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Автор: Grafiati

Опубліковано: 11 вересня 2023

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1

Petti, D., D. Crawford, and N. Chauvin. "Fuels for Advanced Nuclear Energy Systems." MRS Bulletin 34, no. 1 (January 2009): 40–45. http://dx.doi.org/10.1557/mrs2009.11.

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AbstractFuels for advanced nuclear reactors differ from conventional light water reactor fuels and also vary widely because of the specific architectures and intended missions of the reactor systems proposed to deploy them. Functional requirements of all fuel designs for advanced nuclear energy systems include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintenance of a geometry that can be cooled. In all cases, anticipated fuel performance is the limiting factor in reactor system design, and cumulative effects of increased utilization and increased exposure to inservice environments degrade fuel performance. In this article, the current status of each fuel system is reviewed, and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed.

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2

Caciuffo, R., C. Fazio, and C. Guet. "Generation-IV nuclear reactor systems." EPJ Web of Conferences 246 (2020): 00011. http://dx.doi.org/10.1051/epjconf/202024600011.

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In this paper, we provide a concise description of the six nuclear reactor concepts that are under development in the framework of the Generation-IV International Forum. After a brief introduction on the world energy needs, its plausible evolution during the next fifty years, and the constraints imposed by the necessity to address the climate challenges we are facing today, we will present the main features of the innovative nuclear energy systems that hold the promise to produce almost-zero-carbon-emission electricity, heat for chemistry and industrial manufacturing, hydrogen to be used as energy vector, and affordable freshwater. Potential advantages over currently available nuclear systems in terms of increased safety, reduced proliferation risks, economical affordability, sustainability of the fuel cycle, and management of the waste inventory will be critically discussed against the technical challenges that remain to be overcome.

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3

Large, J. H. "Decommissioning of Nuclear Reactor Systems." Proceedings of the Institution of Mechanical Engineers, Part A: Journal of Power and Energy 206, no. 4 (November 1992): 273–80. http://dx.doi.org/10.1243/pime_proc_1992_206_044_02.

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The decision-making process involving the decommissioning of the British graphite-moderated, gas-cooled Magnox power stations is complex. There are timing, engineering, waste disposal, cost and lost generation capacity factors and the ultimate uptake of radiation dose to consider and, bearing on all of these, the overall decision of when and how to proceed with decommissioning may be heavily weighed by political and public tolerance dimensions. These factors and dimensions are briefly reviewed with reference to the ageing Magnox nuclear power stations, of which Berkeley and Hunterston A are now closed down and undergoing the first stages of decommissioning and Trawsfynydd, although still considered as available capacity, has had both reactors closed down since February 1991 and is awaiting substantiation and acceptance of a revised reactor pressure vessel safety case. Although the other first-generation Magnox power stations at Hinkley Point, Bradwell, Dungeness and Sizewell are operational, it is most doubtful that these stations will be. able to eke out a generating function for much longer. It is concluded that the British nuclear industry has adopted a policy of deferred decommissioning, that is delaying the process of complete dismantlement of the radioactive components and assemblies for at least one hundred years following close-down of the plant. In following this option the nuclear industry has expressed considerable confidence that the decommissioning technology required will he developed with passing time, that acceptable radioactive waste disposal methods and facilities will be available and that the eventual costs of decommissioning will not escalate without restraint.

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4

Hiscox, Briana, Benjamin Betzler, Vladimir Sobes, and William J. Marshall. "NEUTRONIC BENCHMARKING OF SMALL GAS-COOLED SYSTEMS." EPJ Web of Conferences 247 (2021): 10033. http://dx.doi.org/10.1051/epjconf/202124710033.

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To demonstrate that nuclear reactors can be built faster and more economically than they have been in the past, the US Department of Energy Office of Nuclear Energy is sponsoring the development of a small nuclear reactor called the Transformational Challenge Reactor (TCR) [1–2]. An important part of the design and licencing process of a new reactor is validation of the software used to analyze the reactor using established reactor physics benchmarks. This paper discusses validation of the neutronics software used to model four preliminary designs of the TCR core [2]. Because the TCR core design uses innovative technology and methods, comparable established benchmarks are limited or do not exist. For this effort, established benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) [3] were considered to be suitable for this design based on analysis using the SCALE/TSUNAMI-computed similarity indices to determine the amount of shared uncertainty between the design and each selected benchmark experiment. This paper addresses the challenges faced in benchmarking a unique reactor for licensing and construction, a task that will become more common as a new generation of innovative nuclear reactors are designed and built.

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5

Saha, Dilip, and John Cleveland. "Natural Circulation in Nuclear Reactor Systems." Science and Technology of Nuclear Installations 2008 (2008): 1. http://dx.doi.org/10.1155/2008/932319.

