Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor under Steady-State and Transient Conditions

Huda, M.Q.; Bhuiyan, S.I.; Chakrobortty, T. K.; Sarker, Mithun; Mondal, Manas

doi:10.13182/nt01-a3205

Cited by 8 publications

(4 citation statements)

References 1 publication

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“…The thermal and hydrodynamic purpose of the design is to safely remove the heat generated in the fuel without producing excessive fuel temperatures or steam void formations and without closely approaching the hidrodynamic Critical Heat Flux (CHF) (Huda et al 2001). As the IPR-R1 TRIGA reactor core power is increased, the heat transfer regime from the fuel cladding to the coolant changes from the single phase natural convection regime to subcooled nucleate boiling.…”

Section: Critical Heat Flux and Dnbrmentioning

confidence: 99%

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor

Mesquita¹,

Palma²,

Costa³

et al. 2012

Nuclear Reactors

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Section: Critical Heat Flux and Dnbrmentioning

confidence: 99%

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor

Mesquita¹,

Palma²,

Costa³

et al. 2012

Nuclear Reactors

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“…The reactor core consists of 100 fuel elements including 5 fuel follower control rods and 2 instrumented fuel elements. These fuel elements are arranged in a concentric hexagonal array within the core shroud [2]. The fuel is a solid, homogeneous mixture of Er-U-ZrH alloy containing 20% by weight of uranium enriched to 19.7% 235 U, and about 0.47% by weight of erbium.…”

Section: Introductionmentioning

confidence: 99%

Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

Rahman¹,

Akond²,

Basher³

et al. 2014

WJNST

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The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.

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“…The fuel elements at the reactor core are cooled by water natural convection. The heat removal capability of this process is great enough for safety reasons at the current maximum 250 kW power level of the reactor (Veloso, 1999 andHuda et al, 2001). However, a heat removal system is provided for removing heat from the reactor pool water, as show in Figure 1.…”

Section: Introductionmentioning

confidence: 99%

Power calibration of the TRIGA mark I nuclear research reactor

Mesquita¹,

Rezende²,

Tambourgi³

2007

J. Braz. Soc. Mech. Sci. & Eng.

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This paper presents the results and methodology used to calibrate the thermal power of the TRIGA Mark I IPR-R1 Research Reactor at the Nuclear Technology Development Centre (CDTN), in Belo Horizonte, Brazil. The TRIGA Mark I is a pool type reactor, cooled by water natural convection. The method used in the calibration consisted in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in this primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The thermal losses from the primary loop were not evaluated since the inlet and outlet temperatures were measured just above the water surface of the reactor pool. The temperature of the water in the reactor pool as well as the reactor room temperature were set as close as possible to the soil temperature to minimize heat leakages. These leakages are mainly due to the conduction through the concrete and metal walls and also due to the evaporation and convection through the water surface of the reactor pool.

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Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor under Steady-State and Transient Conditions

Cited by 8 publications

References 1 publication

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor

Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

Power calibration of the TRIGA mark I nuclear research reactor

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