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Journal articles on the topic 'Boiling water reactors'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 August 2024

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1

Ganda, Francesco, Francisco J. Arias, Jasmina Vujic, and Ehud Greenspan. "Self-Sustaining Thorium Boiling Water Reactors." Sustainability 4, no. 10 (2012): 2472–97. http://dx.doi.org/10.3390/su4102472.

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2

Lin, Chien C. "Hydrogen Water Chemistry Technology in Boiling Water Reactors." Nuclear Technology 130, no. 1 (2000): 59–70. http://dx.doi.org/10.13182/nt00-a3077.

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3

Györke, Gábor, and Attila R. Imre. "Physical-chemical Background of the Potential Phase Transitions during Loss of Coolant Accidents in the Supercritical Water Loops of Various Generation IV Nuclear Reactor Types." Periodica Polytechnica Chemical Engineering 63, no. 2 (2019): 333–39. http://dx.doi.org/10.3311/ppch.12770.

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Loss of coolant accidents (LOCA) are a serious type of accidents for nuclear reactors, when the integrity of the liquid-loop breaks. While in traditional pressurized water reactors, pressure drop can cause flash boiling, in Supercritical-Water Cooled reactors, the pressure drop can be terminated by processes with fast phase transition (flash boiling or steam collapse) causing pressure surge or the expansion can go smoothly to the dry steam region. Modelling the pressure drop of big and small LOCAs as isentropic and isenthalpic processes and replacing the existing reactor designs with a simplified supercritical loop, limiting temperatures for various outcomes will be given for 24.5 and 25 MPa initial pressure. Using the proposed method, similar accidents for chemical reactors and other equipment using supercritical fluids can be also analyzed, using only physical-chemical properties of the given supercritical fluid.

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4

YAMAZAKI, Yukitaka, Katsumi YAMADA, Chikako IWAKI, Shinichi MOROOKA, Hideo SONEDA, and Tomohiro YAGII. "ICONE15-10464 DEVELOPMENT OF LOW PRESSURE LOSS STEAM SEPARATOR FOR BOILING WATER REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_248.

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5

Dokhane, Abdelhamid. "Boiling water reactors as dynamic complex systems." International Journal of Nuclear Energy Science and Technology 4, no. 4 (2009): 275. http://dx.doi.org/10.1504/ijnest.2009.028588.

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6

Hampel, R., A. Traichel, S. Fleischer, and R. Kästner. "Water level in boiling water reactors — Measurement, modelling, diagnostic." Progress in Nuclear Energy 43, no. 1-4 (2003): 121–28. http://dx.doi.org/10.1016/s0149-1970(03)00018-0.

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7

Chen, Yen-Shu, Li-Ying Huang, and Ansheng Lin. "Water inventory calculation for the shutdown boiling water reactors." Nuclear Engineering and Design 408 (July 2023): 112315. http://dx.doi.org/10.1016/j.nucengdes.2023.112315.

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8

Giustini, Giovanni. "Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review." Inventions 5, no. 3 (2020): 47. http://dx.doi.org/10.3390/inventions5030047.

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The boiling process is utterly fundamental to the design and safety of water-cooled fission reactors. Both boiling water reactors and pressurised water reactors use boiling under high-pressure subcooled liquid flow conditions to achieve high surface heat fluxes required for their operation. Liquid water is an excellent coolant, which is why water-cooled reactors can have such small sizes and high-power densities, yet also have relatively low component temperatures. Steam is in contrast a very poor coolant. A good understanding of how liquid water coolant turns into steam is correspondingly vital. This need is particularly pressing because heat transfer by water when it is only partially steam (‘nucleate boiling’ regime) is particularly effective, providing a great incentive to operate a plant in this regime. Computational modelling of boiling, using computational fluid dynamics (CFD) simulation at the ‘component scale’ typical of nuclear subchannel analysis and at the scale of the single bubbles, is a core activity of current nuclear thermal hydraulics research. This paper gives an overview of recent literature on computational modelling of boiling. The knowledge and capabilities embodied in the surveyed literature entail theoretical, experimental and modelling work, and enabled the scientific community to improve its current understanding of the fundamental heat transfer phenomena in boiling fluids and to develop more accurate tools for the prediction of two-phase cooling in nuclear systems. Data and insights gathered on the fundamental heat transfer processes associated with the behaviour of single bubbles enabled us to develop and apply more capable modelling tools for engineering simulation and to obtain reliable estimates of the heat transfer rates associated with the growth and departure of steam bubbles from heated surfaces. While results so far are promising, much work is still needed in terms of development of fundamental understanding of the physical processes and application of improved modelling capabilities to industrially relevant flows.

