Journal articles: 'Nuclear boiling'

1

Wilczek, Frank. "Nuclear and subnuclear boiling." Nature 395, no. 6699 (September 1998): 220–21. http://dx.doi.org/10.1038/26107.

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Guo, Zhong De, and Shu Fang Zhang. "The Pure Heat Conversion Coefficient Analysis Method for Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit." Advanced Materials Research 383-390 (November 2011): 6514–18. http://dx.doi.org/10.4028/www.scientific.net/amr.383-390.6514.

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Pure heat conversion coefficient, which is one important parameter for the advance boiling water reactor nuclear power unit, is defined, and the coefficient reflects energy grade values of heaters. According to the structural characteristics of the thermodynamic system of the advanced boiling water reactor nuclear power plant, four sorts of auxiliary steam-water components are categorized. Via strict deduction and demonstration, the general matrix of the coefficient is deduced, so the thermal economic analysis of advance boiling water reactor nuclear power unit can be done with one of extraction steam efficiency. In this way, a new method of thermal economic quantitative analysis for this unit is offered.

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Giustini, Giovanni. "Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review." Inventions 5, no. 3 (September 14, 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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Kim, Kang Seog, Andrew Ward, Ugur Mertyurek, Mehdi Asgari, and William Wieselquist. "Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor." Journal of Nuclear Engineering 5, no. 3 (August 1, 2024): 260–73. http://dx.doi.org/10.3390/jne5030018.

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The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.

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Podowski, Michael Z., and Raf M. Podowski. "Mechanistic Multidimensional Modeling of Forced Convection Boiling Heat Transfer." Science and Technology of Nuclear Installations 2009 (2009): 1–10. http://dx.doi.org/10.1155/2009/387020.

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Due to the importance of boiling heat transfer in general, and boiling crisis in particular, for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive efforts have been made in the past to develop a variety of methods and tools to evaluate the boiling heat transfer coefficient and to assess the onset of temperature excursion and critical heat flux (CHF) at various operating conditions of boiling channels. The objective of this paper is to present mathematical modeling concepts behind the development of mechanistic multidimensional models of low-quality forced convection boiling, including the mechanisms leading to temperature excursion and the onset of CHF.

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Bang, In Cheol, Jacopo Buongiorno, Lin-Wen Hu, and Hsin Wang. "ICONE15-10030 Measurement of Key Pool Boiling Parameters in Nanofluids for Nuclear Applications." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_11.

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7

Tõke, Jan. "Boiling Patterns of Iso-asymmetric Nuclear Matter." Journal of Physics: Conference Series 420 (March 25, 2013): 012100. http://dx.doi.org/10.1088/1742-6596/420/1/012100.

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Stojanovic, Andrijana, Srdjan Belosevic, Nenad Crnomarkovic, Ivan Tomanovic, and Aleksandar Milicevic. "Nucleate pool boiling heat transfer: Review of models and bubble dynamics parameters." Thermal Science, no. 00 (2021): 69. http://dx.doi.org/10.2298/tsci200111069s.

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Understanding nucleate pool boiling heat transfer and, in particular the accurate prediction of conditions that can lead to critical heat flux, is of the utmost importance in many industries. Due to the safety issues related to the nuclear power plants, and for the efficient operation of many heat transfer units including fossil fuel boilers, fusion reactors, electronic chips, etc., it is important to understand this kind of heat transfer. In this paper, a comprehensive review of analytical and numerical work on nucleate pool boiling heat transfer is presented. In order to understand this phenomenon, existing studies on boiling heat transfer coefficient and boiling heat flux are also discussed, as well as characteristics of boiling phenomena such as bubble departure diameter, bubble departure frequency, active nucleation site density, bubble waiting and growth period and their impact on pool boiling heat transfer.

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Baldwin, Michael, Andre LeClair, Alok Majumdar, Jason Hartwig, Vishwanath Ganesan, and Issam Mudawar. "Modeling of cryogenic heated-tube flow boiling experiments of nitrogen and methane with Generalized Fluid System Simulation Program." IOP Conference Series: Materials Science and Engineering 1301, no. 1 (May 1, 2024): 012158. http://dx.doi.org/10.1088/1757-899x/1301/1/012158.

