Relevant bibliographies by topics / Nucleate boiling degradation

Academic literature on the topic 'Nucleate boiling degradation'

Author: Grafiati
Published: 10 December 2022
Last updated: 28 January 2023

Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles

Select a source type:

Contents

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Nucleate boiling degradation.'

Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.

You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.

Journal articles on the topic "Nucleate boiling degradation"

1

Xia, Da-Hai, Bing Zhou, Jianqiu Wang, Zhiming Gao, Jihui Wang, Jing-Li Luo, and Chen Shen. "Passivation degradation of Alloy 800 on nucleate boiling surface." Corrosion Engineering, Science and Technology 52, no. 5 (March 30, 2017): 391–96. http://dx.doi.org/10.1080/1478422x.2017.1306991.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Ware, A. G., and V. N. Shah. "Age-related degradation of boiling water reactor vessel internals." Nuclear Engineering and Design 133, no. 1 (February 1992): 49–62. http://dx.doi.org/10.1016/0029-5493(92)90088-d.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Pshenichnikov, Anton, Saishun Yamazaki, David Bottomley, Yuji Nagae, and Masaki Kurata. "Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors." Journal of Nuclear Science and Technology 56, no. 5 (March 20, 2019): 440–53. http://dx.doi.org/10.1080/00223131.2019.1592724.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Xi, Mengmeng, Rong Cai, Xiao Chu, Fan Yang, Xu Ran, Zhifang Qiu, Jian Deng, and Wenxi Tian. "Theoretical Study on the Characteristics of Critical Heat Flux in Rectangular Channel of Natural Circulation under Motion Conditions." Science and Technology of Nuclear Installations 2020 (September 2, 2020): 1–18. http://dx.doi.org/10.1155/2020/9390645.

Full text
Abstract:
With the wide application of sea-based reactors, the impact of ocean conditions on the safety performance of reactors has gradually attracted attention. In this paper, by establishing the thermal hydraulic transient analysis model and the critical heat flux (CHF) model of natural circulation system, the CHF characteristics in the rectangular channel of natural self-feedback conditions under ocean conditions are studied. The results show that the additional acceleration field generated by ocean conditions will affect the thermal hydraulic parameters of the natural circulation system, that is, the external macroscopic thermal hydraulic field. On the other hand, the boiling crisis mechanism will be affected, that is, the force on the bubble and the thickness of the liquid film. Within the parameters of the study, ocean conditions have a great impact on CHF of natural circulation, and the maximum degradation of CHF is about 45%. The obtained analysis results are significant to the improvement of design and safety operation of the reactor system.
APA, Harvard, Vancouver, ISO, and other styles
5

Galushin, Sergey, and Pavel Kudinov. "Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code." Science and Technology of Nuclear Installations 2019 (April 17, 2019): 1–18. http://dx.doi.org/10.1155/2019/5310808.

Full text
Abstract:
Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.
APA, Harvard, Vancouver, ISO, and other styles
6

Pshenichnikov, Anton, Masaki Kurata, David Bottomley, Ikken Sato, Yuji Nagae, and Saishun Yamazaki. "New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1: introduction of boiling water reactor mock-up assembly degradation test programme." Journal of Nuclear Science and Technology 57, no. 4 (November 18, 2019): 370–79. http://dx.doi.org/10.1080/00223131.2019.1691070.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Dhir, Vijay K. "Advances in Understanding of Pool Boiling Heat Transfer—From Earth on to Deep Space." Journal of Heat Transfer 141, no. 5 (April 15, 2019). http://dx.doi.org/10.1115/1.4043282.

Full text
Abstract:
In this work, the effectiveness of the numerical simulations in advancing fundamental understanding of bubble dynamics and nucleate pool boiling heat transfer is discussed. The results of numerical simulations are validated with experiments on ground, in parabolic flights and on the International Space Station (ISS). As such validation is carried out when the level of gravity is varied over seven orders of magnitude. It is shown that reduced gravity stretches the length and time scales of the process and generally leads to degradation of rate of heat transfer associated with nucleate boiling.
APA, Harvard, Vancouver, ISO, and other styles
8

Manglik, R. M., and A. D. Athavale. "Pseudoplasticity and Dynamic Interfacial Tension Relaxation Effects on Nucleate Pool Boiling in Aqueous Polymeric Liquids." Journal of Heat Transfer 141, no. 5 (March 27, 2019). http://dx.doi.org/10.1115/1.4042699.

