Procedures for manufacturing an instrumented nuclear fuel element

Durazzo, Michelangelo; Umbehaun, Pedro E.; Torres, Walmir M.; Souza, J.A.B.; Silva, D.G.; Andrade, Delvonei Alves de

doi:10.1016/j.pnucene.2019.01.021

Cited by 9 publications

(6 citation statements)

References 6 publications

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“…• Quantity of prims layers: 25 layers. Based on [1,3,4,7,8], it is possible to assume that the coolant does not experience a bulk phase change (bulk boil) even when the flow is very reduced for a power of 3.5MW, not requiring a more complex approach as necessary for multiphase flows. In this way, it was assumed that only liquid coolant flows through the channel.…”

Section: Methodsmentioning

confidence: 99%

“…Other relevant features of the IFE are summarized as follows [1,2,7,9]: a total of 15 thermocouples (TCs) are present in different channels. The TCs are encapsulated in small aluminum discs made with the 6061 alloy, avoiding the Galvanic corrosion and possible accidents involving leak of fission products to the reactor pool.…”

Section: Figure 1: Schematic View Of the Ife -Main Dimensions And The...mentioning

confidence: 99%

“…The IEA-R1 is a plate type research reactor in which an investigation of the flow parameters that could lead to a such undesirable state in several stead-state or transient operational regimes could be performed in a safe manner. To aid in this investigation, the IEA-R1 has an instrumented fuel element -IFE to which the pressure drop is known, and that is equipped with thermocouples that are used to measure the temperature in different points through the channel length in some channels of the IFE [1][2][3][4][6][7][8][9]. Despite of this, the IFE instrumentation does not allow to access the details of the flow pattern, which could be done by using a CFD (Computational Fluid Dynamics) code [3].…”

Section: Introductionmentioning

confidence: 99%

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Simulation of steady-state and transient loss of cooling accident of a channel in a reactor plate-type fuel element

Monteiro

Maiorino²

2023

Braz. J. Rad. Sci.

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The suitable cooling of fuel elements in a nuclear reactor is an important requirement that must be met to avoid that the fuel temperature rises above the safety limits according to the reactor design. During the reactor operation, there are transients that could lead to a temperature that overcomes these limits, such as those related to the cooling system. The CFD codes are tools that could aid in the understand effort of the phenomenon during such transients, allowing to access details of the flow that are not possible, or are possible only with limitations, by using other kind of codes or experiments. In the present work the results obtained using ANSYS-CFXÒ code for the IEA-R1 reactor during a steady state and transient of slow loss of cooling accident event are presented. The results obtained shown a good agreement with experimental data reported and works that used this reactor as case of study. These results are part of a research in which the main objective is to simulate the flow of the coolant in the fuel element channels during transients. These results would support an initial analysis of the flow during the transition from forced to natural convection that occurs when the coolant flow falls below a settled value and the valve on the bottom of the fuel element opens by gravity, aiming to understand better phenomena involved and the limitations of the models available in the ANSYS-CFXÒ code.

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Section: Methodsmentioning

confidence: 99%

Section: Figure 1: Schematic View Of the Ife -Main Dimensions And The...mentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Simulation of steady-state and transient loss of cooling accident of a channel in a reactor plate-type fuel element

Monteiro

Maiorino²

2023

Braz. J. Rad. Sci.

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“…Thus, all changes addressed in this paper keep the order of execution in its ascending sequence presented in Tables 1-4. Details about all processes can be easily found in the correspondent literature [27] [31].…”

Section: Process Mapping and Data Gatheringmentioning

confidence: 99%

“…This results in the nomenclature of the selected fuel for this work: low enriched uranium silicide and dispersed in aluminum matrix, generally referred to as LEU U 3 Si 2 -Al [29] [30].The final scope delimitation refers to the vector responsible for the availability of nuclear fuel in the research reactor. In this study we considered the plate type fuel element, since this is the type of fuel element most commonly used in nuclear research reactors[31].…”

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confidence: 99%

Studies on Capacity Expansion of Fuel Plants for Nuclear Research Reactors

Negro¹,

Durazzo²,

Mesquita³

et al. 2018

WJNST

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The demand for nuclear fuel for research reactors is rising worldwide. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently and safely. We proposed a specific procedure for increasing production capacity. That procedure was tested with data from a real plant, which produces plate-type fuel elements loaded with LEU U 3 Si 2 -Al fuel. The test was made by means of discrete event simulation, and the results indicated the proposed procedure is efficient in raising production capacity.

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