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Abstract Accurate modeling of cryogenic boiling heat transfer is vital for the development of extended-duration space missions. Such missions may require the transfer of cryogenic propellants from in-space storage depots or the cooling of nuclear reactors. Purdue University in collaboration with NASA has assembled a database of cryogenic flow boiling data points from heated-tube experiments dating back to 1959, which has been used to develop new flow boiling correlations specifically for cryogens. Computational models of several of these experiments have been constructed in the Generalized Fluid System Simulation Program (GFSSP), a network flow code developed at NASA’s Marshall Space Flight Center. The new Purdue-developed correlations cover the full boiling curve: onset of nucleate boiling, nucleate boiling, critical heat flux, and post-critical heat flux boiling. These correlations have been coded into GFSSP user subroutines. The fluids modeled are nitrogen and methane. Predictions of wall temperature are presented and compared to the test data.

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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 (February 19, 2012): 370–76. http://dx.doi.org/10.1007/s10512-012-9506-9.

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11

Nguyen, Ngoc Dat, and Van Thai Nguyen. "Performance Comparison of ANN-Based Model and Empirical Correlations for Void Fraction Prediction of Subcooled Boiling Flow in Vertical Upward Channel." Nuclear Science and Technology 11, no. 4 (January 13, 2023): 07–18. http://dx.doi.org/10.53747/nst.v11i4.335.

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The accurate prediction of void fraction parameter in subcooled boiling flow is very important for nuclear safety since it has significant influences on the mass flow rate, the onset of two-phase flow instability, and the heat transfer characteristics in a nuclear reactor core. Many different models and empirical correlations have been established over a variety of input conditions; however, this classical approach could lead to unsatisfactory prediction due to the uncertainties of model parameter and model forms. To cope with these limitations, Artificial Neural Network (ANN) is a powerful machine learning tool for modeling and solving non-linear and complicated physical problems. Therefore, this work is aim at developing an ANN-based model to predict the local void fraction of subcooled boiling flows. The comparison results of the performance between the ANN-based model and empirical correlations for the void fraction prediction of subcooled boiling in vertical upward channel showed the potential use of ANN-based model in the Computational Fluid Dynamics (CFD) codes to accurately simulate the subcooled boiling phenomena.

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12

Tieszen, S., H. Merte, V. S. Arpaci, and S. Selamoglu. "Crevice Boiling in Steam Generators." Journal of Heat Transfer 109, no. 3 (August 1, 1987): 761–67. http://dx.doi.org/10.1115/1.3248155.

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Experimental results are presented on the influence of confinement (normal to heated surface) on nucleate boiling in forced flow. The forced flow conditions and confinement geometry studied are similar to those found for boiling between a primary-fluid tube and a tube-support plate in steam generators of pressurized-water-reactor nuclear power plants. Visual observations of the boiling process within the confined region (crevice) between the tube and its support plate, obtained by high-speed photography, are related to simultaneous two-dimensional temperature maps of the hot primary-fluid-tube surface. The results demonstrate the existence of three confinement-dependent boiling regimes in forced flow conditions that are similar to those found in pool boiling conditions. These regimes are shown to be associated with the Weber number.

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13

Yang, X., S. Y. Jiang, and Y. Zhang. "Experimental and numerical investigation of sub-cooled boiling, condensation, and void flashing in nuclear heating reactor test loop." Kerntechnik 67, no. 2-3 (April 1, 2002): 90–94. http://dx.doi.org/10.1515/kern-2002-0041.

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Abstract This paper describes experimental and numerical investigations of sub-cooled boiling, condensation, and void flashing in the HRTL-5 test loop, which simulates the primary loop of a 5 MW nuclear heating reactor. A drift-flow model of two-phase with four governing equations was used, in which sub-cooled boiling, condensation, and void flashing have been taken into account. Based on the mathematical model, a program has been developed for analyzing the natural circulation system. As parameters, inlet sub-cooling, system pressure, and heat flux are varied. For comparison, some simplified models, which are designed to reveal the importance of sub-cooled boiling, condensation, flashing in the HRTL-5 test loop, are adopted in the program. The results show: (1) sub-cooled boiling, condensation, and void flashing may have great influence on the distribution of the void fraction and more intense at low system pressure; (2) the calculation of them is correlative and interactive other than independent; (3) for a system with short heated section, long riser, and low pressure, it is possible to reach “boiling out of the core”, where there is almost no void in the heated section, but much in the riser.

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14

Kunugi, Tomoaki, and Yasuo Ose. "Direct Numerical Simulation and Visualization of Subcooled Pool Boiling." Science and Technology of Nuclear Installations 2014 (2014): 1–11. http://dx.doi.org/10.1155/2014/120604.

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A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors). On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated.

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15

Cho, A. "NUCLEAR PHYSICS: Scheme for Boiling Nuclear Matter Gathers Steam at Accelerator Lab." Science 312, no. 5771 (April 14, 2006): 190–91. http://dx.doi.org/10.1126/science.312.5771.190.