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6

Moore, Julian. "Upgrading of nuclear reactor display systems." Electronics and Power 32, no. 10 (1986): 735. http://dx.doi.org/10.1049/ep.1986.0429.

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7

Bonal, Jean-Pierre, Akira Kohyama, Jaap van der Laan, and Lance L. Snead. "Graphite, Ceramics, and Ceramic Composites for High-Temperature Nuclear Power Systems." MRS Bulletin 34, no. 1 (January 2009): 28–34. http://dx.doi.org/10.1557/mrs2009.9.

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AbstractThe age of nuclear power originated with the gas-cooled, graphite-moderated reactor in the 1940s. Although this reactor design had intrinsic safety features and enjoyed initial widespread use, gas-cooled reactor technology was supplanted by higher power density water-cooled systems in the 1960s. However, the next-generation reactors seek enhanced power conversion efficiency and the ability to produce hydrogen, best accomplished with high-temperature gas-cooled systems. Thus, international interest in gas-cooled reactor systems is reemerging. Although the materials systems of these reactors are fairly simple, the reactor environment, particularly its high temperatures and intense irradiation, present extreme challenges in terms of material selection and survivability. This article provides a brief review of materials issues and recent progress related to graphite and ceramic materials for application in gas-cooled nuclear reactor environments. Of particular interest are the drastic, irradiation-induced microstructural evolution and thermophysical property changes occurring as a result of energetic neutron irradiation, which significantly impact the performance and lifetime of much of the reactor core. For “nuclear” graphite, the performance and lifetime not only are closely related to the irradiation environment but also are dramatically affected by the specifics of the particular graphite: manufacturing process, graphitization temperature, composition (amount of coke, filler, etc., depending on where it was mined), and so on. Moreover, the extreme environmental challenges set down by this next generation of fission nuclear plants have driven the development and application of ceramic composites for critical components, pushing beyond upper temperature limits set by metallic alloys used in previous generations of nuclear reactors. The composite material systems of particular interest are continuous carbon-fiber composites and newly developed radiation-resistant silicon carbide fiber composites.

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8

Kondylakis, J. S. "Theoretically and under very special applied conditions a nuclear fission reactor may explode as nuclear bomb." HNPS Proceedings 18 (November 23, 2019): 121. http://dx.doi.org/10.12681/hnps.2558.

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This article/presentation describes a theoretical and applied research in nuclear fission reactor systems. It concerns with theoretical approaches and in very special applied cases consideration where a common nuclear fission reactor system may be considered to explode as nuclear bomb. This research gives critical impacts to the design, operation, management and philosophy of nuclear fission reactors systems. It also includes a sensitivity analysis of a particular applied problem concerning the core melting of a nuclear reactor and its deposit to the bottom of reactor vessel. Specifically, in a typical nuclear fission power reactor system of about 1000 MWe, the nuclear core material (corium) in certain cases can be melted and it may deposited in the bottom of nuclear reactor vessel. So, the nuclear criticality conditions are evaluated for a particular example case(s). Assuming an example composition of melted corium of 98 tones of U238 , 1 tone of U235 , 1 tone Pu239 and 25 tones Fe56 (supporting material) in a 5 m diameter of a finite cylindrical nuclear reactor vessel it is found that it may result in nuclear criticality above the unit. This condition corresponds to Supercritical Fast Nuclear Fission Reactor case, which may under certain very special applied conditions to nuclear explode as nuclear bomb.

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9

Sutopo, Catur Febriyanto, and Arifin M. Susanto. "Kajian pembentukan peraturan mengenai sistem pendingin reaktor dan sistem terkait untuk reaktor berpendingin gas." Jurnal Pengawasan Tenaga Nuklir 1, no. 2 (December 15, 2021): 11–19. http://dx.doi.org/10.53862/jupeten.v1i2.014.

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IN 2021, BAPETEN, AS THE REGULATORY BODY, IS ESTABLISHING A BAPETEN REGULATION REGARDING THE REACTOR COOLANT SYSTEM AND RELATED SYSTEMS, WHICH CURRENTLY ARE NOT YET AVAILABLE. Therefore, it is crucial to establish the BAPETEN Regulation. Based on the reasons, before setting the BAPETEN Regulation, it is necessary to conduct a study that is expected to provide a more comprehensive description and provide recommendations on what things need to be regulated in the BAPETEN Regulation, especially for gas-cooled reactors. The method used in this study is a literature study from various relevant references. The result of this study is that it is essential to require a capacity of the ultimate heat sink, including the spent nuclear fuel storage pool and a minimum period of the ability of the top heat sink in the accident analysis if the decay heat in the storage pool and the residual heat in the reactor core fail simultaneously. On the other hand, it is also necessary to require a margin of uncertainty to evaluate a situation and take corrective action. Likewise, independent and redundant access to the ultimate heat sink is needed to increase reliability. As for gas-cooled reactors, it is required to adapt the terms used. In addition, it is necessary to determine the appropriate definition because some of the terms used in water-cooled reactors have the same terms as gas-cooled reactors but have different functions. Keywords: Regulatory assessment, coolant system, related systems, gas-cooled reactors

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10

Obaidurrahman, Khalilurrahman, and Om Singh. "A comparative study of kinetics of nuclear reactors." Nuclear Technology and Radiation Protection 24, no. 3 (2009): 167–76. http://dx.doi.org/10.2298/ntrp0903167o.

