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Publisher: Elsevier   (Total: 3161 journals)

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Showing 1 - 200 of 3161 Journals sorted alphabetically
A Practical Logic of Cognitive Systems     Full-text available via subscription   (Followers: 9)
AASRI Procedia     Open Access   (Followers: 15)
Academic Pediatrics     Hybrid Journal   (Followers: 33, SJR: 1.655, CiteScore: 2)
Academic Radiology     Hybrid Journal   (Followers: 23, SJR: 1.015, CiteScore: 2)
Accident Analysis & Prevention     Partially Free   (Followers: 94, SJR: 1.462, CiteScore: 3)
Accounting Forum     Hybrid Journal   (Followers: 25, SJR: 0.932, CiteScore: 2)
Accounting, Organizations and Society     Hybrid Journal   (Followers: 34, SJR: 1.771, CiteScore: 3)
Achievements in the Life Sciences     Open Access   (Followers: 5)
Acta Anaesthesiologica Taiwanica     Open Access   (Followers: 7)
Acta Astronautica     Hybrid Journal   (Followers: 411, SJR: 0.758, CiteScore: 2)
Acta Automatica Sinica     Full-text available via subscription   (Followers: 2)
Acta Biomaterialia     Hybrid Journal   (Followers: 27, SJR: 1.967, CiteScore: 7)
Acta Colombiana de Cuidado Intensivo     Full-text available via subscription   (Followers: 2)
Acta de Investigación Psicológica     Open Access   (Followers: 3)
Acta Ecologica Sinica     Open Access   (Followers: 10, SJR: 0.18, CiteScore: 1)
Acta Haematologica Polonica     Free   (Followers: 1, SJR: 0.128, CiteScore: 0)
Acta Histochemica     Hybrid Journal   (Followers: 3, SJR: 0.661, CiteScore: 2)
Acta Materialia     Hybrid Journal   (Followers: 250, 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: 1, SJR: 1.793, CiteScore: 6)
Acta Poética     Open Access   (Followers: 4, SJR: 0.101, CiteScore: 0)
Acta Psychologica     Hybrid Journal   (Followers: 27, 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  
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: 6, SJR: 0.19, CiteScore: 0)
Actualites Pharmaceutiques Hospitalieres     Full-text available via subscription   (Followers: 3)
Acupuncture and Related Therapies     Hybrid Journal   (Followers: 6)
Acute Pain     Full-text available via subscription   (Followers: 14, SJR: 2.671, CiteScore: 5)
Ad Hoc Networks     Hybrid Journal   (Followers: 11, SJR: 0.53, CiteScore: 4)
Addictive Behaviors     Hybrid Journal   (Followers: 16, SJR: 1.29, CiteScore: 3)
Addictive Behaviors Reports     Open Access   (Followers: 8, SJR: 0.755, CiteScore: 2)
Additive Manufacturing     Hybrid Journal   (Followers: 9, SJR: 2.611, CiteScore: 8)
Additives for Polymers     Full-text available via subscription   (Followers: 22)
Advanced Drug Delivery Reviews     Hybrid Journal   (Followers: 147, SJR: 4.09, CiteScore: 13)
Advanced Engineering Informatics     Hybrid Journal   (Followers: 11, SJR: 1.167, CiteScore: 4)
Advanced Powder Technology     Hybrid Journal   (Followers: 16, SJR: 0.694, CiteScore: 3)
Advances in Accounting     Hybrid Journal   (Followers: 8, SJR: 0.277, CiteScore: 1)
Advances in Agronomy     Full-text available via subscription   (Followers: 12, SJR: 2.384, CiteScore: 5)
Advances in Anesthesia     Full-text available via subscription   (Followers: 28, 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: 10, SJR: 0.992, CiteScore: 1)
Advances in Applied Mechanics     Full-text available via subscription   (Followers: 11, SJR: 1.551, CiteScore: 4)
Advances in Applied Microbiology     Full-text available via subscription   (Followers: 22, SJR: 2.089, CiteScore: 5)
Advances In Atomic, Molecular, and Optical Physics     Full-text available via subscription   (Followers: 14, 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: 2, SJR: 0.686, CiteScore: 2)
Advances in Cancer Research     Full-text available via subscription   (Followers: 31, SJR: 3.043, CiteScore: 6)
Advances in Carbohydrate Chemistry and Biochemistry     Full-text available via subscription   (Followers: 8, 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: 3)
Advances in Cellular and Molecular Biology of Membranes and Organelles     Full-text available via subscription   (Followers: 12)
Advances in Chemical Engineering     Full-text available via subscription   (Followers: 27, SJR: 0.156, CiteScore: 1)
Advances in Child Development and Behavior     Full-text available via subscription   (Followers: 10, SJR: 0.713, CiteScore: 1)
Advances in Chronic Kidney Disease     Full-text available via subscription   (Followers: 10, SJR: 1.316, CiteScore: 2)
Advances in Clinical Chemistry     Full-text available via subscription   (Followers: 29, SJR: 1.562, CiteScore: 3)
Advances in Colloid and Interface Science     Full-text available via subscription   (Followers: 19, 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: 15)
Advances in Developmental Biology     Full-text available via subscription   (Followers: 11)
Advances in Digestive Medicine     Open Access   (Followers: 9)
Advances in DNA Sequence-Specific Agents     Full-text available via subscription   (Followers: 5)
Advances in Drug Research     Full-text available via subscription   (Followers: 24)
Advances in Ecological Research     Full-text available via subscription   (Followers: 44, SJR: 2.524, CiteScore: 4)
Advances in Engineering Software     Hybrid Journal   (Followers: 28, SJR: 1.159, CiteScore: 4)
Advances in Experimental Biology     Full-text available via subscription   (Followers: 7)
Advances in Experimental Social Psychology     Full-text available via subscription   (Followers: 44, SJR: 5.39, CiteScore: 8)
Advances in Exploration Geophysics     Full-text available via subscription   (Followers: 1)
Advances in Fluorine Science     Full-text available via subscription   (Followers: 9)
Advances in Food and Nutrition Research     Full-text available via subscription   (Followers: 56, SJR: 0.591, CiteScore: 2)
Advances in Fuel Cells     Full-text available via subscription   (Followers: 16)
Advances in Genetics     Full-text available via subscription   (Followers: 16, SJR: 1.354, CiteScore: 4)
Advances in Genome Biology     Full-text available via subscription   (Followers: 8, SJR: 12.74, CiteScore: 13)
Advances in Geophysics     Full-text available via subscription   (Followers: 6, SJR: 1.193, CiteScore: 3)
Advances in Heat Transfer     Full-text available via subscription   (Followers: 21, SJR: 0.368, CiteScore: 1)
Advances in Heterocyclic Chemistry     Full-text available via subscription   (Followers: 12, SJR: 0.749, CiteScore: 3)
Advances in Human Factors/Ergonomics     Full-text available via subscription   (Followers: 23)
Advances in Imaging and Electron Physics     Full-text available via subscription   (Followers: 2, SJR: 0.193, CiteScore: 0)
Advances in Immunology     Full-text available via subscription   (Followers: 36, SJR: 4.433, CiteScore: 6)
Advances in Inorganic Chemistry     Full-text available via subscription   (Followers: 8, 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: 8, 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: 9)
Advances in Marine Biology     Full-text available via subscription   (Followers: 16, SJR: 0.88, CiteScore: 2)
Advances in Mathematics     Full-text available via subscription   (Followers: 11, SJR: 3.027, CiteScore: 2)
Advances in Medical Sciences     Hybrid Journal   (Followers: 6, SJR: 0.694, CiteScore: 2)
Advances in Medicinal Chemistry     Full-text available via subscription   (Followers: 5)
Advances in Microbial Physiology     Full-text available via subscription   (Followers: 4, SJR: 1.158, CiteScore: 3)
Advances in Molecular and Cell Biology     Full-text available via subscription   (Followers: 21)
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: 3)
Advances in Oncobiology     Full-text available via subscription   (Followers: 1)
Advances in Organ Biology     Full-text available via subscription   (Followers: 1)
Advances in Organometallic Chemistry     Full-text available via subscription   (Followers: 17, 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: 5, SJR: 1.579, CiteScore: 4)
Advances in Pediatrics     Full-text available via subscription   (Followers: 24, SJR: 0.461, CiteScore: 1)
Advances in Pharmaceutical Sciences     Full-text available via subscription   (Followers: 10)
Advances in Pharmacology     Full-text available via subscription   (Followers: 16, SJR: 1.536, CiteScore: 3)
Advances in Physical Organic Chemistry     Full-text available via subscription   (Followers: 8, 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: 9)
Advances in Plant Pathology     Full-text available via subscription   (Followers: 5)
Advances in Porous Media     Full-text available via subscription   (Followers: 5)
Advances in Protein Chemistry     Full-text available via subscription   (Followers: 18)
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: 62)
Advances in Quantum Chemistry     Full-text available via subscription   (Followers: 6, SJR: 0.371, CiteScore: 1)
Advances in Radiation Oncology     Open Access   (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: 5)
Advances in Space Research     Full-text available via subscription   (Followers: 397, SJR: 0.569, CiteScore: 2)
Advances in Structural Biology     Full-text available via subscription   (Followers: 5)
Advances in Surgery     Full-text available via subscription   (Followers: 10, SJR: 0.555, CiteScore: 2)
Advances in the Study of Behavior     Full-text available via subscription   (Followers: 31, SJR: 2.208, CiteScore: 4)
Advances in Veterinary Medicine     Full-text available via subscription   (Followers: 17)
Advances in Veterinary Science and Comparative Medicine     Full-text available via subscription   (Followers: 13)
Advances in Virus Research     Full-text available via subscription   (Followers: 5, SJR: 2.262, CiteScore: 5)
Advances in Water Resources     Hybrid Journal   (Followers: 47, SJR: 1.551, CiteScore: 3)
Aeolian Research     Hybrid Journal   (Followers: 6, SJR: 1.117, CiteScore: 3)
Aerospace Science and Technology     Hybrid Journal   (Followers: 341, 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: 11, SJR: 3.671, CiteScore: 9)
Aggression and Violent Behavior     Hybrid Journal   (Followers: 446, SJR: 1.238, CiteScore: 3)
Agri Gene     Hybrid Journal   (Followers: 1, SJR: 0.13, CiteScore: 0)
Agricultural and Forest Meteorology     Hybrid Journal   (Followers: 17, SJR: 1.818, CiteScore: 5)
Agricultural Systems     Hybrid Journal   (Followers: 32, SJR: 1.156, CiteScore: 4)
Agricultural Water Management     Hybrid Journal   (Followers: 44, SJR: 1.272, CiteScore: 3)
Agriculture and Agricultural Science Procedia     Open Access   (Followers: 2)
Agriculture and Natural Resources     Open Access   (Followers: 3)
Agriculture, Ecosystems & Environment     Hybrid Journal   (Followers: 57, SJR: 1.747, CiteScore: 4)
Ain Shams Engineering J.     Open Access   (Followers: 5, SJR: 0.589, CiteScore: 3)
Air Medical J.     Hybrid Journal   (Followers: 6, SJR: 0.26, CiteScore: 0)
AKCE Intl. J. of Graphs and Combinatorics     Open Access   (SJR: 0.19, CiteScore: 0)
Alcohol     Hybrid Journal   (Followers: 11, SJR: 1.153, CiteScore: 3)
Alcoholism and Drug Addiction     Open Access   (Followers: 9)
Alergologia Polska : Polish J. of Allergology     Full-text available via subscription   (Followers: 1)
Alexandria Engineering J.     Open Access   (Followers: 1, 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: 9, SJR: 0.201, CiteScore: 1)
Alzheimer's & Dementia     Hybrid Journal   (Followers: 50, SJR: 4.66, CiteScore: 10)
Alzheimer's & Dementia: Diagnosis, Assessment & Disease Monitoring     Open Access   (Followers: 4, SJR: 1.796, CiteScore: 4)
Alzheimer's & Dementia: Translational Research & Clinical Interventions     Open Access   (Followers: 4, SJR: 1.108, CiteScore: 3)
Ambulatory Pediatrics     Hybrid Journal   (Followers: 6)
American Heart J.     Hybrid Journal   (Followers: 50, SJR: 3.267, CiteScore: 4)
American J. of Cardiology     Hybrid Journal   (Followers: 54, SJR: 1.93, CiteScore: 3)
American J. of Emergency Medicine     Hybrid Journal   (Followers: 45, SJR: 0.604, CiteScore: 1)
American J. of Geriatric Pharmacotherapy     Full-text available via subscription   (Followers: 10)
American J. of Geriatric Psychiatry     Hybrid Journal   (Followers: 14, SJR: 1.524, CiteScore: 3)
American J. of Human Genetics     Hybrid Journal   (Followers: 34, SJR: 7.45, CiteScore: 8)
American J. of Infection Control     Hybrid Journal   (Followers: 28, SJR: 1.062, CiteScore: 2)
American J. of Kidney Diseases     Hybrid Journal   (Followers: 34, SJR: 2.973, CiteScore: 4)
American J. of Medicine     Hybrid Journal   (Followers: 46)
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: 205, SJR: 2.7, CiteScore: 4)
American J. of Ophthalmology     Hybrid Journal   (Followers: 62, 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: 27, SJR: 2.139, CiteScore: 4)
American J. of Preventive Medicine     Hybrid Journal   (Followers: 28, SJR: 2.164, CiteScore: 4)
American J. of Surgery     Hybrid Journal   (Followers: 38, 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: 6)
Anaerobe     Hybrid Journal   (Followers: 4, SJR: 1.144, CiteScore: 3)
Anaesthesia & Intensive Care Medicine     Full-text available via subscription   (Followers: 62, SJR: 0.138, CiteScore: 0)
Anaesthesia Critical Care & Pain Medicine     Full-text available via subscription   (Followers: 17, SJR: 0.411, CiteScore: 1)
Anales de Cirugia Vascular     Full-text available via subscription  
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: 5, SJR: 4.849, CiteScore: 10)
Analytica Chimica Acta     Hybrid Journal   (Followers: 43, SJR: 1.512, CiteScore: 5)
Analytical Biochemistry     Hybrid Journal   (Followers: 177, SJR: 0.633, CiteScore: 2)
Analytical Chemistry Research     Open Access   (Followers: 11, SJR: 0.411, CiteScore: 2)
Analytical Spectroscopy Library     Full-text available via subscription   (Followers: 11)
Anesthésie & Réanimation     Full-text available via subscription   (Followers: 2)
Anesthesiology Clinics     Full-text available via subscription   (Followers: 23, 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: 190, SJR: 1.58, CiteScore: 3)

