Gas entrainment (GE) from cover gas, which is an inert gas to cover sodium coolant in a reactor vessel, is one of key issue for Sodium-cooled fast reactors (SFRs) design to prevent unexpected effects to core reactivity. In this research series, evaluation method has been investigated for surface dimple depth growth of unstable drifting vortex dimples on the liquid surface in the reactor vessel. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortex on water experiments in a circulating water tunnel. The unstable drifting flow vortexes on the water surface were generated as wake vortexes behind a plate obstacle. Downward flow velocity was induced by bottom slit flow pass along the flow channel. In the previous study, the onset conditions of the gas entrainment were evaluated based on existing non-dimensional numbers method by using the STREAM-VIEWER code. However, the CFD predication accuracy of the detail flow field itself was not clear especially for vortex frequency in the wake flow and detail velocity profiles in the flow channel. In this study, to clarify the accuracy of CFD analysis, Strouhal numbers of vortex frequency and detail flow velocity profiles were compared with experimental data which were measured by Particle Image Velocimetry (PIV) method. As the results, the Strouhal numbers of the vortex frequency behind the plate obstacle reasonably agreed with experimental data. Prediction accuracy for the velocity profiles in the flow channel were also confirmed by comparisons with measured data by the PIV method.

Gas entrainment (GE) from cover gas, which is an inert gas to cover Sodium coolant in a reactor vessel, is one of key issue for Sodium-cooled fast reactors (SFRs) design to prevent unexpected effects to core reactivity. In this study, evaluation method has been investigated for surface dimple depth growth of unstable drifting vortex dimples on the liquid surface in the reactor vessel. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments in the circulating water tunnel. The unstable drifting flow vortexes on the water surface were generated as wake vortexes behind a plate obstacle. Downward flow velocity was induced by bottom slit flow pass along the flow channel. In CFD analysis, flat water surface assumption was used with free-slip model for the development of a conventional design method. The gas core lengths from the surface vortexes were evaluated based on existing non-dimensional numbers method by using the stream-viewer code provided from Japan Atomic Energy Agency (JAEA). In this paper, the results of comparison between experiments and analyses were discussed.

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design. Numerical simulation is a useful method to deal with such a complex phenomenon. We have been developing numerical analysis models of the hot-leg piping using Unsteady Reynolds Averaged Navier-Stokes simulation with Reynolds stress model. In this study, numerical simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment, but it was found some underestimation at other parts, especially in relatively high frequency range. Eigenvalue vibration analysis was also conducted using a finite element method. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by URANS simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.

Thermal striping on the core instrumentation plate (CIP) at the bottom of the upper internal structure (UIS) of an advanced loop-type sodium-cooled fast reactor in Japan (Advanced-SFR) has been numerically investigated. At the top of the core below the CIP, the sodium at high temperature flows out from the fuel subassemblies (FSs) and the sodium at low temperature flows out from the primary control rod (PCR) and backup control rod (BCR) channels, and also the radial blanket fuel subassemblies (RBFSs) at the outer side of the core. In order to predict the thermal striping on the CIP caused by mixing fluids at different temperatures from the FSs, the PCR and the BCR channels, and the RBFSs, a numerical estimation method using a spatial connection methodology between the upper plenum analysis and the local area analysis for the target area has been developed. By using the connection methodology, the numerical simulation considering the influence of the transversal flow in the UIS and the external flow around the UIS in the upper plenum can be performed to improve the accuracy of the estimation results. In this paper, the outline of the spatial connection methodology including data transfer technique from the upper plenum analysis to the local area analysis was described. As a validation process, numerical simulation of the water experiment using the test apparatus named TAFUT which was 1/3-scaled 1/6 partial model of the upper plenum of the Advanced-SFR was performed to confirm applicability of the spatial connection methodology to a practical thermal striping problem. The numerical result of temperature distribution was compared with the measured result in TAFUT experiment. Additionally, mesh sensitivity of the local area analysis model to the numerical results was indicated by using a small and a large area models in order to suggest an appropriate local area analysis model.

