Hot Leg Pipe Research Articles

Experimental investigation of countercurrent flow limitation (CCFL) in a 1/19 scale of VVER-1000 geometry

Countercurrent flow limitation (CCFL) has been experimentally considered in a VVER-1000 hot-leg. The proposed test facility consists of a hot-leg pipe, a reactor vessel simulator (RVs), and a steam generator simulator (SGs). It was designed and constructed based on power-to-volume scaling approach by preserving the Froud number in the hot leg. The scaling ratio was considered 1/19 of the Bushehr NPP (VVER-1000) geometry. The hot leg of the proposed facility provides good optical observation of the gas/liquid interface. The SG simulator has also a side view for optical observation. Experimental results were captured at atmospheric pressure with air/water as fluid. CCFL was obtained by analyzing the pressure differences between RV and SG simulators.This CCFL data was compared with empirical data and correlations available in the literature. With increasing liquid superficial velocity, the gas superficial velocity curve is continuously decreased, which is consistent with Navarro's results. In this study, an experimental correlation is proposed and compared to other correlations. It was very close to the Navarro and Richter correlations.The results of experimental investigations clarified the onset of the CCFL phenomenon in a Bushehr NPP (VVER-1000) hot-leg geometry. Moreover, the achieved patterns and behaviors of the air/water mixture would be used in computational fluid dynamics (CFD) code validations.

Design of type 316L stainless steel 700 °C high-temperature piping

High-temperature design evaluations were conducted on Type 316L stainless steel piping for a 700 °C large-capacity thermal energy storage verification test loop (TESET) under construction at KAERI. The hot leg piping with sodium coolant at 700 °C connects the main components of the loop heater, hot storage tank, and air-to-sodium heat exchanger. Currently, the design rules of ASME B31.1 and RCC-MRx provide design procedures for high-temperature piping in the creep range for Type 316L stainless steel. However, the design material properties around 700 °C are not available in those rules. Therefore, a number of material tests, including creep tests at various temperatures, were conducted to determine the insufficient material properties and relevant design coefficients so that high-temperature design on the 700 °C piping may be possible. It was shown that Type 316L stainless steel can be used in a 700 °C high-temperature piping system of Generation IV reactor systems or a renewable energy systems, such as thermal energy storage systems, for a limited operation time.

Research on the on-line fracture assessment of cracked primary loop pipes in NPPs

The cracked pipes usually need to be replaced or conservatively evaluated in the Nuclear Power Plants (NPPs). There is still a lack of effective methods for a real-time crack propagation and safety evaluation based on massive monitoring transient data. In order to ensure the safe and economic operation of the NPPs, the authors have developed an on-line fracture assessment technology in this paper. The calculation method of total stress suitable for on-line monitoring data, optimizing the calculation method of transient thermal stress and stress intensity factor (SIF) based on the influence function method are discussed. Then, the on-line software for the primary loop hot-leg pipe which is based on the research results is developed. In this paper, an on-line safety assessment method based on the monitoring data which can meet the requirements of the in-service inspection code (RSE-M Code) is given. The influence function is proposed to be established at a fixed time interval which will make the computational complexity reduced significantly. By the parallel calculation method and the characteristics of the algorithm, the origin algorithm along the time axis is optimized to an algorithm that can be calculated to obtain the target parameters at each time point within the specified attenuation time range of the influence function. Then the calculation efficiency is significantly improved. By compiling the data and calculation methods in the code, the calculation method of SIF suitable for on-line monitoring is obtained, and the reliability is also verified by the analyzed case. Based on the technical results obtained from the research and the requirements in the RSE-M Code, the re-development is carried out in the general commercial software platform. The software is developed, and it is demonstrated that it has sufficient analysis accuracy.