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16

Mukherjee, Sayantan, Shikha Ebrahim, Purna Chandra Mishra, Naser Ali, and Paritosh Chaudhuri. "A Review on Pool and Flow Boiling Enhancement Using Nanofluids: Nuclear Reactor Application." Processes 10, no. 1 (January 17, 2022): 177. http://dx.doi.org/10.3390/pr10010177.

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Plasma-facing components (PFCs) are used as the barrier to the beam of high heat flux generated due to nuclear fusion. Therefore, efficient cooling of PFCs is required for safety and smooth operation of a fusion reactor. The Hyper Vapotron (HV) is generally used as the heat exchanger to cool down the PFCs during operation. These heat exchangers use pool and flow boiling mechanisms, and hence, their ability is inherently constrained by critical heat flux (CHF). The boiling of nanofluid is very promising as the working fluid in the HV. The efficiency of the HV increases due to the increase in CHF by applying nanofluids. However, the feasibility of nanofluid cooling in fusion reactors needs proper understanding. This paper reviews the recent developments in the utilization of boiling phenomena in nanofluid as a coolant in the HV. Experiments, theoretical studies, significant achievements, and challenges are analyzed and discussed. Finally, important points are indicated for future research.

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17

Refaey, A. M., S. Elnaggar, S. H. Abdel-Latif, and A. Hamza. "The effect of surfactant concentrations and surface material on heat transfer coefficient in nucleate boiling regime." Kerntechnik 86, no. 5 (October 1, 2021): 365–74. http://dx.doi.org/10.1515/kern-2020-0064.

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Abstract The nucleate boiling regime and two-phase flow are greater importance to the safety analysis of nuclear reactors. In this study, the boiling heat transfer in nuclear reactor is numerical investigated. The computational fluid dynamics (CFD) code, ANSYS Fluent 17.2 is used and the boiling model is employed. The numerical predictions obtained are compared with the experimental data reported by A. Hamza et al. [9]. An experimental test rig is designed and constructed to investigate the effect of cooling water chemistry control and the material of heater surface. CFD software, allows the detailed analysis of the two-phase flow and heat transfer. In this paper, we evaluate the accuracy of the boiling model implemented in the ANSYS Fluent code. This model is based on the heat flux partitioning approach and accommodates the heat flux due to single-phase convection, quenching and evaporation. The validation carried out of surfactant fluid/vapor two-phase flow inside the 2-D cylindrical boiling vessel. A heated horizontal pipe with stainless steel, Aluminum, and Zircalloy surface materials are used to numerically predict the field temperature and void fraction. Different surfactant concentrations ranging from 0, (pure water) to 1500 ppm, and heat fluxes ranging from 31 to 110 kW/m2 are used. The results of the predicted model depict that the addition of SDS Surfactant and increasing the heat flux improves the coefficient of boiling heat transfer for a given concentration. Also, it was found that the increasing of the concentration of aqueous surfactant solution increases the pool boiling heat transfer coefficient. The aqueous surfactant solution SDS improved the heat transfer coefficient of Aluminum, Zircalloy and stainless steel surface materials by 135%.138% and 120% respectively. The results of the numerical model are nearly in agreement with that measured in experimental.

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18

Ma, Xiaojing, and Ping Cheng. "Numerical Simulation of Complete Pool Boiling Curves: From Nucleation to Critical Heat Flux Through Transition Boiling to Film Boiling." Nuclear Science and Engineering 193, no. 1-2 (September 6, 2018): 1–13. http://dx.doi.org/10.1080/00295639.2018.1504566.

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19

SATO, ITARU, JUN SASAKI, HIROSHI SATOH, and KEIJI OKADA. "Effects of Treatment Time and Thickness of Meat on the Removal of Radioactive Cesium from Beef Slices by Boiling and Water Extraction." Journal of Food Protection 82, no. 4 (March 26, 2019): 623–27. http://dx.doi.org/10.4315/0362-028x.jfp-18-427.