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Journal Cover
Nuclear Energy and Technology
Number of Followers: 3  

  This is an Open Access Journal Open Access journal
ISSN (Online) 2452-3038
Published by Elsevier Homepage  [3161 journals]
  • Key results of commissioning activities for the emergency and scheduled
           cooldown system of the AES-2006 unit with the V-392M reactor plant

    • Authors: D.B. Statsura; A.G. Volnov; V.N. Shkalenkov; K.V. Zhirnov; R.M. Topchian
      Pages: 278 - 284
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): D.B. Statsura, A.G. Volnov, V.N. Shkalenkov, K.V. Zhirnov, R.M. Topchian
      The paper describes the combination of safety and normal operation functions adopted in the AES-2006 design, as illustrated by a two-channel structure of active safety systems, showing the major innovative solutions and their differences from earlier designs, such as: – The use of a pump-ejector assembly in the primary circuit emergency and scheduled cooldown and spent fuel pool cooling system; – The emergency and scheduled cooldown and spent fuel pool cooling system is designed as two fully independent channels (each channels consists of two legs, each channel leg has a capacity of 100%); – The redundancy rate is 2 × 200%. Differences in the emergency and scheduled cooldown systems are considered for the V-392M и V-320 designs. Compliance with the acceptance criteria is shown based on test results, including additional tests performed in different modes during the commissioning of the Novovoronezh NPP II Unit 1 with the V-392M reactor plant. A description and a pattern are provided for the full-scale simulator tests to justify the long-term performance of the pump-ejector assembly scheduled for 2017, including an inspection of the water-jet pump nozzle inner surface condition and the water-jet pump service life tests followed by an examination for possible damage and visible defects on the nozzle surface. As a result of the innovations, the overall reliability of safety systems has been improved. The successful completion of the life tests of the pump-ejector assembly of the emergency and planned cooling system made it possible to confirm the reliability and efficiency of the water-jet unit used for the first time in the NPP safety systems.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.001
  • Experience of commissioning of the V-392M reactor plant passive heat
           removal system

    • Authors: K.F. Galiev; S.V. Yaurov; Ye.V. Goncharov; A.S. Volnov
      Pages: 291 - 296
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): K.F. Galiev, S.V. Yaurov, Ye.V. Goncharov, A.S. Volnov
      The paper considers structural features of components in a passive heat removal system (PHRS) used for the first time in Russia in the VVER-1200 reactor. It describes the design deficiencies of the system's air gate valves and regulators detected in the process of pre-commissioning activities and making it impossible for the system to perform its design functions. The deficiencies are explained by unpredictable effects of the working fluid's thermodynamic processes on the movable parts of the equipment. The equipment deficiencies were eliminated through the installation of additional equipment retainers which respond to the system actuation and by additional debugging of components. The retainer allows keeping the air gate valves fully opened and prevents their uncontrolled closure under the action of a rapidly ascending hot air flow inside the PHRS shell. During the commissioning stage of Novovoronezh NPP II's unit 1, following an update of the equipment design, the PHRS proved to be indispensible as the means to remove the reactor core heat in conditions of an unexpected scram and has confirmed its efficiency as part of an integrated test at 75% of the reactor power. The PHRS heat exchanger total heat removing capacity was over 100 MW at the ambient air temperature of –13 °C, and the time for the PHRS to reach full capacity did not exceed 90s, which corresponds to the system's design performance.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.003
  • Testing of the ICMS input data diagnostic system at unit 1 of Novovoronezh
           NPP II

    • Authors: A.V. Semenikhin; Yu.V. Saunin; M.M. Zhuk
      Pages: 297 - 301
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): A.V. Semenikhin, Yu.V. Saunin, M.M. Zhuk
      A real-time in-core monitoring system (ICMS) input data diagnostic system was tested for the first time during the commissioning of Novovoronezh NPP II's unit 1. The system was developed by specialists of “Novovoronezhatomtechenergo”, the Novovoronezh filial of JSC “Atomtechenergo”, with participation of the Kurchatov Institute National Research Center's experts. The purpose of the diagnostic system development is to provide for the continuous monitoring of the input data validity in connection with the fact that new and upgraded ICMS designs include the performance of protective functions based on local in-core parameters. Issues are formulated to be addressed in the development of the ICMS input data diagnostic system, the structure of the developed system is shown and basic data on its operation is given. The paper presents the key system test results which have confirmed that the system is capable to detect in real time the measuring monitoring channels with invalid readings and display respective diagnostic information. The considered ICMS system may serve a prototype for the development of similar systems, to be completed with novel functions, in measuring channels of the NPP monitoring and control systems to enable an integrated real-time analysis of data for identification of the causes for, early detection and prediction of developing defects.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.004
  • New generation refueling machine information and control system

    • Authors: V.V. Korobkin; V.P. Povarov
      Pages: 302 - 306
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): V.V. Korobkin, V.P. Povarov
      The main distinctive feature in the creation of the new nuclear fuel handling complex at the Novovoronezh NPP is the deep integration of the information and control system (ICS) and the electrical equipment complex of the refueling machine (RM) into a single complex. The structure of the multi-level multi-processor and multi-network ICS provides high fault-tolerance and functional safety and fully corresponds to the principle of no single point of failure, even when operating in single-channel mode. Control of the RM during refueling is carried out with the participation of a human operator, who is supported by an intelligent interface. The operator makes decisions on certain actions to control the RM and the complex as a whole, monitors the control process and takes decisions on preventing abnormal or emergency situations. In fact, the ICS is a distributed system that implements the entire control cycle and contains information, processing and control parts. In each part, communication devices realize data connections between the components on the entire graph or on a common bus. The development result is a new generation RM ICS that differs from the existing ones by innovative methodical, algorithmic, hardware–software and design-technological solutions, which reduce the time and increase the safety of nuclear fuel handling in VVER reactors. At the same time, the introduction of the RM ICS increases the installed capacity utilization factor (ICUF) of the power unit by about 1.66%, which ensures additional electricity generation and supply for consumers.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.005
  • Specific features of initial fuel load of the innovative power unit under
           AES-2006 project

    • Authors: A.N. Prytkov; A.B. Tereshchenko; E.I. Golubev; I.A. Boev
      Pages: 307 - 312
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): A.N. Prytkov, A.B. Tereshchenko, E.I. Golubev, I.A. Boev
      Experience of initial fuel loading of VVER-type reactors was analyzed prior to initiation of the “first criticality” phase of Novovoronezh NPP-II unit no. 1 commissioning. The analysis demonstrated a number of negative factors which may develop during the commissioning phase under discussion, for instance those associated with fuel assembly stability. Special measures were undertaken to ensure safe initial fuel loading with simultaneous use of loaded and dummy assemblies. Monitoring of deformation, flushing and verification of dummy fuel assemblies were applied for ensuring safe first fuel loading with simultaneous loading of fuel and dummy assemblies in the reactor. Conventional method of nuclear fuel loading ensuring resistance of partially loaded reactor core against internal and external disturbances of natural and man-inflicted character (in particular, against seismic effects) was refined taking into account the revealed issues and the experience of start-up of new power units by the use in the implementation of initial loading of regular nuclear reactor core with fuel assembly imitators. Simultaneous loading of charged FAs and dummy FAs in the reactor core was used as applied to VVER reactors for the first time. A set of measures was suggested allowing formulating the conclusion about applicability of dummy FAs for joint use with regular FAs. Control of deformations, flushing and inspection of FA imitators for ensuring safe initial core load in the case of joint loading of FAs and dummy FAs in the reactor core were implemented. Additional equipment was implemented for controlling coolant level in the reactor core and concentration of boric acid in the process of initial loading of VVER-1200 reactor core, because low level and absence of coolant circulation in the core do not allow using standard control systems. Effects of calculation parameters and high sensitivity of detectors on the control of neutron flux in the course of implementation of nuclear fuel loading were investigated.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.006
  • Design chemistry implementation experience during the power unit start-up
           and commissioning

    • Authors: S.L. Vitkovsky; A.P. Danilov; M.G. Shchedrin; I.A. Kolyagina
      Pages: 313 - 318
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): S.L. Vitkovsky, A.P. Danilov, M.G. Shchedrin, I.A. Kolyagina
      The article covers the results of the design chemistry implementation during the commissioning of the innovative Unit 1 at the Novovoronezh NPP II equipped with a VVER-1200 reactor. The design chemistry is composed of the requirements for the primary and secondary coolant quality, recirculating coolant water (including essential service water), solutions used in safety systems as well as the requirements for technological tools maintaining their quality. Water chemistry setup operations play a significant role at all the stages of commissioning and low power testing. An analysis is made of essential system reactivation and cleansing stages, preliminary treatment technologies, primary and secondary circuit chemical water treatment, and radioactive water treatment. Some design advantages are highlighted, such as the use of reverse osmosis as one of the stages of water treatment and high-pressure filters on the bypass blowdown cleanup system. Consideration is given to some problematic issues that arose in the course of the start-up operations during the equipment depreservation, radioactive drain water treatment and in the recycling water supply system. The authors also analyze the design flaws and issues that may arise during long-lasting operation and the ways to solve them: (1) to provide a reference technology and equipment for processing radioactive drain water; (2) to exclude flushing with chlorinated hydrocarbons from the technology for depreserving the TG internal surfaces; (3) to apply water treatment with inhibitors providing calcium transport with a value close to 100% to the circulating water supply system with a cooling tower or provide liming of all additional water for southern NPPs to minimize the carbonate index.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.007
  • Results of preoperational inspection at unit 1 of Novovoronezh NPP II

