Relevant bibliographies by topics / Light water reactors Thermal properties

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Light water reactors Thermal properties'

Author: Grafiati
Published: 4 June 2021
Last updated: 17 February 2022

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

Select a source type:

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Light water reactors Thermal properties.'

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 "Light water reactors Thermal properties"

1

Song, Carol. "IRRADIATION EFFECTS ON ZR-2.5NB IN POWER REACTORS." CNL Nuclear Review 5, no. 1 (2016): 17–36. http://dx.doi.org/10.12943/cnr.2016.00010.

Full text
Abstract:
Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institut
APA, Harvard, Vancouver, ISO, and other styles
2

Latta, Ryan, Shripad T. Revankar, and Alvin A. Solomon. "Modeling and Measurement of Thermal Properties of Ceramic Composite Fuel for Light Water Reactors." Heat Transfer Engineering 29, no. 4 (2008): 357–65. http://dx.doi.org/10.1080/01457630701825390.

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

Degueldre, Claude. "Zirconia Inert Matrix for Plutonium Utilisation and Minor Actinides Disposition in Thermal Reactors." Advances in Science and Technology 45 (October 2006): 1907–14. http://dx.doi.org/10.4028/www.scientific.net/ast.45.1907.

Full text
Abstract:
The toxicity of the UO2 spent fuel is dominated by plutonium and minor actinides (MA): Np, Am and Cm, after decay of the short live fission products. Zirconia ceramics containing Pu and MA in the form of an Inert Matrix Fuel (IMF) could be used to burn these actinides in Light Water Reactors. Optimisation of the fuel designs dictated by properties such as thermal, mechanical, chemical and physical must be performed with attention for their behaviour under irradiation. Zirconia must be stabilised by yttria to form a solid solution such as AnzYyPuxZr1-yO2-y where minor actinide oxides are also s
APA, Harvard, Vancouver, ISO, and other styles
4

Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

Full text
Abstract:
Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) o
APA, Harvard, Vancouver, ISO, and other styles
5

Kulikov, G. G., A. N. Shmelev, and V. A. Apse. "Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction." International Journal of Nuclear Energy 2014 (October 16, 2014): 1–18. http://dx.doi.org/10.1155/2014/373726.

Full text
Abstract:
Light materials with small atomic mass (light or heavy water, graphite, and so on) are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb) for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable o
APA, Harvard, Vancouver, ISO, and other styles
6

Aly, Ahmed, Agustin Abarca, Maria Avramova, and Kostadin Ivanov. "EXTENDING CTF MODELING CAPABILITIES TO SFRs AND VALIDATION AGAINST SHRT TESTS." EPJ Web of Conferences 247 (2021): 10034. http://dx.doi.org/10.1051/epjconf/202124710034.

Full text
Abstract:
The utilization of liquid metals as coolants for fast reactors brings several economical and practical advantages that lead to a sustainable future for nuclear energy. Molten sodium is used as a coolant in Sodium Fast Reactors (SFRs). Sodium is relatively cheaper than other metal coolants. It requires lower pumping power, causes less neutron moderation and it is non-corrosive to the fuel cladding. The SFR hexagonal subassemblies are relatively smaller than Light Water Reactors (LWRs) subassemblies. The differences in the geometrical design of SFRs compared to LWRs lead to different physical be
APA, Harvard, Vancouver, ISO, and other styles
7

Ghasabian, M., F. Mofidnakhaei, and S. Talebi. "Effect of gap design pressure on the LWR fuel rods lifetime." Kerntechnik 86, no. 3 (2021): 202–9. http://dx.doi.org/10.1515/kern-2021-0004.

Full text
Abstract:
Abstract The fuel burn-up rate has been raised in recent years to improve the efficiency of nuclear LWRs (light water reactors). Therefore, surveying and estimating changes in fuel properties and structural materials during radiation exposure is of paramount importance. In the present study, the researchers focused on analyzing the role of LWR fuel rod initial gap pressure (initial gas pressure when a fuel rod is fabricated) on the rod’s thermal and mechanical performance. FRAPCON-4.0 steady-state fuel performance code was used to simulate the effect of initial gap pressure on the behavior of
APA, Harvard, Vancouver, ISO, and other styles
8

Vänskä, Emilia, and Tapani Vuorinen. "Effect of cellulase-assisted refining on the thermal degradation of bleached high-density paper." Holzforschung 69, no. 6 (2015): 703–12. http://dx.doi.org/10.1515/hf-2014-0194.

