• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.2 no.4
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• pp.82-87
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• 2003

Flow angle with Swirl in a horizontal circular tube and a cylindrical annuli were experimentally studied for its visualization. This present investigation deals with flow angle, flow visualization studies and vortex core by using oil smoke and a hot wire anemometer for Re = 40,000 and 50000 at X/D = 41, 59 and 71 in a horizontal circular tube. In the swirl air flow, a vortex core was formed at high swirl intensity along the test tube. The flow angle and the vortex core depended on the swirl intensity along the test tube. The results of flow angles with swirl measured by flow visualization and hot wire reasonably agree with those of Sparrow One of the primary objectives of this research was to measure the flow angle with swirl in a cylindrical annuli along the test tube for different Reynolds numbers. The Reynolds number for these measurements ranged from 60,000 to 100,000 with L/D = a to 4.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Journal of computational fluids engineering
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• v.10 no.4 s.31
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• pp.66-71
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• 2005

HANARO, a multi-purpose research reactor, 30 MWth open-tank-in-pool type, is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and a target handling tool is under development for loading and unloading it in a circular flow tube (OR-5) of HANARO. A guide tube is extended from the reactor core to the top of the reactor chimney for easily loading the target under a normal operation of the reactor. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube. The jet flow was suppressed in the guide tube after reducing the inner diameter of a flow restriction orifice installed in the OR-5 flow tube for adding the pressure difference in the flow tube. This paper describes an analytical analysis to calculate the flow distribution in the core of HANARO after suppressing the jet flow of the guide tube. As results, it was confirmed through the analysis results that the flow distribution in the core of HANARO were not adversely affected.

• Journal of computational fluids engineering
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• v.10 no.2
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• pp.1-6
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• 2005

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed af inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve meters (12 m) depth of the reactor pool and cooled by the upward flow that the coolant enters the lower inlet of the plenum, rises up through the grid plate and the core channel and comes out from the outlet of chimney. A fission moly guide tube is extended from the reactor core to the top of the reactor chimney for easily loading a fission moly target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by jet flow. This paper describes an analytical analysis that is the study of the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the flow rate, reduced to about fourteen kilogram per second (14 kg/s) from the original flow rate of sixteen point three kilogram per second (16.3 kg/s) did not show the guide tube jet.

• Journal of computational fluids engineering
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• v.14 no.2
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• pp.59-67
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• 2009

An accurate prediction of the bypass flow is of great importance in the VHTR core design concerning the fuel thermal margin. Nevertheless, there has not been much effort in evaluating the amount and the distribution of the core bypass flow. In order to evaluate the behavior and the distribution of the coolant flow, a unit-cell experiment was carried out. Unit-cell is the regular triangular section which is formed by connecting the centers of three hexagonal blocks. Various conditions such as the inlet mass flow rate, block combinations and the size of bypass gap were examined in the experiment. CFD analysis was carried out to analyze detailed characteristics of the flow distribution. Commercial CFD code FLUENT 6.3 was validated by comparing with the experimental results. In addition, SST model and standard k-$\varepsilon$ model were validated. The results of CFD simulation show good agreements with the experimental results. SST model shows better agreement than standard k-$\varepsilon$ model. Results showed that block combinations and the size of the bypass gap have an influence on the bypass flow ratio but the inlet mass flow rate does not.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.7 no.4
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• pp.709-714
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• 2006

The propose of this study was to investigate the feasibility of using waste lime core to remove phosphorus from wastewater in continuous flow reaction. The phosphorus was found to be removed from municipal wastewater by hydroxyapatite crystallization and precipation. Waste lime core size 1, 2 showed phosphorus removal rate of about 90% during early 11 hrs of run time. In addition, breakpoint time was decreased by increased inflow rate regardless of waste lime core size.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.26 no.4B
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• pp.505-513
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• 2001

지문영상의 분류는 데이터베이스의 용량이 클 경우 검색시간을 효율적으로 단축시킬 수 있는 핵심적인 기술이다. 따라서 본 논문에서 core point 와 flow-line 추적을 이용한 효율적인 지문 영상 분류 기법을 제안한다. 제안한 방법은 특히 압착 날인된 지문 영상의 분류에 적합한 방법으로 크게 2단계로 이루어져 있다. 첫 번째 단계에서는 먼저 Poincare index를 이용하여 core point를 찾아내고 이를 바탕으로 개략적인 분류를 수행한다. 그 다음 두 번째 단계에서는 core point를 중심으로 flow-line을 추적하여 그 결과를 가지고 세부적인 분류를 수행한다. 세부분류 단계에서는 평활화된 블록의 방향정보를 이용한 효과적인 flow-line 추적 알고리즘과 이를 이용한 새로운 분류 방법이 제안된다. 제안한 방법은 회전이나 이동 그리고 약간의 잡음에 강인한 지문 분류 방법으로 지문입력기를 통하여 획득된 700장의 지문 영상에 적용해 본 결과 93.6%의 분류율을 나타내었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.