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9

Pruit, D. W., D. R. Tinkler, and Y. M. Farawila. "ICONE15-10489 An Enhanced Detect-and-Suppress Stability Protection Method for Boiling Water Reactors." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_260.

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10

Chiang, Ren-Tai. "Safety Features of Advanced and Economic Simplified Boiling Water Reactors." Indonesian Journal of Physics and Nuclear Applications 3, no. 1 (2018): 1–6. http://dx.doi.org/10.24246/ijpna.v3i1.1-6.

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The Advanced Boiling Water Reactor (ABWR) and the Economic Simplified Boiling Water Reactor (ESBWR) are two kinds of contemporary, advanced, commercially available nuclear power reactors. Reactor internal pumps in an ABWR improve performance while eliminating the large recirculation pumps in earlier BWRs. The utilization of natural circulation and passive safety systems in the ESBWR design simplifies nuclear reactor system designs, reduces cost, and provides a reliable stability solution for inherently safe operation. The conceptually reliable stability solution for inherently safe ESBWR operation is developed by establishing a sufficiently high natural circulation flow line, which has a core flow margin at least 5% higher than the stability boundary flow at 100% rated power of a conventional BWR, and then by designing a high flow natural circulation system to achieve this high natural circulation flow line. The performance analyses for the ESBWR Emergency Core Cooling System (ECCS) show that: (1) the core remains covered with a large margin and there is no core heat up in the ESBWR for any break size, (2) the long-term containment pressure increases gradually with time, in the order of hours, and the peak pressure is below the design value with a large margin, and (3) the margins depend on the containment volumes and water inventories. These safety design features ensure inherently safe ESBWR operation. Enhanced safety features based on lessons learned from the Fukushima nuclear accident are added in ABWR’s and ESBWR’s safety designs. The major enhancements are the further prevention of station blackout and loss of ultimate heat sink.

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11

Sarott, Flurin-A. "Water Chemistry in Boiling Water Reactors – A Leibstadt-Specific Overview." CHIMIA International Journal for Chemistry 59, no. 12 (2005): 923–28. http://dx.doi.org/10.2533/000942905777675336.

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12

Hazzan, M. J., M. S. Stocknoff, David W. Barcomb, and Timothy Irving. "Radiation Levels During Shutdown in Boiling Water Reactors." Nuclear Technology 69, no. 3 (1985): 249–56. http://dx.doi.org/10.13182/nt85-a33608.

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13

Wachter, O., and G. Brümmer. "Experiences with austenitic steels in boiling water reactors." Nuclear Engineering and Design 168, no. 1-3 (1997): 35–52. http://dx.doi.org/10.1016/s0029-5493(96)01308-8.

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14

Jones, Robin L., Joe D. Gilman, and J. Lawrence Nelson. "Controlling stress corrosion cracking in boiling water reactors." Nuclear Engineering and Design 143, no. 1 (1993): 111–23. http://dx.doi.org/10.1016/0029-5493(93)90279-i.

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15

Postnikov, N. S. "Investigation of chaotic oscillations in boiling-water reactors." Atomic Energy 107, no. 5 (2009): 291–301. http://dx.doi.org/10.1007/s10512-010-9228-9.

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16

Kurskii, A. S., V. M. Eshcherkin, V. V. Kalygin, M. N. Svyatkin, and I. I. Semidotskii. "Boiling water vessel reactors for nuclear district heating." Atomic Energy 111, no. 5 (2012): 370–76. http://dx.doi.org/10.1007/s10512-012-9506-9.