    • Authors: O.V. Urazov
      Pages: 319 - 322
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): O.V. Urazov
      Preoperational inspection for the base metal and welded joint state in components and pipelines at unit 1 of Novovoronezh NPP II is discussed. Aspects involved in the inspection of the reactor vessel welded joints on the anticorrosive cladding side and the steam generator heat-exchange tubes, using innovative test systems and techniques, are considered. For the reactor vessel inspection, a dedicated computational and engineering justification, RTO-KR-UZK-15, [1] was developed, based on which manual ultrasonic testing of the VVER-1200 reactor vessel welded joints was performed from the inside, through the anticorrosive cladding. A procedure [2], developed by experts of Politest Eddy-Current Testing Center Co. Ltd. and the Scientific and Technical Center for Emergency Engineering Operations at NPPs, a branch of JSC “Concern Rosenergoatom”, and used for the first time at a Russian NPP at the stage of preoperational incoming inspection for the base metal and weld state in components and pipelines at unit 1 of Novovoronezh NPP II using Politest-PG, an innovative system for automated eddy-current testing of the steam generator heat-exchange tubes at NPPs with VVER reactors, was employed to inspect the heat-exchange tubes for all steam generators. As the result of the test, the steam generator tubes were found to contain discontinuities with a maximum depth of up to 10% of the rated wall thickness which is tolerable. The inspection results have confirmed the unit to be fit for commercial operation.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.008
  • Experience of commissioning the AES-2006 (V-392M) steam generator blowdown

    • Authors: S.V. Yaurov; K.F. Galiev; A.V. Borovoy; A.S. Volnov
      Pages: 323 - 330
      Abstract: Publication date: December 2017
      Source:Nuclear Energy and Technology, Volume 3, Issue 4
      Author(s): S.V. Yaurov, K.F. Galiev, A.V. Borovoy, A.S. Volnov
      Structural features of the AES-2006 design (V-392M reactor plant) steam generator blowdown and drainage system have been considered. Design peculiarities of components and the system as the whole have been comprehensively analyzed, and the advantages and disadvantages of the circuitry and design solutions used are presented. An apparent advantage of the system's flowchart is the maximum blowdown rate increased to 140 t/h. At the same time, issues have been identified caused by insufficient elaboration of structural elements. Based on the earlier experience of the system commissioning, a modified flowchart has been proposed for the steam generator blowdown and drainage using multi-pass valves with an electric single-turn (EST) actuator drive in the system. The flowchart modification makes it possible to reduce the specific content of metal in the system and to provide an extra space for the maintenance of the system's regenerative heat exchanger in a pressurized shell, to use eight multi-pass valves for the steam generators instead of 36 electromagnetic valves, and to cut the operating and repair costs. Modifications have been proposed to the regenerative heat exchanger design to give it a better performance, including installation of circular partitions in the lower inlet (pressure) and the upper outlet chambers. As the result, this leads to a heat exchanger with three passes in the tube space which helps achieving the required flow rate. The above updates will improve the performance and reliability of the steam generator blowdown and drainage system as far as its design functions are concerned, this to result in the secondary circuit's water chemistry arranged so that to minimize the amount of deposits on the heat-exchange surface of the PGV-1000MKP steam generators.

      PubDate: 2017-12-27T08:44:02Z
      DOI: 10.1016/j.nucet.2017.11.009
  • New generation first-of-the kind unit – VVER-1200 design features

    • Authors: V.G. Asmolov; I.N. Gusev; V.R. Kazanskiy; V.P. Povarov; D.B. Statsura
      Abstract: Publication date: Available online 11 November 2017
      Source:Nuclear Energy and Technology
      Author(s): V.G. Asmolov, I.N. Gusev, V.R. Kazanskiy, V.P. Povarov, D.B. Statsura
      The paper is concerned with the commissioning of the new generation NPP-2006 power unit with the VVER-1200 reactor. A comparison is made between the characteristics of the new NvNPP II-1 and commercial VVER-1000 power units (B-320). Some design and circuit solutions used in the NPP-2006 project were described, which made it possible to increase the installed capacity of the power unit, which was achieved, in particular, by increasing the pressure of the primary circuit by 0.5 MPa and the pressure in the steam generators by 0.6 MPa, and also by increasing the capacity of the main circulation pumps by 2000 m3/h. The main differences in the equipment and composition of passive and active safety systems of the NPP-2006 power unit are considered. A brief description of the safety systems first applied at Russian power units is given, e.g., a two-channel structure of active safety systems with redundant emergency pumps in each channel, a double containment, a core melt localization device, a passive heat removal system, etc. Due to the increased number of BRU-Ks, it was possible to increase their performance from 15 to 3 s, which significantly improved the maneuverability of the power unit in abnormal conditions. The structure of the APCS is considered, which is applied in the NPP-2006 project, using programmable technology based on the TELEPERM XS platform. The peculiarities of the power unit commissioning are analyzed, problematic issues that have arisen at various stages of the construction are revealed, some data on the tests carried out and the results of these tests are given. Finally, an analysis is made of some design drawbacks revealed during the construction and commissioning of the power unit, an evaluation of the project was made, and proposals were formulated to finalize the VVER-1200 project for consideration in the subsequent NPP projects.

      PubDate: 2017-11-16T01:13:32Z
      DOI: 10.1016/j.nucet.2017.10.003
  • Dynamic stability of the VVER-1200 power unit

    • Authors: I.N. Gusev; V.R. Kazanskiy; I.L. Vitkovsky
      Abstract: Publication date: Available online 11 November 2017
      Source:Nuclear Energy and Technology
      Author(s): I.N. Gusev, V.R. Kazanskiy, I.L. Vitkovsky
      The paper presents the results of critical experiments to study the dynamic stability of a power unit with the VVER-1200 reactor conducted as part of the pre-commissioning activities at the pilot operation stage of Novovoronezh NPP II's unit 1. The following dynamic tests were conducted: – trip of one main feedwater pump (MFP) with no standby MFP starting to operate at the power level of 100% N nom, involving a detailed analysis of the variation in the process parameters of such mode and the process dynamics, and an assessment of the test results on a full-scale simulator; – trip of one out of four reactor coolant pump sets (RCPS) in operation at the power level of 100% N nom and the reactor plant safety assessment in the context of the reactor core thermal reliability; – turbine generator (TG) load shedding to the auxiliary level with assessments for the behavior of the key reactor plant characteristics. The paper presents records for transients and safety-related process parameters, and describes the operation of the unit components and essential controls in the dynamic test process. A conclusion is made based on an analysis of the test results that the VVER-1200 unit has a high dynamic stability. The results of the dynamic stability studies for unit 1 of Novovoronezh II make it possible to provide a number of recommendations for further designs, including specifically the following: – accelerated warning protection (AWP) should be used instead of power reduction and limiting for modes with tripped main feedwater pumps; – generator-grid timing devices should be used for modes with the unit operating for auxiliary power supply; – Russian-developed software and hardware tools should be fully switched to in implementing both normal operation and safety control systems, since the adjustment of protection and interlocking algorithms used in the AREVA software and hardware package introduced at Novovoronezh II requires the developer's authorization which involves substantial time and financial expenditures.

      PubDate: 2017-11-16T01:13:32Z
      DOI: 10.1016/j.nucet.2017.10.004
  • Sorption of cations of heavy metals and radionuclides from the aqueous
           media by new synthetic zeolite-like sorbent

    • Authors: A.S. Shilina; V.D. Bakhtin; S.B. Burukhin; S.R. Askhadullin
      Abstract: Publication date: Available online 11 November 2017
      Source:Nuclear Energy and Technology
      Author(s): A.S. Shilina, V.D. Bakhtin, S.B. Burukhin, S.R. Askhadullin
      Expensive conventional sorption materials are applied for implementing control and treatment of water media of nuclear atomic plants. New synthetic alumosilicate zeolite-like adsorbent capable to efficiently purify aqueous media both of NPPs and of other facilities of the nuclear industry objects is suggested. Such sorbent properties as thermal, radiation and chemical resistance allowing performing purification of hot aqueous media without their prior cooling down at NPPs are of particular importance. The sorbent is synthesized from inexpensive raw materials produced by the domestic industry using energy-efficient production methods within single technological process. The product thus obtained has high specific surface ∼1000 m2/g, high thermal and chemical resistance and withstands temperatures up to 650 °С with maintaining its physics-chemical and sorption properties, it is resistant to aggressive media, radiation doses from 5 to 10 MGy produce no effect on its structure and sorption capacity. The sorbent demonstrates high ability for sorption of cations of heavy metals and radionuclides. Sorption capacity amounts for Ni2+ to 140 mg/g, for Cu2+ to 160 mg/g, for Fe3+ to 560 mg/g, for Cr3+ to 110 mg/g, for Cs+ to 2000 mg/g and for Sr2+ to 226 mg/g. Purification factor for 137Сs from solutions of liquid radioactive waste (LRW) is equal to 2.6. Possible mechanism for spent sorbent utilization is to expose it during 2–3 h to temperatures from 700 to 800 °C. After that this material is highly compacted changing the bulk densities of from 0.2 g/cm3 to 2.5 g/cm3 and is vitrified with closure of pores. It is assumed that the new sorbent can be applied by all facilities within lifecycles of which complex water purification process is used.

      PubDate: 2017-11-16T01:13:32Z
      DOI: 10.1016/j.nucet.2017.10.001
  • Protactinium-231 – New burnable neutron absorber

    • Authors: G.G. Kulikov; E.G. Kulikov; A.N. Shmelev; V.A. Apse
      Abstract: Publication date: Available online 11 November 2017
      Source:Nuclear Energy and Technology
      Author(s): G.G. Kulikov, E.G. Kulikov, A.N. Shmelev, V.A. Apse
      Burnable neutron absorbers such as gadolinium and erbium are used for compensating excess reactivity in nuclear reactors. Their daughter nuclides resulting from neutron absorption by erbium and gadolinium do not play important role from the viewpoint of neutronics processes occurring in the reactor core. Selection of such burnable neutron absorber, daughter nuclides of which would favorably impact fission chain reaction, is of significant interest. The aim of the present study is to investigate neutronics properties of 231Pa – the new burnable neutron absorber – and possibilities of its producing in significant quantities. The chain of isotopic transformations starting from 237Np is the analogue of the chain of isotopic transformations started from 231Pa. However, improvement of neutron-multiplication properties in the 237Np-chain can only be achieved in fast neutron spectra while in the case of 231Pa-chain positive neutron balance can be achieved both in fast and thermal neutron spectra. From this viewpoint the chain starting from 231Pa is unique. In addition, 237Np can be produced in nuclear reactors as the result of neutron radiative capture by 235U while significant amounts of 231Pa can only be produced through the threshold (n,2n) and (n,3n)-reactions on 232Th under its irradiation with super high-energy neutrons. Such neutrons with super high energies are practically absent even in fast spectrum reactors, but, however, these neutrons are available in fusion reactors. Breeding of 231Pa in fusion reactors and further use of 231Pa in nuclear power reactors can make it possible to realize potential capabilities of fusion facilities for radical increase of nuclear reactor fuel burn-up. Thus, 231Pa isotope is the new and unique burnable neutron absorber never suggested for the purpose before. Evaluated nuclear data libraries JENDL-4.0 and ENDF/B-V, as well as computer software system SCALE-4.3 were used in the implementation of the present study. The following results were obtained. (1) In contrast to conventional burnable neutron absorbers on the basis of gadolinium and erbium, the protactinium isotope suggested in this paper appears to be more attractive because it allows us not only compensating initial excess reactivity, but, also, ensuring high fuel burn-up due to good multiplication properties of its daughter nuclides. (2) Significant quantities of protactinium could be produced in hybrid fusion–fission reactors acting as sources of neutrons (not sources of energy) with parameters already achieved by the present moment by experimental facilities in USA, Japan, UK.