A new correlation for CCFL at full scale PWR excluding the Hutze effect in UPTF data

The effect of the Hutze (a secondary ECC injection pipe within the hot leg pipe of a Konvoi-type PWR) upon CCFL was experimentally investigated at COLLIDER test facility of the Technical University Munich (TUM). COLLIDER represents a PWR hot leg with a large dimeter of 190 mm and ~¼ scale. A Hutze pipe was built into COLLIDER’s hot leg according to a German Konvoi-type PWR reactor design data. The effect of the Hutze upon CCFL are shown through comparisons with previous investigations at COLLIDER without a Hutze. Visual observations using high-speed recordings are presented at different flow conditions: supercritical conditions at low air velocities, transition into subcritical flow, and the onset of CCFL with comparison against previous recording without a Hutze. Limits of: the onset of hydraulic jumps, onset of CCFL (flooding), and most importantly CCFL characteristics are also shown and compared. The Hutze was found to induce a large impact upon these limits: two hydraulic jumps are observed prior to the onset of CCFL instead of only one (in case without a Hutze), the limits of the onset of hydraulic jumps and onset of CCFL occur at much lower air velocities, and the hysteresis effect (difference between limits of flooding and deflooding) disappears. CCFL characteristics curve is also shifted toward lower values of gas velocities and less liquid flows into the reactor vessel at same air velocities in the case with a Hutze which means a less cooling potential during the reflux condensation mode. Ohnuki correction of the Hutze effect upon CCFL characteristics of 1988 is tested and it seems to work well at small diameter but it overestimates CCFL characteristics for COLLIDER data. The previous agreement between CCFL characteristics obtained from downscaled experiments with D > 50 mm without a Hutze (including COLLIDER data and other large scale facilities) with UPTF data (at full scale) is found to happen because of the Hutze effect which shifts the CCFL characteristics downwards in Wallis diagram and not because of the correct representation of occurring CCFL phenomena. The actual CCFL characteristics for the 1:1 scale without a Hutze are expected to be higher than those reported from data with Hutze at UPTF facility. This indicates that there is still a scale effect even at a large diameter channel, and that most previous downscaled experiments have underestimated CCFL characteristics at full scale. Based upon obtained results, a correlation of CCFL characteristics at full scale excluding the Hutze effect is given. This correlation is of a special interest for the design of new PWR reactors as most new designs does not include a Hutze.

Study on Inelastic Strain-Based Seismic Fragility Analysis for Nuclear Metal Components

The main purpose of this study is to investigate the feasibility of the seismic fragility analysis (FA) with the strain-based failure modes for the nuclear metal components retaining pressure boundary. Through this study, it is expected that we can find analytical ways to enhance the high confidence of low probability of failure (HCLPF) capacity potentially contained in the conservative seismic design criteria required for the nuclear metal components. Another goal is to investigate the feasibility of the seismic FA to be used as an alternative seismic design rule for beyond-design-basis earthquakes. To do this, the general procedures of the seismic FA using the inelastic seismic analysis for the nuclear metal components are investigated. Their procedures are described in detail by the exampled calculations for the surge line nozzles connecting hot leg piping and the pressurizer, known as one of the seismic fragile components in NSSS (Nuclear Steam Supply System). To define the seismic failure modes for the seismic FA, the seismic strain-based design criteria, with two seismic acceptance criteria against the ductile fracture failure mode and fatigue-induced failure mode, are used in order to reduce the conservatism contained in the conventional stress-based seismic design criteria. In the exampled calculation of the inelastic seismic strain response beyond an elastic regime, precise inelastic seismic analyses with Chaboche’s kinematic and Voce isotropic hardening material models are used. From the results of the seismic FA by the probabilistic approach for the exampled target component, it is confirmed that the approach of the strain-based seismic FA can extract the maximum seismic capacity of the nuclear metal components with more accurate inelastic seismic analysis minimizing the number of variables for the components.

Thermal Hydraulic Analysis of Double Ended Hot Leg Pipeline Break Accident in AP1000 Reactor

In the present research, the Loss of Coolant Accident (LOCA), as a postulated accident in nuclear reactors, in its worst condition called large LOCA has been modeled. A specific type of large LOCA in which a totally guillotine break occurs in the hot leg pipe is the Double Ended Hot Leg (DEHL). Once the ‘LOCA’ occurs, the pipe is break and the coolant is lost and as a result, the danger of core melting is expected. In this paper, the single volume method was used to perform the numerical modeling on AP1000 reactor, a pressurized water reactor, which is among the most complex safe reactors that have been made so far. A wide range of safety margins are provided by its safety systems. The passivity is one of the most important safety features of AP1000 by which several simplifications can be made so that safety, reliability, construction, operation, maintenance, investment, protection and plant costs can be enhanced. Analyzing the most dangerous accident in one of the most secure reactors can be considered a valuable and logical work. The MATLAB software was used to perform the numerical analysis. Finally, the results obtained from the analysis were compared with the reports on the safety, security and environmental of AP1000 reactor.

