• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.802-810
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• 2020

Three-dimensional finite element simulations are implemented for the in-pile thermo-mechanical behavior in U-Mo/Al monolithic fuel plates with different thermal creep rates of cladding involved. The numerical results indicate that the thickness increment of fuel foil rises with the thermal creep coefficient of cladding. The maximum Mises stress of cladding is reduced by ~85% from 344 MPa on the 98.0^th day when the creep coefficient of cladding increases from 0.01 to 10.0, due to its equivalent thermal creep strain enlarged by 3.5 times. When the thermal creep coefficient of Aluminum cladding increases from 0 to 1.0, the maximum mesoscale stress of fuel foil varies slightly. At the same time, the peak mesoscale normal stress of fuel foil can reach 51 MPa on the 98.0^th day for the thermal creep coefficient of 10, which increases by 60.3% of that with the thermal creep un-occurred in the cladding. The maximum through-thickness creep strain components of fuel foil differ slightly for different thermal creep coefficients of cladding. The dangerous region of fuel foil becomes much closer to the heavily irradiated side when the creep coefficient of cladding becomes 10.0. The creep performance of Aluminum cladding should be optimized for the integrity of monolithic fuel plates.

• 한국압력기기공학회 논문집
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• 제7권4호
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• pp.44-50
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• 2011

Turbulent mixing of hot and cold coolants is one of the possible causes of high cycle thermal fatigue in piping systems of nuclear power plants. A typical situation for such mixing appears in turbulent flow through a T-junction. Since the high cycle thermal fatigue caused by thermal striping was not considered in the piping fatigue design in several nuclear power plants, it is very important to evaluate the effect of thermal striping on the integrity of mixing tees. In the present work, before conducting detailed evaluation, three thermal striping evaluation methodology suggested by EPRI, JSME and NESC are analyzed. Then, a by-pass pipe connected to the shutdown cooling system heat exchanger is investigated by using these evaluation methodology. Consequently, the resulting thermal stresses and the fatigue life of the mixing tee are reviewed and compared to each other. Futhermore, the limitation of each methodology are also presented in this paper.

• 한국마이크로전자및패키징학회:학술대회논문집
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• 한국마이크로전자및패키징학회 2002년도 추계기술심포지움논문집
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• pp.39-47
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• 2002

Heat is always the root of stress acting upon the electronic package, regardless of the heat due to the device itself during operation or working under the adverse environment. Due to the significant mismatch in coefficient of thermal expansion (CTE) and the thermal conductivity (K) of the packaging components, on one hand intensive research has been conducted in order to enhance the device reliability by minimizing the mechanical stressing and deformation within the package. On the other hand the effectiveness of different thermal enhancements are pursued to dissipate the heat to avoid the overheating of the device. However, the interactions between the thermal-mechanical loading has not yet been address fully. in articular when the temperature gradient is considered within the package. To address the interactions between the thermal loading upon the mechanical stressing condition. coupled-field analysis is performed to account the interaction between the thermal and mechanical stress distribution. Furthermore, process induced defects are also incorporated into the analysis to determine the effects on thermal conducting path as well as the mechanical stress distribution. It is concluded that it feasible to consider the thermal gradient within the package accompanied with the mechanical analysis, and the subsequent effects of the inherent defects on the overall structural integrity of the package are discussed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.134.1-134
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• 2002

• 대한기계학회논문집A
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• 제27권6호
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• pp.987-996
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• 2003

The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis, probabilistic analysis must be performed to show that failure probability Is within the limit. In this study, probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.237-248
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• 2013

Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.