Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention Part II: Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

Chu, T. Y.; Bentz, J.H.; Slezak, Scott E.; Pasedag, W.F.

doi:10.1016/s0029-5493(96)01279-4

Cited by 36 publications

(8 citation statements)

References 6 publications

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“…The CYBL experiment carried out by Chu et al (1997) used a pressure vessel with a ratio of 1:1 to conduct an experimental study on the boiling heat transfer and flow process in the external cooling process. Different from the AP600/1,000 reactor pressure vessel, the lower head of the RPV in the CYBL experiment is an ellipsoid structure.…”

Section: Introductionmentioning

confidence: 99%

Study on the influences of RPV deformation on CHF under IVR conditions

Wen

Gao²,

Zhang³

et al. 2023

Front. Energy Res.

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After the reactor core melt, the thermal impact of high temperature molten mixture to the lower head of reactor pressure vessel (RPV) will cause the shape change of cooling channel constituted by reactor pressure vessel outer surface and other supporting structures. This phenomenon may lead to local heat transfer deterioration and then cause the failure of in-vessel retention (IVR). Therefore, it is necessary to study the reactor pressure vessel deformation under in-vessel retention conditions. This paper proposes a numerical method for the CHF of the outer wall of reactor pressure vessel using ANSYS and FLUENT. The geometry of coolant channel is modeled by ANSYS with the coupling of temperature and mechanical field analysis in ANSYS, and the critical heat flux (CHF) of reactor pressure vessel outer surface is further predicted by FLUENT. With the comparing between test data and calculation, the calculation methods are verified. The results show that the CHF will be decreased by the deformation of PRV caused by the core melt.

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Section: Introductionmentioning

confidence: 99%

Study on the influences of RPV deformation on CHF under IVR conditions

Wen

Gao²,

Zhang³

et al. 2023

Front. Energy Res.

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“…This measure is adopted in low and medium power reactors, such as the AP600, the AP1000 [3,4], the Loviisa nuclear power plant [5], and the KERENA [6] as a design feature for severe accident mitigation, and in the high-power reactors of the APR (Advanced Power Reactor) 1400 and the APR + (Advanced Power Reactor Plus) as an accident management strategy [7][8][9]. Many studies have been performed to evaluate the IVR-ERVC [10][11][12][13], but more efforts are needed to solve the uncertainties of IVR-ERVC evaluation.…”

Section: Introductionmentioning

confidence: 99%

Corium Behavior in the Lower Plenum of the Reactor Vessel Under Ivr-Ervc Condition: Technical Issues

Park¹,

Kang²,

Hong³

et al. 2012

Nuclear Engineering and Technology

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“…[1][2][3][4] This measure has been adopted in some low-power reactors such as the AP600 and the Loviisa nuclear power plants and in the highpower reactors of the AP1000 and the advanced power reactor (APR) 1400 as an accident management strategy for severe-accident mitigation with the aim of retaining the molten core material in-vessel. [5][6][7] However, it is known that the thermal margin between the volumetric heat source in the corium pool of the reactor's lower plenum and the heat transfer rate from the lower reactor vessel wall to the coolant in the reactor's cavity is not sufficient for a high-power reactor like the APR1400, unlike the low-power reactors of the AP600 and the Loviisa nuclear power plants.…”

Section: Introductionmentioning

confidence: 99%

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

Park

Kim

et al. 2008

Journal of Nuclear Science and Technology

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One-dimensional (1D) air-water two-phase natural circulation flow in the ''thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow-one-dimensional'' (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases.

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Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention Part II: Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

Cited by 36 publications

References 6 publications

Study on the influences of RPV deformation on CHF under IVR conditions

Study on the influences of RPV deformation on CHF under IVR conditions

Corium Behavior in the Lower Plenum of the Reactor Vessel Under Ivr-Ervc Condition: Technical Issues

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

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