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17

Tellez, Alberto Quezada, Francisco A. Godínez, Guillermo Fernández-Anaya, Marco A. Polo-Labarrios, and Sergio Quezada García. "Multifractal detrended fluctuation analysis of boiling water reactors." Nuclear Engineering and Design 421 (May 2024): 113106. http://dx.doi.org/10.1016/j.nucengdes.2024.113106.

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18

Sharaievskii, G. "Problems in Validation of the Chornobyl Accident Initiating Event." Nuclear and Radiation Safety, no. 1(69) (February 17, 2016): 20–27. http://dx.doi.org/10.32918/nrs.2016.1(69).03.

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The paper presents validation of known calculation dependencies used in RELAP-5 and other advanced computer codes to predict thermohydraulic anomalies from the standpoint of analyzing effect of initial coolant boiling in the Chornobyl accident on its further progression. The authors show current unsatisfactory efficiency of state-of-the-art computer codes in definition of the initial boiling point for the coolant in water-cooled nuclear reactors. The calculation methodology for improving accuracy in the predicting of dangerous thermal anomaly in reactor channels is under consideration.

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19

Huang, Hai, Chenxi Cao, Yue Wang, Youwei Yang, Jianning Lv, and Jing Xu. "Model-Based Analysis for Ethylene Carbonate Hydrogenation Operation in Industrial-Type Tubular Reactors." Processes 10, no. 4 (2022): 688. http://dx.doi.org/10.3390/pr10040688.

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Hydrogenation of ethylene carbonate (EC) to co-produce methanol (MeOH) and ethylene glycol (EG) offers an atomically economic route for CO2 utilization. Herein, aided with bench and pilot plant data, we established engineering a kinetics model and multiscale reactor models for heterogeneous EC hydrogenation using representative industrial-type reactors. Model-based analysis indicates that single-stage adiabatic reactors, despite a moderate temperature rise of 12 K, suffer from a narrow operational window delimited by EC condensation at lower temperatures and intense secondary EG hydrogenation at higher temperatures. Boiling water cooled multi-tubular reactors feature near-isothermal operation and exhibit better operability, especially under high pressure and low space velocity. Conduction oil-cooled reactors show U-type axial temperature profiles, rendering even wider operational windows regarding coolant temperatures than the water-cooled reactor. The revelation of operational characteristics of EC hydrogenation under industrial conditions will guide further improvement in reactor design and process optimization.

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20

Muñoz-Cobo, J. L., S. Chiva, and A. Escrivá. "Influence of subcooled boiling on out-of-phase oscillations in boiling water reactors." Nuclear Engineering and Design 235, no. 10-12 (2005): 1267–82. http://dx.doi.org/10.1016/j.nucengdes.2005.01.018.

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21

Lin, Chaung, Feng-Ling Jeng, Chi-Szu Lee, and Raghu Raghavan. "Hierarchical Fuzzy Logic Water-Level Control in Advanced Boiling Water Reactors." Nuclear Technology 118, no. 3 (1997): 254–63. http://dx.doi.org/10.13182/nt97-a35366.

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22

Khedr, Ahmed, Martina Adorni, and Francesco d’Auria. "The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios." Nuclear Technology and Radiation Protection 20, no. 1 (2005): 16–22. http://dx.doi.org/10.2298/ntrp0501016k.

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The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

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23

Loberg, John, Michael Österlund, Jan Blomgren, and Klaes-Håkan Bejmer. "Neutron Detection–Based Void Monitoring in Boiling Water Reactors." Nuclear Science and Engineering 164, no. 1 (2010): 69–79. http://dx.doi.org/10.13182/nse09-17.

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24

Sun, Bill K. H., Robert Colley, David G. Cain, and John W. Hallam. "Development of a Postscram Analyzer for Boiling Water Reactors." Nuclear Technology 76, no. 3 (1987): 352–59. http://dx.doi.org/10.13182/nt87-a33920.

