Relevant bibliographies by topics / VVER-1200

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  1. Journal articles
  2. Dissertations / Theses
  3. Book chapters
  4. Conference papers

Academic literature on the topic 'VVER-1200'

Author: Grafiati
Published: 28 June 2021
Last updated: 7 February 2022

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Journal articles on the topic "VVER-1200"

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Tran, Vinh Thanh, Viet Phu Tran, and Thi Dung Nguyen. "A study on the core loading pattern of the VVER-1200/V491." Nuclear Science and Technology 7, no. 1 (September 1, 2021): 21–27. http://dx.doi.org/10.53747/jnst.v7i1.115.

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The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant.However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER- 1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel andmoderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR.
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Le, Dai Dien. "Comparative analysis of reactor coolant pump coastdown transient using VVER-1200 NPP simulator." Nuclear Science and Technology 7, no. 1 (September 1, 2021): 10–20. http://dx.doi.org/10.53747/jnst.v7i1.114.

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Verification has been performed to check the VVER-1200 NPP simulator installed atNuclear training Center, VINATOM by comparing main parameters in nominal power operation with design data given in safety analysis report of VVER-1200/V392M as well as Ninh Thuan FSSAR. A good agreement was found between the VVER-1200 NPP simulator and VVER-1200/V392M. In this study, the reactor coolant coastdown transient is investigated using the VVER-1200 NPP simulator in comparison with SAR documents. The real time feature of the simulator as well as simulated results performed in the simulator through switching off one reactor coolant pump in comparison with VVER-1000 experiments are given. A good agreement between the measured and simulated results shows that the thermal hydraulic characteristics and the control protection systems are modeled in a reasonable way. The analysis gives a good basis for the further studies on the simulator.
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Aver’yanova, S. P., and P. E. Filimonov. "Xenon stability of VVER-1200." Atomic Energy 107, no. 6 (December 2009): 424–28. http://dx.doi.org/10.1007/s10512-010-9246-7.

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Bui, Thi Hoa, Tan Hung Hoang, and Minh Giang Hoang. "Safety Analyses of VVER-1200/V491 reactor for longterm station blackout along with small LOCAs." Nuclear Science and Technology 6, no. 4 (December 30, 2016): 8–17. http://dx.doi.org/10.53747/jnst.v6i4.171.

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Performance of Passive Heat Removal through Steam Generator (PHRS-SG) of VVER-1200/V491 reactor presented in Safety Analysis Report for Ninh Thuan 1 shows that in case of long term station black out (SBO), VVER-1200/V491 reactor can be cooldown and remained in safety mode at least 24 hours based on PHRS-SG performance. Anyway, long term station blackout along with small break in main coolant pipe of VVER-1200/V491 is assumed to be happening as an extension design condition that needs to be investigated. This study focuses on investigation of SBO along with different size of small break of LOCAs with expectation of finding the range of break size that the reactor is still kept in safety mode during 24 hours. During the investigation, some indicators for fuel damage such as the timing of HA1 actuation or mass of coolant inventory discharged are introduced as necessary information contributed to Severe Accident Management Guideline (SAMG).
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Arzhaev, Alexander, Alexey Arzhaev, Valentin Makhanev, Mikhail Antonov, Anton Emelianov, Aleksander Kalyutik, Yury Karyakin, et al. "Possible in-service damages of steam generators at VVER-1000 and VVER-1200 NPP units and their impact on long-term operation." E3S Web of Conferences 209 (2020): 03005. http://dx.doi.org/10.1051/e3sconf/202020903005.

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Specific features of corrosion-mechanical damages of primary circuit header to steam generator vessel branch welds at VVER-1000 NPPs and their impact on safety and economic efficiency during long-term operation are analysed. Measures to avoid the damages for similar zones of VVER-1200 steam generators are discussed.
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Al Malkawi, Rashdan Talal, Sergey B. Vygovsky, and Osama Wasef Batayneh. "Investigation of the impact of steady-state VVER-1000 (1200) core characteristics on the reactor stability with respect to xenon oscillations." Nuclear Energy and Technology 6, no. 4 (November 20, 2020): 289–94. http://dx.doi.org/10.3897/nucet.6.60464.