      PubDate: 2017-11-16T01:13:32Z
      DOI: 10.1016/j.nucet.2017.10.002
  • Metascientific foundations of understanding of status of technology

    • Authors: V.A. Kanke
      Abstract: Publication date: Available online 15 October 2017
      Source:Nuclear Energy and Technology
      Author(s): V.A. Kanke
      The article considers the status of technology as a branch of science. Technology is understood as the totality of all technical theories accepted by the scientific community. The dominant trend in understanding the status of technology is that a demarcation line is drawn between, on the one hand, science and, on the other hand, technology. In fact, there is a kind of alienation of technology from science. Researchers provide various arguments to prove the legitimacy of this alienation. As a matter of fact, there are no arguments against the separation technology from science. This asymmetry is subjected to critical analysis in the article. The author, being guided by the theory of conceptual transduction developed by him, proves that scientifically sound arguments testify to the inclusion of technology in science on the rights of a full-fledged branch in all respects. The alienation of technology from science is baseless. All arguments that downgrade the status of technology as a branch of science are erroneous. The known features characterizing technology, namely, its appeal to the phenomenon of obligation, practical orientation, consideration of the methods for inventing and handling artifacts, in no way do not call into question the scientific status of technology. The decisive mistake of those who support the separation of science from technology is that they do not consider the conceptual and methodological structure of existing branches of science. If one looks at it, it turns out that it is the same for all branches of science, including technology. First, in all cases, scientific theories are ways to manage the concepts of principles, laws and variables by methods of deduction, adduction, induction and abduction. Second, the progress of science is realized as a transition from less developed to more developed theories. Third, each branch of science develops its own potential not in isolation from other branches of science but through its interdisciplinary relations with them. These three features indicate the status of technology as a branch of science. A proper understanding of the status of technology is a factor contributing to its development.

      PubDate: 2017-10-24T19:49:53Z
      DOI: 10.1016/j.nucet.2017.09.001
  • Evaluation of effective threshold displacement energies and other data
           required for the calculation of advanced atomic displacement

    • Authors: A.Yu. Konobeyev; U. Fischer; Yu.A. Korovin; S.P. Simakov
      Abstract: Publication date: Available online 31 August 2017
      Source:Nuclear Energy and Technology
      Author(s): A.Yu. Konobeyev, U. Fischer, Yu.A. Korovin, S.P. Simakov
      Minimum displacement threshold energy, averaged displacement threshold energy, effective displacement energy, and parameters of arc-dpa equations were estimated for 70 materials from Li to U using available experimental data. Obtained data can be used for approximate calculation of the radiation damage rate for materials irradiated with neutrons in the different facilities.

      PubDate: 2017-09-07T07:59:55Z
      DOI: 10.1016/j.nucet.2017.08.007
  • Criteria of return on investment in nuclear energy

    • Authors: V.V. Kharitonov; N.N. Kosterin
      Abstract: Publication date: Available online 24 August 2017
      Source:Nuclear Energy and Technology
      Author(s): V.V. Kharitonov, N.N. Kosterin
      Analytical relationships between the investment performance criteria (net present value (NPV), levelized cost of electricity (LCOE), internal rate of return (IRR), discounted payback period (T PB), and discounted costs (Z)) and basic engineering-economic parameters of nuclear reactors (capital costs K, annual operating costs Y, annual revenue R, NPP construction T C and operation T E periods), characterizing the NPP profitability and competitiveness at the microeconomic level, are defined for the first time. The power function of discounted cash flows was used in calculations. It is shown that the joint analysis of the entire set of investment efficiency criteria (not only LCOE as it is often done) can help avoid contradictions in assessing the NPP project profitability and formulate optimal requirements on the reactor engineering and economic parameters. The obtained analytical expressions provide solutions not only of the traditional «direct problem» (assessing efficiency criteria according to the forecasted capital and operating costs and profit stream) but, which is of equal importance, the solution of the «inverse problem»: assessing restrictions on capital and operating costs, i.e. identifying «investment corridors», based on the desired values of efficiency criteria. The investment risk assessment results obtained by Monte-Carlo method are presented in order to account for the inherent uncertainties in the forecasts of long-term cash flow during the NPP construction and operation required for assessing the efficiency of investments. The calculation results of probability distributions of the investment efficiency (profitability) criteria are presented for the specified ranges of uncertainties the forecasted cash flow. It is shown that the risk of project unprofitability can be quite high. In order to reduce investment risks, it is necessary to justify the changes in basic reactor parameters (decrease in K, Y, T C and increase in R and T E) and uncertainty ranges in the initial data.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.006
  • Comparative analysis of nondestructive ASSAY techniques for 235U and 239Pu
           in structural materials at a high gamma background level

    • Authors: M.Yu. Kalenova; A.V. Ananyev; P.B. Baskov; S.V. Sklyarov; I.V. Kuznetsov
      Abstract: Publication date: Available online 24 August 2017
      Source:Nuclear Energy and Technology
      Author(s): M.Yu. Kalenova, A.V. Ananyev, P.B. Baskov, S.V. Sklyarov, I.V. Kuznetsov
      Potential techniques to identify small quantities of fissionable materials (FM) (0.001% wt.) in conditions of a high gamma background level have been reviewed and compared, and the optimal possibility for the nuclear material (NM) control in spent fuel assemblies (SFA) has been selected. It was found through numerical simulation that a system based on a passive neutron control method could be used to detect FMs indirectly when the spent nuclear fuel (SNF) burn-up and cooling time are known. Two types of detectors have been compared: 3He counters and 235U-based fission chambers. Better application prospects of 3He counters, based on the SNM-18 neutron counter, have been shown and drawbacks of passive control technique have been pointed out. An active neutron control has been found to be the best way to address the problem considered. The system's computational model shows that the signal exceeds the triple background error (both for ambient and intrinsic background from Cm isotopes) more than twelvefold. To improve the signal recording efficiency, the system has been modified to allow for irregularities in the geometrical position of structural materials (SM) in the measuring chamber. The proposed procedure makes it possible to determine in a short time the content of 239Pu, 242Cm, and 244Cm in an SFA. After the quantity of 239Pu is determined, it is possible to estimate the content of other isotopes (Am, U, Np) due to the constancy of the 239Pu mass ratio to the mass of the actinide identified.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.002
  • Statistical analysis of data on failures of the nuclear plant equipment in
           conditions of a non-homogeneous flow of events. Part 2

    • Authors: A.V. Antonov; V.A. Chepurko
      Abstract: Publication date: Available online 24 August 2017
      Source:Nuclear Energy and Technology
      Author(s): A.V. Antonov, V.A. Chepurko
      There are three stages in the process of operation of technical equipment, each with a specific trend of the failure flow parameter (FFP) behavior. During normal operation, the FFP value is approximately constant. In this case, the equipment operation process is presumably time-homogeneous, and the reliability indicators are calculated by classical methods. The FFP decreases with time in the burn-in period and it increases at the stage of aging. This means that the operating times between two successive failures at the burn-in and aging stages are not identically distributed random quantities, and the flow of failures cannot be looked upon as recurrent. Calculation of reliability characteristics shall take into account that the failure flow is time-non-homogeneous. The paper describes a method to estimate the nuclear power plant (NPP) equipment reliability indicators that makes it possible to take into account the potential non-homogeneity of the failure flow. Peculiarities of obtained statistical data on failures are identified. Application of a normalizing flow function model to calculate the required reliability indicators is described. A practical example of an analysis of data on the CPS KNK-56 component failures at Bilibino NPP is provided. Presentation of the statistical data analysis procedure described in [1] is continued.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.001
  • Quadrature formulas for integral equations of kinetics and digital

    • Authors: A.G. Yuferov
      Abstract: Publication date: Available online 23 August 2017
      Source:Nuclear Energy and Technology
      Author(s): A.G. Yuferov
      The aim of this work is to derive quadrature formulas for nuclear reactor kinetic equations in the form of Volterra integral equations of the second kind and reactimeter equations in the form of integral convolution, the kernel of which is a decay function of delayed neutron precursors (DNP) in the non-group form. The expediency of the transition to integral equations is caused by the unification of the direct (calculation of power dynamics) and the reverse (calculation of current reactivity) tasks of reactor kinetics. As a result, the solution is reduced to the calculation of the delayed neutrons integral (DNI). This eliminates the source of computational-experimental discrepancies in estimations of reactivity, which is due to the difference in computational algorithms of direct and inverse problems. The paper describes a general scheme for converting different transport equation approximations to describe the contribution of delayed neutrons by means of an integral convolution without using dynamic equations of the DNP concentration. This conversion reduces the model dimension, simplifies the software implementation, eliminates the stiffness problem of differential kinetic equations and provides the stability of calculations. The model dimension is preserved in the case of several fissile nuclides. The integral form of the equations makes it possible to use the experimental decay function in quadrature formulas, which can be identified in the operating conditions of a nuclear reactor and stored pointwise in a nongroup form without decomposition into the sum of exponentials. This eliminates the need to solve the non-linear problem of identifying group parameters of delayed neutrons and increases the adequacy of modeling. A series of quadrature formulas for the calculation of the DNI are obtained and the corresponding algorithms of a digital reactimeter and numerical simulation of the reactor kinetics are described.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.005
  • A code for 3D calculations of the output characteristics for a single-cell
           thermionic fuel element of thermionic nuclear power plants for different

    • Authors: M.A. Polous; D.I. Solovyev; V.I. Yarygin
      Abstract: Publication date: Available online 23 August 2017
      Source:Nuclear Energy and Technology
      Author(s): M.A. Polous, D.I. Solovyev, V.I. Yarygin
      R&D has been conducted by a cooperation of Rosatom State Corporation's enterprises to build a line of autonomous small nuclear power plants (SNPP) of up to 1 MWel to support government programs for the development of Russia's Arctic region. In terms of heat and electricity supplies in an installed electric power range of 10 to 100 kWel, the most attractive solution is offered by highly autonomous, compact and easy-to-maintain SNPPs with an in-core thermionic system. The key component of a thermionic NPP is a thermionic fuel element (TFE), which structurally integrates fuel and electrogenerating elements. Experimental studies and tests of thermionic plants are complex and expensive, so emphasis in the design of TNPPs is placed on mathematical simulation of physical processes taking place in the TFE. The paper considers the results of a 3D numerical simulation for the thermal and electrical characteristics of a single-cell TFE for a TNPP as part of one of the feasible SNPP designs, based on the procedure developed using COMSOL Multiphysics, an advanced software platform, and called by the authors COMSOL-EGK-SC. Initial data have been formulated to calculate a single-cell TFE, stages are described for the TFE mathematical model development in the COMSOL-EGK-SC software environment, and numerical calculation results for the thermal and electrical performance based on experimental data on the current–voltage characteristics (CVC) of a thermionic converter (TC) and the results of a neutronic calculation for the possible structure of the TNPP core as part of an SNPP are presented.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.004
  • X-ray radiography investigation of structural conditions of Fe-15Cr-35Ni-
           11 W steel irradiated by ion-plasma fluxes

