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In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

In the design of the Advanced Sodium-cooled Fast Reactor in Japan, the mean velocity of the coolant is approximately 9 m/s in the primary hot leg (H/L) piping, which has a diameter of 1.27 m. The Reynolds number in the H/L piping reaches 4.2 × 107. Furthermore, a short elbow, which has Rc/D = 1.0 (Rc: curvature radius, D: pipe diameter), is used in the H/L piping to achieve a compact plant layout and reduces plant construction costs. In the H/L piping, flow-induced vibration (FIV) is a concern due to the excitation force caused by pressure fluctuation in the short elbow. In a previous study, the relation between flow separation and pressure fluctuations in the short elbow was investigated under the specific inlet condition of a flat velocity profile of time-averaged axial velocity and relatively low velocity fluctuation intensity. However, the inlet velocity condition of the H/L in a reactor may be a highly turbulent non-uniform profile owing to the complex geometry in the reactor vessel (R/V). In this report, the influence of inlet velocity condition on the unsteady velocity characteristics in the short elbow was studied. Although the flow around the inlet of the H/L in the R/V could not be simulated completely, the inlet velocity conditions were controlled by installing a perforated plate, which appropriately plugged the flow holes. Then, controlled flow patterns were established at a position 2D upstream of the elbow inlet. Observed flow structures by particle image velocimetry indicated that the inlet velocity profiles affected a circumferential secondary flow, which then affected an area of flow separation at the elbow. It was also found that the velocity fluctuation at low frequency components observed upstream of the elbow could remain in downstream of the elbow though its intensity was attenuated.

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad r...

A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analys...

The Japan Sodium-cooled Fast Reactor (JSFR) is studied as an advanced loop-type sodium-cooled reactor. A selector-valve (SV) mechanism is adopted in the design of JSFR for its failed fuel detection...

In the volume-of-fluid algorithm, the calculations of the interface normal and curvature are crucially important for accurately simulating interfacial flows. However, few methods have been proposed for the high-precision interface calculation on an unstructured grid. In this paper, the authors develop a height function method that works appropriately on an unstructured grid. In the process, the definition of the height function is discussed, and the high-precision calculation method of the interface normal is developed to meet the necessary condition for a second-order method. This new method has highly reduced computational cost compared with a conventional high-precision method because the interface normal calculation is completed by solving relatively simple algebraic equations. The curvature calculation method is also discussed and the approximated quadric curve of an interface is employed to calculate the curvature. Following a basic verification, the developed height function method is shown to successfully provide superior calculation accuracy and highly reduced computational cost compared with conventional calculation methods in terms of the interface normal and curvature. In addition, the height function method succeeds in calculating accurately the slotted-disk revolution problem and the oscillating drop on unstructured grids. Therefore, the developed height function method is confirmed to be an efficient technique for the high-precision numerical simulation of interfacial flows on an unstructured grid.

The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

This paper discusses key issues and highlighted topics in thermal hydraulics in connection with Japan’s current sodium-cooled fast reactor development efforts, with particular focus on design study...

A conceptual design study of Japan sodium-cooled fast reactor (JSFR) is in progress as the “Fast Reactor Cycle Technology Development (FaCT)” project in Japan. A selector-valve type failed fuel detection and location (FFDL) system is applied to the JSFR design that has an upper internal structure (UIS) with a slit above the core and several sampling nozzles for the FFDL are set in the UIS around the slit to detect the fission product (FP) from the subassemblies below the slit. Therefore, mixing process in the UIS of complicated geometry should be known and appropriate arrangement of the sampling nozzles in the UIS is needed. A water experiment using a 1/5-scale model was carried out to investigate the mixing process in the UIS and concentration distribution of FP simulant. Experimental results showed that the sampling nozzles set in the UIS detected the FP simulant concentration within the criteria of FFDL signal detection, even in case of failed subassemblies under the UIS slit. In addition, identificati...

Liquid sodium is used as the coolant in fast reactors (FRs). It is a superior thermal medium, but its chemical reactivity with oxygen and water is very high. After the Monju sodium leak accident, a better physical and chemical understanding of the reaction phenomena was required from the viewpoint of the public acceptance of FRs. The purpose of this study is to understand the oxidation behavior of liquid sodium in detail, because it plays an important role in the continuation of the combustion process. The liquid sodium surface is covered with a thin oxide layer at the start of combustion. Generally, the surface oxide layer on the solid metal prevents further oxidation; however, dendritic sodium oxide is generated, and it grows on the oxide layer. This means that sodium can be supplied to the interface of the reaction without being blocked by the reaction products. As a result, the reaction continues spontaneously. Therefore, reaction products such as dendritic oxide have a significant influence on the ov...

