Core thermal–hydraulic design

Takada, Eiji; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke

doi:10.1016/j.nucengdes.2004.07.009

Cited by 15 publications

(8 citation statements)

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“…The maximum fuel temperature estimated in the HTTR design stage was 1492 • C in the high-temperature test operation at 440 effective full-power days, and it was lower than a thermal design limit of 1495 • C in the normal operation [5]. However, an estimated temperature of 1492 • C was obtained by using hot spot factors with a large safety margin for the consideration of uncertain-ties in the design stage without the HTTR practical operating data, and thus there is no doubt that the estimated temperature includes excessive conservativeness.…”

Section: Introductionmentioning

confidence: 74%

“…Based on the data obtained by the rise-to-power tests of the HTTR up to an outlet coolant temperature of 850 • C [11], the systematic factors of the hot spot factors and analysis conditions were re-evaluated. The systematic factors related to thermal power, axial power distribution, and flow rate distribution; and the analysis conditions related to inlet coolant temperature, primary coolant flow rate, control rod position, and effective full-power days were revised [5,12]. Table 2 shows the revised systematic factors.…”

Section: Estimation Of Maximum Fuel Temperature 41 Revision Of Systmentioning

confidence: 99%

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Evaluation of maximum fuel temperature in HTTR

Inaba

Tochio

Ueta

et al. 2014

Journal of Nuclear Science and Technology

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In order to ensure the thermal integrity of fuel in the high temperature engineering test reactor (HTTR), it is necessary that the maximum fuel temperature in the normal operation is to be lower than a thermal design limit of 1495 • C. In the core thermal and hydraulic design of the HTTR, the maximum fuel temperature was estimated to be 1492 • C, which satisfied the thermal design limit. However, the estimated temperature was derived by using hot spot factors with a large safety margin for the consideration of uncertainties in the design stage without the HTTR practical operating data, and thus there is no doubt that the estimated temperature includes excessive conservativeness. In order to obtain the maximum fuel temperature with appropriate conservativeness, the maximum fuel temperature has been re-evaluated on the basis of the HTTR operating data. In this paper, the random factors of the hot spot factors are revised by using the HTTR first fuel fabrication data, and the new maximum fuel temperature is estimated. As a result, the estimated maximum fuel temperature can be reduced to 1424 • C. The reduction of the maximum fuel temperature leads to a larger thermal margin in nuclear and fuel designs.

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Section: Introductionmentioning

confidence: 74%

Section: Estimation Of Maximum Fuel Temperature 41 Revision Of Systmentioning

confidence: 99%

Evaluation of maximum fuel temperature in HTTR

Inaba

Tochio

Ueta

et al. 2014

Journal of Nuclear Science and Technology

Self Cite

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“…In the safety requirements of HTGR, the maximum fuel temperature was determined based on the criterion that the coating layer of fuel particles must remain on the particles and retain fission products within the particles. The maximum fuel temperature should be under an acceptable level at 1,495 o C during normal operation and not exceed 1,600 o C during any anticipated accident (Maruyama et al, 1994;Takada et al, 2004).…”

Section: Description Of the Reactormentioning

confidence: 99%

Concept of prismatic high temperature gas-cooled reactor with SiC coating on graphite structures

Trinuruk

Obara

2014

Annals of Nuclear Energy

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A 100-MW th prismatic high temperature gas-cooled reactor was designed to be a long-life small reactor. To acquire a passive safety feature, the reactor was mainly improved with regard to graphite oxidation resistance. The concept of applying a silicon carbide coating layer on the surface of the graphite structures in the core was proposed to overcome any serious problem from graphite oxidation during unforeseen situations. However, there was concern that the deviation of neutronic and thermal properties of silicon carbide from graphite could affect the reactor operation and the heat transfer characteristics. Therefore, in this study we investigated the effects of applying a silicon carbide coating layer over the graphite structures from the neutronic and thermal-hydraulic points of view. Silicon carbide coating can lower the effective multiplication factor and shorten the reactor operating cycle, but not significantly. From the viewpoint of thermal-hydraulic operation, silicon carbide has lower thermal conductivity than that of graphite, so the layer of silicon carbide could act as a wall to keep the heat from moving across the layer. Under normal operation, the layer of silicon carbide coating had a less significant impact on the maximum fuel temperature, and the temperature remained lower than

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“…Several studies of bypass flows have been made in the past by applying simplified methods such as flow network calculations and correlations obtained from experiments (Takada et al 2004;Nakano, Tuji, and Tazawa 2008). However, the distribution of temperature in the fuel pins and graphite blocks, as well as coolant outlet temperatures, are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core.…”

Section: Tablesmentioning

confidence: 99%

CFD Analysis of Core Bypass Phenomena

Johnson¹,

Satō²,

Schultz³

2010

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EXECUTIVE SUMMARYThe U.S. Department of Energy is exploring the potential for the very high temperature gas-cooled reactor (VHTR), which will be either a prismatic or pebble-bed type reactor. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow, which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increases in maximum fuel temperatures.Traditionally, simplified methods such as flow network calculations employing experimental correlations have been used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods that can be applied to the reactor core thermal hydraulic design and safety analysis.Computational fluid dynamics (CFD) codes, which have a local physics based simulation capability, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to a whole core length of 10.704 m. The computational domain is divided vertically with an upper reflector, a fuel section, and a lower reflector. Each side of the sector grid can be set as a symmetry boundary. The geometry used in the present study is shown in Figure E A steady Reynolds-averaged Navier-Stokes approach is employed using the commercial code FLUENT, which employs the finite volume method. Prior to the calculations, simple validation and grid sensitivity studies are conducted in order to evaluate the turbulence models and reduce numerical errors. Comparative studies are conducted varying several parameters including gap width, heat generation vii profile, turbulence model, and irradiation shrinkage to investigate the sensitivity of the flow and temperatures in the core to each parameter. In addition, a study wherein 20% bypass flow is achieved, thought to have occurred in earlier gas-cooled reactors, is performed to determine approximate gap widths that would be present and to visualize flow and temperature distributions.Results of a study that varied the gap width from 0 to 5 mm show that the bypass flow provides significant cooling to the graphite block, and that the maximum fuel temperature increases with gap width if the bypass flow is "robbed" from the coolant channels. Bypass flow induces a steep overall tem...

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Core thermal–hydraulic design

Cited by 15 publications

References 0 publications

Evaluation of maximum fuel temperature in HTTR

Evaluation of maximum fuel temperature in HTTR

Concept of prismatic high temperature gas-cooled reactor with SiC coating on graphite structures

CFD Analysis of Core Bypass Phenomena

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