Application of two alternative shutdown severe accident management guideline (SSAMG) entry conditions for CPR1000

Abstract Recently, with the development and application of full-scope level 2 probabilistic safety assessment (PSA) method around the world, severe accident phenomena during shutdown and low power conditions have aroused extensive attention in nuclear industry. And the shutdown severe accident management guideline (SSAMG) is claimed to be developed, and the verification of the traditional and alternative entry conditions is the first consideration in this procedure. Thus in this paper, the feasibility of the hot leg pipe temperature and the modified jakob number are analyzed based on a SBO sequence firstly. Subsequently, verification work is conducted under a SBO sequence with pressurizer manhole open and a SBO sequence along with SBLOCA. The results proved the excellent effectiveness of the two parameters to be used as alternative SSAMG entry conditions. Also, a relational figure is constructed based on the results of diverse sequences with various primary system pressure to provide visualized guidance for operators. What's more, the value of modified jakob number which indicates the SSAMG entry is thought to be in the range of 0.5–1.

On CCFL at a PWR Hot-Leg Pipe Geometry and Comparison Between Results in COLLIDER 1/4th- and Kobe 1/15th-Scaled Facilities

Countercurrent flow limitation (CCFL) in a pressurized water reactor hot-leg pipe geometry with a 190-mm pipe diameter was investigated experimentally and numerically at the COLLIDER test facility of the Technical University Munich in the past 3 years. This paper summarizes the most important CCFL findings learned from the COLLIDER test facility and tries to explain the reasons for obtaining different descriptions, results, and conclusions at different CCFL experimental investigations. The factors that can affect CCFL experimental results are explained in detail including some scale effects. The necessary preconditions to compare two sets of data from different CCFL experimental investigations are discussed in detail. The difference among CCFL-related limits/curves is clarified taking data at the COLLIDER as an example. The limits included the limit of the transition from a supercritical into a subcritical flow (SSTL); the onset of CCFL limit (iCCFL) inside the hot-leg pipe; the onset of CCFL limit (eCCFL) at the entrance of the steam generator; the deflooding limit (CCFLd); the CCFL characteristics curve (CCFLch), which predicts the water delivery rate after the onset of iCCFL; and the onset of hysteresis limit. It will be shown that among these limits only SSTL, CCFLch, and eCCFL are original limits while the rest are derivatives of them. In particular, it will be shown that the iCCFL limit is a combination of the SSTL and CCFLch limits. The effect of scale upon the eCCFL’s mechanism (whether a water accumulation or droplet entrainment at the entrance to the steam generator) is clarified via a comparison to a 50-mm CCFL facility at Kobe University. This paper tests the scalability of interface distribution at quasi-stationary conditions (i.e., points along the CCFLch curve) via a comparison of time-averaged interface distributions obtained at similar inlet conditions ( at the COLLIDER 190-mm and Kobe 50-mm channels. The comparison will show that interface distributions (which are directly linked to the pressure drop and interfacial momentum transfer) cannot be scaled at the bend/riser/entrance region because of the influence of the channel diameter upon occurring CCFL mechanism. Meanwhile, the water level gradient can be similar at the horizontal part, but not the relative water depth.

Interfacial behavior of the air-water counter-current two-phase flow in a 1/30 scale-down of pressurized water reactor (PWR) hot leg

The study on the interfacial phenomena of flooding during the counter-current flow of gas-liquid in a complex pipe was carried-out experimentally. It consists of a combination of the horizontal pipe, elbow and inclined pipes to simulate the hot leg pipe in a PWR. The inner diameter was 25.4 mm as scale down of real hot leg pipe of German Konvoi type reactor. The visual observations were performed using high-speed video cameras. The measured parameters were the liquid film thickness and the pressure difference between reactor pressure vessel (RPV) simulator and steam generator (SG) simulator. The film thickness inside the test pipe was determined by using the processed image of visual data.As a result, three different mechanisms at the onset of flooding in counter-current were clarified, and are strongly affected by liquid superficial velocity. At low liquid superficial velocity (JL), the flooding mechanism is triggered by the presence of wavy interface along the horizontal pipes in the form of a liquid blockage on the elbow. At the medium JL, the slug was detected along the horizontal pipe. However, the slug collapses immediately since the water is insufficient to maintain the blockage. At high JL, the slug occurs and travels along the horizontal pipe until the elbow. Moreover, the effect of the length of horizontal pipe on the interfacial behavior of air water counter current flow also can be explained. As the length of the horizontal pipe decreases, the onset of flooding occurs at a higher gas liquid mass flow rate.