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan sodium-cooled fast reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.

The cover gas entrainment at the free surface of sodium coolant becomes one of the significant issues according to the compact sizing of reactor vessel in the latest reactor design. In the present study, some basic experiments for the gas entrainment due to the surface vortex were performed in order to obtain the fundamental knowledge about the entrained bubble size. Distributions of entrained bubble diameters in several experimental conditions were obtained from bubble images using an image processing technique. Velocity fields around vortices and surface dimple shapes (gas cores) due to surface vortices were measured to grasp those influences on bubble shapes. The result showed that mean equivalent diameters of bubbles were varied from 1.3 to 2.1 mm in the range of present experimental conditions. The bubble sizes were influenced by the thickness of gas core.

In the design of Japan Sodium-cooled Fast Reactor (JSFR), coolant velocity is beyond 9 m/s in the primary hot leg pipe of 1.27 m diameter. The Reynolds number in the piping reaches 4.2 × 10 7 . Moreover, a short-elbow is adopted in the hot leg pipe in order to achieve compact plant layout and to reduce plant construction cost. Therefore, the flow-induced vibration (FIV) arising from the piping geometry may occur in the short-elbow pipe. The FIV is due to the excitation source which is caused by the pressure fluctuation in the pipe. The pressure fluctuation in the pipe is closely related with the velocity fluctuation. As the first step of clarification of the FIV mechanism, it is important to grasp the mechanism of flow fluctuation in the elbow. In this study, water experiments with two types of elbows with different curvature ratios were conducted in order to investigate the interaction between flow separation and the secondary flow due to the elbow curvature. The experiments were conducted with the short-elbow and the long-elbow under Re = 1.8 × 10 5 and 5.4 × 10 5 conditions. The velocity fields in the elbows were measured using a high-speed Particle Image Velocimetry (PIV). The time-series of axial velocity fields and the cross-section velocity fields obtained by the high-speed PIV measurements revealed the unsteady and complex flow structure in the elbow. The flow separation always occurred in the short-elbow while the flow separation occurred intermittently in the long-elbow case. The circumferential secondary flows in clockwise and counterclockwise directions flowed forward downstream of reattachment point alternately in both elbows.

Natural circulation plays a significant role in the decay heat removal function of a sodium-cooled reactor. A recent design of the Japan Sodium-Cooled Fast Reactor (JSFR) fully uses natural circulation for a decay heat removal system (DHRS). A dipped heat exchanger (DHX) is immersed in the reactor upper plenum as the DHRS. The DHX provides cold sodium in the upper plenum during the decay heat removal operation. This cold sodium covers the top of the core under the low-flow-rate conditions of natural circulation. Several water experiments of natural circulation in fast reactors revealed that the cold fluid in the reactor upper plenum might partially and temporally penetrate into the low power core channels, e.g., the radial blanket fuel subassemblies. Sodium experiments were carried out to find the onset conditions and the penetration depth of such partial reverse flow driven by buoyancy force. A blanket subassembly and the upper plenum were modeled in the test section including the axial upper neu...

In a sodium-cooled fast reactor (SFR), inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The sources of the gas bubbles are entrainment and dissolution of the reactor cover gas (argon) at the vessel free surface and emission of the helium gas that is produced as a result of disintegration of B4C control rod material. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, it is a key issue from the design and safety viewpoint and the allowance level is necessary regarding the gas entrainment at the free surface and the gas bubble concentration in the primary system. In the present study, a gas entrainment allowance level at the free surface is discussed and rationalized for the Japanese SFR (JSFR) design. The influence of the gas entrainment is evaluated using the void fraction at the core inlet. Design criteria for the acceptable level of the gas entrainment and gas concentration are proposed in consideration of the background level of gasses in the coolant. For the purpose, a plant dynamics code VIBUL has been developed to apply to the JSFR design to evaluate the concentration distribution of the dissolved gas and the free gas bubble in the JSFR system. Using the plant dynamics code for the bubble behavior, the background level of the free gas (void fraction at the core inlet) has been obtained. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment rate and the entrained bubble radius. Furthermore, the possibility of bubble removal and design requirement of the device is investigated to satisfy the allowance level. It is noted that the background level is already very low in comparison with the induced void reactivity by the void passing the reactor core.

