAMG) can be simplified or even unnecessary for the HTR-PM.Above work also can provide reference for the further study on the safety regulations, standards or guidelines for the design, operation and accident management of the HTGR.
       
  • Thermal-hydraulic design methodology and trade-off studies for a dual-salt
           breed-and-burn molten salt reactor
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Alisha Kasam, Jeong Ik Lee, Eugene ShwagerausAbstractA methodology is developed for thermal-hydraulic analysis and design of a breed-and-burn molten salt reactor (BBMSR). By using separate fuel and coolant molten salts, the BBMSR is proposed to overcome key materials limitations of traditional breed-and-burn and molten salt reactor designs. The BBMSR fuel concept includes an inner wall that divides the ascending and descending flows of naturally convecting fuel salt. A finite-difference model (FDM) is developed to iteratively solve for the temperature and velocity distributions in both sections of the concentric fuel. The FDM is used to perform parametric studies of the effect of fuel geometry and heat generation rate on the heat transfer performance of the fuel. The FDM is then integrated into a design search algorithm that identifies the operational limits for a given BBMSR fuel geometry, within a set of defined constraints. A range of thermal-hydraulic fuel design options are evaluated, and trade-off studies are performed to identify the most promising fuel design space for competitive power production and neutronic efficiency in the BBMSR.
       
  • Penetration flow into a branch pipe causing thermal fatigue at a mixing
           tee
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Koji Miyoshi, Yoichi Utanohara, Masayuki KamayaAbstractFlow structure in a branch pipe of the mixing tee was investigated in order to improve the assessment method for thermal fatigue damage. The mixing tests were performed using a test section which was made of transparent acrylic resin and consisted of a horizontal main pipe and a vertical branch pipe. The penetration flow into the branch pipe was observed under the condition of branch flow with low flow velocity. The flow in the branch pipe was classified into three flow patterns: no penetration; entrained penetration; and impinged penetration. These flow patterns depended on the momentum ratio of the main and branch pipes. The entrained penetration was caused by the entrainment of the main flow into the separation region where the branch flow was bent by the main flow and separated from the branch pipe wall. The impinged penetration was the flow pattern in which the main flow impinged on the branch pipe wall and penetrated into the branch pipe. These penetration flows occurred intermittently and the penetration depth changed rapidly. The maximum penetration depth correlated with the momentum ratio.
       
  • Study on melt stratification and migration in debris bed using the moving
           particle semi-implicit method
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Gen Li, Panpan Wen, Haobo Feng, Jun Zhang, Junjie YanAbstractIn the severe accident of nuclear reactors, the melt stratification and migration in debris bed directly affect the melt heat transfer characteristics and the heat flux distribution on the lower head wall, which in turn affect the follow-up accident progressions including the pressure vessel failure mode, melt leakage, and molten corium concrete interaction. They are the key issues in nuclear reactor severe accident research. In the present study, the MPS method was applied to analyze melt stratification and migration in debris bed by including the models of radiation heat transfer among debris, melt phase change and surface tension. To validate the numerical method, the experiment was performed using the metal tin and ternary nitrate as the core melt simulant. Then, the stratification behavior of oxide and metallic core melt was investigated under varied parameters and conditions. The results indicated that the crust formed at the oxide and metal melt interface could block the stratification, which had important influence to the layered configuration. The melt contact area increased in the conditions of crust formation prior to the stratification completion. The high melt and debris temperatures and surface tension could contribute to the stratification, but the high viscous melt when undergoing phase transition needed a long time. Moreover, the comparison of melt stratification under adiabatic and isothermal boundary conditions showed that the oxide melt solidification trapped some metal melt in the voids. The penetration distance of metal melt in the oxide debris bed increased rapidly in the initial moment and gradually in the following moment, and finally the migration was terminated by melt crust. The melt penetration distance increased as the increase of debris bed porosity and melt and debris temperatures.
       
  • Reactor physics evaluation of the TRIGA LEU fuel in the 20 MW NIST
           research reactor
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Kyle A. Britton, Zeyun WuAbstractThis paper performs a neutronics evaluation of the General Atomics UZrHx low enriched uranium fuel – TRIGA fuel – in the National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this study is to examine the accountability and sustainability of the TRIGA fuel on neutronics aspects when applying it to the NBSR conversion. A feasibility scoping study was previously undertaken with considerations on various fuel dimensions, fuel rod layout configurations, and structure material selections, identifying the best option of deploying the TRIGA fuel to NBSR. Continuing with these efforts, an equilibrium NBSR core using the identified fuel was generated, and a well-round physics assessment was carried out by examining key neutronics performance characteristics of the core. All calculations were completed with MCNP-6, a 3-D Monte Carlo neutron transport code. The same fuel management scheme and fuel cycle length as the existing NBSR was adopted in the equilibrium core generation adopts to retain performance consistencies. The effectiveness of the fuel was examined at four representative burnup states of the fuel cycle. Neutronics performances of the equilibrium core was characterized by the fast and thermal neutron flux level as well as power distribution in the core. Reactor safety related parameters such as kinetics parameters and power peaking factors were also evaluated in the study. All results were compared against the current NBSR fueled with HEU for justifications. The findings in this research prove the viability of the TRIGA fuel for the NBSR conversion, and provide supporting data for future investigations on this subject.
       