Full text
Abstract:
Nucleate pool boiling heat transfer and its ebullient dynamics in polymeric solutions at atmospheric pressure saturated conditions are experimentally investigated. Three grades of hydroxyethyl cellulose (HEC) are used, which have intrinsic viscosity in the range 5.29 ≤ [η] ≤ 10.31 [dl/g]. Their aqueous solutions in different concentrations, with zero-shear viscosity in the range 0.0021 ≤ η0 ≤ 0.0118 [N⋅s/m2], exhibit shear-thinning rheology in varying degrees, as well as gas–liquid interfacial tension relaxation and wetting. Boiling heat transfer in solutions with constant molar concentrations of each additive, which are greater than their respective critical polymer concentration C*, is seen to have anomalous characteristics. There is degradation in the heat transfer at low heat fluxes, relative to that in the solvent, where the postnucleation bubble dynamics in the partial boiling regime is dominated by viscous resistance of the polymeric solutions. At higher heat fluxes, however, there is enhancement of boiling heat transfer due to a complex interplay of pseudoplasticity and dynamic surface tension effects. The higher frequency vapor bubbling train with high interfacial shear rates in this fully developed boiling regime tends to be influenced by increasing shear-thinning and time-dependent differential interfacial tension relaxation at the dynamic gas–liquid interfaces.
APA, Harvard, Vancouver, ISO, and other styles
9

Yamashita, Takuya, Hiroshi Madokoro, and Ikken Sato. "Post-test Analyses of the CMMR-4 Test." Journal of Nuclear Engineering and Radiation Science, June 11, 2021. http://dx.doi.org/10.1115/1.4051443.

Full text
Abstract:
Abstract Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, BWR accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR Crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO2 pellets were installed instead of UO2 pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.
APA, Harvard, Vancouver, ISO, and other styles

Books on the topic "Nucleate boiling degradation"

1

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering. and Oak Ridge National Laboratory, eds. Boiling-water reactor internals aging degradation study: Phase 1. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
2

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering. and Oak Ridge National Laboratory, eds. Boiling-water reactor internals aging degradation study: Phase 1. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
3

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering. and Oak Ridge National Laboratory, eds. Boiling-water reactor internals aging degradation study: Phase 1. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
4

K, Samanta P., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., and Brookhaven National Laboratory, eds. Applications of reliability degradation analysis. Washington, DC: Division [of] Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1996.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
5

K, Samanta P., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., and Brookhaven National Laboratory, eds. Applications of reliability degradation analysis. Washington, DC: Division [of] Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1996.

Find full text
APA, Harvard, Vancouver, ISO, and other styles

Book chapters on the topic "Nucleate boiling degradation"

1

Bahn, C. B., I. S. Hwang, I. H. Rhee, U. C. Kim, and J. W. Na. "Experimental Simulation of Boiling Crevice Chemistry." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 536–43. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch56.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Gordon, Barry, and Susan Garcia. "Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2061–76. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch213.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Yeh, Tsung-Kuang, Ching Chang, Fang Chu, and Chia-Shen Huang. "The Predicted Effectiveness of Noble Metal Treatment at the Chinshan Boiling Water Reactor." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1211–24. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch128.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Heldt, J., and H. P. Seifert. "Stress Corrosion Cracking of Reactor Pressure Vessel Steels Under Boiling Water Reactor Conditions." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 901–10. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch95.

Full text
APA, Harvard, Vancouver, ISO, and other styles
5

Yeh, Tsung-Kuang, and Mei-Ya Wang. "Water Chemistry in the Primary Coolant Circuit of a Boiling Water Reactor during Startup Operations." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2079–89. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch215.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Gordon, Barry, and Susan Garcia. "Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2061–77. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_123.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Gibbs, J. P., R. G. Ballinger, J. H. Jackson, D. Isheim, and H. Hänninen. "Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Boiling Water Reactor Environments." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 745–60. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch78.

