Thermal–hydraulic analysis of HPLWR fuel assembly cluster

Tashakor, S.; Salehi, Ali; Jahanfarnia, Gholamreza; Fard, Ali Abbaspour Tehrani

doi:10.1016/j.supflu.2013.02.025

Cited by 15 publications

(9 citation statements)

References 10 publications

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“…The operating pressure and core inlet/outlet temperatures are 25 MPa and 280/500 •C, respectively. Fuel assembly configuration has been proposed by Tashakor et al [6,7]. A view of the fuel assembly in HPLWR is shown in Fig.…”

Section: Methodsmentioning

confidence: 99%

“…In each region, the average temperature and modified fluid density are calculated accordingly. [6,7].…”

Section: Methodsmentioning

confidence: 99%

“…Lee et al developed a LOCA analysis code for supercritical-pressure light water cooled reactors [5]. Neutronic and Thermal hydralic analysis of HPLWR fuel assembly cluster has been performed by Tashakor et al [6,7]. Tashakor et al [8] presented an axial enrichment profile in advance nuclear energy power plant at supercriticalpressures.…”

Section: Introductionmentioning

confidence: 99%

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Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Tashakor

2018

Arab Journal of Nuclear Sciences and Applications

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Since 2002 a generation IV nuclear energy system development project has been initiated to enhance the future role of nuclear energy systems. Challenging technology goals for generation IV nuclear reactor systems were defined in 4 areas: sustainability, economics, safety and reliability and proliferation resistance. Light water reactors at supercritical pressure, currently under design, are the new generation of nuclear reactors. The high-performance light water reactor (HPLWR) is the European version of the supercritical-pressure water cooled reactor (SCWR). The aim of the present study is to conduct a reliability analysis of heavy water emergency core cooling system (ECCS) (instead of light water ECCS) in HPLWRs during LOCA (loss of coolant accident) through neutronics calculations and applied feedback.. Reliability analysis (neutronic calculations) is performed using WIMS and CITATION codes. Changes in neutronic parameters, such as neutron cross-sections, effective multiplication factor and power peaking factor during LOCA, are discussed by injecting different concentrations of the heavy water coolant by ECCS.

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Section: Methodsmentioning

confidence: 99%

“…In each region, the average temperature and modified fluid density are calculated accordingly. [6,7].…”

Section: Methodsmentioning

confidence: 99%

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Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Tashakor

2018

Arab Journal of Nuclear Sciences and Applications

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“…The decrease in density will reduce moderation, the power will decrease in the upper part of the fuel channels, while the 95 power peak will shift to the lower part of the fuel assembly as described by Tashakor et al. [193]. Five cases were investigated with the only variable parameters being N1 and N2.…”

Section: Heat Flux Distributionmentioning

confidence: 99%

“…During the transition from sub-to supercritical conditions, the fluid density decreases significantly (almost one order of magnitude) and therefore, at the inlet of the fuel assembly, moderation is much greater than at the outlet. Tashakor et al [193,210] performed an evaluation of power distribution along the fuel rods and found that the power distribution in the axial direction has a sinusoidal shape with its peak moved the upstream section of the fuel assembly to around 15% of the total fuel rod length. The other parameters are ca.…”

Section: Introductionmentioning

confidence: 99%

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

Rowinski¹

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Analytical methods and Computational Fluid Dynamics (CFD) examined the safety and performance of advanced nuclear reactors. Thermal and pressure drop evaluations of an innovative annular fuel in the European lead-cooled System ELSY fast reactor were conducted. Annular fuel has superior cooling capacity, and for a square fuel lattice, the pressure drop can be lower than the reference value for standard solid fuel. Consequently, when annular fuel rods replace solid fuel rods of the same dimensions the power rating is upgraded. While the hexagonal lattice cools more effectively, the pressure drop is higher than the standard fuel assembly and incompatible with power augmentation. Numerical investigations of supercritical flow of both carbon dioxide and water in a vertical tube under non-uniform heat flux applied at the wall demonstrated that current correlations for the Heat Transfer Coefficient (HTC) at low enthalpy values are accurate to ±15%. At high enthalpy, only the general trend is replicated due to large property value differences near the wall and in the bulk flow. In these circumstances, the thermal conductivity, specific heat, density and viscosity are 5-8 times lower after the transition from sub-to super-critical conditions. Moreover, the efficiency of different HTC correlation models is coolant dependant, e.g. the Swenson* formula proves superior for carbon dioxide, whereas the Ornatsky model** provides better agreement with water, where a lower maximum wall temperature was obtained when a non-uniform heat flux was applied, as compared to the reference case of uniform heat flux. For supercritical water flow inside a 2  2 fuel rod bundle, non-uniform heat flux increases wall temperature beyond the reference uniform flux. Also, wall temperature peaks occur only at the gaps between the fuel rods.

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Supercritical water-cooled reactor

Zhang

2024

Nuclear Power Reactor Designs

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Thermal–hydraulic analysis of HPLWR fuel assembly cluster

Cited by 15 publications

References 10 publications

Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

Supercritical water-cooled reactor

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