  • Neutron point kinetics model with precursors’ shape function update
           for molten salt reactor
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Rodrigo Costa Diniz, Alessandro da Cruz Gonçalves, Felipe Siqueira de Souza da RosaAbstractThe circulation of liquid fuel is a feature of innovative nuclear reactors, such as Molten-Salt Reactors, and requires a specific neutronics model to take it into account. In this paper we propose a point kinetics model, for a hypothetical one-dimensional reactor, that is based upon and improves one that already exists in the literature. The model is therefore solved numerically and compared to spatial kinetics, which is considered to be the exact solution. The approximations that were done, which best portray the phenomena associated to the problem, are discussed to explain the improvements obtained with the proposed model.
       
  • Fuel options for nuclear ship reactors featuring reactivity swing below
           one dollar
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Mengqi Bai, Benjamin A. Lindley, Tim AbramAbstractEnvironmental protection against air pollution and climate change has drawn great and increasing public attention in recent decades. Nuclear power involving no pollutant and CO2 emissions provides a more sustainable way for commercial shipping activities. Nine fuel options are simulated to find the combinations of key parameters that make the core to have a lifetime reactivity swing below one dollar. Such a low reactivity swing, when combined with a negative feedback reactivity coefficient, may facilitate passive load follow and passive safety, which means the reactivity can be controlled through temperature or thermal expansion of the reactor core. These fuel options are compared regarding reactor physics and thermal hydraulics, under certain constraints associated with ship reactors. WIMS code is used for burnup, perturbation and parameter calculating. MONK and SERPENT codes are used to verify the WIMS model. It is found that ship-based lead-cooled fast reactor can be designed to have superior core safety and good economics. Different fuel options show diverse characteristics in terms of power output limits, refuelling intervals, natural circulation cooling capabilities and other safety related parameters.
       
  • Application of BEPU method to loss of RHR event during mid-loop operation:
           Uncertainty quantification of RELAP5 model
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Toshihide TorigeAbstractIn recent years, best estimate plus uncertainty (BEPU) method has become more important, and the guidelines on the process have been developed. However, there are not many specific reports on the probability distribution of the model uncertainty that greatly affects the evaluation result. Also, there are not many application cases of the BEPU method for the events of the PWR mid-loop operation which is suggested to be important in terms of probabilistic risk assessment (PRA). Therefore, this report describes the results of the quantification of the RELAP5/MOD3.2 model uncertainties related to the loss of residual heat removal (RHR) event during the PWR mid-loop operation. Based on the experimental literatures, the probability distributions of the model uncertainty were developed, which contributes to statistical analyses and research on the model reliability.
       
  • A review of HTGR graphite dust transport research
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Qi Sun, Wei Peng, Suyuan Yu, Kaiyuan WangAbstractGraphite dust is the important contents of source term for safety analysis of high temperature gas-cooled reactors (HTGR). The spherical fuel element circulation in a pebble bed reactor causes many interactions between the fuel elements and other graphite components that inevitably leads to graphite dust production. Micron size graphite particles then move with the helium gas and deposit on various surfaces and in flow dead zones in the primary loop, which complicates equipment maintenance and repair and affects the heat transfer. In addition, the graphite dust is quite porous, so some radioactive fission products will adhere to the dust, which leads to radioactive fission products being distributed on the surfaces of the primary loop. Graphite dust carrying radioactive fission products can also leak into the environment during break accidents leading to radioactive pollution of the environment. Thus, studies are needed for the graphite dust transport in HTGRs. This paper reviews the research on the generation, distribution, radioactivity, deposition, resuspension and coagulation of graphite dust in a pebble bed high temperature reactor. The results show that most of the graphite dust is produced by mechanical wear, while chemical reactions can become an important source during an ingress accident. The graphite dust particles generally have sizes on the order of microns and carry radioactive substances. The graphite dust flows along with the helium in the primary loop and adheres to equipment surfaces. Local turbulent diffusion and large temperature gradients cause the graphite dust to deposit on the surfaces, while gravitational settling has a dominant effect in dead-end zones. In case of accidents or other transients, the dust deposited on the surfaces can become resuspended which will sharply increase the dust concentration, leading to uncertainties about the subsequent operating characteristics. In addition, coagulation and growth of the graphite dust particles due to thermophoresis and electric field forces is also a matter of concern.
       
  • Damage performance based seismic capacity and fragility analysis of
           existing concrete containment structure subjected to near fault ground
           motions
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Song Jin, Jinxin GongAbstractContainment structures fill the critical role of providing the final barrier against release of radioactive materials. Near fault earthquake pose great threaten to the function of existing containment structure. There is a growing necessity of performing a detailed seismic capacity and fragility analysis of existing containment structure. This paper presents seismic capacity and fragility analysis of existing containment structure subjected to near fault ground motions from damage performance based perspective. To capture uncertainties associated with the ground motions, incremental dynamic analysis technique is conducted, and three damage levels are proposed based on tensile damage factor provided in concrete damaged plasticity (CDP) model. Fragility curves of containment structure under minor damage, moderate damage and severe damage is obtained with MATLAB optimization program. Results of this study reveal that tensile damage of containment structure within 8 m of the bottom of the containment is most vulnerable locations in containment structure. The bottom location of the containment structure is dominant by bending cracks, and the lower location of the containment structure (except the bottom location) is governed by diagonal shear cracks. With the increase of ground motion intensity, shear behavior of the containment structure becomes more obvious.The HCLPF seismic capacity of this containment structure under minor damage, moderate damage and severe damage is 0.452 g, 0.631 g and 0.790 g, respectively.
       
