• Journal of computational fluids engineering
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• v.10 no.4 s.31
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• pp.1-11
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• 2005

A numerical study of the evaluation of turbulence models for thermal striping phenomenon is performed. The turbulence models chosen in the present study are the two-layer model, the shear stress transport (SST) model and the V2-f model. These three models are applied to the analysis of the triple-jet flow with the same velocity but different temperatures. The unsteady Reynolds-averaged Navier-Stokes (URANS) equation method is used together with the SIMPLEC algorithm. The results of the present study show that the temporal oscillation of temperature is predicted by the SST and V2-f models, and the accuracy of the mean velocity, the turbulent shear stress and the mean temperature is a little dependent on the turbulence model used. In addition, it is shown that both the two-layer and SST models have nearly the same capability predicting the thermal striping, and the amplitude of the temperature fluctuation is predicted best by the V2-f model.

• 한국전산유체공학회:학술대회논문집
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• 2005.04a
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• pp.142-147
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• 2005

A numerical study of evaluation of turbulence models for thermal striping phenomenon is performed. The turbulence models chosen in the present study are the two-layer model, the shear stress transport (SST) model and the V2-f model. These three models are applied to the analysis of the triple jet flow with the same velocity but different temperature. The unsteady Reynolds-averaged Navier-Stokes (URANS) equation method is used together with the SIMPLE algorithm. The results of the present study show that the temporal oscillation of temperature is predicted only by the V2-f model, and the accuracy of the mean velocity, the turbulent shear stress and the mean temperature is a little dependent on the turbulence model used. The the two-layer model and the SST model shows nearly the same capability of predicting the thermal striping and the amplitude of the temperature fluctuation is predicted best by the V2-f model.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3928-3942
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• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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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.

• Journal of computational fluids engineering
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• v.19 no.4
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• pp.29-36
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• 2014

A computational study of a thermal striping in the upper plenum of PGSFR(Prototype Generation-IV Sodium-cooled Fast Reactor) being developed at the KAERI(Korea Atomic Energy Research Institute) is presented. The LES(Large Eddy Simulation) approach is employed for the simulation of thermal striping in the upper plenum of the PGSFR. The LES is performed using the WALE (Wall-Adapting Local Eddy-viscosity) model. More than 19.7 million unstructured elements are generated in upper plenum region of the PGSFR using the CFX-Mesh commercial code. The time-averaged velocity components and temperature field in the complicated upper plenum of the PGSFR are presented. The time history of temperature fluctuation at the eight locations of solid walls of UIS(Upper Internal Structure) and IHX(Intermediate Heat eXchanger) are additionally stored. It has been confirmed that the most vulnerable regions to thermal striping are the first plate of UIS. From the temporal variation of temperature at the solid walls, it was possible to find the locations where the thermal stress is large and need to assess whether the solid structures can endure the thermal stress during the reactor life time.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.837-843
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• 1998

This paper presents the thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the secondary piping of LMFR using Green's function method and standard FEM. The thermohydraulic loading conditions used in the present analysis are simplified sinusoidal thermal loads and the random type data thermal load. The thermomechainical fatigue damage was evaluated according to ASME code subsectionNH. The analysis results of fatigue for the sinusoidal and random load cases show that fatigue failure would occur at a geometrically discontinuous location during 90,000 hours of operation The fracture mechanics analysis showed that the crack would be initiated at an early stage of the operation. The fatigue crack was evaluated to propagate up to 5 ㎜ along the thickness direction during the first 944 and 1083 hours of operation for the sinusoidal and the random loading cases, respectively.

• The KSFM Journal of Fluid Machinery
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• v.10 no.5
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• pp.9-19
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• 2007

Fluid temperature fluctuations in a mixing tee pipe were numerically analyzed by LES model in order to clarify internal turbulent flows and to develope an evaluation method for high-cycle thermal fatigue. Hot and cold water with an temperature difference $40^{\circ}C$ were supplied to the mixing tee. Fluid temperature fluctuations in a mixing tee pipe is analysed by using the computational fluid dynamics code, FLUENT, Temperature fluctuations of the fluid and pipe wall measured as the velocity ratio of the flow in the branch pipe to that in the main pipe was varied from 0.05 to 5.0. The power spectrum method was used to evaluate the heat transfer coefficient. The fluid temperature characteristics were dependent on the velocity ratio, rather than the absolute value of the flow velocity. Large fluid temperature fluctuations were occurred near the mixing tee, and the fluctuation temperature frequency was random. The ratios of the measured heat transfer coefficient to that evaluated by Dittus-Boelter's empirical equation were independent of the velocity ratio, The multiplier ratios were about from 4 to 6.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.147-154
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• 2006

Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modelling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code_Saturne, developped at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code_Aster). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mock-up, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3874-3897
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• 2023

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High-Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next-generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds-averaged Navier-Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.