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The article presents a method for obtaining an analytical expression for the criterion of stability of a VVER-1000 (1200) reactor with respect to xenon oscillations of the local power in the core, containing an explicit dependence of the criterion ratio coefficients on the arbitrary axial neutron field distribution in steady states of the core. Based on the data of numerical experiments using a full-scale model of the Kalinin NPP power units, the authors present the results of checking the validity of this expression for the reactor stability criterion with respect to xenon oscillations for different NPPs with VVER-1000 (1200) reactors.
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Gusev, Igor Nikolaevich, Vladimir Ruslanovich Kazanskiy, and Igor Leonidovich Vitkovsky,. "Dynamic stability of the VVER-1200 power unit." Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 2017, no. 3 (October 2017): 22–32. http://dx.doi.org/10.26583/npe.2017.3.02.

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Gusev, I. N., V. R. Kazanskiy, and I. L. Vitkovsky. "Dynamic stability of the VVER-1200 power unit." Nuclear Energy and Technology 3, no. 4 (December 2017): 270–77. http://dx.doi.org/10.1016/j.nucet.2017.10.004.

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Kovács, Dorina, and Dávid Kemény. "Investigation of VVER-1200 reactor pressure vessel’s material." IOP Conference Series: Materials Science and Engineering 903 (August 26, 2020): 012051. http://dx.doi.org/10.1088/1757-899x/903/1/012051.

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Petrovski, A. M., T. N. Korbut, E. A. Rudak, and M. O. Kravchenko. "Accounting of the vver-1200 overload influence for fission products activities calculating." Proceedings of the National Academy of Sciences of Belarus, Physical-Technical Series 64, no. 4 (January 11, 2020): 491–96. http://dx.doi.org/10.29235/1561-8358-2019-64-4-491-496.

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Current work is aimed at the analysis of the fission products decay influence during fuel reloading, when calculating the accumulated fission products activity for the VVER-1200 reactor fuel campaign. The Bateman problem solution based technique was used for calculations, within the framework of the two fissile nuclides approximation. The fission products producing process for the VVER-1200 reactor stationary campaign is considered, taking into account the reactor shutdown periods for refueling and without taking them into account (instant reload approximation). It was shown, that the instant reload approximation for fission products activity calculations gives the similar accurate result, as calculations with taking into account the shutdown periods. The results can be used to significantly simplify the calculations of fission product activity accumulation in nuclear power reactors.
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Dissertations / Theses on the topic "VVER-1200"

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Valíček, Roman. "Koncepční návrh sekundárního a terciálního okruhu pro nový jaderný zdroj." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2021. http://www.nusl.cz/ntk/nusl-443224.

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The aim of this thesis is to create a conceptual design of the secondary and tertiary circuit of a new nuclear power plant in our country, which is preceded by a brief search of particular generations of nuclear reactors. The VVER-1200 reactor became the model for the new nuclear power plant. There was designed a heat diagram containing primary components in this work. After that, there was the balance calculation and calculation of parameters of the main devices of the proposed scheme such as condenser, low-pressure and high-pressure regeneration exchangers and steam generator.
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Book chapters on the topic "VVER-1200"

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"Experimental Results on VVER-1200 Vibration Acoustics." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 475–524. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0014.

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"Vibration Diagnostic Features." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 91–117. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0003.

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"MCC Vibrations Modeling." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 119–28. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0004.

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"Long Lines with Lumped Complex Resistance." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 271–306. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0009.

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"MCP Vibrations." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 437–73. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0013.

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"Wave Equation." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 165–211. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0006.

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"Elastic Dissipative Medium. Attenuation of Acoustic Waves." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 307–64. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0010.

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"Mutual Oscillations of Reactor Vessel and Reactor Core Barrel." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 525–47. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0015.

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"ASW Globality–Locality." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 365–403. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0011.

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"MCC Acoustics in Lumped Parameters." In Vibration Acoustics Applied to VVER-1200 Reactor Plant, 129–64. WORLD SCIENTIFIC, 2021. http://dx.doi.org/10.1142/9789811234675_0005.

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Conference papers on the topic "VVER-1200"

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Hasan Tanim, Md Mehedi, Md Feroz Ali, Md Asaduzzaman Shobug, and Shamsul Abedin. "Analysis of Various types of Possible Fault and Consequences in VVER-1200 using PCTRAN." In 2020 International Conference for Emerging Technology (INCET). IEEE, 2020. http://dx.doi.org/10.1109/incet49848.2020.9153969.

