• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• 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.52 no.3
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.22 no.5
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• pp.554-560
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• 2021

CRUD (chalk river unidentified deposits) is a porous material deposited on the surface of nuclear fuel during nuclear power plant operation. The CRUD is composed of metal oxides, such as iron, nickel, and chromium. It is essential to investigate the effects of the CRUD layer on the wall heat transfer between the nuclear fuel surface and the coolant in the event of a nuclear accident. CRUD only negatively affects the temperature of the nuclear fuel due to heat resistance because the effects of the CRUD layer on two-phase boiling heat transfer are not considered. In this study, the physical property models for the porous CRUD layer were developed and implemented into the SPACE code. The effects of boiling heat transfer models on the peak cladding temperature and quenching were investigated by simulating a reflood experiment. The calculation results showed some positive effects of the CRUD layer.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.490-491
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• 2005

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• 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.