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25

MATHIEU, CHARLES E. "THE APPLICATION OF BOILING WATER REACTORS TO SHIP PROPULSION." Journal of the American Society for Naval Engineers 72, no. 3 (2009): 503–8. http://dx.doi.org/10.1111/j.1559-3584.1960.tb02394.x.

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26

Liu, Maolong, Nejdet Erkan, Yuki Ishiwatari, and Koji Okamoto. "Passive depressurization accident management strategy for boiling water reactors." Nuclear Engineering and Design 284 (April 2015): 176–84. http://dx.doi.org/10.1016/j.nucengdes.2014.12.020.

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27

Olvera-Guerrero, Omar Alejandro, Alfonso Prieto-Guerrero, and Gilberto Espinosa-Paredes. "A non-linear stability monitor for boiling water reactors." Annals of Nuclear Energy 135 (January 2020): 106983. http://dx.doi.org/10.1016/j.anucene.2019.106983.

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28

John, T. M., and Om Pal Singh. "The Interpretation of Neutron Noise in Boiling Water Reactors." Nuclear Science and Engineering 89, no. 4 (1985): 322–29. http://dx.doi.org/10.13182/nse85-a18624.

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29

Lin, Chien C. "The Radiolytic Gas Production Rate in Boiling Water Reactors." Nuclear Science and Engineering 99, no. 4 (1988): 390–93. http://dx.doi.org/10.13182/nse88-a23567.

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30

Merkulov, Viktor, Nikolay Didenko, Djamilia Skripnuk, and Sergey Kulik. "Analysis of small modular reactor technologies and socio-economic aspects of their application in the Russian Arctic in the era of digital transformation." E3S Web of Conferences 402 (2023): 10011. http://dx.doi.org/10.1051/e3sconf/202340210011.

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Small modular reactor technologies and social, economic, and technological aspects of their application in the Russian Arctic are considered in the article. An overview of the key factors influencing an implementation of small modular reactor plants in remote regions with a decentralized power grid is presented. The main directions of small modular reactor design activities of the key Russian centers of atomic research and development are given. An overview of current Russian small modular reactor technologies including pressurized water reactors, boiling water reactors, reactors installed on floating nuclear power plants, high-temperature gas-cooled reactors, and liquid metal cooled reactor is conducted. Economic, social, ecological, and digital aspects of applications of small modular reactor in the Russian Arctic are considered. A detailed survey of areas of small modular reactor application including extractive, processing, industrial energy-intensive facilities, and power and heat supply of cities is also given. The importance of digital twins of small modular as an essential element in the development and maintenance of complex engineering products and industrial facilities throughout the entire life cycle is discussed in the article. Conclusions about key advantages and prospects of an application of small modular reactors in the Russian Arctic are made.

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31

Zhao, Jiyun, Pradip Saha, and Mujid S. Kazimi. "Core-Wide (In-Phase) Stability of Supercritical Water-Cooled Reactors—II: Comparison with Boiling Water Reactors." Nuclear Technology 161, no. 2 (2008): 124–39. http://dx.doi.org/10.13182/nt08-a3918.

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32

Pivovarov, V. "BOILING WATER REACTOR WITH TIGHT LATTICE OF FUEL RODS - DIRECT-CIRCUIT WATER-COOLED FAST REACTOR." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 2 (2019): 107–16. http://dx.doi.org/10.55176/2414-1038-2019-2-107-116.

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The original concept of the boiling water reactor core with the reduced moderation of neutrons is proposed, in which a negative void reactivity effect is provided not by increase leakage of neutrons in the axial direction, but by an another physical principle. Instead of the traditional core flattening, a special heterogeneous arrangement is proposed, in which, along with tight lattice fuel assemblies (fuel rod diameter is 13.5 mm, the distance between the fuel rods is 1.3 mm) containing uranium-plutonium (MOX) fuel, there are fuel assemblies with uranium-thorium fuel (UO2+ThO2) with a small (~1 %) initial content of 233U or 235U and an increased water-fuel ratio (fuel rod diameter is 12.6 mm, the distance between the fuel rods is 2.2 mm). Uranium-thorium assemblies provide a negative component of the reactivity effect during dehydration of the core. The results of the calculation of the reactor with a capacity of 3000 MW (t) showed the possibility of achieving a breeding ratio of fuel within 0.96-1.0 with a negative void reactivity effect (-0.2 %). The main advantages of the proposed concept are a directcircuit scheme, medium technological parameters close to traditional boiling reactors, allowing the use of available construction materials and equipment.