  • Fracture studies on reactor pressure vessel subjected to pressurised
           thermal shock: A review
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): K. Thamaraiselvi, S. VishnuvardhanAbstractStructural integrity of a Reactor Pressure Vessel (RPV) plays a decisive role in the operating life of a nuclear power plant. A potential threat to the structural integrity of an RPV occurs as a result of the exposure of RPV to Pressurised Thermal Shock (PTS). Pressurised thermal shock is a thermo-mechanical load on the RPV wall induced by steep temperature gradients and the structural load created by internal pressure of the fluid within the RPV. The combined stress from the PTS event can cause crack formation or in severe cases the total structural failure of the reactor pressure vessel. This paper reviews various stages involved in the structural integrity assessment of RPV subjected to PTS where focus is given to fracture analysis which is pivotal in the structural integrity assessment of RPVs against the PTS. Significant contributions made in the fracture assessment of RPV steel under PTS is discussed. Experimental investigations carried out to evaluate the fracture behaviour of RPV steel subjected to PTS is presented. Effect of constraint on the evaluated fracture parameters is also studied.
       
  • Framework for dynamic analysis of radioactive material transport packages
           under accident drop conditions
    • Abstract: Publication date: 15 April 2020Source: Nuclear Engineering and Design, Volume 360Author(s): Xian-Xing (Lambert) Li, Catherine Wang, Jim SatoAbstractRadioactive material transport packages for certain forms of radioactive materials are designed to withstand a 9-meter accident drop onto an essentially rigid surface in accordance with the IAEA regulations SSR-6. This study presents a framework for dynamic analysis of transport packages under accident drop and impact conditions. To evaluate structural behavior and performance of the transport packages with large deformation occurring in a very short duration, material models incorporated in the dynamic analysis framework are established with four key aspects: 1) true stress and strain are used to describe characteristics of large deformation of materials; 2) effect of strain rate is included to reveal dynamic response of materials to high-energy impact loads; 3) material constitutive models are established to capture realistic mechanical behaviour of materials under highly triaxial stress states; and 4) stress triaxiality-dependent damage and fracture criteria of materials are incorporated to limit the plastic deformation to material fracture and to safeguard the containment boundary of the Dry Storage Container (DSC) that contains the used (i.e., irradiated) fuel or other forms of radioactive materials. Critical material and model parameters and properties associated with the material constitutive models, true stress-strain relationships, strain rate effect, and damage and fracture criteria are appropriately determined based on experimental results or theoretically derived based on observed material behavior in elementary stress states. The implementation of the proposed analysis framework is demonstrated by a case study for a Type B(U) transport package, with results to predict the maximum true plastic strain, strain rate, and damage condition of the materials of the package. Since the proposed analysis framework is comprehensive enough to cover necessary aspects to complete an advanced analysis, and the critical material and model parameters are properly determined within this framework that can be readily implemented, the proposed analysis framework is capable of addressing high-energy accident drop scenarios and impact conditions of radioactive material transport packages and containers to meet the nuclear licensing requirements.
       
  • 3-D visualization of AGR fuel channel bricks using Structure-from-Motion
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Kristofer Law, Graeme West, Paul Murray, Chris LynchAbstractThis paper outlines a new framework for applying Structure-from-Motion (SfM) to challenging, feature-poor environments such as those observed during AGR fuel channel inspection. Deriving structural information from Advanced Gas-cooled Reactor (AGR) inspection footage is challenging due to several key issues: lack of discriminative salient features within the channel, inconsistency in lighting during the inspection process, lack of textural information within the channel and noise from the inspection equipment. This presents difficulties to techniques such as SfM due to its reliance on finding and reliably tracking a set of robust features from multiple viewpoints. This paper introduces the first use of an incremental 3-D reconstruction framework which can produce reconstructions of footage obtained within a nuclear reactor. It approaches this issue by introducing a novel correspondence searching methodology which can operate within feature-poor environments by utilising a constrained, iterative threshold matching technique to obtain robust feature matches. This paper demonstrates the approach using two datasets: laboratory footage obtained from an experimental setup emulating a small sub-section of the channel and in-core inspection footage of AGR fuel channels.
       
  • An advanced method for quick detection of heavy water leak in steam
           generators of pressurized heavy water reactors
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): B.N. Dileep, P.M. Ravi, Managanvi Sangamesh, N. KarunakaraA user friendly and rugged online system has been developed based on Cerenkov photon counting technique for heavy water leak detection in steam generators of pressurized heavy water reactor. The system has the capability to identify presence of fission and activation products in the secondary coolant of steam generator, an indication of leak from primary coolant, by detecting Cerenkov emission produced by hard beta emitting radionuclides in water. In this method the steam sample from steam generator is cooled during its transit through sampling line and the condensate is passed online through a polythene plastic flow cell placed between two matched bi-alkali photomultiplier tubes with coincident counting circuitry to observe the Cerenkov photon count rate. The minimum detectable leak rate for the newly developed system, when used in a steam generator of a 220 MWe reactor, was determined to be 3.6 kg h−1. Upon designing, construction and characterization, the system was successfully employed to detect and identify leak in steam generator of a reactor.Graphical abstractGraphical abstract for this article
       
  • Status of research and development of learning-based approaches in nuclear
           science and engineering: A review
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Mario Gomez-Fernandez, Kathryn Higley, Akira Tokuhiro, Kent Welter, Weng-Keen Wong, Haori YangAbstractNuclear technology industries have increased their interest in using data-driven methods to improve safety, reliability, and availability of assets. To do so, it is important to understand the fundamentals between the disciplines to effectively develop and deploy such systems. This survey presents an overview of the fundamentals of artificial intelligence and the state of development of learning-based methods in nuclear science and engineering to identify the risks and opportunities of applying such methods to nuclear applications. This paper focuses on applications related to three key subareas related to safety and decision-making. These are reactor health and monitoring, radiation detection, and optimization. The principles of learning-based methods in these applications are explained and recent studies are explored. Furthermore, as these methods have become more practical during the past decade, it is foreseen that the popularity of learning-based methods in nuclear science and technology will increase; consequently, understanding the benefits and barriers of implementing such methodologies can help create better research plans, and identify project risks and opportunities.
       