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The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors), heavy water reactors (pressurized heavy water reactors), and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.

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11

Wootton, Mark James, John D. Andrews, Adam L. Lloyd, Roger Smith, A. John Arul, Gopika Vinod, M. Hari Prasad, and Vipul Garg. "Risk modelling of ageing nuclear reactor systems." Annals of Nuclear Energy 166 (February 2022): 108701. http://dx.doi.org/10.1016/j.anucene.2021.108701.

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12

Sinha, Ratan Kumar. "Advanced Nuclear Reactor Systems – An Indian Perspective." Energy Procedia 7 (2011): 34–50. http://dx.doi.org/10.1016/j.egypro.2011.06.005.

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13

Postnikov, N. S. "Synthesis of nuclear reactor relay regulation systems." Atomic Energy 80, no. 6 (June 1996): 394–401. http://dx.doi.org/10.1007/bf02415579.

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14

Ion, Sue. "Challenges to deployment of twenty-first century nuclear reactor systems." Proceedings of the Royal Society A: Mathematical, Physical and Engineering Sciences 473, no. 2198 (February 2017): 20160815. http://dx.doi.org/10.1098/rspa.2016.0815.

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The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

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15

Erickson, Anna, and Christopher Stewart. "Monitoring of nuclear reactors with antineutrinos: comparison of advanced reactor systems." Journal of Physics: Conference Series 1216 (April 2019): 012018. http://dx.doi.org/10.1088/1742-6596/1216/1/012018.

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16

Suryono, T. J., Sudarno, S. Santoso, and R. Maerani. "Modelling of FPGA-based Reactor Protection Systems of an Experimental Power Reactor." Journal of Physics: Conference Series 2048, no. 1 (October 1, 2021): 012038. http://dx.doi.org/10.1088/1742-6596/2048/1/012038.

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Abstract The reactor protection system of nuclear power plants including an experimental power reactor which will be built by Indonesia is a safety system that actuates the control rods to be inserted in the reactor core to absorb the neutron to stop the fission reaction and then shut down the reactor (reactor trip). The reactor protection system (RPS) is actuated when the level of signals from the sensors of important components in the reactors deviates from the setpoint determined in the bi-stable processor of the RPS. RPS for the experimental power reactor has 3 redundant channels for reliability and to minimize fake signals from the sensors due to electrical noise. It can be done by selecting the channels in local coincidence logic in the RPS by voting 2 of 3 channels which are eligible to generate actuation signals to trip the reactor. Recently, the RPSs are based on the programmable logic controller (PLC). However, now the trend changes to FPGA-based RPS because of its simplicity and reliability. This paper investigates the model of the FPGA-based RPS for an experimental power reactor and the functionality of each component of the model. The results show that the model can represent the functionality of RPS for the experimental power reactor.

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17

Khan, Salah Ud-Din, Zeyad Almutairi, and Meshari Alanazi. "Techno-Economic Assessment of Fuel Cycle Facility of System Integrated Modular Advanced Reactor (SMART)." Sustainability 13, no. 21 (October 26, 2021): 11815. http://dx.doi.org/10.3390/su132111815.

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The economic assessment of advanced nuclear power reactors is very important, specifically during the early stages of concept design. Therefore, a study was performed to calculate the total cost estimation of fuel cycle supply for a system modular advanced reactor (SMART) by using the Generation-IV economic program called G4-ECONS (Generation 4 Excel-based Calculation of Nuclear Systems). In this study, the detailed description of each model and methodology are presented including facility, operations, construction matrix, post-production model, and fuel cycle cost estimation model. Based on these models, six Generation-III+ and Generation-IV nuclear reactors were simulated, namely System 80+ with benchmark data, System 80+ with uranium oxide (UOx) and mixed oxide (MOx) fuel assemblies, fast reactor, PBMR (Pebble Bed Modular Reactor), and PWR (Pressurized Water Reactor), with partially closed and benchmarked cases. The total levelized costs of these reactors were obtained, and it was observed that PBMR showed the lowest cost. The research was extended to work on the SMART reactor to calculate the total levelized fuel cycle cost, capital cost, capital component cost, fraction of capital spent, and sine curve spent pattern. To date, no work is being reported to calculate these parameters for the SMART reactor. It was observed that SMART is the most cost-effective reactor system among other proven and advanced pressurized water-based reactor systems. The main objective of the research is to verify and validate the G4-ECONS model to be used for other innovative nuclear reactors.