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33

D’Auria, F., N. Aksan, and H. Glaeser. "Physical Phenomena in Nuclear Thermal Hydraulics and Current Status." Tecnica Italiana-Italian Journal of Engineering Science 65, no. 1 (2021): 1–11. http://dx.doi.org/10.18280/ti-ijes.650101.

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116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

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34

Macdonald, Digby D., George R. Engelhardt, and Andrei Petrov. "A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part I, Assessment of Radiolysis Models." Corrosion and Materials Degradation 3, no. 3 (2022): 470–536. http://dx.doi.org/10.3390/cmd3030028.

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A critical review is presented on modeling of the radiolysis of the coolant water in nuclear power reactors with emphasis on ITER. The review is presented in two parts: In Part I, we assess previous work in terms of compliance with important chemical principles and conclude that no model proposed to date is completely satisfactory, in this regard. Thus, some reactions that have been proposed in various radiolysis models are not elementary in nature and can be decomposed into two or more elementary reactions, some of which are already included in the models. These reactions must be removed in formulating a viable model. Furthermore, elementary reactions between species of like charge are also commonly included, but they can be discounted upon the basis of Coulombic repulsion under the prevailing conditions (T < 350 °C) and must also be removed. Likewise, it is concluded that the current state of knowledge with respect to radiolytic yields (i.e., G-values) is also unsatisfactory. More work is required to ensure that the yields used in radiolysis models are truly “primary” yields corresponding to a time scale of nanoseconds or less. This is necessary to ensure that the impact of the reactions that occur outside of the spurs (ionizing particle tracks in the medium) are not counted twice. In Part II, the authors review the use of the radiolysis models coupled with electrochemical models to predict the water chemistry, corrosion potential, crack growth rate in Type 304 SS, and accumulated damage in the coolant circuits of boiling water reactors, pressurized water reactors, and the test fusion reactor, ITER. Based on experience with fission reactors, the emphasis should be placed on the control of the electrochemical corrosion potential because it is the parameter that best describes the state of corrosion in coolant circuits.

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35

Morreale, A. C., M. J. Brown, and S. M. Petoukhov. "PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES." AECL Nuclear Review 4, no. 1 (2015): 53–65. http://dx.doi.org/10.12943/anr.2014.00035.

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The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within AECL to evaluate modern severe accident modelling codes to the NRU reactor case to enhance the understanding of accident progression and predict the system damage and radiation release consequences of a severe accident, which is a very low probability event. The NRU is smaller and operates at a lower power than the large scale power reactors (e.g., pressurized heavy water reactors, pressurized water reactors, and boiling water reactors) that these codes were designed to analyze. Additionally, the NRU has a unique design different from the power reactors and several features relevant to severe accidents including filtered venting, large passive heat sinks, and a dispersion fuel design of uranium-silicide in an aluminum matrix. The major severe accident analysis codes available to AECL and their applicability to the NRU are explored in this paper. In addition, a preliminary strategy for employing the most applicable codes to the NRU for the purposes of severe accident modelling is proposed.

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36

Takagi, Junichi, and Kenkichi Ishigure. "Thermal Decomposition of Hydrogen Peroxide and Its Effect on Reactor Water Monitoring of Boiling Water Reactors." Nuclear Science and Engineering 89, no. 2 (1985): 177–86. http://dx.doi.org/10.13182/nse85-a18191.

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37

FORD, F. Peter. "Quantitative Prediction of Environmentally Assisted Cracking in Boiling Water Reactors." Proceedings of the Asian Pacific Conference on Fracture and Strength and International Conference on Advanced Technology in Experimental Mechanics 1.01.203 (2001): 27–39. http://dx.doi.org/10.1299/jsmeatemapcfs.1.01.203.0_27.