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ABSTRACT Since the Fukushima nuclear accident, radioactive contamination of foods has become a serious concern among the Japanese people. In the present study, the decontamination effects of boiling and water extraction on beef slices containing radioactive cesium were examined, focusing on the effects of beef thickness and treatment time. Boiling effectively decontaminated the beef slices depending on the treatment time and the thickness of the beef. The time needed for 50% decontamination (T½) was estimated at 0.25, 0.89, 2.0, and 20 min for beef slices 1, 2, 4, and 10 mm thick, respectively. Water extraction also decontaminated the beef, but its efficiency was far less than that of boiling; the T½ of water extraction was 6.8, 24, and 187 min for beef slices 2, 4, and 10 mm thick, respectively. The T½ increased nearly in proportion to the square of the thickness of the beef slice; the thinner the slice, the more quickly it was decontaminated. The boiling treatment was approximately 7 to 10 times more effective than water extraction based on the T½ value. Foods currently distributed in the Japanese market do not need to be decontaminated; however, this information will be helpful for people who are anxious about radioactive contamination caused by the nuclear accident. HIGHLIGHTS

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Chiang, Ren-Tai. "Overview on Steady-state Nuclear Methods for BWR Nuclear Core Design and Analysis." ASEAN Journal on Science and Technology for Development 35, no. 3 (December 24, 2018): 223–27. http://dx.doi.org/10.29037/ajstd.514.

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An overview on nuclear methods for boiling water reactors (BWR) core design and analysis is provided based on the ANS Standard 19.3. The steady-state BWR nuclear methods, composed of neutron cross section library generation method, lattice physics method and core physics method, are systematically reviewed and associated computer codes in common use for BWR core design and analysis are listed. Veriﬁcation and validation, the two complementary aspects in determining the range of applicability of the calculation system, are discussed extensively. The biases and uncertainties for the predictions from the calculation system over its demonstrated range of applicability are also discussed.

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21

Chuang, T. J., and Y. M. Ferng. "Experimentally investigating boiling characteristics in the transition boiling for the downward facing heating." Annals of Nuclear Energy 91 (May 2016): 148–55. http://dx.doi.org/10.1016/j.anucene.2016.01.004.

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22

Lahaye, Marc, Cyrille Rochas, and Wilfred Yaphe. "A new procedure for determining the heterogeneity of agar polymers in the cell walls of Gracilaria spp. (Gracilariaceae, Rhodophyta)." Canadian Journal of Botany 64, no. 3 (March 1, 1986): 579–85. http://dx.doi.org/10.1139/b86-074.

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The heterogeneity of agar polymers in the cell walls of Gracilaria tenuistipitata, G. eucheumoides, and G. blodgettii was demonstrated by sequential extraction of the algae in water at 22 °C; in boiling 80, 60, 40, and 20% ethanol–water solutions; and in water at 100 and 121 °C. Agar polymers in each extract were analyzed by 1H and 13C nuclear magnetic resonance spectroscopy. The agar polymers extracted from the three algae with water at 22 °C had a high concentration of repeat units substituted by alkali-labile sulfate. Methylated agar polymers were extracted with boiling 80 and 60% ethanol from G. tenuistipitata and G. eucheumoides, the former having a high content of 6-O-methyl-D-galactose and a low concentration of 2-O-methyl-3,6-anhydro-L-galactose and the latter consisting mainly of 2-O-methyl-3,6-anhydro-L-galactose. Sulfated agar polymers (alkali-stable sulfate) were extracted from G. tenuistipitata and G. blodgettii with boiling 20% ethanol and water (100 and 121 °C). Floridean starch was detected by 13C nuclear magnetic resonance in the extracts from G. tenuistipitata and G. eucheumoides.

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23

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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24

Bankoff, S. G., and T. E. Rehm. "Convective Boiling in Narrow Concentric Annuli." Journal of Engineering for Gas Turbines and Power 112, no. 4 (October 1, 1990): 607–13. http://dx.doi.org/10.1115/1.2906213.

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An experimental apparatus was used to simulate the annulus formed by a single tube passing through a tube support plate (TSP) in the steam generator of a PWR nuclear power plant. It was found that the extent of thin film evaporation, compared to nucleate boiling, is larger for smaller annuli, with a transition at 0.203 mm gap width to a local dryout/rewet condition. A model was developed that predicts an increasing extent of surface dryout/rewet prior to critical heat flux (CHF). A CHF correlation was developed by modification of a pool boiling CHF correlation, and the Chen (1966) correlation was modified to allow prediction of the two-phase heat transfer coefficient.

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Dong, Wu-han, Ming Gao, Zhong-xiang Shen, and Li-xin Zhang. "Study on boiling heat transfer mechanism based on microlayer evaporation theory." Journal of Physics: Conference Series 2280, no. 1 (June 1, 2022): 012034. http://dx.doi.org/10.1088/1742-6596/2280/1/012034.