    • Authors: V.G. Malynkin; E.V. Platonova
      Abstract: Publication date: Available online 23 August 2017
      Source:Nuclear Energy and Technology
      Author(s): V.G. Malynkin, E.V. Platonova
      It was found that structural-phase transformations induced by radiation in the highly doped heat resistant Fe-15Cr-35Ni-11 W alloy under the effects of treatment with ion-plasma beams differ from the transformations in steels of types 0 × 18H10T and 0 × 16N15M3B widely used in nuclear power engineering. Presence of these differences was established by performing X-ray radiography analysis, which demonstrated that additional reflections on the X-ray patterns of irradiated samples of Fe-15Cr-35Ni-11 W alloy appear from the side of large angles relative to the reflections for the initial solid solution. Detailed X-ray diffraction studies carried out by the authors showed that additional peaks appeared from the side of smaller angles in the X-ray diffraction patterns of iron-chromium alloys of type 0 × 18 (10–30) H additionally doped with Ti, Mo, Nb, Al to the amount of 1–3% and irradiated with ion-plasma beams. In both cases the phase thus formed is of isomorphic matrix type and is thermally metastable and, in contrast to 0 × 18H10T steel, Fe-15Cr-35Ni-11 W alloy undergoes softening. The analysis of published data on the possible causes inflicting similar structural-phase transformations in materials subjected to intensive ion-plasma treatment was performed. Concentrations of crystalline lattice stacking faults in Fe-15Cr-35Ni-11 W alloy and in 0 × 18H10T steel in the deformed state were determined by X-ray diffraction analysis. It was found that concentration of structural stacking faults in this state is 4 times higher for 0 × 18H10T steel, which indicates the lower stacking fault energy in this steel. Conclusion was made that the observed effects are associated with the mechanism of radiation-induced plastic deformation. Structural-phase changes in Fe-15Cr-35Ni-11 W alloy are associated with deformation by twinning, in contrast to 0 × 18H10T steel, where the observed transformations are due to slip deformation.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.08.003
  • Technology of thermal welding with ultrasonic weld joint treatment as
           applied to NPP formworks

    • Authors: S.I. Minin
      Abstract: Publication date: Available online 16 August 2017
      Source:Nuclear Energy and Technology
      Author(s): S.I. Minin
      Increasing the strength, reliability and durability of welded structures is an important problem in welding engineering. A significant influence is exerted by residual stresses. Uneven heating of the product during welding causes its uneven temperature deformation. The product material solidity prevents the free temperature deformation of its individual parts, resulting in the formation of stresses and plastic deformation of some parts of the joint metal during welding, and after cooling, welding stresses and deformations remain in the product. The paper describes an innovative technology of thermal welding with ultrasonic weld joint treatment as applied to aluminum sliding formworks used in NPP construction. Ultrasonic treatment in the process of welding greatly increases the strength of formwork weld joints by reducing their residual stresses, grain size and degassing. The structure and properties of aluminum welds become identical to the base metal. Thermal welding with ultrasonic weld joint treatment will improve the reliability of welded joints and increase their time in service. The paper presents the results of theoretical and experimental studies of ultrasonic effects on the welded joints and heat-affected zone.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.005
  • Genetic algorithms for nuclear reactor fuel load and reload optimization

    • Authors: A.V. Sobolev; A.S. Gazetdinov; D.S. Samokhin
      Abstract: Publication date: Available online 16 August 2017
      Source:Nuclear Energy and Technology
      Author(s): A.V. Sobolev, A.S. Gazetdinov, D.S. Samokhin
      Approaches are examined in the present paper to the application of genetic algorithms for optimization of initial reactor load and subsequent reloading and reshuffling of fuel assemblies in the nuclear reactor core. The issues associated with selection of the optimization criterion, which was chosen to be the nuclear fuel burnup depth, are discussed. The burnup depth is estimated after the fuel assembly is unloaded from the core, i.e. after residence in the reactor core during 3 fuel irradiation campaigns. An important aspect determining the efficiency of the use of the genetic algorithm in the problem under examination is that the neutronics calculation of the reactor core is to be performed in sufficient details allowing "feeling" the change in the location of the fuel assemblies relative to each other. The use of low-precision instrument results in the uselessness of the proposed approach to the optimization of reactor core loading. The opposite extreme, i.e. the excessive degree of details, is associated with significant increase of expended computer CPU time. In the presented paper, the TRIGEX [1,2] application software package was used in the analysis of neutronics characteristics of the reactor core providing acceptable degree of details and capable to demonstrate sensitivity of the results to the changes in the reactor load arrangement. The genetic algorithm incorporates the use of at least two basic procedures—selection and mutation. One of the most important issues in the application of the genetic algorithm is the definition of the basic concepts, namely the concepts of mutation, crossing, and specimen. The answers to these questions as applicable to the problem under discussion are provided in the present paper. In addition, the main recommendations for the organization of crossing and mutation procedures are also given. The efficiency of use of the developed model of the genetic algorithm is demonstrated by the test example of a BN type reactor. The results of the test run demonstrated that the use of the proposed approach allows searching for optimal reactor load mapping for each separate core reshuffling operation. The main objective of the performed study was to demonstrate the applicability and efficiency of the new up-to-date approach to solving the problem of fuel loading into a nuclear reactor.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.002
  • Development of approaches to estimation of risk parameters

    • Authors: M.A. Yeliseyeva; K.N. Malovik
      Abstract: Publication date: Available online 16 August 2017
      Source:Nuclear Energy and Technology
      Author(s): M.A. Yeliseyeva, K.N. Malovik
      To a great extent, safety is ensured in the design and operation of hazardous production facilities (HPF) through identifying, analyzing and predicting the risk of accidents (failures), involving, where possible, a more complete quantitative risk estimation in determination of the HPF condition [1], which is the responsibility of the Federal Service for Environmental, Technological and Nuclear Supervision (Rostechnadzor). Among the HPFs, where multifactor risks exist at the stage of design, a special place is occupied by nuclear power installations, shelf development facilities, oil and gas platforms, as well as critical infrastructure facilities as the assets essential for the healthy state of society and the national economy in conditions of impacts from the catastrophic risk factors [2–4]. The issues involved in the estimation and prediction of hazards from unfavorable situations, emergencies, accidents and failures are considered in [2,3,5–7] where the safety of HPFs is defined by two major factors: probability of an unfavorable event (situation) and the damage from such event, using different risk identification methods, including recent advances in the asymptotic theory of the probability of extreme values. To solve the risk estimation problems, issues involved in the estimation of risk parameters have been considered with different options of the HPF state graphical space interpretation. Peculiarities of estimating the risk sensitivity and the risk degree have been described and the evolution of approaches to the estimation of risk in the HPF design and operation has been shown. Big data analysis methods for risk management have been proposed.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.001
  • Analysis of attractiveness of nuclear materials as applied to the on-site
           fuel cycle of inherently safe fast reactors

    • Authors: E.M. Lvova; A.N. Chebeskov
      Abstract: Publication date: Available online 14 August 2017
      Source:Nuclear Energy and Technology
      Author(s): E.M. Lvova, A.N. Chebeskov
      As of the present moment a fairly well-established concept of “attractiveness of nuclear materials” is widely used in scientific publications. This term implies that nuclear materials which are involved in the civil fuel cycle may be used for fabricating primitive nuclear explosive devices or even nuclear weapons. This concept serves as an instrument for comparative analysis of various nuclear materials as pertains to their possible diversion for unauthorized application. Attractiveness of nuclear materials is determined in the first place by the neutronics properties inherent to these materials. These properties include the capability of the material under examination to initiate self-sustained chain reaction because otherwise this material will be absolutely unattractive for the above-mentioned purposes. Besides that, the main properties and important characteristics of nuclear materials influencing their attractiveness are the intrinsic neutron background and heat release. The present paper presents the analysis of fuel compositions involved in the fuel cycle of inherently safe BR-1200 fast reactors (BREST-1200) incorporating on-site NFC infrastructure in terms of their attractiveness. The object of investigation are the elementary systems in the form of spheres containing nuclear materials of the BR-1200 fast reactor fuel cycle both without neutron reflectors and surrounded with such reflectors made of different materials. Here, critical conditions are defined for each system for which the main properties characterizing the attractiveness of nuclear materials are calculated taking into account the reflector material and thicknesses.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.003
  • Experimental verification of neutron inelastic scattering cross section
           for iron

    • Authors: B.V. Zhuravlev; N.N. Titarenko
      Abstract: Publication date: Available online 14 August 2017
      Source:Nuclear Energy and Technology
      Author(s): B.V. Zhuravlev, N.N. Titarenko
      Results of experimental verification of results on neutron inelastic scattering cross section for iron obtained in measurements of spectra of neutrons undergoing inelastic scattering at incident neutron energies of 6.0, 7.0 and 8.0 MeV and their calculations within the framework of statistical theory of nuclear reactions and direct interactions are presented. Description is given of the methodology of measurements and simulation calculations. Spectra of neutrons undergoing inelastic scattering obtained for iron in the process of experiments are analyzed in comparison with calculated data. New measurements of neutron inelastic scattering spectra and their analysis within the framework of contemporary modeling understanding allowed formulating proposal on introduction of corrections in the indigenous BROND-2.2 library of recommended evaluated neutron data and insignificant corrections in the most recent version of the library BROND-3.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.006
  • The results of the transmutation of fission fragments in the spectrum of
           neutrons of thermal and fast reactors

    • Authors: N.V. Ivanov; Yu.A. Kazansky; G.V. Karpovich
      Abstract: Publication date: Available online 14 August 2017
      Source:Nuclear Energy and Technology
      Author(s): N.V. Ivanov, Yu.A. Kazansky, G.V. Karpovich
      Radioactivity of spent nuclear fuel (SNF) discharged from nuclear reactor core is determined during the first 100 years by fission fragments (FF), after that the main contribution in the SNF activity is made by actinides. Existing scenarios of SNF handling are based on the transmutation of minor actinides (MA) into fission fragments accomplished in fast reactors. Scenarios of transmutation of fission fragments in thermal and fast neutron spectra and time-dependent radiation characteristics are examined in the present study. Nuclide composition of fission fragments is taken from the results of simulation of burnup of 439GT fuel assembly (TVSA type) for VVER-1000 nuclear reactor during 3 years performed using MCU-5 software complex. The obtained data were used for determining starting nuclide composition for different cooling-down times prior to the beginning of transmutation (irradiation in neutron fluxes) to be input in the ORIGEN2 code. The following three options of irradiation of fission fragments are presented: transmutation without cooling down, cooling down fission fragments during 4 years prior to irradiation, cooling down fission fragments during 30 prior to irradiation. Duration of irradiation was selected to be equal to 3 and 15 years. Efficiency of transmutation was determined using time-dependent “transmutation factor” equal to the ratio of radioactivity of nuclides in the process of transmutation and after its completion to their radioactivity without transmutation. The calculated values of transmutation factors proved to be noticeable only during irradiation in reactor core: these values reached 5–10 and were dependent only on the duration of fission fragment cooling down prior to the beginning of transmutation. After removal of fission fragments from neutron flux transmutation factor decreased to unity within several years. After one hundred more years after irradiation in neutron spectrum of thermal reactor transmutation factor reduces to 0.8–0.5 depending on the duration of the transmutation process. Slight growth of transmutation factor to the values of 1.2–1.8 was observed after irradiation in fast reactor spectrum within the time interval of 200–1000 years and after 1000 years following this its reduction to the value of 0.9–0.7 was noted. The main conclusion is that purposeful incineration of fission fragments is senseless because only insignificant gain in radioactivity (a little less than by the factor two) is achieved after 1000 years. The indifference of fission fragments with regard to transmutation can be partially explained by the fraction of stable nuclides which increases with extension of the period of fission fragments cooling down. Upon completion of the cycle of fuel use it contains approximately 15% of stable nuclides among fission fragments, and after 30 years of cooling down the fraction of stable isotopes reaches 85%.