Full text
Abstract:
Abstract Pulp was treated with cellulase, and the chemical, physical, and optical properties of the refined pulps in paper sheets were measured in terms of the degree of polymerization of cellulose, tensile strength, elongation, burst strength, light scattering, and brightness. The sheets were thermally treated for 20 and 60 min at 225°C in the presence of 1% and 75% (v/v) water vapor. The cellulase treatment intensified the fibrillation of fibers and reduced the specific energy consumption during refining. It was demonstrated based on the water retention value that the refining modified the w
APA, Harvard, Vancouver, ISO, and other styles
9

Li, Xiuping, Yuchun Zhai, Peihua Ma, and Rongxiang Zhao. "Preparation and Photocatalysis of Nano-Zn/Ce Composite Oxides." Australian Journal of Chemistry 67, no. 4 (2014): 657. http://dx.doi.org/10.1071/ch13448.

Full text
Abstract:
Metal oxide photocatalysts often lead to partial or complete mineralization of organic pollutants. On irradiation with UV-visible light, metal oxides catalyze redox reactions in the presence of air and O2 and water. Using ascorbic acid as a new combustion agent, ZnO was rapidly synthesized. Nano-Zn/CeO2 composites were prepared by a heterogeneous-precipitation method using (NH4)2CO3 as precipitation agent. X-ray diffraction, field-emission scanning electron microscopy, transmission electron microscopy, Fourier-transform infrared spectrometry, ultraviolet spectrophotometry, and differential the
APA, Harvard, Vancouver, ISO, and other styles
10

Yu, Tao, Hong Mei Wang, and Xin Tan. "New Reactor Fabricated Using Light Leakage Fiber for Azo Dye Degradation." Advanced Materials Research 716 (July 2013): 235–39. http://dx.doi.org/10.4028/www.scientific.net/amr.716.235.

Full text
Abstract:
Lanthanum-doped anatase TiO2coatings, which are composed of assemble crystalline of 50 nm diameter particles, when the percentage of dopant is 0.5 wt%, have been successfully fabricated by solgel dip-coating process on light leakage silica fiber (LSF) which length is 15cm and diameter is 125μm. This was achieved by adjustment of the lanthanum-doped solgel parameters such as molar ratio of precursors in lanthanum-doped TiO2-sols, the ratio of titanium tetrabutoxide to polyvinyl alcohol, dip-coating velocity, drying duration in air, thermal treatment and number of cyclical time of the process. T
APA, Harvard, Vancouver, ISO, and other styles
More sources

Dissertations / Theses on the topic "Light water reactors Thermal properties"

1

Baskar, Abishek. "Feasibility study on Thermal Anemometry at LWR conditions." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-298521.

Full text
Abstract:
Dryout and Departure from Nucleate boiling (DNB) are utmost thermal-hydraulic concerns for the safety of LWRs. The behavior of two-phase flows at these conditions is still not fully understood. There is at least a need for a good local velocity and void fraction database at these conditions. This database can be exploited by CFD codes, thereby leading to understanding and predicting DNB and boiling crisis. Since these conditions occur in LWR at pressures greater than 70 bar and temperatures above 285 $^oC$, most instrumentations fail at these conditions. So there is a need for developing or op
APA, Harvard, Vancouver, ISO, and other styles
2

Peréz, Mañes Jorge [Verfasser], and R. [Akademischer Betreuer] Stieglitz. "Development of CFD Thermal Hydraulics and Neutron Kinetics Coupling Methodologies for the Prediction of Local Safety Parameters for Light Water Reactors / Jorge Peréz Mañes. Betreuer: R. Stieglitz." Karlsruhe : KIT-Bibliothek, 2013. http://d-nb.info/1045663654/34.

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

Grundmann, Ulrich, Ulrich Rohde, Siegfried Mittag, and Sören Kliem. "DYN3D version 3.2 - code for calculation of transients in light water reactors (LWR) with hexagonal or quadratic fuel elements - description of models and methods -." Forschungszentrum Dresden, 2010. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-28604.