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Abstract Based on dynamic microlayer evaporation model. Heat transfer mechanism of nuclear boiling and microlayer thickness under single bubble on heated surface. It has a lot to do with the temperature of the wall. Therefore, this paper studies the mechanism of bubble microlayer and makes comparative analysis with its experiments. It is helpful to master nuclear boiling heat transfer more comprehensively. In the study of individual bubble microlayer thickness. The effects of time and radius on wall temperature are considered in this paper. The theoretical formula of initial thickness of microlayer is deduced. The theoretical formula is compared with the experimental data of microlayer evaporation under the nucleation bubble. It was found that the experimental data fell within ±25% of the formula range. It is consistent with the previous experimental research.

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Lipka, Maciej, Gawel Madejowski, Rafal Prokopowicz, and Krzysztof Pytelt. "Approximate model for evaluation of thermal-hydraulic transients associated with rapid power increase in research nuclear reactor." Nuclear Technology and Radiation Protection 35, no. 4 (2020): 310–15. http://dx.doi.org/10.2298/ntrp2004310l.

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A simple model, for the estimation of changes in the nuclear fuel element cladding temperature as well as the conditions of the formation of the onset of nucleate boiling, is proposed. The results of this estimation are sufficient to assess the effect of the transient with the peak of the reactor's power on the device's safety, without the need for time-consuming thermal calculations. The basic parameters determined using the proposed model are the maximum wall temperature of the device in a hot spot, the time constant of the wall temperature change, and the course of changes in the onset of nucleate boiling ratio in time. The model may be used for investigating the thermal behavior of thin heat-generating and water-cooled elements (such as fuel elements or uranium irradiation targets) during rapid power rise. The results of the temperature estimation with the proposed model were tested considering the hot spot in the MR-6 type nuclear fuel. The SN code with coupled kinetics and thermal-hydraulics, developed in the MARIA reactor was used to validate the results. The maximum cladding temperature, the thermal time constant and the onset of nucleate boiling ratio parameter simulated by the SN code and the proposed scheme appeared to be very consistent.

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John, T. M., and O. P. Singh. "Coolant boiling noise in LMFBRs." Annals of Nuclear Energy 12, no. 1 (January 1985): 45–47. http://dx.doi.org/10.1016/0306-4549(85)90007-6.

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28

Sorokin, V. V. "Boiling and crisis of boiling of a two-phase liquid in spherical microfuel fills." Atomic Energy 106, no. 1 (January 2009): 17–25. http://dx.doi.org/10.1007/s10512-009-9125-2.

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Avramenko, A. A., A. I. Tyriniv, N. P. Dmitrenko, and M. M. Kovetska. "INFLUENCE OF UNSTEADY CONDITIONS ON HEAT EXCHANGE DURING A SHARPY TRANSITION TO FILM BOILING." Thermophysics and Thermal Power Engineering 46, no. 3 (February 2, 2022): 23–32. http://dx.doi.org/10.31472/ttpe.3.2022.2.

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Boiling is one of the main physical processes, which that take place in heat exchange equipment designed for various purposes. The problem of removing large thermal loads from the heated surface is important for nuclear energy, chemical industry, metallurgy, electronics and other areas where intense heat is released. Boiling processes in process equipment perform important protective functions and can control its effectiveness. According to the boiling curve, with increasing temperature power, the flow passes through five regions, starting from the single-phase region of free convection and ending with the region of developed film boiling. The purpose of this article is an analytical study of heat transfer at spontaneous transition to the film boiling (explosive type of boiling), taking into account the unsteady nature of this process. In order to achieve the aim of this research, two analytical approaches were used, namely, the symmetry method and the Laplace method. As a result of mathematical transformations, expressions for the nonstationary temperature distribution and the Nusselt number are obtained. The obtained expressions make it possible to analyze the dynamics of non-stationary heat exchange processes. The results of analytical and numerical modeling were also compared. It was found that the results of the self-similar solution have a better comparison with numerical data compared to the results according to the Laplace method.

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Bucci, M., M. Zupančič, and I. Golobič. "Multi-scale boiling heat transfer investigation on micro-thin aluminum heaters." Journal of Physics: Conference Series 2766, no. 1 (May 1, 2024): 012128. http://dx.doi.org/10.1088/1742-6596/2766/1/012128.