  • A new approach for crack detection and sizing in nuclear reactor cores
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Michael G Devereux, Paul Murray, Graeme M. WestAbstractRemote Visual Inspection (RVI) of reactors in nuclear power plants allows station operators to assess the health and condition of their plant. In the UK, most nuclear stations are of the Advanced Gas-cooled Reactor (AGR) design. During planned periodic outages, a representative portion of each AGR core is inspected using specialist tools equipped with various sensors including a video camera for RVI. If cracks are observed in the core during data capture, a stitched image of the region needs to be created so that the crack can be analysed and sentenced (classifying the crack morphology, location, orientation and size) before the station is returned to service, provided return to service is justified. Currently, the crack analysis and sizing activities are conducted manually by expert analysts in a laborious process. In this paper, we present a new image processing approach capable of automating aspects of the crack analysis process. Specifically, we describe a set of techniques for quickly and accurately detecting the presence of cracks in AGR fuel channel inspection images. We also present a method for detecting circular channel features known as trepanned holes whose dimensions are known and can thus be used for scaling. The results of applying the proposed techniques are evaluated on image data from real AGR fuel channels and are shown to produce comparable results to those obtained manually. The advantage of the proposed approach is that it is fast, robust and more repeatable than the existing manual approach.
       
  • The corrosion of aluminum alloy and release of hydrogen in nuclear reactor
           emergency core coolant: Implications for deflagration and explosion risk
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Junlin Huang, Derek Lister, Shunsuke UchidaAbstractHydrogen evolution (HE) accompanying the corrosion of aluminum alloy in the sump water formed after a loss-of-coolant accident (LOCA) influences the safety of reactor containments due to the deflagration and explosion risk of the air-hydrogen mixture. In experiments examining the corrosion of Al alloy 6061 in borated solutions simulating the water chemistry of post-LOCA sump water, HE rates were evaluated by analyzing the potentiodynamic cathodic polarization curves from rotating cylinder electrodes, and an empirical formula predicting the HE rates as a function of solution pH was proposed based on the evaluation results. At pH 7, the dominant cathodic reaction during free corrosion was found to be the oxygen reduction reaction and the HE rate was slow; however, the HE reaction became increasingly significant as the solution pH progressively increased to 11. The HE sources on the alloy surface were inferred to be mainly the bulk matrix, which was covered with a less protective hydroxide film in more alkaline solution, as well as the trenches formed by the intense anodic dissolution of the matrix around electrochemically more noble Fe-bearing intermetallic particles. Local alkalization due to preferential oxygen reduction on these intermetallic particles may also contribute to the formation of such trenches and HE; however, this mechanism should have been mostly suppressed by the buffering capacity of the borated solutions.
       
  • Numerical and experimental campaigns for lead solidification modelling and
           testing
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Manuela Profir, Vincent Moreau, Tomáš MelicharAbstractThe Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in pool configuration is investigated in the framework of the SESAME project. This paper focuses on the coupling between numerical simulations and experimental activity with the objective to make CFD a valid tool in support to the design of safe and innovative Gen-IV nuclear reactors. The attention is focused on the possible occurrence of lead solidification phenomenon. The aim is to demonstrate that the overall modelling of HLM complex systems can include solidification phenomenology by reproducing small/medium scale experiment, making comparison with experimental results, improving the numerical setting (post-test) and evaluating time scales, limitations and computational costs. A dedicated experimental campaign on the SESAME-Stand facility has been performed by the Research Centre Rez (CVR), relying on the use of CFD support, with the specific objective to build a series of datasets suited also for the CFD modelling validation.The numerical simulations of the SESAME-Stand experimental facility performed in STAR-CCM + are described. The model uses liquid lead as working fluid in the pool and air in the cooling channel. By increasing the mass flow rate in the cooling air channel, the solidification process is initiated and the freezing front propagates in the pool until reaching an internal obstacle. The issues encountered in the pre-test simulations have been fully overcome by means of systematic investigations and all the improvements have been applied in the post-test model. A decisive modelling aspect was the correct implementation of the thermal radiation which plays an important role in the cooling process. The numerical model allowed to reach an increased number of initial steady states and accessible transients, according to the experimental matrix. The comparison with the experimental results shows similar temperature configurations and comparable lead frozen fractions in the steady state cases and good agreement in the fast transient solidification-remelting cases.
       
  • Lead coolant modeling in system thermal-hydraulic code HYDRA-IBRAE/LM and
           some validation results
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): N.A. Mosunova, V.M. Alipchenkov, N.A. Pribaturin, V.F. Strizhov, E.V. Usov, P.D. Lobanov, D.A. Afremov, A.A. Semchenkov, I.A. LarinAbstractThe paper overviews the challenges associated with the development of a system thermal-hydraulic code applied for lead coolant by the example of a Russian system thermal-hydraulic code HYDRA-IBRAE/LM. It presents the PIRT describing relevant thermal-hydraulic phenomena that are to be simulated to model operational transients and accidents of Russian lead-cooled reactor BREST-OD-300. The paper also mentions the points in favor of using three-fluid model as a basic one. The results of evaluation and selection of lead coolant closure relations, that could be used in system thermal-hydraulic code, are presented. The paper demonstrates that heat transfer in the lead coolant may be calculated using closure relations obtained for the other liquid metal coolants introducing correction factors depending on the value of contact thermal resistance caused, for example, by oxide films present at the surface. The paper presents recommended properties of the lead coolant and summarizes experimental research findings of the Russian project PRORYV that can be used to validate system thermal-hydraulic codes. The paper discusses validation results for HYDRA-IBRAE/LM code based on experiments carried out using a BREST-OD-300 steam generator model and those with lead coolant cooldown due to argon injection.
       