Full text
Abstract:
DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balanc
APA, Harvard, Vancouver, ISO, and other styles
4

Grundmann, Ulrich, Ulrich Rohde, Siegfried Mittag, and Sören Kliem. "DYN3D version 3.2 - code for calculation of transients in light water reactors (LWR) with hexagonal or quadratic fuel elements - description of models and methods -." Forschungszentrum Rossendorf, 2005. https://hzdr.qucosa.de/id/qucosa%3A21687.

Full text
Abstract:
DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balanc
APA, Harvard, Vancouver, ISO, and other styles
5

Rohde, Ulrich, Yaroslav Kozmenkov, Valeri Pivovarov, and Yurij Matveev. "WTZ mit Russland - Transientenanalysen für Kernreaktoren - Abschlussbericht." Helmholtz-Zentrum Dresden-Rossendorf, 2011. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-72002.

Full text
Abstract:
Der Reaktordynamikcodes DYN3D wurde in der neu entwickelten Mehrgruppen-Version DYN3D-MG für die Anwendung auf wassergekühlte Reaktoren alternativ zu industriellen DWR und SWR ertüch-tigt. Es wurde die Anwendbarkeit für den graphitmoderierten Druckröhrenreaktor EGP-6 (KKW Bilibi-no), eine Konzeptstudie eines fortgeschrittenen Siedewasserreaktors mit schnellem Neutronenspekt-rum (RMWR) und das Reaktorkonzept RUTA-70 zur Wärmeversorgung nachgewiesen. Beim RUTA-Reaktor geht es vor allem um die Modellierung des Naturumlaufs des Kühlmittels bei niedrigen Sys-temdrücken. Zur Validierung wurden Exper
APA, Harvard, Vancouver, ISO, and other styles
6

Rohde, Ulrich, Yaroslav Kozmenkov, Valeri Pivovarov, and Yurij Matveev. "WTZ mit Russland - Transientenanalysen für Kernreaktoren - Abschlussbericht." Forschungszentrum Dresden-Rossendorf, 2010. https://hzdr.qucosa.de/id/qucosa%3A22136.

Full text
Abstract:
Der Reaktordynamikcodes DYN3D wurde in der neu entwickelten Mehrgruppen-Version DYN3D-MG für die Anwendung auf wassergekühlte Reaktoren alternativ zu industriellen DWR und SWR ertüch-tigt. Es wurde die Anwendbarkeit für den graphitmoderierten Druckröhrenreaktor EGP-6 (KKW Bilibi-no), eine Konzeptstudie eines fortgeschrittenen Siedewasserreaktors mit schnellem Neutronenspekt-rum (RMWR) und das Reaktorkonzept RUTA-70 zur Wärmeversorgung nachgewiesen. Beim RUTA-Reaktor geht es vor allem um die Modellierung des Naturumlaufs des Kühlmittels bei niedrigen Sys-temdrücken. Zur Validierung wurden Exper
APA, Harvard, Vancouver, ISO, and other styles
7

Mai, Anh T. "Thermal hydraulic and fuel performance analysis for innovative small light water reactor using VIPRE-01 and FRAPCON-3." Thesis, 2011. http://hdl.handle.net/1957/26964.

Full text
Abstract:
The Multi-Application Small Light Water Reactor (MASLWR) is a small natural circulation pressurized light water reactor design that was developed by Oregon State University (OSU) and Idaho National Engineering and Environmental Laboratory (INEEL) under the Nuclear Energy Research Initiative (NERI) program to address the growing demand for energy and electricity. The MASLWR design is geared toward providing electricity to small communities in remote locations in developing countries where constructions of large nuclear power plants are not economical. The MASLWR reactor is designed to operate f
APA, Harvard, Vancouver, ISO, and other styles
8

Bowser, Christopher Jordan. "RELAP5-3D modeling of ADS blowdown of MASLWR facility." Thesis, 2012. http://hdl.handle.net/1957/30438.