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18

Luo, Run, Shripad T. Revankar, and Fuyu Zhao. "Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems." Applied Sciences 11, no. 17 (September 3, 2021): 8179. http://dx.doi.org/10.3390/app11178179.

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The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.

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19

M., Rehab, and Gamal M. "Robust H2 Control of the Nuclear Reactor Systems." International Journal of Computer Applications 176, no. 2 (October 17, 2017): 33–39. http://dx.doi.org/10.5120/ijca2017915549.

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20

Jahshan, S. N., and T. Kammash. "Multimegawatt Nuclear Reactor Design for Plasma Propulsion Systems." Journal of Propulsion and Power 21, no. 3 (May 2005): 385–91. http://dx.doi.org/10.2514/1.5456.

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21

Chionis, D., A. Dokhane, H. Ferroukhi, and A. Pautz. "Application of causality analysis on nuclear reactor systems." Chaos: An Interdisciplinary Journal of Nonlinear Science 29, no. 4 (April 2019): 043126. http://dx.doi.org/10.1063/1.5083905.

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22

Hamada, Yasser Mohamed. "Liapunov's stability on autonomous nuclear reactor dynamical systems." Progress in Nuclear Energy 73 (May 2014): 11–20. http://dx.doi.org/10.1016/j.pnucene.2013.12.012.

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23

Sequeira, César A. C., and Diogo M. F. Santos. "Diffusion-controlled processes in nuclear reactor oxide systems." Nuclear Engineering and Design 241, no. 12 (December 2011): 4903–8. http://dx.doi.org/10.1016/j.nucengdes.2011.08.059.

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24

Pereguda, A. I., and A. A. Petrenko. "Ensuring required reliability for nuclear reactor protection systems." Soviet Atomic Energy 67, no. 6 (December 1989): 859–63. http://dx.doi.org/10.1007/bf01124957.

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25

Petrov, V. P. "Improving the reliability of nuclear-reactor scram systems." Soviet Atomic Energy 71, no. 6 (December 1991): 1047–49. http://dx.doi.org/10.1007/bf01123165.

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26

Cheng, X., Y. H. Yang, Y. Ouyang, and H. X. Miao. "Role of Passive Safety Systems in Chinese Nuclear Power Development." Science and Technology of Nuclear Installations 2009 (2009): 1–7. http://dx.doi.org/10.1155/2009/573026.

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Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near-term and medium-term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large-scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

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27

Bousbia-Salah, Anis, Areeya Jirapongmed, Tewfik Hamidouche, John White, Francesco D'Auria, and Martina Adorni. "Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor." Nuclear Technology and Radiation Protection 21, no. 1 (2006): 3–12. http://dx.doi.org/10.2298/ntrp0601003b.

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RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

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28

Heidet, Florent, Nicholas R. Brown, and Malek Haj Tahar. "Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles." Reviews of Accelerator Science and Technology 08 (January 2015): 99–114. http://dx.doi.org/10.1142/s1793626815300066.

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This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

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29

Hill, I. "IDENTIFICATION OF REACTOR PHYSICS BENCHMARKS FOR NUCLEAR DATA TESTING: TOOLS AND EXAMPLES." EPJ Web of Conferences 247 (2021): 10028. http://dx.doi.org/10.1051/epjconf/202124710028.

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Measurements of reactor physics quantities aimed at identifying the reactivity worth of materials, spectral ratios of cross-sections, and reactivity coefficients have ensured reactor physics codes can accurately predict nuclear reactor systems. These measurements were critical in the absence of sufficiently accurate differential data, and underpinned the need for experiments through the 50s, 60s, 70s and 80s. Data from experimental campaigns were routinely incorporated into nuclear data libraries either through changes to general nuclear data libraries, or more commonly in the local libraries generated by a particular institution or consortium interested in accurately predicting a specific nuclear system (e.g. fast reactors) or parameters (e.g. fission gas release, yields). Over the last three decades, the model has changed. In tandem access to computing power and monte carlo codes rose dramatically. The monte carlo codes were well suited to computing k-eff, and owing to the availability of high quality criticality benchmarks and these benchmarks were increasing used to test the nuclear data. Meanwhile, there was a decline in the production of local libraries as new nuclear systems were not being built, and the existing systems were considered adequately predicted. The cost-to-benefit ratio of validating new libraries relative to their improved prediction capability was less attractive. These trends have continued. It is widely acknowledged that the checking of new nuclear data libraries is highly skewed towards testing against criticality benchmarks, ignoring many of the high quality reactor physics benchmarks during the testing and production of general-purpose nuclear data libraries. However, continued increases in computing power, methodology (GPT), and additional availability reactor physics experiments from sources such as the International Handbook of Evaluated Reactor Physics Experiments should result in better testing of new libraries and ensured applicability to a wide variety of nuclear systems. It often has not. Leveraging the wealth of historical reactor physics measurements represents perhaps the simplest way to improve the quality of nuclear data libraries in the coming decade. Resources at the Nuclear Energy Agency can be utilized to assist in interrogating available identify benchmarks in the reactor physics experiments handbook, and expediting their use in verification and validation. Additionally, high quality experimental campaigns that should be examined in validation will be highlighted to illustrate potential improvements in the verification and validation process.