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Abstract Nucleate boiling is a highly efficient heat transfer mode distinguished by the liquid-vapor phase change, which occurs through the formation, growth, and detachment of vapor bubbles from a heated surface. Its crucial role in various industrial applications, such as nuclear power plant operation and effective heat management in small electronic devices, has driven significant research efforts. However, despite extensive research dedicated to boiling investigations, there are still substantial knowledge gaps that hinder our ability to accurately predict heat removal rates. These knowledge gaps arise from the complex nature of small-scale boiling phenomena, which are further complicated by their strong dependence on operating conditions and the interactions between walls and fluids. In an effort to address some of these gaps, we conducted multi-scale investigations during pool boiling of de-ionized water on micro-thin aluminum heaters. We captured bubble dynamics through multiple synchronized diagnostic sources, including high-speed backlit imaging to track bubble growth, synchronized high-speed infrared thermometry to capture the corresponding thermal footprint on the boiling surface, and in-house developed fast-response micro-thermocouples to measure temperature at multiple locations within the fluid. Our study reveals peculiar aspects of heat transfer mechanisms occurring at single bubble level (low heat fluxes) and in fully developed nucleate boiling regimes (high heat fluxes).

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Ceceñas-Falcón, Miguel, and Robert M. Edwards. "Stability Monitoring Tests Using a Nuclear-Coupled Boiling Channel Model." Nuclear Technology 131, no. 1 (July 2000): 1–11. http://dx.doi.org/10.13182/nt00-a3100.

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32

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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33

Jaunet, Y., M. Bucci, M. Zupančič, J. Sebilleau, C. Colin, and I. Golobič. "Study of nucleate boiling growth regime on thin surfaces." Journal of Physics: Conference Series 2766, no. 1 (May 1, 2024): 012135. http://dx.doi.org/10.1088/1742-6596/2766/1/012135.

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Abstract In a context of energy transition, heat exchanges are a key for energy sobriety. Among those exchanges, nucleate boiling is the preferred heat transfer method for high heat flux. Vaporisation plays an important role in the nuclear industry, but also in thermal machines such as cooling systems in data centers. In spite of decades studying boiling thermodynamics, modelling heat transfers in the nucleate boiling is still a major difficulty because of its numerous parameters. In the 1960s, scientists developed promising models called heat flux partitioning models. These models have been improved over the years, but they need closure laws about bubbles dynamics and nucleation site density. For the purpose of improving those models, we took part in an international collaboration (ANR TraThI) to study interface heat transfer. Project focuses on boiling with a controlled nucleation site density with a strong emphasis on studying isolated bubbles in water pool boiling, and later addressing multiple bubble interactions in pool and flow boiling. This work focuses on the bubble growth regimes on thin metallic foil observed in a water pool boiling experiment, with a distinction between microlayer and contact line growth regime. Pulsed nanosecond laser was used to create active nucleation site, while high-speed infrared thermography and shadowgraphy were implemented to record transient wall temperatures and bubble dynamics, respectively. This work shows that our experimental configuration induced two different bubble growth regimes, depending on the imposed heat flux and the recent past at the nucleation site and its vicinity. Study provides a framework for further in-depth investigations with different experimental configurations.

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Qing, Ning, Lawrence D. Colebrook, John T. Edward, Allan Kon, and Francis L. Chubb. "Reactions of α-phenylglycinamide with some carbonyl compounds. Formation of 5,7-diisopropyl-8,8-dimethyl-2-oxo-3-phenylimidazolidino-[1,2-c]-tetrahydro-[1,3]-oxazine, and determination of structure and stereochemistry by nuclear Overhauser effect difference measurements." Canadian Journal of Chemistry 67, no. 10 (October 1, 1989): 1560–64. http://dx.doi.org/10.1139/v89-238.

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α-Phenylglycinamide 5 reacts with ketones to give either the Schiff base 3 or the 4-imidazolidone 4, but reacts with boiling isobutyraldehyde to give the bicyclic compond 6. The structure and stereochemistry of 6 have been established by 1H nuclear Overhauser effect difference measurements, supported by molecular mechanics calculations. Keywords: structure, stereochemistry, nuclear Overhauser effect, molecular mechanics, α-phenylglycinamide reactions.

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35

Schindler, M., and LW Jiang. "Epidermal growth factor and insulin stimulate nuclear pore-mediated macromolecular transport in isolated rat liver nuclei." Journal of Cell Biology 104, no. 4 (April 1, 1987): 849–53. http://dx.doi.org/10.1083/jcb.104.4.849.

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Fluorescence photobleaching was used to measure the effect of epidermal growth factor (EGF), insulin, and glucagon on the nuclear transport of fluorescent-labeled dextrans across the nuclear pore complex. EGF and insulin were found to stimulate transport approximately 200%, while boiling these polypeptide growth factors greatly diminished this enhancement activity. Glucagon demonstrated no enhancement effect. The nuclear transport enhancement effects were observed at EGF and insulin concentrations that elicit the various physiological responses, e.g., nanomolar range.