Full text
APA, Harvard, Vancouver, ISO, and other styles
8

Cowan, Robert L., Juan Varela, and Susan E. Garcia. "The Effect of On-Line Noble Metal Addition on the Shut down Dose Rates of Boiling Water Reactors." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2023–36. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch210.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

Yeh, Tsung-Kuang, and Mei-Ya Wang. "Water Chemistry in the Primary Coolant Circuit of a Boiling Water Reactor During Startup Operations." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2079–89. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_124.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Gibbs, J. P., R. G. Ballinger, J. H. Jackson, D. Isheim, and H. Hänninen. "Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Boiling Water Reactor Environments." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 745–62. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_47.

Full text
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "Nucleate boiling degradation"

1

Kedzierski, M. A. "Effect of Al2O3 Nanolubricant on a Turbo-BII R134a Pool Boiling Surface." In ASME 2012 Third International Conference on Micro/Nanoscale Heat and Mass Transfer. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/mnhmt2012-75024.

Full text
Abstract:
This paper quantifies the influence of Al2O3 nanoparticles on the pool boiling performance of R134a/polyolester mixtures on a Turbo-BII-HP boiling surface. An Al2O3 nanolubricant (a lubricant containing dispersed nano-size particles) was made by suspending nominally 10 nm diameter Al2O3 particles in a synthetic polyolester to roughly a 1.0% volume fraction. The nanoparticles caused, on average, a 12% degradation in the boiling heat transfer relative to that for R134a/polyolester mixtures without nanoparticles for the three lubricant mass fractions that were tested. The degradation was nearly constant for heat fluxes between 20 kW/m2 and 120 kW/m2. It was speculated that the boiling heat transfer degradation was primarily due to a combination of (1) film boiling in the reentrant cavity rendering the nucleate boiling enhancement mechanism of the nanoparticles ineffective and (2) a reduction in bubble frequency due to the increased surface wetting as caused by the nanoparticles. In addition, these degradation factors might be mitigated with increased nanoparticle loading.
APA, Harvard, Vancouver, ISO, and other styles
2

Graham, Jacob, Angelo Hawa, and Patricia Weisensee. "Evolution of Heat Transfer in Pool Boiling in Contaminated Water." In ASME 2020 18th International Conference on Nanochannels, Microchannels, and Minichannels collocated with the ASME 2020 Heat Transfer Summer Conference and the ASME 2020 Fluids Engineering Division Summer Meeting. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icnmm2020-1041.

Full text
Abstract:
Abstract Boiling heat transfer serves as an efficient mechanism to dissipate large amounts of thermal energy due to the latent heat of phase change. In academic studies, typically ultra-pure deionized (DI) water is used to avoid contamination. However, in industrial and commercial settings, the working fluid might be contaminated with sediments, dust, salts, or organic matter. Long-term boiling processes in non-DI water cause substantial build-up of a stable layer of deposit that dramatically reduces the heat transfer coefficient. Therefore, heating applications in a contaminated medium demand strategies to prevent such fouling. Here, we studied the use of lubricant infused surfaces (LIS) and their ability to possibly minimize the deposition of calcium sulfate. Aluminum samples were infused with Krytox 102 oil and the heat transfer coefficient was investigated at a vertical and horizontal surface orientation. Fouling effects were introduced by pool boiling for 7.5 hours in a 6.97 mM calcium sulfate solution at constant heat flux. Heat flux curves for both plain aluminum and LIS were calibrated before contamination. Initially, the LIS was unable to support a nucleate phase and transitioned directly from liquid convection to film boiling heat transfer. Upon partial degradation of the lubricant layer during long-run experiments, nucleate boiling ensued. Over 7.5 hours, the heat transfer coefficient of each sample (Al and LIS) degraded between 5.4% and 7.9% with no significant correlation with either lubricant treatment or surface orientation. Post boiling profilometry was conducted on each sample to characterize the thickness and distribution of the calcium sulfate layer. In these experiments, the plain aluminum surface outperformed the LIS at both orientations in minimizing calcium layer thickness. The LIS oriented vertically outperformed the LIS oriented horizontally.
APA, Harvard, Vancouver, ISO, and other styles
3

Moreno, Gilbert, Steven J. Oldenburg, Seung M. You, and Joo H. Kim. "Pool Boiling Heat Transfer of Alumina-Water, Zinc Oxide-Water and Alumina-Water+Ethylene Glycol Nanofluids." In ASME 2005 Summer Heat Transfer Conference collocated with the ASME 2005 Pacific Rim Technical Conference and Exhibition on Integration and Packaging of MEMS, NEMS, and Electronic Systems. ASMEDC, 2005. http://dx.doi.org/10.1115/ht2005-72375.