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan Atomic Energy Agency, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor during the core disruptive accident. Thermal-hydraulics evaluations in FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, thermal-hydraulics in FAIDUS are investigated by using a subchannel analysis code ASFRE, which is applicable to a wire-wrapped fuel pin bundle with a distributed resistance model and a simplified turbulence mixing model. At first, the distributed resistance model was validated by comparison of pressure drop coefficients with experimental data obtained in water experiments with simulated FAs under the condition of wide-range Reynolds number. And then, the turbulence mixing model was validated by comparison of temperature distribution in the pin bundle with experimental data obtained in sodium experiments with simulated FAs. After the applicability of ASFRE to FAs was confirmed through these validations, thermal-hydraulic analyses of a FA with 271 fuel pins without the inner duct and a FAIDUS with 255 fuel pins were conducted. The obtained results indicate that no significant asymmetric temperature distribution occurs in a FAIDUS as a FA without an inner duct. In addition, the temperature distribution of FAIDUS with 255 fuel pins under the low flow rate condition tended to be the same as that of a FA with 271 fuel pins due to the local flow acceleration and the flow redistribution caused by the buoyancy force.

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration of a hot-leg piping is one of main concerns in the design. The flow field in the hot-leg piping is affected by flow disturbance at the inlet, so it is important to evaluate flow field including the upper plenum. In this study, we analyzed unsteady fluid flow by using an integrated model of the upper plenum and the hot-leg piping system. Unsteady Reynolds averaged Navier-Stokes simulation with Reynolds stress model was used for the numerical simulation. The results were compared with experiment results of 1/3 scaled-model of hot-leg piping with the inlet conditions of rectified, swirling and deflected flows as well as simulation results of 1/3 scaled-model of hot-leg piping with rectified flow. In general, the simulation results obtained by using the integrated model show a similar tendency with the experiment results of deflect flows in the downstream region from the elbow exit. The coupling effect of swirling and deflected flows seems to be not significant although further investigation is needed.

Wall pressure measurements and flow visualization were conducted for a 90 degree elbow with an axis curvature radius the same as its inner diameter (125 mm, R c /D = 1). Reynolds numbers 320,000 and 500,000, based on the inner diameter and bulk velocity, were examined. A deflected inflow, having an almost constant velocity slope, was introduced in the present study. The velocity at the inside was 20% faster than the bulk velocity in the plane one diameter upstream of the elbow inlet. Ensemble averaged pressure distributions showed that no difference of normalized pressure could be found in cases of Reynolds numbers of 320,000 and 500,000. Comparisons with a uniform inlet flow case [1] proved that a low-pressure region at the intrados of the elbow was weakened and that a high-pressure region outside strengthened in the deflected inflow case. The present case had a characteristic pressure distribution that the pressure downstream of the elbow increased at the inside until two diameters downstream from the elbow exit. Flow visualization concluded that the corresponding pressure increase was caused by a collision of a strengthened secondary flow convected from the extrados. The unsteady pressure distribution in the present case showed that a circumferential extent of a strongly fluctuating region in the inside and downstream of the elbow decreased, comparing with the uniform inlet flow case.[1] Power spectral density functions of pressures exhibited that the fluctuation having the Strouhal number (based on the inner diameter and bulk velocity) of 0.6 existed in the downstream region of the elbow, which is 0.1 larger than that of the uniform inflow case.

This paper is intended to validate the numerical simulation tool, which is Unsteady Reynolds Averaged Navier-Stokes equation (U-RANS) approach with the Reynolds Stress Model using a commercial computational fluid dynamics code, by applying to the flow through a single short-elbow in the 1/10 and 1/3 scale water experiments simulating the Japan Sodium-Cooled Fast Reactor (JSFR) hot-leg piping. An additional objective of this paper is to investigate the effect of outlet condition at which the coolant overflows a partition wall in the upper part of an intermediate heat exchanger in the JSFR design. The numerical results were in good agreement with the 1/10 and 1/3 scale experimental data indicating time-averaged velocity distributions, flow field visualization, and power spectral densities of pressure fluctuation. These comparisons can conclude that the U-RANS numerical simulation tool was validated with its applicability to a single short elbow flow. The numerical simulation has also shown that the unsteady flow fields in the short elbow flow, which was characterized by a cyclic secondary flow and the subsequent horseshoe vortex. In this study, the effect of the outlet condition was also examined through the numerical simulation. At the outlet of the pipe, the simulation modeled the partition wall in the upper part of the intermediate heat exchanger, which has never been simulated in the experiments. The numerical simulation results were compared between with and without the intermediate heat exchanger at the pipe outlet in terms of the time-averaged velocity distribution, pressure fluctuation power spectral density, and so on. In the result, no significant difference between them was observed, so that it can be said that the effect of the outlet condition is negligibly small.