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38

Espinosa-Paredes, Gilberto, Alfonso Prieto-Guerrero, Alejandro Núñez-Carrera, and Rodolfo Amador-García. "Wavelet-Based Method for Instability Analysis in Boiling Water Reactors." Nuclear Technology 151, no. 3 (2005): 250–60. http://dx.doi.org/10.13182/nt05-a3647.

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39

Farawila, Yousef M., and Douglas W. Pruitt. "Critical Power Response to Power Oscillations in Boiling Water Reactors." Nuclear Science and Engineering 143, no. 3 (2003): 211–25. http://dx.doi.org/10.13182/nse03-a2331.

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40

Forsberg, Charles W. "Passive Emergency Cooling Systems for Boiling Water Reactors (PECOS-BWR)." Nuclear Technology 76, no. 1 (1987): 185–92. http://dx.doi.org/10.13182/nt87-a33909.

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41

Tiftikci, Ali, and Mehmet Türkmen. "Monte Carlo model of annular flow in boiling water reactors." Progress in Nuclear Energy 123 (May 2020): 103307. http://dx.doi.org/10.1016/j.pnucene.2020.103307.

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42

Vook, R. W., T. V. Rao, T. Swirbel, J. Bucci, and W. Meyer. "Thin films for radiation control in boiling water nuclear reactors." Proceedings, annual meeting, Electron Microscopy Society of America 44 (August 1986): 520–21. http://dx.doi.org/10.1017/s0424820100144115.

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Boiling water nuclear reactors (BWR's) experience radioactive film buildup on the inner walls of their out-of-core stainless steel (S.S.) cooling water pipes. These films consist of various oxides of Fe, Cr, and Ni, and contain small amounts of radioactive Co-60. As a result the pipes must be decontaminated or replaced periodically. Efforts are currently being made to passivate these S.S. surfaces so as to reduce the rate of radiation buildup. In the present work, the effects of various protective metallic thin film coatings on the morphology of the radioactive oxide film grown in a simulated BWR test loop are reported.

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43

Burte, D. P., and S. G. Vaidya. "Parametrization for optimization of reload patterns for boiling water reactors." Annals of Nuclear Energy 20, no. 4 (1993): 237–49. http://dx.doi.org/10.1016/0306-4549(93)90079-5.

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44

Wang, Haoyu, Andrew Longman, J. Thomas Gruenwald, James Tusar, and Richard Vilim. "Machine-Learning Analysis of Moisture Carryover in Boiling Water Reactors." Nuclear Technology 205, no. 8 (2019): 1003–20. http://dx.doi.org/10.1080/00295450.2019.1583957.

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45

Weeks, John R., Brijesh Vyas, and Hugh S. Isaacs. "Environmental factors influencing stress corrosion cracking in boiling water reactors." Corrosion Science 25, no. 8-9 (1985): 757–68. http://dx.doi.org/10.1016/0010-938x(85)90009-5.

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46

Sikorska, Daria, Julia Brzozowska, Agata Pawełkiewicz, Mateusz Psykała, Przemysław Błasiak, and Piotr Kolasiński. "Convective Heat Transfer in PWR, BWR, CANDU, SMR, and MSR Nuclear Reactors—A Review." Energies 17, no. 15 (2024): 3652. http://dx.doi.org/10.3390/en17153652.

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Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors.

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Paramanantham, SalaiSargunan S., Thanh-Hoang Phan, and Warn-Gyu Park. "Numerical analysis of bubble condensation behavior under high-pressure flow conditions." Proceedings of the Institution of Mechanical Engineers, Part C: Journal of Mechanical Engineering Science 234, no. 18 (2020): 3725–41. http://dx.doi.org/10.1177/0954406220916496.