      PubDate: 2017-08-31T23:59:18Z
      DOI: 10.1016/j.nucet.2017.07.004
  • Swelling of improved 16Cr–15Ni–2Mo–Mn–Ti–V–B steel under dose
           rates from 1×10−8 to 1.7×10−6 dpa/s

    • Authors: E.A. Kinev; V.L. Panchenko
      Abstract: Publication date: Available online 7 June 2017
      Source:Nuclear Energy and Technology
      Author(s): E.A. Kinev, V.L. Panchenko
      Radiation-induced swelling negatively affects the operability of structural units of fast breeder reactor (FBR) cores. Therefore, search for new and improvement of existing steels for reducing swelling is an important task. Since 2003, type 16Cr–15Ni–2Mo–Mn–Ti–V–B steel has shown a significant increase in radiation resistance due to its improved composition and heat treatment. Specialists of the JSC INM studied swelling of 16Cr–15Ni–2Mo–Mn–Ti–V–B steel with improved composition, data on the maximum swelling temperature, average swelling rate within typical coolant temperature ranges, as well as fast reactor dose rate were obtained. The obtained results are based on swelling measurements using hydrostatic weighing and transmission microscopy. Errors in hydrostatic measurements were examined with involvement of metallography data and selection of immersion liquid. It was revealed that the average swelling rate of improved 16Cr–15Ni–2Mo–Mn–Ti–V–B steel at the maximum swelling temperature is within the range of 0.04–0.14%/dpa. Shifting of this temperature from 460 to 520°C with increase of the maximum damaging dose from 60 to 80dpa (1.3× 10− 6 and 1.7×10− 6 dpa/s, respectively), is observed. At doses below 10dpa and temperatures below 400°C the average swelling rate may reach 0.04%/dpa. At temperatures of about 600°C and irradiation doses below 50dpa the swelling rate does not exceed 0.01%/dpa during the whole period of observation.

      PubDate: 2017-06-10T16:38:07Z
      DOI: 10.1016/j.nucet.2017.05.005
  • Improving the general and ecological image of nuclear power

    • Authors: A.L. Suzdaleva
      Abstract: Publication date: Available online 7 June 2017
      Source:Nuclear Energy and Technology
      Author(s): A.L. Suzdaleva
      The purpose of this publication is to familiarize a wide range of experts with effective ways to improve the image of nuclear power installations in Russia. The negative attitude towards such installations is explained not by the danger actually posed by them but by the insufficient effectiveness of the activities for the formation of public opinion and by the already formed implicit memory. There is a traditionally negative stereotype people have about increased dangers caused by nuclear power plants. It is suggested that passive information struggle between the advocates of and opponents to the evolution of nuclear power should be replaced by active efforts to destroy the negative stereotype existing in public consciousness. The objective of active image-making is to form people's psychological attitude regarding the importance of nuclear power evolution as a life improving factor. Ways for the practical application of active image-making methods have been proposed. It is recommended to conduct an integrated analysis of the population's mass frustrations and deprivations with respect to the moral, economic and environmental aspects of social life. A conclusion has been made on the necessity of the state's participation in improving the image of nuclear power installations.

      PubDate: 2017-06-10T16:38:07Z
      DOI: 10.1016/j.nucet.2017.05.013
  • Numerical investigation of Prandtl number effect on heat transfer and
           fluid flow characteristics of a nuclear fuel element

    • Authors: R.K. Abdul Razak; Samee Mohammed; M.K. Ramis
      Abstract: Publication date: Available online 1 June 2017
      Source:Nuclear Energy and Technology
      Author(s): R.K. Abdul Razak, Samee Mohammed, M.K. Ramis
      This paper investigates the heat transfer and fluid flow characteristics of liquid metal coolants (such as Sodium, Sodium potassium, Bismuth, Lead, and Lead–bismuth) flowing over a nuclear fuel element having non-uniform internal energy generation numerically using finite difference method. The Full Navier Stokes Equations governing the flow were converted into stream function-Vorticity form and solved simultaneously along with energy equation using central finite difference scheme. For the two dimensional steady state heat conduction and Stream-Function Equation, the discretization was done in the form suitable to solve using ‘Line-by-Line Gauss-Seidel’ solution technique whereas the discretization of Vorticity transport and energy equations were done using Alternating Direction Implicit (ADI) scheme. After discretization the systems of equations were solved using ‘Thomas Algorithm’. The complete task was done by writing a computer code. The results were obtained in the form of variation of Maximum temperature in the fuel element (hot spots) and its location, mean coolant temperature at the exit .The parameters considered for the study were aspect ratio of fuel element, Ar , conduction-convection parameter Ncc , total energy generation parameter Qt , and flow Reynolds number ReH . The results obtained can be used to minimize the Maximum temperature in the fuel element (hot spots).

      PubDate: 2017-06-05T15:39:50Z
      DOI: 10.1016/j.nucet.2017.04.002
  • Monte Carlo evaluation of neutron irradiation damage to the VVER-1000 RPV

    • Authors: Seyed Fazel Ghazi Ardekani; Kamal Hadad
      Abstract: Publication date: Available online 1 June 2017
      Source:Nuclear Energy and Technology
      Author(s): Seyed Fazel Ghazi Ardekani, Kamal Hadad
      Neutron irradiation damage is of the most critical damage mechanisms in nuclear power plant's pressure vessel. Neutrons transfer their kinetic energy to target atoms of RPV that start to jump creating vacancies and interstitials known as Frenkel Pair (FP). The FPs are responsible for formation of defected clusters and microstructural modifications (e.g. phase reactions, segregations). These effects deteriorate physical and mechanical properties of the RPV steels among which are increasing the hardness and decreasing the embrittlement which results in limiting the life-time of RPV. The most sensitive location in the RPV related neutron irradiation damages is in the beltline region and adjacent to the reactor core. Welds and their heat affected zones (HAZs) in this region are of particularly importance due to higher probability flaws compared with the base metal. In this paper, Monte Carlo simulation of the detailed reactor core have led to identifying the areas maximum neutron flux in the RPV. Then the SPECTER and SRIM Monte Carlo codes are used to evaluate the neutron spectral-averaged DPA values for the beltline region of RPV.

      PubDate: 2017-06-05T15:39:50Z
      DOI: 10.1016/j.nucet.2017.04.001
  • Forecasting reliability of ShADR-32M coolant flow rate sensors

    • Authors: A.I. Pereguda; V.I. Belozerov
      Abstract: Publication date: Available online 31 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.I. Pereguda, V.I. Belozerov
      Complexity of technical systems, such as NPPs, implicates the necessity to undertake efforts for assessment of reliability of equipment, especially critical equipment influencing the reactor operation safety. Therefore, tasks associated with investigation of regularities of variations of parameters of equipment and the processes of their approaching to the conditions of equipment failures, and with the development of methodologies and algorithms for obtaining quantitative reliability indicator values with respect to progressive (parametric) failures become important. Such task is addressed in the present paper as applicable to ShAFDR-32M coolant flow rate sensors of RBMK-1000 reactor. Analysis of statistical data obtained during planned diagnostic measurements of two determining parameters of functionality of ShAFDR-32M coolant flow rate sensors (minimum values of negative half wave of the amplitude and standard deviation for the rotation period of the flow rate sensor ball) allowed developing the mathematical model of sensor parametric reliability. The process which is the superposition of elementary regeneration process and the stochastic process with independent increments will be understood as the mathematical model of coolant flow rate meter. Investigation of mathematical model of coolant flow rate sensor functioning reliability allowed obtaining in closed form the correlations between the average time of the flow rate meter operation before crossing of the preset boundary by each of the determining parameters and the probability of failure-free operation of the flow rate meter in asymptotic setting without introducing any assumptions with regard to the laws of distribution of random values. The results obtained can be easily generalized to embrace the case when the dimension of the vector of determining parameters is larger than two. Results of the investigation are applied in the calculations of quantitative indicators of parametric reliability of flow rate sensors.

      PubDate: 2017-05-31T13:45:43Z
  • Consideration of economic risks in a comparative analysis of nuclear
           technologies with different maturity levels

    • Authors: A.A. Andrianov; Yu.A. Korovin; I.S. Kuptsov; V.M. Murogov; O.N. Andrianova
      Abstract: Publication date: Available online 30 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.A. Andrianov, Yu.A. Korovin, I.S. Kuptsov, V.M. Murogov, O.N. Andrianova
      Less developed reactor technologies are characterized by a high uncertainty level of the key performance indicators, as compared to more mature options, due to lack of information on the design, operation and cost data, etc., while the expected performance of such systems is normally more attractive in comparison with more mature options. Evidently, a greater uncertainty leads to higher economic risks involved in the deployment of the respective technology. Evaluating comparatively the competitiveness and performance of reactor technologies at different maturity levels requires taking into account economic risks to balance the judgments regarding the expected performance of the considered options. A reliable basis for this is formed by the economic risk theory. Evaluation of risk indicators requires calculation of characteristics for probabilistic distributions of economic performance indicators and systemic use of statistical approaches based on Monte Carlo methods. Demonstration analysis results for the risk indicator evaluation have been discussed as applied to different economic performance indicators based on the example of a comparative analysis for two hypothetical light water reactor technologies to be considered in the selection of the most attractive option. Use of economic risk indicators for the comparative evaluation of reactor technologies appears to be helpful to decision makers not familiar with the technical characteristics and performance measures of reactor technologies but informed about the economic risk concepts. Such methodology may be employed efficiently to interpret the ranking results in a multi-criteria comparative evaluation of less and more mature reactor technologies.

      PubDate: 2017-05-31T13:45:43Z
      DOI: 10.1016/j.nucet.2017.05.012
  • Prediction of the moderator temperature field in a heavy water reactor
           based on a cellular neural network

    • Authors: S.O. Starkov; Y.N. Lavrenkov
      Abstract: Publication date: Available online 26 May 2017
      Source:Nuclear Energy and Technology
      Author(s): S.O. Starkov, Y.N. Lavrenkov
      Reactors with heavy water coolants and moderators have been used extensively in today's power industry. Monitoring of the moderator condition plays an important role in ensuring normal operation of a power plant. A cellular neural network, the architecture of which has been adapted for hardware implementation, is proposed for use in a system for prediction of the heavy water moderator temperature. A reactor model composed in accordance with the CANDU Darlington heavy water reactor design was used to form the training sample collection and to control correct operation of the neural network structure. The sample components for the adjustment and configuration of the network topology include key parameters that characterize the energy generation process in the core. The paper considers the feasibility of the temperature prediction only for the calandria's central cross-section. To solve this problem, the cellular neural network architecture has been designed, and major parts of the digital computational element and methods for their implementation based on an FPLD have also been developed. The method is described for organizing an optical coupling between individual neural modules within the network, which enables not only the restructuring of the topology in the training process, but also the assignment of priorities for the propagation of the information signals of neurons depending on the activity in a situation analysis at the neural network structure inlet. Asynchronous activation of cells was used based on an oscillating fractal network, the basis for which was a modified ring oscillator. The efficiency of training the proposed architecture using stochastic diffusion search algorithms is evaluated. A comparative analysis of the model behavior and the results of the neural network operation have shown that the use of the neural network approach is effective in safety systems of power plants.