Full text
Abstract:
Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool wat
APA, Harvard, Vancouver, ISO, and other styles
9

Chang, Cheng-Chieh, and 張正杰. "The Physical, Thermal Isolation and Light Blocking Properties ofSilicon Containing Water-Borne Polyurethane / Clay Nanocomposites Coated Nylon Fabrics." Thesis, 2005. http://ndltd.ncl.edu.tw/handle/79692587183087632310.

Full text
Abstract:
碩士<br>逢甲大學<br>紡織工程所<br>93<br>The main purpose is to synthesize different clay content silicon containing anionic water-borne polyurethane / clay nanocomposites, which is coated on nylon taffeta fabrics with different thickness. The effect of thermal properties, mechanical properties and its light blocking is investigated. In consideration of the heat properties, the coated fabrics exhibit two degradation areas. Tdi have a maximum value at 3% clay content and increase with thickness. From the mechanical properties, it is found that both the breaking load and the percentage of elongation of
APA, Harvard, Vancouver, ISO, and other styles
10

Hallee, Brian Todd. "Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method." Thesis, 2013. http://hdl.handle.net/1957/37872.

Full text
Abstract:
The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Acciden
APA, Harvard, Vancouver, ISO, and other styles

Books on the topic "Light water reactors Thermal properties"

1

Lele, H. G. "H2OPROP," Computer code for determination of light water properties. Bhabha Atomic Research Centre, 1999.

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

Hämäläinen, A. Applying thermal hydraulics modeling in coupled processes of nuclear power plants. VTT Technical Research Centre of Finland, 2005.

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

Coordinated Research Programme Meeting on Investigation of Fuel Element Cladding Interaction with Water Coolant in Power Reactors (1986 Bhabha Atomic Research Centre). Proceedings of Coordinated Research Programme Meeting on Investigation of Fuel Element Cladding Interaction with Water Coolant in Power Reactors and Topical Meeting on Water Chemistry in Nuclear Energy Systems, Bhabha Atomic Research Centre, Bombay 400085, November 24-28, 1986. The Centre, 1986.

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

Chopra, O. K. Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems. Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1991.

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

Assessment of thermal embrittlement of cast stainless steels. Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1994.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
6

France. Commissariat à l'énergie atomique., ed. Proceedings of the seminar on thermal performance of high burn-up LWR fuel: 3-6 March 1998, Commissariat à lÉnergie Atomique (CEA) Cadarache, France. Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1998.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
7

Schroeder, Daniel V. An Introduction to Thermal Physics. Oxford University Press, 2021. http://dx.doi.org/10.1093/oso/9780192895547.001.0001.

Full text
Abstract:
Thermal physics deals with collections of large numbers of particles—typically 10<sup>23</sup> or so. Examples include the air in a balloon, the water in a lake, the electrons in a chunk of metal, and the photons given off by the sun. We can't possibly follow every detail of the motions of so many particles. So in thermal physics we assume that these motions are random, and we use the laws of probability to predict how the material as a whole ought to behave. Alternatively, we can measure the bulk properties of a material, and from these infer something about the particles it is made of. This
APA, Harvard, Vancouver, ISO, and other styles
8

U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Engineering and System Technology. and Sandia National Laboratories, eds. Experimental results pertaining to the performance of thermal igniters. Division of Engineering and Systems Technology, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1989.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
9

Hulsman, James William. The evaluation of an alternative scattering kernel for hydrogen bound in water. 1986.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
10

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering. and Argonne National Laboratory, eds. Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems. Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.

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

Book chapters on the topic "Light water reactors Thermal properties"

1

Mori, Hideo, Yoshinori Hamamoto, Koichiro Ezato, Kazuyuki Takase, and Takeharu Misawa. "Thermal Hydraulics." In Supercritical-Pressure Light Water Cooled Reactors. Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55025-9_3.

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

Lucas, T., R. G. Ballinger, H. Hanninen, and T. Saukkonen. "Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch91.

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

Lucas, T., R. G. Ballinger, H. Hanninen, and T. Saukkonen. "Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors. Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_54.

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

Adamson, R. B., and P. Rudling. "Properties of zirconium alloys and their applications in light water reactors (LWRs)." In Materials Ageing and Degradation in Light Water Reactors. Elsevier, 2013. http://dx.doi.org/10.1533/9780857097453.2.151.