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Heat transfer during subcooled flow boiling has a pivotal role in pressurized water reactors; it also occurs in boiling water reactors prior to the onset of saturated nucleate boiling. We examined the condensation behavior of vapor bubbles in the subcooled liquid phase using the fully compressible two-phase homogeneous mixture method, solved by an implicit dual-time preconditioned method. The continuous surface force method was applied to determine the surface tension between the phases in the simulation. To predict the empirical coefficient, we conducted a sensitivity study using Lee’s mass transfer model. For nuclear applications, we simulated high-pressure vapor–water conditions under higher mass flow conditions. The comparison of the numerical simulation and experimental results showed that the proposed model accurately predicted the condensation behavior of the bubble. Additionally, we investigated single bubble condensation behavior at different operating pressures, subcooling temperatures, bubble diameters, and bulk velocities. We also investigated the effects of high-pressure condensation on bubble shape. At lower subcooling temperatures, the condensation rate increased as pressure increased; however, at higher subcooling temperatures, pressure had no significant impact on the condensation rate.

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Wang, Mei-Ya, and Tsung-Kuang Yeh. "Evaluation of Early Hydrogen Water Chemistry on Corrosion Mitigation in Boiling Water Reactors." Nuclear Science and Engineering 186, no. 2 (2017): 180–89. http://dx.doi.org/10.1080/00295639.2016.1273014.

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Yefimov, Olexander, Mykola Pylypenko, Larysa Tiutiunyk, Tetyana Harkusha, Tetyana Yesipenko, and Anastasiia Motovilnik. "Construction Materials of Active Zones of New Generation Nuclear Reactors." NTU "KhPI" Bulletin: Power and heat engineering processes and equipment, no. 1-2 (August 7, 2023): 43–46. http://dx.doi.org/10.20998/2078-774x.2023.01.07.

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The materials of the article consider the analysis of construction materials of active zones of new generation nuclear reactors. The analysis reflects general ideas about the development of reactor technologies: in the 1950s and 1960s, the first generation of reactors was created; in the early 1970s, the operation of industrial reactors began - reactors of the second generation: pressurized water reactors (WWER, PWR), boiling water reactors (RBMK, BWR), heavy water reactors (CANDU), as well as gas-cooled reactors (AGR). Further development of some types of reactors made it possible to create reactors of the third generation in the 1980s. Priority when choosing directions of development in the category of revolutionary projects should have proposals capable of bringing a new quality to solving the problems of the nuclear energy industry of the future. Promising reactors have advantages in economy, safety, reliability and non-proliferation of nuclear materials. The effectiveness and reliability of structural materials are determined by the totality of changes in the characteristics of the materials as a result of the entire complex of phenomena occurring in them in the field of irradiation, in connection with the changing parameters and operating conditions. The use of high-purity metals as initial components of new structural materials and the development or optimization of their smelting technologies should ensure the required level of characteristics and properties of products made from them. The implementation of these concepts should be ensured by the development of new structural materials: ferritic-martensitic and austenitic steels, nickel and other new alloys.

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Xue, He, and Tetsuo Shoji. "Quantitative Prediction of EAC Crack Growth Rate of Sensitized Type 304 Stainless Steel in Boiling Water Reactor Environments Based on EPFEM." Journal of Pressure Vessel Technology 129, no. 3 (2006): 460–67. http://dx.doi.org/10.1115/1.2748827.

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The quantitative prediction of environmentally assisted cracking (EAC) or stress corrosion cracking (SCC) is essential in order to predict service life and also the structural integrity and safety assessment of light water reactors. During the last 3 decades many of the research results obtained on the quantitative prediction of the EAC crack growth rate have been based on linear fracture mechanics. In order to investigate EAC behavior in the high strain zone of important structures in light water reactors, the approach taken in this paper is one in which quantitative calculations of the EAC crack growth rate, incorporating the SCC deformation /oxidation model and the elastic-plastic finite element method (EPFEM), are carried out. This approach can be used for the quantitative prediction of EAC crack growth rate in both the low and high strain zones of key structures in light water reactors. The crack growth behavior of sensitized type 304 stainless steel with a 1T-CT specimen in simulated boiling water reactor (BWR) environments is analyzed based on this approach. The effect of several environmental, material, and mechanical parameters on the EAC crack growth rate of nickel based alloys in high-temperature aqueous environments is also discussed.

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