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the “Fast Reactor Cycle Technology Development (FaCT)” project in Japan. JSFR adopts a selector-valve mechanism for a failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, appropriate sampling method for indentifying under-slit failed fuel subassemblies has been developed by water experiments.

Temperature fluctuation in fluid causes high cycle thermal fatigue in structure materials according to temperature distributions and time variations. A mixing tee is one of typical geometries where temperature fluctuation occurs. In the nuclear reactors and general plants, an elbow is often used near the mixing tee and it brings biased velocity distribution and also the secondary flow. In this study, influences of upstream elbow in the main pipe were studied in a water experiment of mixing tee with the elbow. Temperature distribution in the mixing tee was measured by a movable thermocouple tree and velocity field was measured by a high speed PIV. The temperature fluctuation at the mixing tee with the upstream elbow had the large component at low frequency in comparison with the straight case. The effect of the upstream elbow is significant to evaluate the high cycle thermal fatigue in the mixing tee because of larger importance of low frequency fluctuation on the point of fluid temperature-stress conversion.

An innovative sodium-cooled fast reactor has been investigated as part of the fast reactor cycle technology development project (FaCT). Thermal stratification after a scram is one of the main thermal loads of a reactor vessel (R/V). R/V has an upper inner structure (UIS), which consists of perforated horizontal plates and control rod guide tubes, and has a slit in the radial direction for fuel handling. The UIS slit causes an asymmetric flow pattern in R/V. A water experiment using a 1/10-scale model was carried out. A steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. This means that the jet through the UIS slit entrains the bottom of the stratification interface. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at the dipped plates where a fuel handling machine was inserted during a fuel exchange operation, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 21% smaller than that in the case of the higher plug position. This reduction of the temperature gradient was sufficient to maintain the structural integrity of the R/V wall against the thermal stratification.

Water experiments were carried out for thermal hydraulic aspects of thermal striping in a mixing tee, which has main to branch diameter ratio of 3. Detailed temperature and velocity fields were measured by a movable thermocouple tree and particle image velocimetry. Flow patterns in the tee were classified into three groups; wall jet, deflecting jet, and impinging jet, which had their own temperature fluctuation profiles, depending on a momentum ratio between the main and branch pipes. Non-dimensional power spectrum density (PSD) of temperature fluctuation showed a unique profile, when the momentum ratio was identical. Numerical simulation based on finite difference method showed alternative vortex development, like Karman vortex series, behind the jet from the branch pipe in the wall jet case. The prominent frequency of the temperature fluctuation in the calculation was 0.2 of St number based on the branch pipe diameter and in good agreement with the experimental results. Mixing behavior in the tee was characterized by the relatively large vortex structures defined by the diameters and the velocities in the pipes.

An innovative sodium-cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems associated with this design is gas entrainment at the free surface in the R/V. Horizontal plates are set below the free surface in order to prevent the gas entrainment. A water experiment was performed using a partial model built to 1/1.8 scale. The objective was to investigate the occurrence of gas entrainment under certain conditions and the mechanism of the gas entrainment. It was found that there were two types of gas entrainment phenomenon, and the conditions for their occurrence were far different from the rated condition in the reactor. One type of gas entrainment occurred at the wake region of flow around the cold leg pipe created due to the larger horizontal velocity in the R/V. The other type of gas entrainment occurred at the region between the hot leg pipe and the R/V wall when the coolant level was low and the downward velocity was large. The mechanisms of the gas entrainment at the two regions were clarified by the detailed measurement of transient flow velocity field.

The WATLON water experiment about fluid mixing in a tee pipe is calculated with a finite element volume method and a large eddy simulation (LES) approach, with the TRIO-U code. Its unstructured tet...

A compact sodium-cooled reactor is an important candidate as a fast breeder reactor (FBR) and has been investigated in the feasibility study of FBR cycle. Due to the compact sizing of the reactor vessel, gas entrainment at the free surface of sodium coolant becomes one of the significant issues for reactor design, and it is required to clarify the criterion of gas entrainment at free surface and the tolerance. In the present study, some visualization experiments were performed in a water-air system focusing on the gas entrainment due to surface vortex and its transient phenomena. Influences of horizontal velocity were clarified by the visualization. The gas entrainment due to the surface vortex occurs intermittently. Time trends of circulation and length of gas core for the intermittent surface vortices were measured by the particle image velocimetry and visualization. It was found that the gas core length extends with time delay to the increase of circulation around the vortex.

