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Relevant bibliographies by topics / Fuel centerline temperature

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Academic literature on the topic 'Fuel centerline temperature'

Author: Grafiati

Published: 4 June 2021

Last updated: 2 February 2022

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Journal articles on the topic "Fuel centerline temperature"

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KAWAMURA, Hiroshi, and Hiroei ANDO. "Fuel centerline temperature response of LWR fuel rods on reactor scram." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 31, no. 7 (1989): 852–60. http://dx.doi.org/10.3327/jaesj.31.852.

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Horhoianu, G., D. V. Ionescu, and E. I. Pauna. "In-reactor measurements of fuel centerline temperature variation during power change." Kerntechnik 75, no. 3 (2010): 81–85. http://dx.doi.org/10.3139/124.110073.

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Nguyen, Van Tung, Trong Hung Nguyen, Thanh Thuy Nguyen, and Duy Minh Cao. "Predicting behavior of AP-1000 nuclear reactor fuel rod under steady state operating condition by using FRAPCON-4.0 software." Nuclear Science and Technology 8, no. 2 (2021): 43–50. http://dx.doi.org/10.53747/jnst.v8i2.90.

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This paper reports the results on the predictions of behavior of AP-1000 nuclear reactorfuel rod under steady state operating condition by using FRAPCON-4.0 software. The predictive items were the temperature distribution in the fuel rod, including fuel centerline temperature, fuel pellet surface temperature, gas temperature, cladding inside and outside temperature, oxide surface and bulk coolant temperature; and gap conductance and thickness.The predictive items also include deformation of fuel pellets, fission gas release and rod internal pressure, cladding oxidation and hydration. The predi

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Kuang, Cui Peng, Yuan Zhou Li, Shi Zhu, and Shao Hua Mao. "Research on the Combustion Characteristics of Small-Scaled Ethanol Pool Fire of Different Aspect Ratios." Advanced Materials Research 347-353 (October 2011): 1161–65. http://dx.doi.org/10.4028/www.scientific.net/amr.347-353.1161.

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Four groups of small-scaled ethanol pool fire experiment with different aspect-ratio(s) is undertaken, to gauge the mass loss rate of fuel as well as the plume centerline temperature distribution. Comparison of plume centerline temperature is made between the theoretical values estimated by Heskestad plume model and experimental results, which indicates that: with the increasing of s, the difference between theoretical values and experimental values tend to grow greater; and when s≈1, theoretical values and experimental values cohere well.

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Wang, Chang Jian. "Simulation of Heptane Jet Fire at Low Atmosphere Pressure." Advanced Materials Research 516-517 (May 2012): 1070–73. http://dx.doi.org/10.4028/www.scientific.net/amr.516-517.1070.

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Due to safety consideration of storage and transportation of liquid fuel at low atmospheric pressure region, the influence of low atmospheric pressure on heptane jet fire was numerically investigated, based on LES and mixture-fraction combustion model. Injection heptane diameters satisfy Rosin-Rammler distribution. The simulation shows that, low atmospheric pressure has an evident effect on jet fire. It extends the fire length and shortens the lift-off height. The centerline temperature rises to the maximum value more rapidly and then it decays more slowly. The maximum centerline temperature i

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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.

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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

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Chandramouli, Deepthi, and Shripad T. Revankar. "Development of Thermal Models and Analysis of UO2-BeO Fuel during a Loss of Coolant Accident." International Journal of Nuclear Energy 2014 (August 26, 2014): 1–9. http://dx.doi.org/10.1155/2014/751070.

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Small fraction of high conductivity BeO in UO2 fuel significantly improves thermal conductivity and also affects the overall performance of the fuel during steady state operation and during transients. In this study, performance of UO2-BeO composite under transient conditions such as loss of coolant accident (LOCA), using FRAPTRAN (fuel rod analysis program transient), was carried out. The subroutines in FRAPTRAN code that calculate key thermophysical properties such as thermal conductivity, specific heat capacity, and specific enthalpy were modified to account for the presence of the BeO in U

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Isnaini, Muhammad Darwis, and Muhammad Subekti. "VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 18, no. 1 (2016): 41. http://dx.doi.org/10.17146/tdm.2016.18.1.2367.

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The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR) using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves

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Salman, Ahmed M., Ibrahim A. Ibrahim, Hamada M. Gad, and Tharwat M. Farag. "Effects of Air Temperature on Combustion Characteristics of LPG Diffusion Flame." Materials Science Forum 1008 (August 2020): 128–38. http://dx.doi.org/10.4028/www.scientific.net/msf.1008.128.