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

Tinoco, Hernan. "CFD as a Tool for the Analysis of the Mechanical Integrity of Light Water Nuclear Reactors." In Nuclear Reactor Thermal Hydraulics and Other Applications. InTech, 2013. http://dx.doi.org/10.5772/52691.

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

Pioro, Igor L. "Application of Supercritical Fluids in Thermal- and Nuclear-Power Engineering." In Handbook of Research on Advancements in Supercritical Fluids Applications for Sustainable Energy Systems. IGI Global, 2021. http://dx.doi.org/10.4018/978-1-7998-5796-9.ch017.

Full text
Abstract:
Supercritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide, and helium are considered and discussed. Also, particularities of heat transfer at Supercritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. Supercritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs, and SCW in SCWRs) and/or working fluids in power cycles (helium, mixture of nitrogen (80%) and helium (20%), nitrogen and carbon dioxide in Brayton gas-turbine cycles, and SCW/“steam” in Rankine cycle).
APA, Harvard, Vancouver, ISO, and other styles
7

Pioro, Igor, Mohammed Mahdi, and Roman Popov. "Application of Supercritical Pressures in Power Engineering." In Advanced Applications of Supercritical Fluids in Energy Systems. IGI Global, 2017. http://dx.doi.org/10.4018/978-1-5225-2047-4.ch013.

Full text
Abstract:
SuperCritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide and helium are considered and discussed. Also, particularities of heat transfer at SuperCritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. SuperCritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs; and SCW in SCWRs) and/or working fluids in power cycles (helium; mixture of nitrogen (80%) and helium [20%]; nitrogen, and carbon dioxide in Brayton gas-turbine cycles; and SCW “steam” in Rankine cycle).
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "Light water reactors Thermal properties"

1

Zhou, Wenzhong, and Rong Liu. "Enhanced Thermal Conductivity UO2-BeO Fuels Fabrication Methods and Their Thermal Performance in Light Water Reactors." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30647.

Full text
Abstract:
An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. Two fabrication methods to produce high thermal conductivity UO2-BeO oxide nuclear fuels were summarized. These two processing routes generated pellets with two different microstructures. The characteristics and microstructures of the fuel are determined for use in FEM thermal models, and the relevant thermal properties for UO2-BeO fuels by
APA, Harvard, Vancouver, ISO, and other styles
2

Samuel, J., G. Lerchl, G. D. Harvel, and I. Pioro. "Investigation of ATHLET System Code for Supercritical Water Applications." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30136.

Full text
Abstract:
SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV nuclear-reactor concepts. They are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. However, codes that have been used to study the phenomena analytically have not been validated for supercritical water. The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and a
APA, Harvard, Vancouver, ISO, and other styles
3

Villamere, Bryan, Leyland J. Allison, Lisa Grande, Sally Mikhael, Adrianexy Rodriguez-Prado, and Igor Pioro. "Thermal Aspects for Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75990.

Full text
Abstract:
SuperCritical Water-cooled Reactors (SCWRs) are a Generation IV nuclear reactor concept. Two main SCWR design concepts are Pressure-Vessel (PV) type and Pressure-Tube (PT) type reactors. SCWRs would use light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). A reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is that a SCW NPP will have a thermal efficiency of 45 to 50%, a remarkable improvement from the current 30–35%. SCWRs have another added benefits such as a simplified flow circuit in which steam generators,
APA, Harvard, Vancouver, ISO, and other styles
4

Liu, Rong, Jie-Jin Cai, Wen-Zhong Zhou, and Ye Wang. "Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81237.

Full text
Abstract:
ThO2 has been considered as a possible replacement for UO2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO2 fuel. In this study, the thermal, mechanical properties of Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel were reviewed with updated properties and compared with UO2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by u
APA, Harvard, Vancouver, ISO, and other styles
5

Chen, Y., B. Alexandreanu, and A. S. Rao. "Cracking Behavior of a Decommissioned Material in Light Water Reactor Environment." In ASME 2020 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/pvp2020-21141.