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30

JÓŹWIK, Roman. "THE USE OF NUCLEAR ENERGY FOR MILITARY AND CIVILIAN PURPOSES SAFETY IN THE NUCLEAR POWER INDUSTRY." Journal of Science of the Gen. Tadeusz Kosciuszko Military Academy of Land Forces 185, no. 3 (June 1, 2017): 106–23. http://dx.doi.org/10.5604/01.3001.0010.5127.

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The purpose of the article was to gather the basic information about the mechanism be-hind nuclear energy formation and the types of reactors, already built worldwide or poten-tially planned for construction in the near future, and to present the history of the begin-nings of nuclear power in Poland. The issues of the safety of reactors, independent safety assurance systems and systems for emergency shutdown of a reactor are discussed in more detail. The problem of responsibility for the safety of nuclear equipment is also ex-amined, including the relevant authority and method for such safety inspection. The initia-tives taken in Poland in connection with the programme for the nuclear power industry are also described.

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31

Wicks, Frank. "50 Years of Nuclear Power." Mechanical Engineering 129, no. 11 (November 1, 2007): 36–39. http://dx.doi.org/10.1115/1.2007-nov-4.

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This article highlights the Atomic Age that announced itself to the world with the destruction of two Japanese cities in 1945. After the first bomb fell, on Hiroshima in August, mankind suddenly realized that it possessed a new technology of unprecedented destructive power. In 1948, Untermyer transferred to the Argonne National Laboratory near Chicago. The lab traced its origins to Enrico Fermi, who with Leo Szilard had been first to demonstrate a nuclear chain reaction only six years earlier. Argonne was the first national laboratory with the mission of developing nuclear power for peaceful purposes. Untermyer left General Electric (GE) in 1964 to find the National Nuclear Equipment Corp. to design and manufacture equipment for the nuclear industry. He was awarded several more patents. GE’s Economic Simplified Boiling Water Reactor also makes use of passive systems. According to GE, the reactor has more than 72 hours of passive running capability, and its simplified systems make it cheaper to build and run.

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M., Rehab, Gamal M., and Ibrahim E. "Robust H∞ Controller Design for the Nuclear Reactor Systems." International Journal of Computer Applications 178, no. 4 (November 15, 2017): 15–19. http://dx.doi.org/10.5120/ijca2017915808.

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Končar, Boštjan, Eckhard Krepper, Dominique Bestion, Chul-Hwa Song, and Yassin A. Hassan. "Two-Phase Flow Heat Transfer in Nuclear Reactor Systems." Science and Technology of Nuclear Installations 2013 (2013): 1–2. http://dx.doi.org/10.1155/2013/587839.

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Wang, Junye. "Design method of flow distribution in nuclear reactor systems." Chemical Engineering Research and Design 91, no. 4 (April 2013): 595–602. http://dx.doi.org/10.1016/j.cherd.2012.10.003.

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Wood, C. J., and M. Pick. "Conference Report: Water Chemistry of Nuclear Reactor Systems ? 2004." Nuclear Future 1, no. 3 (May 2005): 10. http://dx.doi.org/10.1680/nuen.1.3.10.66246.

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Corradini, M. L. "NEW REACTOR TECHNOLOGY: SAFETY IMPROVEMENTS IN NUCLEAR POWER SYSTEMS." Health Physics 93, no. 5 (November 2007): 547–59. http://dx.doi.org/10.1097/01.hp.0000282045.96877.b4.

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37

Chen, Fang, Xi-Lin Dong, Yan Tang, An-Chi Huang, Mei-Lin Zhang, Qing-Chun Kang, Zhong-Jun Shu, and Zhi-Xiang Xing. "Thermal Characteristic Analysis of Sodium in Diluted Oxygen via Thermogravimetric Approach." Processes 10, no. 4 (April 5, 2022): 704. http://dx.doi.org/10.3390/pr10040704.