A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing causes thermal fatigue in structural components, is of importance for integrity of nuclear reactors and also general plants. Sodium cooled fast reactor had also several incidents of coolant leakage due to the high cycle thermal fatigue. A sodium experiment of parallel triple-jet configuration was performed to evaluate transfer characteristics of temperature fluctuation from fluid to structure. The non-stationary heat transfer characteristics could be represented by a heat transfer coefficient, which was constant in time and independent of the frequency of temperature fluctuation.

An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

In the design study for Japan Sodium-Cooled Fast Reactor (JSFR), fully natural circulation system is adopted as the decay heat removal system. We have been developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions. Since the method is currently based on conservative assumptions and data, there is room for further rationalization of the safety margin which can be achieved by conducting best estimate analyses with confidence and with quantified uncertainty of results. This paper describes a development of PIRT (Phenomena Identification and Ranking Table) for JSFR under natural circulation decay heat removal operations and the sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the core hot spot evaluation method.

A quantitative evaluation on the thermal striping, in which temperature fluctuation due to convective mixing imposes thermal fatigue on structures, is of importance for reactor safety. As for convective mixing, a water experiment was performed on vertical, parallel triple-jet : a cold jet at the center and hot jets in both sides. Three kinds of calculations based on the finite difference method for the experiment were carried out. Two types of turbulence models were used in the calculations, namely k-e two-equation turbulence model (k-e Model) and low Reynolds number turbulence stress and heat flux equation models (LRSFM). Furthermore, a quasi-direct numerical simulation (DNS) was performed. The DNS could simulate the time-averaged temperature field. The prominent frequency in temperature fluctuation obtained by the LRSFM was in good agreement with that in the experiment. The profile of power spectrum density of temperature fluctuations calculated by the DNS was close to the experimental results.

Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer.

This study was intended to examine sodium entrainment behavior in the case that a hole was formed on a tube wall in the steam generator of a fast breeder reactor and high pressure and high temperature water jetted out into sodium. Flow visualization experiments of an air jet in liquid were performed. The test vessel was 270 mm wide, 5 mm depth and 300 mm high. The air jet was blown vertically upward into stagnant liquid in the test vessel from a rectangular cross-section nozzle of 1 mm wide, 5 mm depth and 20 mm long which was located at the bottom of the test vessel. A flow state of the jet in the liquid was recorded with a high speed video camera at the fastest 150,000 frame/s. The test liquid was water and kerosene. Filament-like ears and wisps pulled out from the wavy interface were noticed on the interface between liquid and the air jet. The ears and the wisps were broken off and entrained into the air jet. The droplets broke up to small entrainments. This process seemed quite similar to the entrainment process in the annular dispersed flow in a pipe. Entrainment was initiated at a little bit downstream from the nozzle outlet. The entrainment inception point moved downstream as the air jet velocity increased. Axial directional entrainment velocity increased as the air jet velocity increased and the entrainment proceeded downstream. Transversal directional entrainment velocity was much slower than the axial directional entrainment velocity. The variation of the entrainment velocity in the transversal direction was not so prominent. The entrainments produced at the interface of the air jet moved to gather at the center portion of the air jet as those were accelerated.Copyright © 2014 by ASME

A sodium-cooled fast breeder reactor is now at the developing stage in Japan. One concern for safety is cover gas entrainment into the sodium coolant. The gas entrainment rate into liquid by the vortex formed on the free surface was examined experimentally. Four kinds of test fluid were used; water at 25 °C, water at 60 °C, 20 cSt silicone oil and kerosene. Gas was air. The flow state of gas entrainment was visually observed by using a high speed video camera. The gas entrainment rate into liquid was measured. Following conclusions were obtained. When exit velocity was low, bubble-type gas entrainment occurred. As the exit velocity increased, the gas entrainment type turned from the bubble type to a vortex type and gas entrainment rate considerably increased. The relation between gas entrainment rate and liquid velocity was mainly affected by the viscosity of liquid. As viscosity became large, higher exit velocity was required to get the same gas entrainment rate. The effect of surface tension on the gas entertainment rate was minor or little. No systematic trend by the surface tension was noticed in the gas entrainment rate. Present results of the onset of the bubble type gas entrainment are well expressed with both the modified Baum and the modified Takahashi et al. correlation although the modified Takahashi et al. correlation provides a little better prediction than the modified Baum correlation. A flow state at the outlet piping has significant effect on the gas entrainment rate. The dimension of the outlet piping may become important to consider the gas entrainment rate in the vortex type region.Copyright © 2014 by ASME

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 4.2×107. Moreover, a short-elbow which has Rc/D = 1.0 (Rc: 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 2D 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.