Full text
Abstract:
Abstract The performance of structural materials is critical for the safe and economic operation of light water reactors. During power operation, reactor core internal materials are exposed to aggressive corrosive coolant environment, vigorous thermal/mechanical loading, and intensive neutron irradiation. Such severe service conditions can activate and enhance a wide range of degradation processes, leading to deteriorated material properties and service performance. To ensure the structural integrity and functionality of nuclear reactor components, material degradation and damage mechanisms mu
APA, Harvard, Vancouver, ISO, and other styles
6

Song, Jian, Limin Liu, Simiao Tang, et al. "Implementation of Liquid Metal Properties in RELAP5 MOD3.2 for Safety Analysis of Sodium-Cooled Fast Reactors." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66009.

Full text
Abstract:
Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the
APA, Harvard, Vancouver, ISO, and other styles
7

Xinyuan, Shang, Zhang Shaoyang, and Zhang Aimin. "Modeling and Thermal Calculation of Fuel Rod With SiCf/SiC Cladding for LWRs." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67078.

Full text
Abstract:
SiC fiber composite material as fuel cladding is proposed to increase the power density and maximum allowable fuel burnup in light water reactors. Empirical models about thermal properties of the SiC material are developed as a function of operating temperature and neutron fluence. A fuel rod modeling code frapcon2-SiC based on frapcon2 is compiled to predict the performance of SiC cladding when operating. Comparison of the behavior between the SiCf/SiC cladding and Zr-4 cladding in different thickness reveals that higher temperature will get due to the poor thermal conductivity of the SiC. As
APA, Harvard, Vancouver, ISO, and other styles
8

Chen, Yiren, Wei-Ying Chen, Chi Xu, et al. "Fracture Toughness and Deformation Behavior of Cast Austenitic Stainless Steels After Thermal Aging." In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-65768.

Full text
Abstract:
Cast austenitic stainless steels (CASSs) are used in the cooling system of light water reactors (LWRs) for components with complex shapes, such as pump casings, valve bodies, coolant piping, etc. The CF grades of CASS alloys are the cast equivalents of 300-series stainless steels (SSs) and show excellent mechanical properties and corrosion resistance. In contrast to the fully austenitic microstructure of wrought SSs, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite and are vulnerable to thermal aging embrittlement. The service performance of CASS alloys is of c
APA, Harvard, Vancouver, ISO, and other styles
9

Chen, Y., W.-Y. Chen, A. S. Rao, et al. "Fracture Resistance of Cast Austenitic Stainless Steels." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60736.

Full text
Abstract:
Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is
APA, Harvard, Vancouver, ISO, and other styles
10

Pascoe, Caleb, Ashley Milner, Hemal Patel, et al. "Thermal Aspects of Using Uranium Dicarbide Fuel in an SCWR at Maximum Heat-Flux Conditions." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29974.

Full text
Abstract:
There are 6 prospective Generation-IV nuclear reactor conceptual designs. SuperCritical Water-cooled nuclear Reactors (SCWRs) are one of these design options. The reactor coolant in SCWRs will be light water operating at 25 MPa and up to 625°C, actually at conditions above the critical point of water (22.1 MPa and 374°C, respectively). Current Nuclear Power Plants (NPPs) around the world operate at sub-critical pressures and temperatures achieving thermal efficiencies within the range of 30–35%. One of the major advantages of SCWRs is increased thermal efficiency up to 45–50% by utilizing the
APA, Harvard, Vancouver, ISO, and other styles

Reports on the topic "Light water reactors Thermal properties"

1

J. E. Daw, J. L. Rempe, and D. L. Knudson. Thermal Properties of Structural Materials Found in Light Water Reactor Vessels. Office of Scientific and Technical Information (OSTI), 2009. http://dx.doi.org/10.2172/974795.

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

Mikkelson, Daniel Mark, Shannon M. Bragg-Sitton, Cristian Rabiti, J. Michael Doster, and Konor L. Frick. Ranking Thermal Energy Storage Technologies for Integration with Light Water Reactors. Office of Scientific and Technical Information (OSTI), 2019. http://dx.doi.org/10.2172/1583122.

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

D.M. McEligot, K. G. Condie, G. E. McCreery, et al. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors. Office of Scientific and Technical Information (OSTI), 2005. http://dx.doi.org/10.2172/911892.

Full text
APA, Harvard, Vancouver, ISO, and other styles
You might also be interested in the extended bibliographies on the topic 'Light water reactors Thermal properties' for particular source types:
Journal articles
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