Full text
Abstract:
This investigation conducts pool boiling experiments under saturated conditions (Tsat = 60 °C) using nanofluids as the coolants. Three different nanofluids were tested including zinc oxide (ZnO)-water, aluminum oxide (Al2O3)-water and aluminum oxide (Al2O3)-water+ethylene glycol (ethylene glycol solution). At saturation (Tsat = 60°C), the pool boiling performance of the two pure water based nanofluids were similar. The maximum CHF enhancement as compared to the predicted Zuber’s [1] CHF evaluated at an equivalent saturation temperature is ∼180% for Al2O3-water nanofluids and ∼240% for ZnO-water nanofluids. In both cases, no degradation in the boiling heat transfer rate was observed for lower nanoparticle concentrations. However, higher nanoparticle concentrations demonstrate nucleate boiling heat transfer degradation at high heat fluxes. The dispersion of Al2O3 nanoparticles in various ethylene glycol solutions is also found to enhance CHF by as much as ∼130%. A significant difference in the diameter of individual grains/particles (27 ± 16.3 nm) and the volume weighted average diameter of particles in solution (155 ± 80 nm) indicates that the Al2O3-water nanofluids consist primarily of nanoparticle agglomerates. Gravimetric fractionation of the nanofluid produced nanofluids with particle/particle aggregate average diameters that ranged from 69–346 nm. Over the size range tested, there was no significant CHF dependence on the average particle diameter.
APA, Harvard, Vancouver, ISO, and other styles
4

Geisler, Karl J. L., and Avram Bar-Cohen. "Surface Effects on Confinement-Driven Pool Boiling Enhancement in Vertical Parallel-Plate Channels." In ASME 2005 Summer Heat Transfer Conference collocated with the ASME 2005 Pacific Rim Technical Conference and Exhibition on Integration and Packaging of MEMS, NEMS, and Electronic Systems. ASMEDC, 2005. http://dx.doi.org/10.1115/ht2005-72666.

Full text
Abstract:
Evidence of confinement-driven boiling heat transfer enhancement in vertical channels is very well documented in the literature and much has been observed about its nature and behavior. However, the majority of the available correlations is empirically-based and they tend to be very restricted in their range of applicability and portability. In order to further elucidate the effect of this type of geometrical confinement on boiling heat transfer, an experimental study has been performed on vertical, rectangular parallel-plate channels immersed in the dielectric liquid FC-72. The enhancement of nucleate boiling performance with decreased channel spacing was found to depend on the type of heater employed but could not be explained by the surface roughness. On the other hand, degradation of the Critical Heat Flux (CHF) limit with decreasing channel spacing was found to be independent of the surface and to be well predicted by a correlation available in the literature.
APA, Harvard, Vancouver, ISO, and other styles
5

Okawa, Tsuyoshi, and Naoyuki Yomori. "Sensitivity Degradation Characteristics of Incore Neutron Detector for Heavy Water Reactor, Fugen NPP." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22234.