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As the main reactor type of the fourth-generation nuclear power systems, sodium-cooled fast reactors are now designed and built worldwide. A sodium pool cooling circulation process is indispensable in a sodium-cooled fast reactor. However, the sodium pool fire design is the basis of accidents in sodium-cooled fast reactors. The fire hazard caused by the sodium–oxygen reaction and fast reactor safety have attracted extensive attention. Dry powder is widely used as an effective fire-extinguishing agent to control sodium fire. The sodium will burn in an oxygen-depleted atmosphere when using dry powder to cover fire. In this study, the change law of thermogravimetry of melted sodium is studied by thermogravimetric analysis (TGA) and the apparent activation energy (Ea) is obtained, which has a linear relationship with the oxygen concentration. The results can provide a reference for improving the engineering design standards of sodium fire suppression systems and can also be incorporated into simulation software to improve the accuracy of fire suppression simulations.

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Skalozubov, V. I., O. A. Dorozh, V. A. Kondratyk, S. I. Kosenko, and V. I. Konshin. "APPROACHES TO MODELING CONDITIONS OF THERMOACOUSTIC INSTABILITY IN NON-EQUILIBRIUM TWO-PHASE COOLANT OF NUCLEAR REACTORS." Thermophysics and Thermal Power Engineering 49, no. 2 (June 11, 2023): 95–102. http://dx.doi.org/10.31472/ttpe.2.2023.11.

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The consequences of thermoacoustic instability of the coolant in the active zone of nuclear reactors can be high-amplitude, high-frequency dynamic loads on the internal structures and a violation of the tightness of the TVEL shells. However, until now, there are no reactor control/diagnostic systems and operational instructions for managing accidents in conditions of thermoacoustic instability of the coolant in the reactor core. The main reason for this situation is the lack of substantiated methods for modeling the criteria and conditions for the occurrence of thermoacoustic instability in the active zone. The purpose of the work is the development of a criterion method for modeling the conditions of thermoacoustic instability of the coolant in the active zone of the reactor to substantiate the appropriate reactor control systems and symptom-oriented emergency instructions. An original method of determining the criteria and conditions of thermoacoustic instability of the coolant in the active zone depending on the determining parameters of the thermodynamic state of the reactor plant has been developed.. Based on the thermodynamic approach, which takes into account the level of completion of interphase heat and mass transfer processes in acoustic pressure waves, the criteria and area of thermoacoustic instability of the coolant in the reactor core are determined. The established area of thermoacoustic instability of the coolant in the reactor core was verified on the basis of known experimental data obtained at the experimental installation, which meets the criteria of thermodynamic similarity to the core of the VVER-1000 core. Based on the developed criterion method, the main provisions and requirements for the relevant reactor control/diagnostic systems and symptom-oriented instructions for managing accidents in conditions of thermoacoustic instability of the coolant in the active zone of the reactors are defined.

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Cabellos, Oscar, Christophe Demazière, Sandra Dulla, Nuria Garcia-Herranz, Rafael Miró, Rafael Macian, Máté Szieberth, Emma Buchet, Suzi Maurice, and Samy Strola. "GRE@T-PIONEeR: Teaching the nuclear data pipeline using innovative pedagogical methods." EPJ Web of Conferences 284 (2023): 19001. http://dx.doi.org/10.1051/epjconf/202328419001.

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GRE@T-PIONEeR - GRaduate Education Alliance for Teaching the PhysIcs and safety Of NuclEar Reactors - is a project funded by the Euratom – Horizon 2020 Framework Programme which aims at developing and providing specialised and advanced courses in computational and experimental reactor physics at the graduate level (MSc and PhD levels) and post-graduate level, as well as the staff members working in the nuclear industry. One of the work packages of GRE@T-PIONEeR is devoted to developing a specific course on the nuclear data pipeline processes and to present the role of nuclear data to play in calculations of innovative reactor systems. This course covers all steps in the nuclear data life cycle, starting from the measurements to their validation and final use in nuclear reactor calculations. Beyond the technical contents of the courses being developed, the paper describes the use of innovative pedagogical methods and active learning techniques, such as flipped classes, aimed at promoting student learning.

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Xie, Hongyun, Haixia Gu, Chao Lu, and Jialin Ping. "Online Simulation of Nuclear Power Plant Primary Systems." Science and Technology of Nuclear Installations 2020 (December 15, 2020): 1–9. http://dx.doi.org/10.1155/2020/8819239.