Droplets entrainment ratio in a PWR hot-leg pipe geometry

Droplets entrainment was found to contribute to the occurrence of countercurrent flow limitation (CCFL) at COLLIDER test facility which has a PWR hot-leg pipe geometry with a large diameter value of 190 mm. Previous investigations showed that droplets entrainment starts to become observable at medium water inlet velocities (In Wallis parameters Jw,in∗0.5>0.2). In current article, the occurrence of droplets entrainment at COLLIDER test facility is discussed along with a detailed descriptions of visual observations that corresponds to entrainment mechanisms at the junction between the hot-leg and the steam generator. These mechanisms are compared to typical mechanisms found in literature for annular flows. The significance of COLLIDER large-diameter and pipe geometry against rectangular and narrow channel geometry is explained. Experimental measurements of the droplet entrainment ratio at COLLIDER test facility (using air/water at atmospheric pressure) are introduced and discussed. Measurements cover a range of air superficial velocities between 10.6 and 15.2 m/s, and water superficial velocities between 0.05 and 0.12 m/s. A correlation of the critical air velocity at which the droplets entrainment start to occur is given. An empirical correlation of the entrainment ratio was developed for obtained data. The correlation shows a stronger dependency upon the water inlet velocity in comparison to correlations found in literature for annular flows. A comparison between the prediction of available entrainment ratio correlations in literature and COLLIDER experimental data is made. The comparison shows the need of a new correlation for the droplets entrainment that takes place at the entrance of a hot-leg pipe geometry. The obtained correlations are useful to predict the onset of entrainment and to quantify its possible contribution to the reduction of water delivery into the reactor core during the occurrence of an SBLOCA accident. The last result becomes of a special importance knowing that CCFL caused by droplets entrainment occurs at lower air velocities than those at which a typical CCFL occurs (caused by large roll-waves at the bend), and that droplets entrainment does not require a transition from a supercritical into a subcritical flow condition in order to take place.

Image-processing of time-averaged interface distributions representing CCFL characteristics in a large scale model of a PWR hot-leg pipe geometry

Countercurrent Flow Limitation (CCFL) was experimentally investigated in a 1/3.9 downscaled COLLIDER facility with a 190mm pipe’s diameter using air/water at 1 atmospheric pressure. Previous investigations provided knowledge over the onset of CCFL mechanisms. In current article, CCFL characteristics at the COLLIDER facility are measured and discussed along with time-averaged distributions of the air/water interface for a selected matrix of liquid/gas velocities. The article demonstrates the time-averaged interface as a useful method to identify CCFL characteristics at quasi-stationary flow conditions eliminating variations that appears in single images, and showing essential comparative flow features such as: the degree of restriction at the bend, the extension and the intensity of the two-phase mixing zones, and the average water level within the horizontal part and the steam generator. Consequently, making it possible to compare interface distributions obtained at different investigations. The distributions are also beneficial for CFD validations of CCFL as the instant chaotic gas/liquid interface is impossible to reproduce in CFD simulations. The current study shows that final CCFL characteristics curve (and the corresponding CCFL correlation) depends upon the covered measuring range of water delivery. It also shows that a hydraulic diameter should be sufficiently larger than 50mm in order to obtain CCFL characteristics comparable to the 1:1 scale data (namely the UPTF data). Finally, the study shows that the change of the flow condition inside the hot-leg is not only related to the water and air inlet velocities, but is also dependent upon the existent interface distribution within the hot-leg, and that several CCFL cases of identical inlet flow conditions can exist with different interface distribution and pressure difference. The last result is of a special importance to the investigation of this phenomenon during SBLOCA accidents, since the entire phenomenon is driven by pressure difference between the steam generator and reactor vessel, as well as by gravity. This result show also that CCFL characteristics cannot be investigated using 1D codes, as the interface distribution within the hot-leg during a SBLOCA accident will depend upon flow history or previous interface distribution. Current investigations support the effort to provide more knowledge over CCFL in order to extrapolate results obtained in downscaled models into the 1:1 scale.