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Morozov, A. V., O. V. Remizov, and A. S. Soshkina. "Experimental Study of Feed Water Level Decreasing Effect on VVER Steam Generator Model Operation in Condensation Mode." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16470.

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The essential technological distinctions of the AES-2006 project with VVER-1200 reactor, equipped with passive safety systems, determine some specifics in the character of accidents with the coolant leaks from the reactor primary circuit and the refusal of active part of the emergency core cooling system. In the case of an accident with depressurization of the reactor primary circuit, the system of passive heat removal (PHRS) ensures the transition of steam generators (SG) into the mode of steam condensation. As a result, a condensate comes to the core, providing its additional cooling. But in the case of beyond design basis accident the rupture of second circuit pipelines or PHRS steam-condensation path is possible. At such type of accidents the feed water level in SG vessel will be decrease, what can results in the deterioration of its condensation power. For experimental investigation of the condensation mode of VVER steam generator, a large scale HA2M-SG test facility was constructed in IPPE. The test facility incorporates VVER reactor SG model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator, cooling by process water and buffer tank, equipped by steam supply system from the IPPE heat power plant. The facility main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. Experiments at the HA2M-SG test facility have been performed at the pressure 0.36 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. On the base of the results of these experiments the correlation of condensation power from feed water level in SG model was obtained. The results of carried out tests make it possible to draw a conclusion about sufficiently large stability of VVER steam generator working in condensation mode to feed water level decreasing.
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Efanov, A. D., S. G. Kalyakin, A. V. Morozov, O. V. Remizov, A. A. Tsyganok, V. N. Generalov, V. M. Berkovich, and G. S. Taranov. "Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48842.

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In new Russian NPP with VVER-1200 reactor (V-392M reactor plant) in the event of an accident being due to the rupture of the reactor primary circuit and accompanied by the loss of a.c. sources, provision is made for the use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. To experimental investigation of the condensation mode of operation of VVER steam generator, a large scale HA2M-SG test rig was constructed. The test rig incorporates: tank-accumulator, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger simulator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure – 1.6 MPa, temperature – 200 Celsius degrees. Experiments at the HA2M-SG test rig have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond basis accident. The report presents the test procedure and the basic obtained test results.
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Bezlepkin, Vladimir, Sergey Semashko, Sergey Alekseev, Marina Ivanova, Teimuraz Vardanidze, and Yuriy Petrov. "Improvement of the System for Passive Heat Removal Through Steam Generators (SG PHRS) on NPP With VVER-1200 in the Light of “Fukushima” Accident." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30240.

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As a result of catastrophic events on the nuclear power plant “Fukushima” the European organizations on regulation of nuclear power (ENSREG) initiated wide-scale measures for complex designs revision of already operating and under construction European and Russian NPPs. Inspection was made about resistance of power units to external influences of the natural character, being accompanied by multiple failures of safety systems. Within these works stress tests for constructed power units of LAES-2 and the Baltic NPP were executed. The structure of these checks included the settlement analysis of a condition of NPP at accident with loss of all AC power supply sources which results are presented in report materials. Accident calculations with a full blackout were executed on the best-estimated heat-hydraulics code KORSAR/GP for justification of power unit preservations in the intact condition within 72 hours from the accident beginning by means of SG PHRS. The system is developed for feed of the SG PHRS tanks and the fuel pool for working capacity extension the SG PHRS and power unit preservation in a stable condition more than 72 hours from the accident beginning. Use of system for feed of tanks the SG PHRS and the fuel pool allows to increase significantly resistance of the NPP to external influences of the natural character and to increase time of preservation of the blackout power unit in a stable condition more than 5 days.
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Akter, Sangida, Md Sadman Anjum Joarder, Md Ghulam Zakir, Altab Hossain, Md Abdur Razzak, and Md Shafiqul Islam. "Comparative Analysis of Thermal Hydraulic Parameters of AP-1000 and VVER-1200 Nuclear Reactor for Turbine Trip Concurrent with Anticipated Transient Without SCRAM (ATWS)." In 2021 International Conference on Automation, Control and Mechatronics for Industry 4.0 (ACMI). IEEE, 2021. http://dx.doi.org/10.1109/acmi53878.2021.9528212.

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Groudev, Pavlin P., Antoaneta E. Stefanova, and Petya I. Vryashkova. "MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30941.