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Real-time Simulation (RTS) has long been used in the nuclear power industry for operator training and engineering purposes. And, online simulation (OLS) is based on RTS and with connection to the plant information system to acquire the measurement data in real time for calibrating the simulation models and following plant operation, for the purpose of analyzing plant events and providing indicative signs of malfunctioning. OLS has been applied in certain industries to improve safety and efficiency. However, it is new to the nuclear power industry. A research project was initiated to implement OLS to assist operators in certain critical nuclear power plant (NPP) operations to avoid faulty conditions. OLS models were developed to simulate the reactor core physics and reactor/steam generator thermal hydraulics in real time, with boundary conditions acquired from plant information system, synchronized in real time. The OLS models then were running in parallel with recorded plant events to validate the models, and the results are presented.

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Severyn, Valery, and Elena Nikulina. "APPLICATION OF INFORMATION TECHNOLOGY FOR MODELING THE CONTROL DYNAMICS OF A NUCLEAR REACTOR ZONING ON THE VERTICAL AXIS." Journal of Automation and Information sciences 5 (September 1, 2021): 45–56. http://dx.doi.org/10.34229/1028-0979-2021-5-4.

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The structure of information technology for modeling control systems, which includes a block of systems models, a module of integration methods and other program elements, is considered. To analyze the dynamics of control of a nuclear reactor, programs of mathematical models of a WWER-1000 nuclear reactor of the V-320 series and its control systems in the form of nonlinear systems of differential equations in the Cauchy form have been developed. For the integration of nonlinear systems of differential equations, an algorithm of the system method of the first degree is presented. A mathematical model of a WWER-1000 reactor as a control object with division into zones along the vertical axis in relative variables of state is considered, the values of the constant parameters of the model and the initial conditions corresponding to the nominal mode are given. Using information technology for ten zones of the reactor, the system integration method was used to simulate the dynamics of control of a nuclear reactor. Graphs of neutron and thermal processes in the reactor core, as well as changes in the axial offset when the reactor load is dumped under the influence of the movement of absorbing rods and an increase in the concentration of boric acid, are plotted. The analysis of dynamic processes of reactor control is carried out. The programs of integration methods and models of the WWER-1000 reactor of the V-320 series are included in the information technology to optimize the maneuvering modes of the reactor.

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Peakman, Aiden, and Bruno Merk. "The Role of Nuclear Power in Meeting Current and Future Industrial Process Heat Demands." Energies 12, no. 19 (September 25, 2019): 3664. http://dx.doi.org/10.3390/en12193664.

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There is growing interest in the use of advanced reactor systems for powering industrial processes which could significantly help to reduce CO 2 emissions in the global energy system. However, there has been limited consideration into the role nuclear power would play in meeting current and future industry heat demand, especially with respect to the advantages and disadvantages nuclear power offers relative to other competing low-carbon technologies, such as Carbon Capture and Storage (CCS). In this study, the current market needs for high temperature heat are considered based on UK industry requirements and work carried out in other studies regarding how industrial demand could change in the future. How these heat demands could be met via different nuclear reactor systems is also presented. Using this information, it was found that the industrial heat demands for temperature in the range of 500 ∘ C to 1000 ∘ C are relatively low. Whilst High Temperature Gas-cooled Reactors (HTGRs), Very High Temperature Reactors (VHTRs), Gas-cooled Fast Reactors (GFRs) and Molten Salt Reactors (MSRs) have an advantage in terms of capability to achieve higher temperatures (>500 ∘ C), their relative benefit over Liquid Metal-cooled Fast Reactors (LMFRs) and Light Water Reactors (LWRs) is actually smaller than previous studies indicate. This is because, as is shown here, major parts of the heat demand could be served by almost all reactor types. Alternative (non-nuclear) means to meet industrial heat demands and the indirect application of nuclear power, in particular via producing hydrogen, are also considered. As hydrogen is a relatively poor energy carrier, current trends indicate that the use of low-carbon derived hydrogen is likely to be limited to certain applications and there is a focus in this study on the emerging demands for hydrogen.

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Castagna, Christian, Daniela Cancila, and Antonio Cammi. "Adoption of ACPS in Nuclear Reactor Analysis." ACM SIGAda Ada Letters 41, no. 1 (October 28, 2022): 69–73. http://dx.doi.org/10.1145/3570315.3570320.

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Nuclear power plants are a paramount example of critical cyber-physical systems. Some of the current researches in nuclear reactor analysis concern the degree of acceptable uncertainty of the whole system. Some difficulties arise from weaving fields of different disciplines, such as computer science (e.g. embedded software and hardware), mechatronics (e.g. sensors and actuators) and physics (e.g. neutronics and thermal-hydraulics). To complicate the scenario further, each field demands different disciplines, competencies and different teams. In this short paper, we highlight the importance of the cross-fertilization of different disciplines in the nuclear reactor domain and we investigate emerging methods to control uncertainty in the nuclear reactor design.