  • Three-dimensional transient analysis of coupled RCS-containment integral
           system
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Li Ge, Zijiang Yang, Jianqiang Shan, Huaqi Li, Dong LiuAbstractThe integral analysis of reactor cooling system (RCS) and containment is becoming increasingly important for advanced reactors, especially reactors with passive safety technology. In this study, a new three-dimensional containment thermal–hydraulic analysis method is established using the equations and solutions similar to the reactor cooling system program for a strong coupling calculation of the containment and RCS. The integral program was verified by simulating the gravity drain line break accident of the PUMA test facility. The simulation results indicate that the coupled RCS-containment integral analysis program can be applied for analyzing the thermal–hydraulic coupling of the containment and RCS. The integral program was applied to simulate the LOCA for the RCS-containment integral system of AP1000. The results indicate that the integral program can be applied to analyse the AP1000 LOCA.
       
  • Simulation of coolant mixing in a BWR spent fuel storage pool and flood
           chamber
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): T. HöhneAbstractThe waiting times in case of failure of the cooling system of the spent fuel storage pool were determined with the three-dimensional numerical calculation tool ANSYS CFX. With the calculated variant, it is assumed that the swivel gate is opened when required. For the decay heat of 1.5 MW, waiting times until the spent fuel storage pool has been heated to 60 °C or 80 °C were calculated and the temperature offset between the spent fuel storage pool and the storage chamber pool was determined.The calculations showed that the coolant from the flood chamber and the storage chamber, which is located above the lower edge of the open swivel gate, mixes ideally with the water from the spent fuel pool. The results of the CFD analysis can be used for the cross-code verification of models in integral codes.
       
  • Porous structure of cylindrical packed beds with aspect ratios between 1
           and 2
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Charl G. du ToitAbstractThe paper provides a comprehensive summary of the analysis of the porous structure of cylindrical packed beds with aspect ratios between 1 and 2. For the first time a proper geometrical explanation is given for the shape of the curve for the overall porosity as a function of the aspect ratio. Errors found in the published equations for the lateral coordinates of the spheres in packed beds with aspect ratios between 1.866 and 2.0 are corrected. A geometrical explanation for the presence and the behaviour of the inflection point (“porosity surge wave”) observed in the curves for the radial variation in the porosity is also given for the first time. A systematic analysis of the axial variation in the porosity is also presented for the first time and based on the geometry analytical equations are derived for the minimum and maximum axial porosity as a function of aspect ratio. The results can serve as a benchmark for experimental and numerical studies on the porous structure of small aspect ratio cylindrical packed beds.
       
  • A criticality study on the LA-1 accident using Monte Carlo methods
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Mikolaj OettingenThe paper presents a criticality analysis of the first fatal nuclear accident with unexpected criticality of the fissionable system. The operator of the experiment, Harry Daghlian, received a radiation dose of about 5.1 Sv and died 28 days later due to acute radiation syndrome. The paper features a numerical reconstruction of the experimental set-up and the environment in the laboratory at the time of the accident, based on publicly available documentation. The Monte Carlo Continuous Energy Burnup Code was used in all criticality calculations. The study is a follow-up of the previous criticality analysis of the Louis Slotin accident (LA-2), in which the same plutonium core was used in a fatal experiment. In the paper, I present the numerical model developed, the influence of the system components on the criticality state, the approach to criticality in function of the tungsten-carbide reflector mass, as well as a benchmarking study using JEFF3.1 and ENDF/B-VII.1 nuclear data libraries.Graphical abstractGraphical abstract for this article
       
  • D 2 O + H 2 O +coolant&rft.title=Nuclear+Engineering+and+Design&rft.issn=0029-5493&rft.date=&rft.volume=">Neutronic feasibility of civil marine small modular reactor core using
           mixed D 2 O + H 2 O coolant
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Syed Bahauddin Alam, Bader Almutairi, Dinesh Kumar, Shakhawat H. Tanim, Safwan Jaradat, Cameron S. Goodwin, Kirk D. Atkinson, Geoffrey T. ParksAbstractIn an effort to decarbonize the marine sector, there are growing interests in replacing the contemporary, traditional propulsion systems with nuclear propulsion systems. The latter system allows freight ships to have longer intervals before refueling; subsequently, lower fuel costs, and minimal carbon emissions. Nonetheless, nuclear propulsion systems have remained largely confined to military vessels. It is highly desirable that a civil marine core not to use highly enriched uranium, but it is then a challenge to achieve long core lifetime while maintaining reactivity control and acceptable power distributions in the core. The objective of this study is to design a civil marine core type of single batch small modular reactor (SMR) with low enriched uranium (LEU) (
       
  • Modeling of fuel-cladding stresses in porous UC/SIC fuel pins
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Andrey Maximenko, Oleg Izhvanov, Eugene A. OlevskyAbstractMaximum stress development in nuclear fuel pins during their exploitation in the gas-cooled Gen-IV reactors is an important parameter for analysis of fuel-cladding compatibility, especially in the case of testing of new fuel or cladding materials. This article is devoted to the stress evaluation in the combination of uranium carbide fuel and silicon carbide cladding. Swelling fuel exerts the largest pressure on the cladding when initial gap between fuel and cladding has been already closed. Numerical modeling predicts rapid stress increase in this regime. To alleviate these stresses annular-shaped kernel-based bi-modal porous pellets have been designed in the General Atomics Co. with small pores in the spherical agglomerates (kernels) and comparatively large pores between kernels. A new numerical approach is developed to take into account specific features of the mechanical behavior of this new fuel. The modeling predicts stress decrease in direct proportion to porosity of the fuel, but the most significant stress reduction is expected with the upper bound increase of reactor power accompanied by the temperature increase in the pellet. Because of the temperature non-uniformity, stress relaxation in the pellet is sensitive not only to the mean value of porosity but also to its distribution along the pellet radius. Comparatively cold regions near the cladding must have higher porosity than high-temperature core regions for efficient stress relaxation.
       
