Conference Agenda

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Session Overview
Session
8.05-3: Experimental & Numerical Studies - III
Time:
Tuesday, 17/Mar/2020:
3:30pm - 5:00pm

Session Chair: Imran Afgan, University of Manchester, United Kingdom
Session Chair: Yacine Addad, Khalifa University, United Arab Emirates
Location: R-3014

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Presentations

Investigation of thermal oscillation induced by dryout in printed circuit steam generator

Jin Su Kwon1, Sung Gil Shin1, Jeong Ik Lee1, Sang Ji Kim2, Yacine Addad3

1Korea Advanced Institute of Science and Technology; 2Korea Atomic Energy Research Institute; 3Khalifa University of Science, Technology and Research (KUSTAR)

To enhance the compactness and inherent safety of a Small Modular Reactor (SMR) using the pressurized water as a working fluid, a new concept of steam generator that can replace the existing helical-type steam generator is being investigated. As an alternative steam generator, a Printed Circuit Steam Generator (PCSG), a kind of Printed Circuit Heat Exchanger (PCHE), is designed and evaluated. It can provide a high heat transfer capability due to micro-sized flow channels and an outstanding structural rigidity that comes from the diffusion bonding process. Most SMRs employ a once-through type steam generator instead of the pool-type steam generator commonly used in large-sized nuclear power plants. It means the steam generator installed in an SMR is designed to produce superheated steam through the boiling region including nuclear boiling, dryout, and post-dryout heat transfer regimes. The dryout occurs where the liquid film in contact with the heated wall disappears and it induces a large axial wall temperature variation due to deterioration of heat transfer. The dryout front fluctuates due to the nature of boiling, which causes periodic thermal stresses in the heated tube. Therefore, research on the thermal oscillation induced by dryout becomes imperative since it can significantly deteriorate the service life of the steam generator. This study describes the PCSG design code and presents the validation with the TRACE code. Based on the analysis results, an experimental facility is designed and presented.



Flow Boiling CHF on Sudden Expansion Channel under Atmospheric Pressure

Yong Jin Kim1, Sang-Ki Moon2, Soon Heung Chang1, Yong Hoon Jeong1

1Korea Advanced Institute of Science and Technology; 2Korea Atomic Energy Research Institute

Deformation of cladding can occur during severe accident condition due to rapid pressure and temperature change. Deformed cladding changes flow and heat transfer characteristics, so coolability of core with deformed cladding was experimentally investigated. Until now, critical heat flux(CHF), which is one of criterion for heat transfer regime transition, on core with deformed cladding was not researched. In this study, CHF on deformed flow path with sudden contraction and sudden expansion of flow path was studied experimentally. Experiment was carried out in KAIST flow boiling experimental loop which was consist of a pump, a preheater, a condenser and a surge tank and test section was heated by DC rectifier with 75kW capacity. SUS 316 was chosen as material for test section with inlet diameter as 10mm and total heated length as 400mm. Tube with inner diameter as 4 mm with three different blockage length as 50, 100 and 150 mm was welded to simulate deformed flow path. Inlet mass flux range was from 50 to 250kg/m2s with inlet subcooling as 50K. Measured CHF value was compared with CHF on plain tube and CHF correlations for low pressure and low mass flux. In this study, location of CHF was changed by the mass flux condition. For low mass flux condition, CHF occurred at test section outlet with CHF value comparable with plain tube data and CHF correlations. For high mass flux condition, CHF occurred at sudden expansion point of test section at lower heat flux. From the numerical simulation of sudden expansion flow, it was concluded that CHF transition occurred due to local bubble stagnation due to balance between drag and buoyancy force of downward flow in sudden expansion flow.



The formation of the gap vortex street formation in square arranged rod bundles

Tomasz Kwiatkowski1, Afaque Shams2

1National Centre for Nuclear Research (NCBJ); 2Nuclear Research and Consultancy Group (NRG)

In the present study, a wide range of steady and unsteady Reynolds Averaged Navier-Stokes (RANS) computations have been performed to reproduce the gap vortex street in a square bare rod bundle configuration. In general, the considered geometric design is based on a well-known Hooper experiment, which is a bare rod bundle with the pitch-to-diameter ratio of P/D = 1.107. This article is divided into three main parts. Till now, most of the studies were performed with the use of a cross-sectional domain due to the limitations of turbulent models and available computer resources. Thus, in the first part, the influence of the cross-section size domain is assessed. A qualitative and quantitative comparison of the macroscopic flow pulsation generated in the full cross-section domain and cross-sectional domain by means of velocity magnitude, power spectral density, wavelength and amplitude of the oscillations is presented. In the second part, an extensive study has been performed to assess the appearance of the gap vortex in the laminar flow regime. In the third part, the influence of the gap width on the gap vortex street has been assessed. The obtained results have shown the prominent impact on the flow characteristics, such as wavelength and the dominant frequency of the pulsation.



Micro-Integral Effect Tests of URI-LO as Reactor Innovation Platform

Kyung Mo Kim, Ji Yong Kim, In Cheol Bang

Ulsan National Institute of Science and Technology

The 4th industrial revolution technologies represented by big data processing technology, artificial intelligence, drone, and 3D printing are widely applied to engineering fields for the benefits in terms of safety and economy. Application of the above-mentioned technologies to nuclear power plants will remarkably reduce the potential risk factor (human error) with enhancing the construction and operation efficiencies. Although there are many research activities advancing safety features and considering implementation of the 4th industrial revolution technologies, test beds evaluating feasibility of some new innovation concepts are insufficient because conventional integral effect test loops are too large and heavy to adopt new ideas like operating power plants. Therefore, micro-integral effect test facility with a small and light scale was developed to overcome such limits of large scale test beds. The UNIST reactor innovation loop, URI-LO which is a scale-down model (1/12 diameter ratio and 1/8 height ratio) of the APR-1400 was designed based on the three-level scaling method. The URI-LO is designed to have enough simulatability of three major accidents of reactor coolant pump seizure accident, station blackout, and loss of feedwater accident of the reference power plant. The refrigerant, FC-72 with a lower boiling point (~56 oC) is used as a working fluid to simulate the operating conditions of the reference power plant with relatively low pressure and temperature conditions. A series of performance tests on innovation concepts such as adoptabile 3D printed components and artificial intelligence diagnostics with acoustic emission sensors are carried out at the URI-LO with merits of its compact design, visibility with some transparent piping systems and components and simulatability on reactor accident scenarios.



 
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