Flow-induced vibration of hot-leg pipings is one of concerns for the design of Japan Sodium-cooled Fast Reactor (JSFR) which is now being developed. The flow field in the hot-leg pipings is supposed to be affected by flow disturbances at the entrance, so it is important to evaluate flow fields including the upper plenum. In this study, a simulation model of the upper plenum and the hot-leg piping system of JSFR was developed. Unsteady fluid flow analyses were then conducted by unsteady Reynolds averaged Navier-Stokes simulation (URANS) with Reynolds stress model. The appropriateness of the calculated results was discussed by comparing available scale model test results. Furthermore, a prototype model for vibration analysis of the hot-leg piping was developed. In the model, the transient pressure data predicted by the URANS were used as input data for the vibration analysis. The number of element was significantly reduced from that of CFD model by considering the correlation length of stress fluctuation. In addition, a stress mapping tool from a CFD model to a model for vibration analysis was created.

In design of the Japan Sodium-cooled Fast Reactor (JSFR), mean velocity of the coolant is approximately 9 m/s in the primary hot leg (H/L) piping which diameter is 1.27 m. The Reynolds number in the H/L piping reaches 4.2×10 7 . Moreover, a short-elbow which has R c /D = 1.0 ( R c : Curvature radius, D : Pipe diameter) is used in the hot leg piping in order to achieve compact plant layout and reduce plant construction cost. In the H/L piping, flow-induced vibration (FIV) is concerned due to excitation force which is caused by pressure fluctuation on the wall closely related with the velocity fluctuation in the short-elbow. In the previous study, relation between the flow separation and the pressure fluctuations in the short-elbow was revealed under the specific inlet condition with flat distribution of time-averaged axial velocity and relatively weak velocity fluctuation intensity in the pipe. However, the inlet velocity condition of the H/L in a reactor may have ununiformed profile with highly turbulent due to the complex geometry in reactor vessel (R/V). In this study, the influence of the inlet velocity condition on unsteady characteristics of velocity in the short-elbow was studied. Although the flow around the inlet of the H/L in R/V could not simulate completely, inlet velocity conditions were controlled by installing the perforated plate with plugging the flow-holes appropriately. Then expected flow patterns were made at 2 D upstream position from the elbow inlet in the experiments. It was revealed that the inlet velocity profiles affected circumferential secondary flow and the secondary flows affected an area of flow separation at the elbow, by local velocity measurement by the PIV (particle image velocimetry). And it was found that the low frequent turbulence in the upstream piping remained downstream of the elbow though their intensity was attenuated.

Thermal striping phenomena caused by mixing of fluids at different temperature is one of the most important issues in design of Fast Breeder Reactors (FBRs), because it may cause high-cycle thermal fatigue in structure and affect the structural integrity. A numerical simulation code MUGTHES has been developed to investigate thermal striping phenomena and to estimate high cycle thermal fatigue in FBRs. In this study, numerical simulation for the WATLON experiment which was the water experiment of a T-junction piping system (T-pipe) conducted in JAEA was carried out to validate the MUGTHES and to investigate the relation between the mechanism of temperature fluctuation generation and the unsteady motion of large eddy structures. In the numerical simulation, the large eddy simulation (LES) approach with standard Smagorinsky model was employed as eddy viscosity model to simulate large-scale eddy motion in the T-pipe. The mesh as the same with the previous study as reference, the finer mesh and the coarser mesh arrangements were employed to estimate the Grid Convergence Index (GCI) for uncertainty quantification in the validation process. The modified method of the GCI estimation based on the least squire version could successfully quantify uncertainty. Through the numerical simulations, it was indicated that the fine mesh arrangement could improve the temperature distribution in the wake. It could be found that the thermal mixing phenomena in the T-pipe were caused by the mutual interaction of the necklace-shaped vortex around the wake from in the front of the branch jet, the horseshoe-shaped vortex and the Karman’s vortex motions in the wake.

To survey core internals simply, new microscopic observation techniques were developed. These techniques were involved the use of “Gel electrode” and “underwater microscope”. A Gel electrode can etch the surface of core internals without a watertight reservoir that makes an etching environment. The underwater microscope can be used to observe surface figures of core internals directly. These techniques were applied to the cracks detected at the shroud support of Tokai II Power Station for investigation of the cause of crack initiation. It became clear that initiation points of the detected cracks were in a nickel-base weld metal.