      PubDate: 2017-05-31T13:45:43Z
      DOI: 10.1016/j.nucet.2017.05.008
  • Experimental studies into the dependences of the axial lead coolant pump
           performance on the impeller cascade parameters

    • Authors: A.V. Beznosov; A.V. Lvov; P.A. Bokov; T.A. Bokova; V.A. Razin
      Abstract: Publication date: Available online 26 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.V. Beznosov, A.V. Lvov, P.A. Bokov, T.A. Bokova, V.A. Razin
      The paper presents results of experimental studies into the dependences of the axial lead coolant pump performance (delivery, head, efficiency) on the impeller cascade parameters, including the number of blades, the cascade blade angle and the cascade solidity. The studies were conducted as applied to conditions of small and medium sized plants based on lead cooled fast neutron reactors with horizontal steam generators. The designs of such plants are now in the process of elaboration at Nizhny Novgorod State Technical University (NNSTU). The studies were conducted at NNSTU's FT-4 test facility at a lead coolant temperature of 440–500°C. In the process of investigations, the number of blades in the form of flat plates was 3, 4, 6 and 8, the cascade blade angle was in a range of 9–43°, and the cascade solidity (0.6–1.2) was changed by changing the blade section chord length. The shaft speed of the NNSTU's NSO-01 pump, onto which changeable impellers were installed, was changed in steps of 100 rev/min in an interval of 600–1100 rev/min. The blade diameter was about 200mm, and the maximum lead coolant flow rate in the course of the tests reached ∼2000t/h. The performance of 27 impellers was investigated. It is recommended that the investigation results should be used in design of axial HLMC pumps.

      PubDate: 2017-05-31T13:45:43Z
      DOI: 10.1016/j.nucet.2017.05.009
  • Computer modeling of thermal processes involving calcium, strontium and
           cesium during radioactive graphite heating in the carbon dioxide

    • Authors: N.M. Barbin; I.A. Sidash; D.I. Terentyev; S.G. Alekseyev
      Abstract: Publication date: Available online 26 May 2017
      Source:Nuclear Energy and Technology
      Author(s): N.M. Barbin, I.A. Sidash, D.I. Terentyev, S.G. Alekseyev
      The design of nuclear reactors did not provide for decommissioning solutions, and there were no safe technologies to handle irradiated reactor graphite. Decommissioning of uranium–graphite reactors is a combination of complex tasks involving the selection of appropriate methods and techniques for the radioactive graphite handling. Computer modeling of the reactor graphite reprocessing by heating in a carbon dioxide environment makes it possible to determine the behavior of radioactive elements. The behavior of Ca, Sr and Cs during radioactive graphite heating in the carbon dioxide atmosphere was studied through computer modeling. It has been found that calcium is present as vapors of Ca, CaO, CaCl and CaCl2, as Ca+ and CaO+ ions, and as condensed forms of CaCO3, CaCl2 and CaO. Strontium is present as vapors of Sr, SrO, SrCl and SrCl2, as Sr+ and SrO+ ions, and as condensed forms of SrCl2, SrCO3 and SrO. Cesium is present as vapors of Cs and CsCl, as Cs+ ions, and as a condensed phase of CsCl2. Basic reactions have been identified and their respective equilibrium constants have been determined. The data obtained has shown that formation of calcium, strontium and cesium chloride vapors takes place at temperatures of 573 to 973K. A temperature increase to 1373K leads to thermal ionization of cesium chloride and to formation of ionized cesium. As the temperature increases to 2273K, thermal ionization of strontium and cesium is observed and ionized calcium and strontium form.

      PubDate: 2017-05-31T13:45:43Z
      DOI: 10.1016/j.nucet.2017.05.006
  • Assessment of the critical condition for the operation of an IBR reactor
           with a subcritical unit in an equilibrium mode

    • Authors: A.I. Brezhnev; A.V. Gulevich O.F. Kukharchuk O.G. Fokina
      Abstract: Publication date: Available online 25 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.I. Brezhnev, A.V. Gulevich, O.F. Kukharchuk, O.G. Fokina
      There considered a system consisting of a fast neutron batch pulsed IBR type reactor and a subcritical unit (neutronically thermal). The reactor is fitted with a reactivity modulator, which provides short-term “transfer” of the system from a deep subcritical to a prompt supercritical state and back. The system is in a deep subcritical state in the intervals between pulses. Such a system is capable to operate in an equilibrium (static) mode only when a critical condition is fulfilled for the kinetic parameters describing its operation. The neutron kinetics is described as part of a two-point approximation. It is assumed that the change in the reactor reactivity at the pulse generation time takes place periodically according to a parabolic law and the reactor is deeply subcritical in the intervals between pulses. Numerical simulation of the critical condition is extremely time-consuming, and analytical representation is almost impossible due to the need for solving ordinary differential equations with variable coefficients. There proposed a methodology for approximate estimating of parameters of a coupled “batch pulsed reactor–subcritical unit” system operating in an equilibrium mode. Analytical relations have been obtained in a quadrature form to calculate the “critical” condition of such a system in the approximation of “frequently recurring” pulses, when the decay of the delayed neutron precursors in the interval between pulses can be ignored. Calculations of the “critical condition” are illustrated by an example of a laser system consisting of an IBR-type batch pulsed reactor and a subcritical neutron multiplication unit, in which fission energy is converted into laser emission energy. Critical parameters of the system were estimated using analytical relations, as well as direct numerical calculations based on the STIK program that models the neutron kinetics in the considered system in a two-point approximation. It has been shown that the results of direct calculations and the estimates based on analytical relations matched good.

      PubDate: 2017-05-26T08:07:13Z
  • Design and experimental assessment of thermal stratification effects on
           operational loading of surge line of Unit 5, Novovoronezh NPP

    • Authors: V.P. Povarov; O.V. Urazov; M.B. Bakirov; V.I. Levchuk
      Abstract: Publication date: Available online 18 May 2017
      Source:Nuclear Energy and Technology
      Author(s): V.P. Povarov, O.V. Urazov, M.B. Bakirov, V.I. Levchuk
      One of the key tasks in substantiating NPP service life extension consists in detailed examination of all factors affecting the remaining lifespan of critical NPP components. Particular attention must be paid to studying the phenomenon of thermal stratification (TS) which is the effect of coolant lamination into the “cold” and “hot” layers in horizontal pipelines when flows with different temperatures propagate at slow velocities. Special importance of this issue is explained by the fact that cyclic loads caused by TS contribute in accumulation of metal damage due to thermal fatigue and can provoke formation and accelerated growth of defects. This study is dedicated to implementing complex analysis of coolant TS observed in horizontal sections of the surge line (SL) on WWER-1000 units from the viewpoint of assessment of TS effects on the stressed-deformed conditions of metal and accumulation of cyclic damages. The experimental data on the distribution of temperature fields in horizontal SL sections, as well as cyclic loading history during several reactor fuel residence campaigns were recorded by the on-line diagnostic monitoring system put into operation on Unit 5 Novovoronezh NPP. Certain distinguishing features of TS in SL of Unit 5 were identified as the result of data analysis depending on operational modes. The most significant TS effects were observed in the control cross-section located within the first horizontal section from the pressurizer. Calculation and experimental assessment of effects of thermal loading factors on the stress-deformed state of SL demonstrated that effects of thermal stratification and thermal fatigue significantly affect the operational loading of the pipeline. It is noted that zones with maximum accumulated damage determined according to the results of calculations coincide with places where factual operational defects were detected. Procedure involving treatment of SL welded joints by the method of surface plastic deformation is suggested as the compensatory measure aimed at extending the SL residual lifespan.

      PubDate: 2017-05-21T00:38:21Z
      DOI: 10.1016/j.nucet.2017.05.001
  • Experimental estimation of the effect of contact condensation of
           steam–gas mixture on VVER passive safety systems operation

    • Authors: A.V. Morozov; A.R. Sakhipgareev
      Abstract: Publication date: Available online 17 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.V. Morozov, A.R. Sakhipgareev
      Results of experimental study of the effects of contact condensation of steam–gas mixture on the operation of VVER passive safety systems and steam generator in emergency condensation mode are presented. Contact condensation takes place when subcooled liquid is supplied in the accumulator tank of VVER reactor auxiliary passive core flooding system (the HA-2 system) in the presence of accumulated non-condensable gases. Water supplied to hydro accumulators can be used for increasing the operating time of VVER steam generator in the emergency condensation mode and for ensuring core cooling during longer time. Low liquid outlet velocity (less than 1m/s) caused by the necessity to ensure safety systems operation in passive mode constitutes the distinguishing feature of the investigated processes. Experiments were conducted on the test facility with parameters typical for the primary cooling loop of the reactor facility 24h after the accident initiation for different concentrations of gas in the steam–gas mixture. Nitrogen and helium, which replaces hydrogen for the purposes of safety assurance, were used as the non-condensable gases. It was established according to the results of experiments that the increase of concentration of non-condensable gas within the volume of HA-2 hydro accumulator model up to 45% leads to the reduction of intensity of contact condensation of steam from the steam–gas mixture by ∼29% in the experiment with nitrogen and by ∼57% in the experiment with helium. The obtained experimental data can be used for numerical simulation of emergency processes in the VVER reactor facility during operation of passive safety systems taking into account the removal of steam–gas mixture from steam generator by supplying the subcooled liquid jet into the volume HA-2 accumulator tanks.

      PubDate: 2017-05-21T00:38:21Z
      DOI: 10.1016/j.nucet.2017.05.002
  • Experimental investigation of heat and mass exchange processes during
           operation of VVER steam generator in emergency condensing mode

    • Authors: A.S. Shlyopkin; A.V. Morozov; D.S. Kalyakin; A.S. Soshkina
      Abstract: Publication date: Available online 17 May 2017
      Source:Nuclear Energy and Technology
      Author(s): A.S. Shlyopkin, A.V. Morozov, D.S. Kalyakin, A.S. Soshkina
      GE2M-PG test facility was constructed at the JSC “SSC RF-IPPE” for assessment of efficiency of VVER steam generator and investigation of effects of initial conditions of the accident on the processes of heat exchange in the steam generator (SG) tube bundle assembly. The facility was used in two series of experiments with and without removal of the steam–gas mixture (SGM) from steam generator cold collector. Functionality of the steam generator model in condensation mode with different concentrations of non-condensable gases in steam at the SG model inlet and different flow rates of steam–gas mixture removed from cold collector was investigated during the first phase of experiment. Effects of main factors of operational modes on the efficiency of heat exchange processes in the steam generator tube bundle assembly was investigated during the second phase of experiments conducted without removal of steam–gas mixture. The obtained results can be applied for verification of computer codes used in calculation simulation of emergency processes in VVER reactor facility.

      PubDate: 2017-05-21T00:38:21Z
      DOI: 10.1016/j.nucet.2017.05.003
  • Analisys of safety system pumps conditions based on their testing results

    • Authors: S.T. Leskin; V.I. Slobodchuk; A.S. Shelegov; D.Yu. Kashin
      Abstract: Publication date: Available online 17 May 2017
      Source:Nuclear Energy and Technology
      Author(s): S.T. Leskin, V.I. Slobodchuk, A.S. Shelegov, D.Yu. Kashin
      The algorithm for analyzing the conditions of emergency system pumps based on their periodical testing results is presented in the paper. The method and the algorithms are based on the presentation of the testing results in the space of the principal components. Such an approach allows representing the pump conditions in a convenient form. The parameter variation measured from the beginning of the test until the steady state conditions are achieved, i.e., the dynamic section of the curve for each parameter, is used for the analysis. Comparing the behavior curves of different technological parameters as a time function of a particular pump for different tests one can see that some sections of these curves do not change from test to test. This clearly means that these sections are not informative relative to extraction of the information concerning the defect development. These sections must be classified as some kind of “noise” and must be excluded as providing little information. On the contrary, the sections with abnormal behavior of technological parameters are more informative, and we accept these sections for further analysis. As a measure of the system uncertainty entropy H(X) is used. This new parameter is defined by the relationship H ( X ) = − ∑ i = 1 N p i log p i , where pi is the probability of the ith state of the system; N is the total number of the states of the system. The entropy allows describing the probabilistic variability of the measured data. The entropy has the maximum value if all the states of the system are equiprobable. We can use this feature of the entropy to choose the more informative time intervals of the dynamic behavior of the technological parameters. The smaller is the entropy, the more probable certain states of the system are. Thus, the most informative are those time sections, which have the maximum entropy value, i.e., the time sections for which the maximum variability of the measured data is observed. Using this approach, a matrix is constructed based on the time intervals with maximum entropy – the so-called matrix of informative criteria. To describe the conditions of the pump using different technological parameters measured in the course of the testing we need to normalize the values of the parameters by the root-mean-square deviations of the parameters. The normalized data are then used for the transformation of the original data matrix on the basis of the most informative criteria using statistical method known as the Karhunen–Loeve transform, which is also known as the principal components method. The approach was applied to processing the testing results of the emergency system pumps of the Kalinin NPP (Russia). Interesting results are obtained.