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In the present study, the combustion characteristics of LPG gaseous fuel diffusion flame at elevated air temperatures were experimentally investigated. An experimental test rig was manufactured to examine a wide range of operating conditions. The investigated parameters are the air temperatures of 300, 350, 400, 450, and 500 K with constant percentage of nitrogen addition in combustion air stream of 5 % to give low oxygen concentration of 18.3 % by mass at constant air swirl number, air to fuel mass ratio, and thermal load of 1.5, 30, and 23 kW, respectively. The gaseous combustion characteris

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Shafiqul Islam, Md, and Abid Hossain Khan. "Thermal-hydraulic analysis of fuel rod of a TRIGA Mark II research reactor." International Journal of Engineering & Technology 9, no. 1 (2020): 69. http://dx.doi.org/10.14419/ijet.v9i1.30035.

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In this work, the feasibility of employing “Single Flow Channel Analysis” technique for obtaining the thermal-hydraulic behavior of a TRIGA Mark II research reactor has been studied. Two different simulation methods have been investigated for this purpose; one in which there is no variation in volumetric heat generation along the fuel axis and the other in which there is variation. A hot rod factor of 1.70 has been taken. Results obtained from simulation methods have been compared with both theoretical results and experimental data provided by the manufacturers. Results show that data generate

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Dissertations / Theses on the topic "Fuel centerline temperature"

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Abdalla, Ayman. "Sensitivity analysis of fuel centerline temperatures in SuperCritical water-cooled reactors (SCWRs)." Thesis, 2012. http://hdl.handle.net/10155/292.

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SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today’s conventional NPPs, which have thermal efficiencies between 30 ‒ 35%, SCW NPPs will have thermal efficiencies within a range of 45 ‒ 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current

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Peiman, Wargha. "Thermal aspects of high efficiency channel with conventional and alternative fuels in SuperCritical water-cooled reactor (SCWR) applications." Thesis, 2011. http://hdl.handle.net/10155/158.

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Chosen as one of six Generation‒IV nuclear-reactor concepts, SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 ‒ 50% owing to reactor‘s high outlet temperatures. A generic pressure-channel (or pressure-tube)SCWR operates at a pressure of 25 MPa with inlet- and outlet-coolant temperatures of 350°C and 625°C. Consequently, the sheath and fuel centerline temperatures are higher in SCWRs than those of the current nuclear reactors. Previous studies have shown that the sheath and fuel centerline temperatures could exceed the design and

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Books on the topic "Fuel centerline temperature"

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Tseng, Che-Chung. FRAPCON-2 predictions of the fuel centerline temperatures in IFA-431 and IFA-432 assemblies. Institute of Nuclear Energy Research, 1985.

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Book chapters on the topic "Fuel centerline temperature"

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Saurav, Suman, M. Muthuganesh, P. K. Chaurasia, and S. Murugan. "Comparative Study of Analytical and Simulation Method in Determination of Fuel Centerline and Clad Surface Temperature of Metallic Nuclear Fuel." In Lecture Notes in Mechanical Engineering. Springer Singapore, 2020. http://dx.doi.org/10.1007/978-981-15-7779-6_1.

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Conference papers on the topic "Fuel centerline temperature"

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Abdalla, Ayman, Wargha Peiman, Igor Pioro, and Kamiel Gabriel. "Sensitivity Analysis of Fuel Centerline Temperature in SCWRs." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-54530.

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The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher

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Kulshreshtha, Digvijay B., and S. A. Channiwala. "Experimental Investigations on Temperature Distribution Along the Liner Wall and Centerline of Small Capacity Gas Turbine Combustion Chamber for Fuel Rich to Fuel Lean Air/Fuel Ratios at Equivalence Ratio of 1." In ASME 8th Biennial Conference on Engineering Systems Design and Analysis. ASMEDC, 2006. http://dx.doi.org/10.1115/esda2006-95664.

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The development of the combustion chamber for 20kW gas turbine unit using kerosene type fuel has been undertaken keeping in view the basic requirements of a good combustion chamber, namely, high combustion efficiency, low pressure loss, smooth ignition, wide stability limits, size and shape compatible with engine envelop, low emissions of smoke, unburned fuel and gaseous pollutant species, durability and ease of maintenance. A sophisticated experimental test rig has then been developed to investigate over a wide range of air/fuel ratios for the temperature profiles at the few axial and liner w

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Peiman, W., Eu Saltanov, L. Grande, I. Pioro, B. Rouben, and K. Gabriel. "Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-54596.

