Using an IIST SBLOCA Experiment to Assess RELAP5/MOD3.2

Lee, Chien-Hsiung; Huang, I-Ming; Chang, Chen-Wei; Liu, Tay-Jian; Ferng, Yuh-Ming

doi:10.13182/nt99-a2957

Cited by 7 publications

(2 citation statements)

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“…The experiment of IIST facility was performed in order to simulate a 2% cold-leg break (the break area is 2% of the scaled cold-leg cross-section area) with total HPI failure [11]. This break is located in loop 2 of IIST facility, which is one of the two loops that do not have a pressurizer.…”

Section: Iist Facility and Sbloca Experimentsmentioning

confidence: 99%

Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

Yang

Wang

Lin

et al. 2012

Science and Technology of Nuclear Installations

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In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data.

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Section: Iist Facility and Sbloca Experimentsmentioning

confidence: 99%

Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

Yang

Wang

Lin

et al. 2012

Science and Technology of Nuclear Installations

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“…In particular, in the event of an accident that the water level in a reactor falls under low-flow velocity or pool conditions, the boiling two-phase flow pattern is complicated due to the rapid increase in void fraction and vapor bubble coalescence. For example, void-fraction and rod-temperature distributions during the boil-off process in the reactor core were obtained by system-integral tests conducted by Anoda et al (1990) and Lee et al (1999) and individual effect tests conducted by Anklam et al (1982), Pearson et al (1984), Aksan et al (1993), Kondo et al (1993) and Qazi et al (1994). An axial void fraction distribution is usually acquired by differential-pressure gages as a volume-averaged value.…”

Section: Introductionmentioning

confidence: 99%

Multi-dimensional void fraction measurement of transient boiling two-phase flow in a heated rod bundle

Arai

Furuya

Kanai

et al. 2015

Mechanical Engineering Journal

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A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied to a boiling two-phase flow experiment under the atmospheric pressure condition assuming at an accident or in a spent fuel pool in a boiling water reactor (BWR). The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. The wire electrodes of 0.2 mm in diameter are arranged in lattice patterns between the rod bundle, while the electric conductance value in a region near one wire and another corresponds to local void fraction in the central-subchannel region. The local void fractions at 32 points (= 6×6-4) can be obtained as a cross-sectional distribution. The local void fractions near the rod surface at 100 points (= 4×25) can be also estimated by the conductance value in a region between one wire and one rod. The devised sensors are installed at five height levels along the axis to acquire two-phase flow behavior. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The temporal resolution of void fraction measurement is 1600 frames (cross-sections) per second. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various inlet flow velocity, rod bundle power and inlet subcooling conditions. The experimental results were presented by the axial and cross-sectional distributions of void fraction, phasic velocity and bubble-chord length.

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