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Guo, Dingqing, Jinkai Wang, Chao Chen, Dongqin Xia, Nuo Yong, and Daochuan Ge. "Preliminary Study on Risk Identification and Assessment Framework for Fusion Radioactive Waste Management." Science and Technology of Nuclear Installations 2022 (May 17, 2022): 1–10. http://dx.doi.org/10.1155/2022/4870208.

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Fusion reactors are expected to be safer, more environmentally friendly, and to have a lower nuclear proliferation risk, compared with other nuclear energy systems. However, it is widely recognized that a large amount of radioactive materials will be produced by a fusion reactor. Therefore, it is important to fully understand the overall radiation risk level of fusion radioactive wastes (radwaste) compared with existing nuclear energy systems. Studies on the treatment of the fusion radwaste have been currently focused on three ultimate options: clearance, recycling, and disposal by activation assessment of radioactive materials from the operation and decommissioning of fusion reactors. However, the radiation risk in the management of fusion radwaste, especially in the final disposal, was seldom studied. Based on the comparative analysis of fusion radioactive waste with ITER and fission reactors (e.g., pressurized water reactor, PWR), this paper tries to discuss how to determine the radiation risk in the process of fusion radwaste management on the premise of the current feasible industrial technology. On this basis, a risk assessment framework for repository disposal under normal degradation and external events is proposed.

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Valenti, Michael. "A Next-Generation Reactor." Mechanical Engineering 120, no. 08 (August 1, 1998): 68–71. http://dx.doi.org/10.1115/1.1998-aug-5.

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This article highlights that Electricité de France’s (EDF) N4 nuclear technology has increased European standards of power, efficiency, and safety beyond previous limits. EDF, the Paris-based French national utility, has developed and is operating the N4 reactor, capable of generating 1450 megawatts, at its power plant in Chooz. The N4 was designed to put public safety concerns to rest while providing more power than the previous generation of 1,300-megawatt EDF reactors. This installation represents the next step in French, European, and possibly the world’s nuclear power. Chooz A began operations in 1967 as the first pressurized water reactor (PWR) in France, originally based on the 185-megawatt synchronized PWR Yankee Rowe plant in Massachusetts. Typical reactor safety systems analyze a problem after it occurs. Such a procedure involves painstaking historical, reconstruction that is time-consuming, often difficult to interpret, and less reliable as time passes.

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Abdelfattah, Hany, Said A. Kotb, Mohamed Esmail, and Mohamed I. Mosaad. "Adaptive Neuro-Fuzzy Self Tuned-PID Controller for Stabilization of Core Power in a Pressurized Water Reactor." International Journal of Robotics and Control Systems 3, no. 1 (November 6, 2022): 1–18. http://dx.doi.org/10.31763/ijrcs.v3i1.710.

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There has been a lot of interest in generating electricity using nuclear energy recently. This interest is due to the features of such a source of energy. The main part of the nuclear energy system is the reactor core, especially the most widely used Pressurized Water Reactor (PWR). This reactor is the hottest part of the nuclear system; security risks and economic possibilities must be considered. Controlling this reactor can increase the security and efficiency of nuclear power systems. This study presents a dynamic model of the (PWR), including the reactor's core, the plenums of the upper and lower, and the connecting piping between the reactor core and steam generator. In addition, an adaptive neuro-fuzzy (ANFIS) self-tuning PID Controller for the nuclear core reactor is presented. This adaptive controller is used to enhance the performance characteristics of PWR by supporting the profile of the reactor power, the coolant fuel, and hot leg temperatures. The suggested proposed ANFIS self-tuning controller is estimated through a comparison with the conventional PID, neural network, and fuzzy self-tuning controllers. The results showed that the proposed controller is best over traditional PID, neural network, and fuzzy self-tuning controllers. All simulations are throughout by using MATLAB/SIMULINK.

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Degweker, S. B. "Reactor noise in accelerator driven systems." Annals of Nuclear Energy 30, no. 2 (January 2003): 223–43. http://dx.doi.org/10.1016/s0306-4549(02)00051-8.

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Degweker, S. B. "Reactor noise in accelerator driven systems." Annals of Nuclear Energy 30, no. 16 (November 2003): 1701. http://dx.doi.org/10.1016/s0306-4549(03)00139-7.

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Gulevich, A. V., and O. F. Kukharchuk. "Methods for calculating coupled reactor systems." Atomic Energy 97, no. 6 (December 2004): 803–11. http://dx.doi.org/10.1007/s10512-005-0066-0.

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Balabin, V., M. Kamnev, N. Kuchin, O. Tyurikov, and A. Khizbullin. "Efficiency of emergency depressurization systems for marine nuclear reactor containments." Transactions of the Krylov State Research Centre 4, no. 386 (November 15, 2018): 107–16. http://dx.doi.org/10.24937/2542-2324-2018-4-386-107-116.

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