Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors

Jo, Daeseong; Park, Jonghark; Chae, Hobyung

doi:10.1016/j.pnucene.2013.11.006

Cited by 34 publications

(5 citation statements)

References 10 publications

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“…This analysis was conducted for the reactor operating at 2 MW with 20 bundles of fuel elements including four pieces of control rod bundles, with the cooling fluid flow rate entering the channel varied from 1.25 to 3.5 m 3 /h, inlet temperature of 44.5 °C, and pressure of 1.9970 bar. The magnitude of the heat flux boiling (qONB) was calculated by using Rohsenow equation (9), the critical heat flux (qDNB) calculated by using Mirshak equation (10) and the flow stability (Sf) calculated by using Whittle and Forgan equation (11). To identify the boiling process, an Onset Nucleate Boiling Ratio (ONBR) was defined as the ratioof boiling heat flux (qONB) to actual heat flux.…”

Section: Resultsmentioning

confidence: 99%

Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

Rahardjo¹,

Wardhani²

2017

Atom Indo.

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The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety analysis of the converted TRIGA reactor is presented by considering the Dynamic Nucleate Boiling Ratio (DNBR) criterion, Onset Nucleate Boiling Ratio (ONBR) limit, and cooling fluid flow stability at various cooling fluid flow rate.The numerical analyses were performed using the HEATHYD program on the hottest channels of reactor core.The combination of heat transfer and fluid flow analysis were conducted for reactor operation at 2 MW with 20 fuel element bundles and four control rod bundles. Incoming fluid flow to the cooling channel was fixed at 44.5 °C temperature and 1.9970 bar pressure, and its flow rate was varied from 1.25 to 3.5 m 3 /h. By inputting these values, as well as the total power of fuel elements per bundle, the wall temperature distribution of the plate fuel element, cooling fluid temperature distribution, and pressure losses in the channels were obtained for the analysis of CHF limit, boiling limit, and flow stability. It was shown that no boiling occurred for the cooling fluid flow rate range of 2.4 to 3.5 m 3 /h, and even at the cooling fluid flow rate of 1.25 m 3 /h where some bubbles occurred, the DNBR was higher than the critical limit (more than 23) while the flow stability criterion in some channels were slightly less than 1 (unstable). At the cooling fluid flow rate of 1.4 m 3 /h, however, the flow became stable in all channel. The results showed that even though some bubbles start to occur, the plate-fueltype 2 MW TRIGA reactor can safely operate in the terms of CHF limit and flow stability.

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Section: Resultsmentioning

confidence: 99%

Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

Rahardjo¹,

Wardhani²

2017

Atom Indo.

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“…The wall temperature at the onset of nucleate boiling is calculated from the correlation of Bergles and Rohsenow, Jo et al (2014),…”

Section: The Onset Of Nucleate Boiling Ratiomentioning

confidence: 99%

“…To evaluate the critical heat flux, the Kaminaga Correlation given in the work ofJo et al (2014), is used, with 𝐴 respectively the flow area and the heated area.For downward flows If 𝐺 * ≥ 𝐺 * , then 𝑞 = min (𝑞 , , 𝑞 , ) If 𝐺 * < 𝐺 * , then 𝑞 = max (𝑞 , , 𝑞 , ) For upward flows If 𝐺 * ≥ 𝐺 * , then 𝑞 = min (𝑞 , , 𝑞 , )If 𝐺 * < 𝐺 * , then 𝑞 = max (𝑞 , , 𝑞 , )…”

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confidence: 99%

Lumped parameter model for the safety assessment of the IAEA 10MW benchmark reactor during a protected loss of flow accident

Hamed

2023

Energy Thermofluids Eng.

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The loss of flow accident in the IAEA 10Mw benchmark reactor is treated in this study through a coupled solution of the point kinetic and thermal hydraulic models. By employing the lumped parameter technic, a one-dimensional thermal hydraulic model is derived and solved beside the point kinetic model with the continuous reactivity feedback effect of the coolant and fuel temperatures. This coupled solution, allows us, to determine, the transient variation of the maximum fuel and coolant temperatures, furthermore, to the safety criteria (ONBR, OFIR, CHFR and BOR) which must be greater than the imposed safety limits, to avoid any nuclear reactor safety issue. Finally, the obtained results of our simulation were validated after a satisfactory comparison with the results of other codes system and developed computer codes.

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“…The fuel element and fluid channel studied in present work are choosed according to the general test reactor (Jo et al, 2014), it include two types: 1) the large plate used in the real reactor, its size is 2 mm × 60 mm × 600 mm, the channel gap is 2.5 mm; 2) The small plate used in the RERTR irradiation experiment (Meyer et al, 2012), its size is 1.27 mm × 25.4 mm × 101.47 mm, the channel gap is also 2.5 mm. Under the guidance of RERTR (Research reactor low concentration Program), Argonne National Laboratory has carried out a series of experimental studies on the performance of plate fuel elements (Meyer et al, 2012).…”

Section: Geometry Modelmentioning

confidence: 99%

Study on Flow Boiling Characteristics in Rectangle Channel After Formation of Blisters

et al. 2021

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Plate-type fuel elements is one of the first fuel structure choice for the novel integrated PWR, however, blisters will appear on the cladding induced by irradiation and fission. In this work CFD method was used to investigate the subcooled boiling characteristic of the water in rectangle channel with round and pillow blisters, the modified RPI model was also proposed, we can draw conclusions as follows: In the channel with round blister, as the blisters will increase the local flow resistance and more fluid will flow through center of the channel. Boiling occurred only in the area near the edges, nearly no vapor appeared at the center of the channel. The boiling region in channel with pillow shape blisters is wider and concentrated between two pillow blisters and downstream of the non-blisters side. The dry out area are both in the downstream region of blisters for the two types of channels.

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Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors

Cited by 34 publications

References 10 publications

Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

Lumped parameter model for the safety assessment of the IAEA 10MW benchmark reactor during a protected loss of flow accident

Study on Flow Boiling Characteristics in Rectangle Channel After Formation of Blisters

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