Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

doi:10.1007/s10891-017-1641-7

Cited by 12 publications

(4 citation statements)

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“…1) is an aerodynamic open loop. It was designed for research of local values of flow velocity and pressure, as well as mixing of coolant in models of main equipment of nuclear reactors [10][11][12][13]. The stand includes: high-pressure fan, receiver tank, pipeline system, experimental model, measuring devices.…”

Section: Experimental Facilitymentioning

confidence: 99%

Visualization of coolant flow in the model of a nuclear reactor pressure vessel

Dobrov¹,

Doronkov²,

Ryazanov³

et al. 2020

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Section: Experimental Facilitymentioning

confidence: 99%

Visualization of coolant flow in the model of a nuclear reactor pressure vessel

Dobrov¹,

Doronkov²,

Ryazanov³

et al. 2020

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“…Another study of the hydrodynamic aspects at the fuel assembly-level has also been carried out by manufacturing a prototype of KLT-40S fuel device with the same geometry scale [7]. An experimental system was made by adding instrumentation and control aspects to the experimental model.…”

Section: Introductionmentioning

confidence: 99%

The Study of Floating Nuclear Power Plant Reactor Core Neutronic Parameters Using Scale 6.1 Code

Fajri¹,

Agung²,

Harto³

2020

Int. J. Adv. Sci. Eng. Inf. Technol.

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KLT-40S nuclear reactor is a small modular floating nuclear power plant made by Russia as a conventional light water reactor (LWR) problem solution nowadays, such as high overnight cost, long commissioning period, and lack of flexibility in supplying a small load of electricity and supplying electricity to isolated areas. With those characteristics, the KLT-40S is suitable to be applied to isolated areas with a small electrical load like archipelagic states such as Indonesia. Based on that reason, Indonesia needs to assess the KLT-40S floating nuclear power plant feasibility study through explorative research. One of those is assessing the reactor core neutronic parameter. In this research, the reactor core modelling is done by using the KENO VIA and T-6DEPL module in SCALE 6.1 code package. Several variations of reactor operating parameters such as fuel composition and configuration, fuel temperature, moderator temperature, and moderator void fraction had been done in this research. The aim was to get several neutronic parameters to confirm the core feasibility from operational and inherent safety perspectives. Those neutronic parameters are fuel cycle length and reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Based on this research result, the fuel configuration that produces 28 months of cycle length is the fuel base of dispersed UO 2 in the silumin matrix with 18,6 % 235 U enrichment. Both of the two fuel bases used in this research have inherent safety characteristics, which are shown by the negative value of the reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Dispersed UO 2 in the silumin matrix fuel base has better inherent safety characteristics than the UO 2 ceramic metal fuel base.

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“…The most accurate method of obtaining these data can be an aerodynamic experiment using an enlarged model of the fuel assembly [12][13][14][15]. However, taking into account the active use of CFD programs for thermal-hydraulic calculations in nuclear industry [16][17][18][19][20], the authors of this work carried out a numerical simulation of the coolant flow in the fragment of a fuel assembly using Ansys CFX. The results of the calculations were compared with the experimental data in order to evaluate the capabilities of the selected program for carrying out similar studies.…”

Section: Introductionmentioning

confidence: 99%

Experimental studies and numerical simulation of coolant hydrodynamics in the inlet area of nuclear reactor fuel assembly

et al. 2018

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An innovative core with an increased energy resource was used when designing RITM-200 reactor unit. The paper presents the results of experimental and numerical simulation of hydrodynamic processes occurring in the inlet region of the RITM reactor fuel assembly model. Computational mesh of fuel assembly, containing ~ 22 million elements was created using Ansys ICEM CFD. The values of relative axial velocities in several cross sections at the inlet to the bundle of fuel elements are obtained. At the inlet to the fuel rods bundle the velocity field is not uniform, due to the complex geometry of the fuel assembly. The obtained results of CFD-simulation can be used to determine the input boundary conditions for subchannel programs of the core thermal-hydraulic analysis. This allows taking into account uneven flow rate distribution in subchannels due to the complex geometry of the fuel assembly inlet region.

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Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Cited by 12 publications

References 0 publications

Visualization of coolant flow in the model of a nuclear reactor pressure vessel

Visualization of coolant flow in the model of a nuclear reactor pressure vessel

The Study of Floating Nuclear Power Plant Reactor Core Neutronic Parameters Using Scale 6.1 Code

Experimental studies and numerical simulation of coolant hydrodynamics in the inlet area of nuclear reactor fuel assembly

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