  • Characterization and prediction of flow-conditions in the hot-leg of PWR
           during loss of coolant accident
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Kumar Samal, Suman GhoshNumerical and computational attempts are made to investigate and characterize the flow-condition in the hot-leg of a Pressurized Water Reactor (PWR) during the loss of coolant accident (LOCA). Finite Volume-based Volume of Fluid (VOF) model is employed for transient numerical simulations of the two-phase hydrodynamics in hot-leg during LOCA. The turbulence effect is captured using ‘k-w’ model. Effect of individual fluid-flowrates and stored water (initial water-level), on the developed counter-current two-phase flow-structures in hot-leg are extensively studied. Variations of the spatial distribution of phases with time are evaluated. Flow-structures in the hot-leg are characterized and parameterized in terms of statistical parameters extracted from the time variation of pressure drop (PD) and volume fraction (VF) across the flow domain. Computational intelligence-based methodologies are developed to capture the complex nonlinear relationship between obtained flow-conditions (occurrence or absence of plugging/blocking) in the hot-leg and the extracted statistical parameters. A computational approach is also developed to find the dependency of occurrence or absence of plugging on the physical working conditions. Plugging is found to be most responsive to the gas-flowrate. Plugging/blocking is easily occurred at high gas-flowrate. It may occur even at moderate gas-flowrate when initial stored water-level in hot-leg crosses the threshold limit. It is found that the developed methodologies are able to perfectly capture the relationship between the flow-condition (occurrence or absence of plugging) in hot-leg and the said statistical parameters. The relation between the flow-condition in the hot-leg and the working conditions is also perfectly correlated by the developed computational approach.Graphical abstractGraphical abstract for this article
       
  • Development and verification of PWR-core nuclear design code system
           NECP-Bamboo: Part III: Bamboo-Transient
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Yunzhao Li, Tao He, Boning Liang, Hongchun Wu, Jiahe Bai, Jiewei YangAbstractTo analyze the rapid change of Pressurized Water Reactor (PWR) core states in the time scale of seconds, a transient analysis code named Bamboo-Transient has been developed to provide neutronics and thermal-hydraulics coupled simulation with distinguished prompt and delayed neutrons. These transient processes can be initialized by control rod movements, core inlet coolant temperature perturbation and/or inlet coolant flow perturbation. Different from the existing codes, it adopts the Predictor-Corrector Improved Quasi-Static Method (PC-IQS) to ensure the computational efficiency, and employs the heterogeneous Variational Nodal Method as its three-dimensional neutron diffusion solver. Three advantages can be provided by HetVNM due to the elimination of the transverse integration approximation. Firstly, control rod cusping effect can be eliminated by directly treating heterogeneous cross sections and discontinuity factors. Secondly, physical adjoint neutron flux can also be obtained by constructing and solving the multi-group adjoint equations with discontinuous factors. Thirdly, the detailed nodal precursor density distribution can be directly provided by the detailed nodal fluxes from VNM to ensure the kinetics calculation accuracy. In this paper, these methods and the Bamboo-Transient code were verified by the NEACRP rod ejection accident and uncontrolled control rod withdraw accident in BEAVRS reactor core.
       
  • Optimal Main Heat Transport system configuration for a nuclear power plant
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Avinash J. Gaikwad, Naresh K. Maheshwari, K. Obaidurrahman, Aniket Gupta, Santosh K. PradhanAbstractSelection of an optimal Main Heat Transport (MHT) system configuration for a nuclear power plant (NPP) is an activity of high significance, as it influences the safety and operational aspects. This paper brings out the methodology for selection of optimal MHT configuration of NPPs which provides enhanced safety along with ease of operation. The criterion for selection of NPP-MHT configuration have been developed for enhancing safety and ease of operation. As a case study, an innovative MHT configuration has been arrived for a channel type Boiling Water Reactor (BWR) which further improves safety. Some of the multiple failure Design Extension Conditions (DEC) scenarios without SCRAM which lead to severe core damage, for commonly deployed MHT configurations, do not encounter any core heat up with this optimal MHT configuration, for a prolonged duration with Emergency Core Cooling System (ECCS). To demonstrate the effect of MHT system configuration on safety, the Natural Circulation Advanced Boiling Water Reactor (NCABWR) has been considered in this study. Though NCABWR is considered for demonstration purpose, similar methodology can be used for devising or conceptualizing an optimal MHT system configuration for any other NPP. The NCABWR is a four inter-connected loop channel type BWR with natural circulation based passive MHT cooling at all power levels. Following the recent DEC events, more emphasis is laid on very low or no radiological impact in the public domain and no requirement of evacuation in the spirit of “more good than harm”. New NPPs like NCABWR come closer to these requirements. With the mostly passive safety systems, NCABWR can avoid and mitigate DEC. The MHT system configuration for such NPPs need to be selected optimally based on the safety enhancement. In this study, optimal NCABWR MHT configuration has been further evolved based on all the Postulated Initiating Events (PIEs) including multiple failures, such as Anticipated Transients without Scram (ATWS), 200% inlet header break Loss of Coolant Accident (LOCA) without reactor trip etc. Several MHT configurations were conceptualized based on comprehensive overall design, layout requirements and reactor specific PIEs like a LOCA with a postulated break in the four partitioned ECCS header compartment. LOCA in one of the four ECCS header compartment required a mandatory interconnection among all the four MHT loops through the Common Reactor Inlet Header (CRIH) without any partition. This analysis and related issues signaled towards requirement of development of an optimal MHT configuration. The objective of the present study is to configure the MHT components and the ECCS in such an optimal formation so as to provide an enhanced performance in terms of heat removal during adverse conditions. This paper brings out clearly the important aspects needed to be considered while selecting the MHT configuration for the NPPs with enhanced safety even for ATWS-DEC scenarios, without compromising the ease of operation. The development of advanced accident tolerant fuels with improved performance, an improved/optimal MHT NPP configuration is inevitable to ensure safety.
       