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Avramova, Maria, and Diana Cuervo. "Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases." Science and Technology of Nuclear Installations 2013 (2013): 1–12. http://dx.doi.org/10.1155/2013/508485.

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Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

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Fujita, Nobuyuki, and David A. Rice. "Core Boiling During Midloop Operation." Nuclear Technology 93, no. 1 (January 1991): 36–46. http://dx.doi.org/10.13182/nt91-a34516.

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38

Rahnema, Farzad, Dan Ilas, and Shivakumar Sitaraman. "Boiling Water Reactor Benchmark Calculations." Nuclear Technology 117, no. 2 (February 1997): 184–94. http://dx.doi.org/10.13182/nt97-a35324.

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Melikhov, Vladimir, Oleg Melikhov, Sergey Yakush, and Oleg Konovalov. "Comparative analysis of film boiling correlations for steam explosion problem." E3S Web of Conferences 411 (2023): 01064. http://dx.doi.org/10.1051/e3sconf/202341101064.

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The article considers the existing models and correlations for describing heat transfer during film boiling. Film boiling is an important thermophysical process that determines the course of interaction of the melt of reactor core materials with the coolant, which can potentially occur during a severe accident at a nuclear power plant with a pressurized water-cooled reactor (VVER/PWR). The values of the Nusselt number predicted by these models are compared. A fairly significant dispersion of calculated parameters has been established, making it difficult to unambiguously choose one or another correlation. However, in the range of parameters typical for the interaction of a high-temperature melt of reactor core materials with a coolant, several correlations give quite close values, which allows them to be recommended for use in calculation codes for modeling thermal-hydraulic processes during a severe accident at a nuclear power plant with VVER/PWR type.

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Mennerdahl, Dennis. "KRITZ-1-Mk CRITICAL MEASUREMENTS AT TEMPERATURES FROM 20 °C TO 250 °C." EPJ Web of Conferences 247 (2021): 09028. http://dx.doi.org/10.1051/epjconf/202124709028.

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Benchmarks are needed to validate methods to account for temperature-dependence of nuclear data. An evaluation of 37 KRITZ-1-Mk critical water height measurements, together with associated iso-reactivity temperature effects and coefficients, is released with the 2019 Handbook of the International Reactor Physics Experiment Evaluation Project (IRPhEP). The KRITZ zero-power research reactor, operated between 1969 and 1975 in Studsvik (Sweden), was contained in a pressure vessel, allowing full size fuel assemblies or fuel rods in light water at temperatures up to 250 °C without boiling. Preliminary results were published in 1971 and 1972 for four series of altogether 37 measurements with Marviken (Boiling Heavy Water Reactor) UO2 fuel rods, each containing a 235U isotopic mass fraction of 1.35 %. Temperature was the predictor variable, while critical water height was the response variable. Each series was characterized by the fuel rod lattice design and by the soluble boron concentration in water. The KRITZ measurements were focused on temperature-dependence (differences). High measurement correlations reduced the ?k uncertainties, typically from 195 pcm to 40 pcm for a large temperature change. Thermal expansion of fuel and reactor components was not measured. Detailed and simple benchmarks include estimated thermal expansion as a simplification. Benchmark calculation results using JEFF-3.3 nuclear data reduce the large biases observed for older libraries but a remarkable positive temperature trend is observed for series 4. In 2019, Studsvik Nuclear released information on KRITZ-1-Mk and on other KRITZ-1 and KRITZ-2 critical measurements with Boiling Water Reactor fuel assemblies and fuel rod clusters.

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Kozhemyakin, V. V., and N. D. Koshkin. "Evaluation of the possibility of using vapor-liquid injectors in installations with a liquid metal coolant." Transactions of the Krylov State Research Centre S-I, no. 1 (December 8, 2021): 169–70. http://dx.doi.org/10.24937/2542-2324-2021-1-s-i-169-170.

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The possibility of using vapor-liquid injectors in a nuclear steam generating plant with a liquid metal coolant is analyzed. In order to ensure the boiling of liquid metal at acceptable temperatures, the option of reducing the metal pressure was considered.

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Rzehak, Roland, and Eckhard Krepper. "CFD for Subcooled Flow Boiling: Parametric Variations." Science and Technology of Nuclear Installations 2013 (2013): 1–22. http://dx.doi.org/10.1155/2013/687494.