Full text
Abstract:
Fugen nuclear power plant is a 165MWe, heavy water-moderated, boiling light water-cooled, pressure tube-type reactor developed by JNC, which is the world’s first thermal neutron power reactor to utilize mainly Uranium and Plutonium mixed oxide (MOX) fuel. Fugen has been loaded a total of 726 MOX fuel assemblies since the initial core in 1978. Each incore neutron detector assembly of Fugen composed of four Local Power Monitors (LPM) is located at sixteen positions in the area of heavy water moderator in the core and monitors its power distribution during operation. The thermal neutron flux of Fugen is relatively higher than that of Boiling Water Reactor (BWR), therefore LPM, which is comprised of a fission chamber, degrades more quickly than that of BWR. An Improved Long-life LPM (LLPM) pasted inner surface wall of the chamber with 234U/235U at a ratio of 4 to 1 had been developed through the irradiation test at Japan Material Test Reactor (JMTR). The 234U is converted to 235U with absorption of neutron, and compensates the consumption of 235U. LPM has been loaded to the initial core of Fugen since 1978. JNC had evaluated its sensitivity degradation characteristics through the accumulated irradiation data and the parametric survey for 234σa and 235σa. Based on the experience of evaluation for sensitivity degradation, JNC has applied shuffling operation of LPM assemblies during an annual inspection outage to reduce the operating cost. This operation realizes the reduction of replacing number of LPM assemblies and volume of radioactive waste. This paper describes the sensitivity degradation characteristics of incore neutron detector and the degradation evaluation methods established in Fugen.
APA, Harvard, Vancouver, ISO, and other styles
6

Kiss, E. "Component Reliability Considerations for New Designs and Extended Operation of Boiling Water Reactor (BWRs)." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48864.

Full text
Abstract:
To achieve high reliability for new designs and extended operation of Reactor Pressure Vessels and Internals it is mandatory to apply the technical knowledge gained during operation of the existing Plants to assure that sufficient “Margin” is built into the new design. This paper discusses the importance of four key structural degradation mechanisms that have been shown by operational experience to affect the reliability of the BWR. These are: 1) Stress Corrosion Cracking (IGSCC) of Stainless Steel and Nickel-based Alloys; 2) Irradiation Assisted SCC (IASCC) of Stainless Steel and Nickel-based Alloys; 3) Irradiation Embrittlement of RPV low alloy Steel; 4) Corrosion Assisted Fatigue of Carbon and Low Alloy Steel. While the focus of this paper is the BWR, the mechanisms discussed are equally applicable to the PWR, although the water chemistry effects and mitigations will be different.
APA, Harvard, Vancouver, ISO, and other styles
7

Li, Yongkui, Yoshiyuki Kaji, and Takahiro Igarashi. "Study of Weld Residual Stress Field in the Girth Seam H6A of Core Shroud of Boiling Water Reactor." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29269.

Full text
Abstract:
Many accidents have occurred in nuclear power plants due to the intergranular stress corrosion cracking (IGSCC) in the heat affected zone (HAZ) of welded joint of the core shroud of boiling water reactors (BWRs) in past years. The IGSCC is considered to be caused by the synergistic roles of corrosion environment, neutron irradiation and the welding residual stress. After several decades, the degradation of Type 316L low carbon stainless steel used in the core shroud occurs due to the neutron irradiation and thermal cycles. The degradation can be referred to the irradiation hardening, segregation of the local chemical composition at grain boundaries and swelling. The synergistic effects of those eventually lead to the initiation and propagation of the irradiation-assisted stress corrosion cracking (IASCC) in core shroud for long operation. The HAZ of the girth seams H6a in the core shroud are sensitive to the stress corrosion cracking. We are focusing on the weld residual stress field around the girth seam H6a in the core shroud as weld. The analysis work adopted different approaches in ABAQUS to simulate the weld residual stress, and they are Static General Analysis (SGA) and Fully Coupled Temperature-Displacement Analysis (FCTDA) respectively. The former is much simple to finish the progress while cannot obtain much accurate results at the boundaries of beads due to the discontinuous temperature field in the model. The later analysis gave the much accurate results comparing with the experimental results. The axial stress field in the crossing section of the wall of the core shroud was also clarified.
APA, Harvard, Vancouver, ISO, and other styles
8

Wu, Wen, and Barclay G. Jones. "Simulation of Bubble Dynamics in Sub-Cooled Boiling on Fuel Clad in PWRs." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22682.