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SuperCritical Water-cooled nuclear Reactor (SCWR) designs are one of six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the

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Grande, Lisa, Wargha Peiman, Sally Mikhael, et al. "Thermal Aspects of Using Uranium Nitride in SuperCritical Water-Cooled Nuclear Reactors." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29790.

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SuperCritical Water-cooled nuclear Reactors (SCWRs) utilize a light-water coolant pressurized to 25 MPa with a channel inlet temperature of 350°C and outlet temperature of 625°C. Previous studies have indicated that uranium dioxide (UO2) nuclear fuel may not be suitable for SCWR use, because the maximum fuel centerline temperature might exceed the industry accepted limit of 1850°C. This research paper explores the use of uranium nitride (UN) as an alternative fuel option to UO2 at SuperCritical Water (SCW) conditions. A generic 1200-MWel Pressure-Tube (PT) -type reactor cooled with SCW was use

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Peiman, W., I. Pioro, and K. Gabriel. "Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16410.

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SuperCritical Water-cooled nuclear Reactor (SCWR) is one of the six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution

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Xin, Jin, Xiaohan Liu, and Xiaoyan Wei. "A New Numerical Method for Fuel Temperature Calculation." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66248.

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For most fuel rod codes, the time independent heat conduction equation, which is a steady heat conduction equation, is applied in fuel temperature calculation. However, it can affect the fuel temperature prediction in II condition, which the linear power has much change in some seconds. For improving the fuel temperature prediction in II condition, this paper gives a new numerical method, which combines classical thermal conduction integration method and the difference applied in time partial derivative. For guaranteeing the numerical method’s stability and convergence rate, the multi-dimensio

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Zahrádka, Tomáš, and Radek Škoda. "Cost Saving When Using Enhanced Conductivity Nuclear Fuel Containing BeO in WWER-1000 Reactors." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30901.

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Current pressurized water reactors utilize sintered UO2 that has a number of advantages and disadvantages. Uranium Dioxide’s low thermal conductivity results in a large thermal gradient within the fuel pellet corresponding to higher centerline temperatures compared to other potential fuel forms. These gradients result in non-uniform thermal expansion leading to large internal stresses resulting in cracking of the pellet and fuel-clad interaction, which can lead to loss of the integrity of the fuel pin. Higher fuel temperatures also increase the release of fission gases. Fuels with higher therm

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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.

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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

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Schriener, Timothy M., and Mohamed S. El-Genk. "Neutronics Optimization of UN Fuel Pin Assemblies for a Sodium-Cooled, Small Modular Reactor." In ASME 2011 Small Modular Reactors Symposium. ASMEDC, 2011. http://dx.doi.org/10.1115/smr2011-6631.

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This paper presents preliminary results of neutronics and thermal-hydraulics design analysis of a sodium cooled, small modular reactor (SMR). The reactor’s nominal thermal power is 150 MWth at sodium inlet and exit temperatures of 630 and 780 K. The reactor core is comprised of three rings of shrouded hexagonal assemblies of 19.8% enriched UN fuel pins and a hexagonal assembly of enriched B4C pins in the central cavity for a coarse reactivity control. The objectives are to provide enough excess reactivity for achieving a refueling cycle > 5 year, maintaining a more even coolant flow in the

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Patel, Hemal, Ashley Milner, Caleb Pascoe, et al. "Thermal Aspects of Using Thorium Dioxide as Alternative Nuclear Fuel in SuperCritical Water-Cooled Reactors." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29975.

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SuperCritical Water-Cooled nuclear Reactors (SCWRs) are one of six choices for Generation IV (Gen IV) reactor concepts. These reactors use light water as a coolant and operate at a pressure of 25 MPa, inlet temperatures 280–350°C and an outlet temperature up to 625°C. Operating at these elevated temperatures and pressures are beneficial due to: 1) increased gross thermal efficiency of SCW Nuclear Power Plants (NPPs) (from 30%–35% of the current NPPs to 45%–50%) and 2) decreased capital and operational costs. Use of SCW as a reactor coolant will permit a direct-cycle steam circuit. SCWRs elimin

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