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This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.
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Arzhaev, Alexander, Alexey Arzhaev, Valentin Makhanev, Mikhail Antonov, Anton Emelianov, Aleksander Kalyutik, Kirill Arzhaev, and Ilya Denisov. "About leak detection systems in the framework of LBB concept application at Russian NPPs." In International Conference "Computing for Physics and Technology - CPT2020". Bryansk State Technical University, 2020. http://dx.doi.org/10.30987/conferencearticle_5fce27742eb478.63520709.

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Application of "leak before break" concept to reactor coolant circuit is obligatory for RF NPPs and the success depends also on fulfilment of requirements to the leak detection systems specified in RF national standard. During 1999-2020 requirements to the leak detection systems were permanently improved in regulatory documents. The most important changes in requirements have been done according to the Federal law on ensuring the uniformity of measurements. Paper gives comparative analyses of these evolutionary changes of requirements as well as details of their implementation during design and manufacturing of leak detection systems to supply NPP Units with VVER-440/1000/1200 and RBMK-1000 reactor facilities. Recently approved in the Russian Federation, the Federal norms and rules (FNR) in the field of atomic energy use ensure the continuity of the general requirements for reactor coolant leak monitoring (detection) systems (LDS) at nuclear power plants (NPPs) in relation to the previously valid regulatory documents.
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Sidorov, Valery G., Vladimir Bezlepkin, Sergej Alekseev, Sergey Semashko, Igor Ivkov, and Vladimir Kukhtevich. "Experimental and Computational Studies of LNPP-2 Passive System for Severe Accident Management." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29857.

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The project of nuclear station LNPP-2 with a reactor power plant VVER type by electrical power 1200 MVt involves a number of new design solutions to increase of parameters of safety. The passive containment heat removal system and heat removal system via steam generators is including of number of such solutions. Passive heat removal system via steam generators (PHRS/SG) is assigned for remove of residual heat of reactors core to final heat absorber (atmosphere) through a secondary circuit at DEC accident. The system PHRS/SG duplicates cooling-down system via SG to final heat absorber in case of impossibility of realization of its design functions. Containment heat removal system (PHRS/C) is assigned for remove of residual heat from containment in accidents with heat-transfer emissions from primary circuit. PHRS/C duplicates functions of a spray system to reduce of pressure under containment in case of spray system failure. In the substantiation of passive security systems the complex in SPbAEP of computational and experimental analysis was executed, the main results of which are shown in the present report.
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Kopytov, I. I., S. G. Kalyakin, V. M. Berkovich, A. V. Morozov, and O. V. Remizov. "Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75942.

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The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.
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Vuorinen, Asko. "Conceptual Design of a 4 × 300 MW Modular Nuclear Plant." In ASME 2011 Small Modular Reactors Symposium. ASMEDC, 2011. http://dx.doi.org/10.1115/smr2011-6529.

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The Finnish companies have built four medium size nuclear power plants. In addition they have constructed two nuclear icebreakers and several floating power plants. The latest 1650 MWe nuclear power plant under construction Olkiluoto-3 has had many problems, which have raised the costs of the plant to €3500/kWe from its original estimate of €2000/kWe and constriction schedule from four to eight years. It is possible to keep the costs down and schedule short by making the plant in shipyard and transport it to site by sea. The plant could be then lifted to its place by pumping seawater into the channel. This kind of concept was developed by the author in 1991, when he was making his thesis of modular gas fired power plants in Helsinki University of Technology. The modular construction of nuclear plants has made in a form of two nuclear icebreakers, which Wa¨rtsila¨ Marine has built in Helsinki Shipyard. The latest modular nuclear plant was launched in 2010 in St Petersburg shipyard. One of the benefits of modular construction is a possibility to locate the plant under rock by making the transportation channels in tunnels. This will give the plant external protection for aircraft crash and make the outer containment unnecessary. The water channels could also be used as pressure suppression pools in case of venting steam from the containment. This could reduce the radioactive releases in case of possible reactor accidents. The two 440 MW VVER plants build in Finland had construction costs of €1600 /kWe at 2011 money. The author believes that a 1200 MW nuclear plant with four 300 MW units can be constructed in five years and with €3300/kW costs, where the first plant could be generating power within 40 months and next units with 6 month intervals.
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