  • Design of core catchers for sodium cooled FBRs – Challenges
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Vidhyasagar Jhade, Prabhat Kumar Shukla, A. Jasmin Sudha, Anil Kumar Sharma, E. Hemanth Rao, Sanjay Kumar Das, G. Lydia, D. Ponraju, B.K Nashine, P. SelvarajAbstractThe whole core meltdown scenario in sodium cooled fast reactors is considered under design extension criteria and has a very low probability of occurrence. To mitigate such a hypothetical severe accident in fast breeder reactors, a core catcher has been provided to accommodate the core debris within the primary containment boundary. In sodium cooled fast breeder reactor (FBR), an in-vessel core catcher, is provided to receive and disperse the fuel debris arising out of core meltdown during Hypothetical Core Disruptive Accident (HCDA). The core catcher prevents hot debris from reaching and damaging the main vessel. It also enables adequate heat transfer by natural convection to keep debris in stable conditions. The present study focuses on design of core catcher with sacrificial barriers and techniques to improve natural cooling of debris on the core catcher for FBRs. Studies are carried out towards the selection of suitable sacrificial material for core catcher, numerical analysis on multilayer and thermal-hydraulic analysis on multi jets core catcher concepts for selection of best suitable technique for debris bed coolability. The core catcher concept with sacrificial barrier and cooling pipes is found to possess enhanced coolability of debris bed. Experimental studies on compatibility of sacrificial ceramic material with sodium indicate strong dependence on microstructure. It has also been found that the multilayer core catcher with MgO as delay bed of 50 mm thickness is adequate to reduce the bottom temperature of core catcher below the design safety limit. Judicious mix of all these findings is expected to have unique design to accommodate whole core meltdown in future fast reactors.
       
  • Computer code analysis of irradiation performance of axially heterogeneous
           mixed oxide fuel elements attaining high burnup in a fast reactor
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Tomoyuki Uwaba, Keisuke Yokoyama, Junichi Nemoto, Ikuo Ishitani, Masahiro Ito, Michel PelletierAbstractCoupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements attaining peak burnups of 98–124 GWd/t in a fast reactor were conducted. Postirradiation experiments (PIEs) of the fuel elements revealed local concentration of Cesium (Cs) near the interfaces between MOX fuel and blanket columns including internal blankets. The PIEs also showed an increase in the cladding diameters of the fuel elements. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cesium-Uranium oxide was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were therefore caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.
       
  • The critical sticking velocity of non-spherical graphite particles: A
           numerical study and validation
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Zhu Fang, Yiyang Zhang, Mingzhe Wei, Shumiao Zhao, Libin Sun, Xinxin WuAbstractFor high temperature gas-cooled reactors (HTGR), the accumulation of the micro-sized graphite dust in the primary loop is a major concern during a potential accident such as the water ingress and loss of coolant accidents. The critical sticking velocity and the restitution coefficient are important for estimating the deposition rate of graphite particles. However, as far as the authors know, the experimental data of the graphite particle–wall collision is limited. Besides, the traditional theoretical models based on approximately spherical particles are not applicable to disk-like graphite particles. In this study, we use the finite element method (FEM) to simulate the particle–wall collision process at a microcosmic scale. The surface adhesion is determined by the van der Waals force between the two contact surfaces. The viscoelastic damping behavior of the material is described by the Maxwell standard model. At lower incident velocities, the effect of the surface adhesion becomes more important, which not only directly contributes to the energy loss by the adhesion work but also indirectly contributes by increasing the strain rate near the interface. For spherical particles, the method shows a good agreement with the experimental data. Then the FEM model is extended to study wall collisions of micro-sized dike-like graphite particles and both the restitution coefficient curve and the critical sticking velocity are obtained. The results show that the critical sticking velocity increases with the decreases of the aspect ratio and the particle size. A correlation for predicting the critical sticking velocity of disk-like graphite particles is proposed. Our work could provide a realistic particle–wall model for predicting the graphite particle deposition rate using Eulerian-Lagrangian CFD method.
       