      PubDate: 2017-05-21T00:38:21Z
      DOI: 10.1016/j.nucet.2017.05.004
  • Hydrozirconium reaction in heterogeneous compositions

    • Authors: V.K. Milinchuk; E.R. Klinshpont; V.I. Belozerov; A.V. Zagorodnyaya
      Abstract: Publication date: Available online 26 April 2017
      Source:Nuclear Energy and Technology
      Author(s): V.K. Milinchuk, E.R. Klinshpont, V.I. Belozerov, A.V. Zagorodnyaya
      The research results presented in this article show the behavior of a hydrozirconium reaction for hydrogen generation at temperatures below 100°C in heterogeneous compositions containing zirconium and chemical activators (e.g., hydrated sodium metasilicate, sodium water glass, or quicklime). The hydrogen yield increases with a temperature increase up to 95°C and is about 0.1–0.2l per 1g of zirconium. Zirconium processing with γ-radiation as well as exposure to acidic and neutral aqueous media increases the hydrogen yield by about 1.2 times. A hydrozirconium reaction is caused by the chemical activators removing the passivating protective zirconium oxide ZrO2 layer from the metal surface. The possibility of a hydrozirconium reaction occurrence should be considered in the organization of technical measures to ensure hydrogen explosion protection at NPPs.

      PubDate: 2017-04-30T15:43:39Z
      DOI: 10.1016/j.nucet.2017.03.003
  • Thickness optimization of Sn–Pb alloys for experimentally measuring mass
           attenuation coefficients

    • Authors: Taranjot Kaur; Jeewan Sharma; Tejbir Singh
      Abstract: Publication date: Available online 26 April 2017
      Source:Nuclear Energy and Technology
      Author(s): Taranjot Kaur, Jeewan Sharma, Tejbir Singh
      An attempt has been made to experimentally investigate the optimum thickness in order to measure mass attenuation coefficients for some Sn–Pb alloy systems at incident photon energies 122, 511 and 662keV. The Sn–Pb alloys were synthesized with different compositions and different thicknesses using melt-quench technique and cast iron mould. The physical parameters such as mass, thickness, density have been measured for all the prepared alloys. Further, the transmitted photon spectra of Cs-137, Co-57 and Na-22 radioactive isotopes were recorded using GAMMARAD5 (scintillator detector) of dimensions 76mm×76mm with and without inserting different alloy samples between the radioactive isotopes and detector. The experimental results so obtained were compared with the theoretical ones of WinXCom and optimum thickness for measuring mass attenuation coefficients for the selected alloys has been recommended.

      PubDate: 2017-04-30T15:43:39Z
      DOI: 10.1016/j.nucet.2017.02.001
  • The technology of thermal welding of the circulation piping of NPPS
           containing the influence of ultrasound

    • Authors: S.I. Minin; A.I. Trofimov; M.A. Trofimov
      Abstract: Publication date: Available online 27 March 2017
      Source:Nuclear Energy and Technology
      Author(s): S.I. Minin, A.I. Trofimov, M.A. Trofimov
      The authors propose a thermal welding technology with ultrasound treatment for NPP circulation pipe-lines. This technology can significantly increase the strength of welded connections by reducing residual stresses, grain size and welded joint degassing. Exposure to ultrasound increases the welding speed and reduces the current consumption, thus resulting in energy saving. The paper presents the results of theoretical and experimental studies of ultrasonic effects on the welded joints and heat-affected zone. As is known, bearing capacity of welded joints is significantly lower than that of the base metal. This is due to the emergence of internal and residual stresses in the process of welding which are added to operating stresses, thus resulting in the destruction of the weld joint metal. Currently, it is common practice to reduce residual stresses in welded connections of NPP circulating pipelines and equipment by means of the thermal tempering and deformation methods. Thermal and deformation methods may reduce residual stresses in the HAZ but do not eliminate structural instability and physical and chemical heterogeneity, resulting in internal stresses and microcracks in the weld metal. Specialists of the Obninsk Institute for nuclear power engineering have developed a technology for thermal welding with ultrasonic treatment during the welding process, as a result of which the metal structure becomes fine-grained and homogeneous. Internal stresses are excluded and residual stresses are relieved in the heat-affected zone. The role of individual factors of ultrasonic field in creating certain structural changes in the metal depends on the crystallization conditions. In different areas of the crystallizing melt, the effect of any of the ultrasonic field factors may dominate. For example, the dispersion of crystals may occur in the two-phase zone, and the acoustic streams and stirring may be only in the liquid phase. If the reduction of grain size and elimination of the columnar structure are due to the ultrasonic dispersion, the change in phase distribution and the dendritic process of elimination are mainly determined by changes in the temperature gradient in the melt and stirring. The reasons for dispersion are cavitation, viscous friction forces, oscillatory and radiation pressure. An increase in the rate of nucleation of crystallization centers is also associated with these parameters.

      PubDate: 2017-04-02T01:30:16Z
      DOI: 10.1016/j.nucet.2017.03.001
  • Fukushima Daiichi accident as a stress test for the national system for
           the protection of the public in event of severe accident at NPP

    • Authors: V.A. Kutkov; V.V. Tkachenko
      Abstract: Publication date: Available online 27 March 2017
      Source:Nuclear Energy and Technology
      Author(s): V.A. Kutkov, V.V. Tkachenko
      It is proposed that the circumstances of the Fukushima Daiichi nuclear accident on 11 March 2011 in Japan should be used as the framework for the stress test of the national system for the protection of public in the beyond design extension conditions at NPP. Stress tests of the public protection strategy show to what extent the national system is stable under the most unfavorable NPP conditions and give an understanding of the potential vulnerabilities and the ways to resolve them. A definition of the Fukushima stress test model has been provided, and the actions undertaken by Japanese authorities under the conditions of the Fukushima Daiichi accident have been considered as the response to this stress test. The stress test has revealed major vulnerabilities in the strategy for the protection of public in the event of an accident at an NPP, which was successfully proven many times by over a hundred exercises at different levels. The stress test showed that the principal vulnerability of protection strategy being in use in Japan in 2011 was the reliance on computer systems in the assessment of the emergency exposure for decision-making during the emergency response phase. It is proposed, that the Fukushima stress test should be used to identify the vulnerabilities in the Russian Federation's strategy for the protection of public in the event of a nuclear accident and to use the lessons learnt from the test results to perfect this strategy.

      PubDate: 2017-04-02T01:30:16Z
      DOI: 10.1016/j.nucet.2017.03.007
  • Analysis of VVER-1000 main circulation pump condition in operation

    • Authors: S.T. Leskin; V.I. Slobodchuk; A.S. Shelegov
      Abstract: Publication date: Available online 25 March 2017
      Source:Nuclear Energy and Technology
      Author(s): S.T. Leskin, V.I. Slobodchuk, A.S. Shelegov
      The paper presents a method and algorithms for detecting abnormal conditions of the main circulation pumps (MCP) of NPPs with VVER-1000 reactors based on their in-process testing results. The methodological basis for the algorithms is the presentation of the nuclear power plant equipment as a complicated system described by the N-dimensional vector in the space of its conditions. A large number of process parameters describing the equipment condition using the Karhunen–Loeve transform is reduced to a much smaller number of informative criteria and presented in the form convenient for an analysis. The effectiveness of the method has been demonstrated in detecting the MCP abnormal behavior at the power units of Kalinin and Novovoronezh NPPs. The method and algorithms developed for monitoring the VVER-1000 MCP condition make it possible to detect an abnormality based on the pump operating data long before it is detected by the regular control systems.

      PubDate: 2017-03-25T21:36:58Z
      DOI: 10.1016/j.nucet.2017.03.002
  • Estimation of damage risks for the weld assembly between the header and
           the steam generator nozzle in a WWER NPP

    • Authors: N.N. Netyaga; S.P. Saakyan; V.P. Povarov
      Abstract: Publication date: Available online 25 March 2017
      Source:Nuclear Energy and Technology
      Author(s): N.N. Netyaga, S.P. Saakyan, V.P. Povarov
      Tasks involved in determination of the NPP equipment failure risks in the process of long-term operation are of extreme importance nowadays, specifically as the NPP units in the Russian Federation are nearing the end of their design lifetimes (25–30 years) and the life of newly built NPPs is increased to 60 years. The end of the effective NPP lifetimes and the increase in the operating life for newly built NPPs dictate the need for new approaches to be developed to ensure safe and reliable operation of critical thermomechanical components and pipelines. One of such components is the weld assembly between the hot header and the steam generator shell (WJ 111), in which formation and high-rate development of in-service defects is possible; the defects being systemic and affecting the operating safety of the entire power unit. Statistics is presented on failures in the weld assembly between the SG hot/cold headers and the DN-1200 piping for WWER reactors. Classification is proposed for the occurrence of limiting failure states for the SG weld assembly. The risk has been calculated for a power unit in the event one of the limiting states occurs. The losses resulting from damage to the SG header—nozzle weld assembly in the WJ 111 area is substantial. The probability of intolerable defects in the WJ 111 area remains high, so measures are required to control their formation time and development process.

      PubDate: 2017-03-25T21:36:58Z
      DOI: 10.1016/j.nucet.2017.03.004
  • Experience of integrated measurements using heterogeneous systems at
           different startup stages of a WWER-1200 power unit

    • Authors: V.I. Pavelko; M.T. Slepov; V.U. Khayretdinov
      Abstract: Publication date: Available online 24 March 2017
      Source:Nuclear Energy and Technology
      Author(s): V.I. Pavelko, M.T. Slepov, V.U. Khayretdinov
      A major feature specific to commissioning of NPPs in Russia and of Russian-designed NPPs in foreign countries is a great deal of measurement involved at different pre-commissioning stages, which differs from foreign experience where emphases is placed on computational justification and the scope of in-situ commissioning testing is smaller. The use of various measuring systems in pre-commissioning operations requires the involvement of a large number of personnel from different organizations, infrequently with no data acquisition coordination among them. This leads to both measuring channels and the information gathered being multiply duplicated (assembly, installation, adjustment), which is one of the sources for its misrepresentation, and to a greatly increased cost of operations. At the same time, most modern power units are equipped with test systems (TS) [1–4] consisting of heterogeneous measuring channels. The current TS commissioning practice required systems to be adjusted and started up only after the 100% unit power level was reached which did not make it possible to use test systems at pre-commissioning and power ascension stages. This study is an attempt to create an integrated information system composed of heterogeneous local systems to enable the largest possible number of standard channels to be used simultaneously with temporarily installed channels for pre-commissioning measurements to provide sound and high-quality information on the state of the unit.

      PubDate: 2017-03-25T21:36:58Z
      DOI: 10.1016/j.nucet.2017.03.005
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