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We investigate the present capabilities of CFD for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Very similar modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant nondimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, liquid temperature, and bubble size. Robust predictive capabilities of the modeling require that it is validated for a wide range of parameters. It is known that a careful calibration of correlations used in the wall boiling model is necessary to obtain agreement with the measured data. We here consider tests under a variety of conditions concerning liquid subcooling, flow rate, and heat flux. It is investigated to which extent a set of calibrated model parameters suffices to cover at least a certain parameter range.

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ARAKI, Hitoshi, Kazuo HAGA, and Koichiro NAKAMOTO. "Sodium boiling detection by acoustic method." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 28, no. 2 (1986): 176–84. http://dx.doi.org/10.3327/jaesj.28.176.

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Chen, X. N., F. Gabrielli, A. Rineiski, and T. Schulenberg. "Boiling water cooled travelling wave reactor." Annals of Nuclear Energy 134 (December 2019): 342–49. http://dx.doi.org/10.1016/j.anucene.2019.06.037.

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Kim, Seung Jun, Russell C. Johns, Junsoo Yoo, and Emilio Baglietto. "Progress Toward Simulating Departure from Nucleate Boiling at High-Pressure Applications with Selected Wall Boiling Closures." Nuclear Science and Engineering 194, no. 8-9 (May 4, 2020): 690–707. http://dx.doi.org/10.1080/00295639.2020.1743579.

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Končar, Boštjan, and Borut Mavko. "Simulation of Boiling Flow Experiments Close to CHF with the Neptune_CFD Code." Science and Technology of Nuclear Installations 2008 (2008): 1–8. http://dx.doi.org/10.1155/2008/732158.

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A three-dimensional two-fluid code Neptune_CFD has been validated against the Arizona State University (ASU) and DEBORA boiling flow experiments. Two-phase flow processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function has been implemented in the Neptune_CFD code aiming to improve the prediction of flow parameters in the near-wall region. The capability of the code to predict the boiling flow regime close to critical heat flux (CHF) conditions has been verified on selected DEBORA experiments. To predict the onset of CHF regime, a simplified model based on the near-wall values of gas volume fraction was used. The results have shown that the code is able to predict the wall temperature increase and the sharp void fraction peak near the heated wall, which are characteristic phenomena for CHF conditions.

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Mimouni, Stephane, William Benguigui, Solène Fleau, Arnaud Foissac, Mathieu Guingo, Mickael Hassanaly, Jérôme Lavieville, et al. "Dispersed Two-Phase Flow Modelling for Nuclear Safety in the NEPTUNE_CFD Code." Science and Technology of Nuclear Installations 2017 (2017): 1–41. http://dx.doi.org/10.1155/2017/3238545.

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The objective of this paper is to give an overview of the capabilities of Eulerian bifluid approach to meet the needs of studies for nuclear safety regarding hydrogen risk, boiling crisis, and pipes and valves maintenance. The Eulerian bifluid approach has been implemented in a CFD code named NEPTUNE_CFD. NEPTUNE_CFD is a three-dimensional multifluid code developed especially for nuclear reactor applications by EDF, CEA, AREVA, and IRSN. The first set of models is dedicated to wall vapor condensation and spray modelling. Moreover, boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. The paper aims at presenting the generalization of the previous DNB model and its validation against 1500 validation cases. The modelling and the numerical simulation of cavitation phenomena are of relevant interest in many industrial applications, especially regarding pipes and valves maintenance where cavitating flows are responsible for harmful acoustics effects. In the last section, models are validated against experimental data of pressure profiles and void fraction visualisations obtained downstream of an orifice with the EPOCA facility (EDF R&D). Finally, a multifield approach is presented as an efficient tool to run all models together.

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Bang, In-Cheol, and Ji-Hwan Jeong. "NANOTECHNOLOGY FOR ADVANCED NUCLEAR THERMAL-HYDRAULICS AND SAFETY: BOILING AND CONDENSATION." Nuclear Engineering and Technology 43, no. 3 (June 25, 2011): 217–42. http://dx.doi.org/10.5516/net.2011.43.3.217.

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GopaL, M., and P. Pratapachandran Nair. "A New Oftimal Control Strategy for a Nuclear Boiling Water Reactor." IEEE Transactions on Nuclear Science 32, no. 2 (1985): 1180–89. http://dx.doi.org/10.1109/tns.1985.4333572.

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Rao, T. V., R. W. Vook, W. Meyer, and C. Wittwer. "Protective coatings for radiation control in boiling water nuclear power reactors." Journal of Vacuum Science & Technology A: Vacuum, Surfaces, and Films 5, no. 4 (July 1987): 2701–5. http://dx.doi.org/10.1116/1.574723.

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Journal articles on the topic 'Nuclear boiling'

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