Thermal striping phenomenon caused by mixing of fluids at different temperatures is one of the most important issues in design of fast breeder reactors (FBRs), because it may cause high-cycle thermal fatigue in structure. Authors have been developed a numerical simulation code MUGTHES to investigate thermal striping phenomena in FBRs and to give transient data of temperature in the fluid and the structure for an evaluation method of the high-cycle thermal fatigue problem. MUGTHES employs the boundary fitted coordinate (BFC) system and deals with three-dimensional transient thermal-hydraulic problems by using the large eddy simulation (LES) approach and artificial wall conditions derived by a wall function law. In this paper, numerical simulations of MUGTHES in T-junction piping system appear. Boundary conditions for the simulations are chosen from an existing water experiment in JAEA, named as WATLON experiment. The wall jet condition in which the branch pipe jet flows away touching main pipe wall on the branch pipe side and the impinging jet condition in which the branch pipe jet impinges on the wall surface on the opposite side of the branch pipe are selected, because significant temperature fluctuation may be induced on the wall surfaces by the branch pipe jet behavior. Numerical results by MUGTHES are validated by comparisons with measured velocity and temperature profiles. Three dimensional large-scale eddies are identified behind of the branch pipe jet in the wall jet case and in front of the branch pipe jet in the impinging jet case, respectively. Through these numerical simulations in the T-pipe, generation mechanism of temperature fluctuation in thermal mixing process is revealed in the relation with the large-scale eddy motion.

A water experiment is performed to investigate thermal striping phenomena in a T-pipe junction which is a typical geometry of fluid mixing. The flow velocity ratio and temperature difference were experimental parameters. The jet form was classified into four patterns; (1) impinging jet , (2) deflecting jet , (3) re-attachment jet and (4) wall jet according to the inflow condition. The parameter experiments showed that the jet form could be predicted by a momentum ratio between the two pipes. The thermochromic liquid crystal sheet showed that a cold spot was formed at the wall surface in the main pipe in the cases of the impinging jet and the wall jet . From the temperature measurement in the fluid, temperature fluctuation intensity was high along the edge of the jet exiting from branch piping. A database of temperature fluctuation and frequency characteristics was established for an evaluation rule of thermal striping in a T-pipe junction.

Turbulent statistics near a structural surface, such as a magnitude of temperature fluctuation and its frequency characteristic, play an important role in damage progression due to thermal stress. A Large Eddy Simulation (LES) has an advantage to obtain the turbulent statistics especially in terms of the frequency characteristic. However, it still needs a great number of computational cells near a wall. In the present paper, a two-layer approach based on boundary layer approximation is extended to an energy equation so that a low computational cost is achieved even in a large-scale LES analysis to obtain the near wall turbulent statistics. The numerical examinations are carried out based on a plane channel flow with constant heat generation. The friction Reynolds numbers ( Re τ ) of 395 and 10,000 are investigated, while the Prandtl number (Pr) is set to 0.71 in each analysis. It is demonstrated that the present method is cost-effective for a large-scale LES analysis.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: evaluating variations in plastic strain increments and evaluating variations in the width of elastic core. To verify the validity of these criteria, calculations were performed for several typical models in C-TDF [2]. This paper shows the results of a simple cylinder model. Cyclic plastic analyses were performed applying sustained internal pressure and alternating linear temperature distribution through the wall. Analyses were performed with various load ranges to evaluate the precise ratchet limit and its behavior across the limit. Both pressure and thermal stress were given parameters. In the analyses, Elastic-Perfectly-Plastic (EPP) material was used and also strain hardening material for comparison. The ratchet limit in the Code [3] is based on Miller’s theoretical analysis [4] for a cylinder assuming a uni-axial stress state, whereas real vessels are in multi-axial stress state. By our calculations, we also examined the ratchet limit in real vessels. The results show that for the cylinder in a multi-axial stress state, the ratchet limit rises 1.2 times the ratchet limit by the Code. The evaluation results show that variations in equivalent plastic strain increments can be used for ratchet criterion and ratcheting can be assessed by confirming the presence of elastic core in the second cycle.

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1],[2] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: (1) Evaluating variations in plastic strain increments, and (2) Evaluating the width of the area in which Mises equivalent stress exceeds 3S m . To verify of these criteria, we selected notched cylindrical vessel models as prime elements. To evaluate the effect of the local peak stress distribution on these criteria, cylindrical vessels with a semicircular notch on the outer surface were selected for this analysis. We used two notch configurations for our analysis, and the stress concentration factor for the notches was set to 1.5 and 2.0. We conducted elastic-plastic analysis to evaluate the ratchet limit. Sustained pressure and alternating enforced longitudinal displacements which causes secondary stress were used as parameters for the elastic-plastic analysis. We found that when no ratchet was observed, the equivalent plastic strain increments decreased and the area in which Mises equivalent stress exceeds 3S m are below the certain range.