Full text
Abstract:
The crud deposition on nuclear fuel assembly cladding generally increases the resistance to heat transfer, which may result in deterioration of thermal performance, degradation of the fuel cladding, and an axial power shift, i.e. Axial Offset Anomaly (AOA). Crud formation continues to elude prediction. An operational difficulty, of not being able to accurately determine power safety margin, then arises. In some cases, this condition has required decreasing the core power by as much as thirty percent, hence, resulting in considerable loss of revenue for the utility. The specific purpose of this study is to examine bubble dynamics, flow characteristics of the surrounding fluid, and its impact on the formation of the curd. The presence of a bubble on the clad surface affects the flow field around it , particularly in forming a stagnant flow region behind the bubble. The temperature difference between the bubble and the bulk coolant surrounding it causes vaporization at the bubble-clad interface and condensation at its apex. Pure water is thereby moved into the bubble through vaporization resulting in the concentration of solutes in the water at the bubble/wall surface region, which may cause their precipitation on and/or attachment to the clad surface, thereby initiating crud deposition. We investigate analytically and numerically, the growth of a bubble in the boundary layer and the influence of the bubble on the flow. Because of the small bubble size, a spherical model of the bubble is selected for our research. A two-step calculation is applied to this model. In the first step, bubble growth is estimated analytically with omission of the effect of the bulk fluid velocity, a reasonable approximation. In the second step, the flow field around the stationary bubble is obtained through numerical methods. Some parameters in PWR operating condition have been determined approximately e.g. size of the bubble, boundary layer thickness, flow velocity and drag forces on the bubble.
APA, Harvard, Vancouver, ISO, and other styles
9

Damerell, Paul S., and Todd A. Spears. "The Joint Owners’ Group Program on MOV Periodic Verification." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22599.

Full text
Abstract:
To address long-term motor operated valve (MOV) performance, the Babcock & Wilcox, Boiling Water Reactor, Combustion Engineering and Westinghouse Groups (B&WOG, BWROG, CEOG and WOG) teamed in 1997 to form the Joint Group (JOG) MOV Periodic Verification (PV) Program. This program is nearing completion, with 98 of the 103 operating U.S. reactor units participating. The goal of the program is to provide a justified approach for periodically testing MOVs, that addresses potential degradation. The program defines an interim approach that specifies periodic tests without flow and differential pressure (DP), at a frequency determined by the MOV’s risk significance and margin. To justify this approach, each participating plant is also DP testing 2 valves per unit. Each valve is tested three times over five years, with at least one year between tests. The data are evaluated jointly to confirm or adjust the initial guidance. The majority of the tests are complete and conclusions are coming into focus. For gate valves, when the valve factor is initially low, increases can occur between one test and a later test. One common way that the valve factor becomes low is disassembling and reassembling the valve. The data show that, following valve disassembly and reassembly, the valve factor tends to be reduced, and it tends to increase in subsequent service. Outside of the valves disassembled and reassembled, some gate valves have low valve factors apparently because the valves are not stroked under DP conditions in service. For butterfly valves, there have been no observations of degradation in bearing friction coefficient. A few valves with bronze bearings in raw (untreated) water service have shown significant variations in friction, but they tend to be a mixture of increases and decreases with no pattern of degradation. Globe valves, both unbalanced and balanced, tend to show a constant valve factor with no indication of degradation.
APA, Harvard, Vancouver, ISO, and other styles
10

Kojima, Nobuo, Koji Nishino, Keisuke Sakemura, and Kengo Kobayashi. "The Solenoid Valve for Main Steam Relief Valve Adapted Under Severe Accident." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16030.

Full text
Abstract:
Abstract MSSRVs in Boiling Water Reactor (BWRs) can be mandatory opened through a solenoid valve, when a nuclear reactor raises unusual pressure. The solenoid valve consists of a valve main part which forms a flow channel, and a pilot part which controls a flow direction. In the accident of the Fukushima Daiichi nuclear power plant, MSSRVs ware over the design specification of these solenoid valves, and were not able to operate. One of the reason, there is degradation of a sealant and a coil of a solenoid valve. As one of the measures for rebooting the BWRs, the development and verification of a solenoid valve which were applied to the SA condition are required. Since we developed and verified it applied to the SA.condition, we report here.
APA, Harvard, Vancouver, ISO, and other styles
We offer discounts on all premium plans for authors whose works are included in thematic literature selections. Contact us to get a unique promo code!

To the bibliography