  • Development of coupled neutronics and fuel performance analysis
           capabilities between Serpent and TRANSURANUS
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Heikki Suikkanen, Ville Rintala, Arndt Schubert, Paul Van UffelenMonte Carlo reactor physics code Serpent is coupled with the fuel performance code TRANSURANUS for steady state analyses including fuel depletion calculations. A two-way coupling is developed, such that the codes exchange data with each other and the solution is obtained as the result of iterations between the two codes. The coupling scheme is external, based on data exchange via output files, such that no source-level modifications are needed on either of the coupled codes. A separate driver program is written to run the coupled calculation including exchange of data, monitoring and determination of convergence. Data provided by the fuel performance code includes the radial temperature distribution and the radius changes in the axial slices of a fuel pin while the reactor physics code provides the linear power and fast neutron flux along with the form of the radial power density in the corresponding axial slices of the pin. The coupling is demonstrated in a calculation case based on a fuel performance benchmark for a burnable absorber rod with gadolinium. The results compared with stand-alone fuel performance calculations demonstrate the capabilities of the coupled calculation system to enhance fuel performance analyses via higher-fidelity neutronic solver capable of providing an accurate neutron flux solution for the rod of interest.Graphical abstractGraphical abstract for this article
       
  • Mechanism of loop seal clearance occurring at U-shaped pipe with air-water
           two-phase flow
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Jun-young Kang, Muhammed Mufazzal Hossen, Byoung-Uhn Bae, Yeon-Sik Kim, Jae Bong Lee, Yong-Chul Shin, Woo Jin Jeon, Kyoung-Ho KangAbstractMain purposes of the LOCAL (LOop seal Clearance separate effect test of ATLAS) facility are to investigate the effect of cross-over leg (COL) diameter at the loop seal clearance (LSC), to evaluate the two-phase flow regime through high speed visualization, and to suggest the mechanistic models for the LSC prediction at Small Break-Loss of Coolant Accident (SB-LOCA) in Light Water Reactors (LWRs). In this paper, we introduce a key mechanism of the onset of the LSC occurring at U-shaped COL with air-water two-phase flow condition. The onset of LSC is coupled with the flow regime transition at vertical upward and horizontal section of COL. Slug flow with single Taylor bubble results in decrease of pressure difference at vertical upward section and causes rapid flow regime transition at horizontal section from the stratified to the slug. The critical velocity of gas leading to flow regime transition at horizontal section can explain the onset of LSC and high speed visualization can provide the clear understanding of LSC phenomena at the local point of view.
       
  • Best-estimate analysis for a MSGTR accident of CANDU-6 plants using the
           MAAP-ISAAC code
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Chul-Kyu Lim, Sang-Koo Han, Sook-Kwan Kim, Dong-Sik Jin, Bong-Jin Ko, Yong-Ho Hong, Kwang-Il Ahn, Soo-Yong Park, Seok-Won Hwang, Paul McMinn, Chan-Young PaikAbstractIn CANDU-6 type reactors, a multiple steam generator tube rupture (MSGTR) accident is characterized as a reactor building (RB) bypass scenario. Although the probability of an MSGTR accident in the CANDU-6 type Wolsong plants is very low, a direct release of radioactive nuclides to the environment can cause severe radiation exposure to residents around the plants. For this reason, a best-estimate analysis has been carried out based on the currently available best-practice knowledge in the Wolsong plants. The present study is divided into three steps. First, the existing parameter file of generic CANDU-6 plants for the MAAP-ISAAC 4.03 code was revised to incorporate its design-specific features for the Wolsong plants. A comparative analysis between the existing and modified parameter files showed that the progress of a severe accident was delayed or mitigated in the case of the modified parameter file. Secondly, variables related to the steam generator decontamination factors (DFs) were improved. It was confirmed that applying the steam generator DFs based on the ARTIST test significantly reduced the release fraction of each nuclide element to the environment. Finally, severe accident management guidance (SAMG) developed for the Wolsong plants was used to manage the progression of severe accidents and to limit the release of fission products during a MSGTR accident. A mitigation action of closing the main steam safety valves (MSSVs) after meeting a SAMG entry condition can significantly reduce the amount of released fission products compared with simply supplying feedwater to a broken steam generator (SG). To meet the domestic regulatory requirement for the release of Cs-137 to the environment (100 TBq), it was identified that an operator should supply feedwater to the broken SG within 1.5 h after meeting the SAMG entry conditions. The best-estimate analysis methodology and relevant accident management strategy for the mitigation of severe accidents, which have been developed through the study, can be applied to all similar CANDU plants and provide valuable insights into Level 2 probabilistic safety assessment and development of relevant SAMG.
       
  • Numerical investigation of flow blockage accident in SFR fuel assembly
    • Abstract: Publication date: 1 April 2020Source: Nuclear Engineering and Design, Volume 359Author(s): Xiang Chai, Lei Zhao, Wenjun Hu, Yun Yang, Xiaojing Liu, Jinbiao Xiong, Xu ChengAbstractSodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declare. Due to the good thermo-physical properties of sodium, SFR has drawn a lot of attentions and there are considerable design experiences related to SFR. In the design of SFR, the pitch to diameter ratio (P/D) is kept small. Debris may accumulate in the flow channels and result in the flow blockage accident due to the corrosion of structural material. If single or several sub-channels are partially or totally occupied by debris, flow area will be suddenly reduced and a recirculation region can be expected downstream the blockage. Coolant temperature in this region can be greatly increased which may threaten the integrity of fuel assembly. In this paper, RANS method is employed to evaluate the impaired heat transfer process governed by blockages formed in a sodium-cooled 61-pin wire-wrapped rod bundle. The simulation results were firstly validated against the experimental data obtained from the literature. A good agreement demonstrated the capability of the employed approach to evaluate turbulent properties of sodium flow as well as the impaired heat transfer characteristics caused by blockages. With the validated method, the influence from blockages are systematically evaluated in which blockage size and its location are varied. Simulation results clearly indicate that a sharp increase of cladding and coolant temperature near the blockage can be expected even if the blockage area is not quite large. The local effect from blockage is also identified in the simulation. The variance of cladding and coolant temperature tends to vanish downstream the blockage and it is not feasible to detect the existence of blockage based on cladding temperature near the outlet even if the blockage area is sufficiently large.
       
 
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