The Welded Joint Strength Reduction Factors of Modified 9Cr–1Mo Steel for the Advanced Loop-Type Sodium Cooled Fast Reactor

Creep strength enhanced ferritic (CSEF) steels including ASME Gr.91 are widely used in fossil power plants. In the advanced loop-type sodium-cooled fast reactor (SFR), modified 9Cr–1Mo steel (ASME Gr.91) is going to be adopted as a structural material. Modified 9Cr–1Mo steel was registered in the Japan Society of Mechanical Engineers (JSME) code as a new structural material for SFRs in the year 2012. The creep-rupture curve of the base metal of this steel was standardized using region splitting analysis method. According to this method, creep-rupture data were divided into two regions, high-stress and low-stress regimes, and those regions were individually evaluated by regression analyses with the Larson–Miller parameter (LMP). The difference in the creep failure mechanisms between the high-stress and low-stress regions was considered in this method. The boundary between these regions was half of the 0.2% proof stress of the base metal at the corresponding temperature. In the modified 9Cr–1Mo steel welded joint, creep strength may markedly degrade, especially in the long-term region. This phenomenon is known as “type-IV” damage due to creep voids and cracks in the fine-grained heat-affected zone (HAZ). There is no precedent for indicating the obvious creep strength degradation of welded joints under SFR temperatures (550 °C or less). Although obvious strength degradation of the welded joints has not yet been observed at 550 °C, it is fair to assume that the strength degradation will occur due to very long-term creep. Therefore, considering strength degradation due to “type-IV” damage is necessary. This paper proposes the creep-rupture curve and the welded joint strength-reduction factor (WJSRF). The creep-rupture curve of the welded joint was proposed by employing a second-order polynomial equation with LMP using region splitting analysis method, which is used for the base metal as well. The WJSRFs were proposed on the basis of design creep-rupture stress strength. The resulting allowable stress was conservative compared with that prescribed in ASME code and the Japan domestic regulation for thermal plants. In addition, the design of the hot-leg pipe in SFR was reviewed considering the WJSRFs.

Experimental investigation of countercurrent flow limitation (CCFL) in a large-diameter hot-leg geometry: A detailed description of CCFL mechanisms, flow patterns and high-quality HSC imaging of the interfacial structure in a 1/3.9 scale of PWR geometry

Countercurrent flow limitation (CCFL) is experimentally investigated in large-diameter hot-leg geometry. COLLIDER test facility consists of reactor vessel simulator (RVs), hot-leg pipe, and a steam generator simulator (SGs) and was constructed to be a 1/3.9 scale model of a real PWR geometry. The facility is completely transparent providing excellent conditions for optical observation of the gas/liquid interface. Experiments were performed at atmospheric pressure using air/distilled water as flow mediums. High-quality high speed recording (HSC) was implemented to acquire the air/water interface which provided a unique and detailed look into the interface structure within the entire hot-leg geometry. This allowed a precise and a high-qualitative identification and description of the flow patterns and mechanisms by which the onset of CCFL occurs. A systematic experimental investigation of the onset of CCFL and deflooding was carried out for water inlet velocities in the range Jw,in*0.5=0.085−0.305, ΔJw,in*0.5=0.01. Air velocity was increased and decreased stepwise to identify the onset of CCFL and deflooding limits precisely. Two onset of CCFL mechanisms were identified and labeled as: bend-CCFL and ADE-CCFL. The results were arranged into four regions according to the inlet water velocity Jw,in*0.5. The four regions differentiate by sub-mechanisms that appear during bend-CCFL and ADE-CCFL. A detailed description of the successive events and mechanisms by which the onset of CCFL during the increase of air velocity, and deflooding occur is provided. The description is supported by high-quality images of the interface structure. The resulting experimental onset of CCFL curve and deflooding curve are provided and discussed. The experimental investigation clarifies the onset of CCFL phenomenon in a large-diameter hot-leg pipe geometry and provides valuable images of the air/water interface for future CFD validation. The current study shows that from a safety point of view, the onset of ADE-CCFL precedes the onset of bend-CCFL as Jw,in*0.5>0.185; its intensity and the amount of backflow water cannot be ignored for Jw,in*0.5>0.27.

Application of the Probabilistic Fatigue Evaluation of Flow-Induced Vibration for Hot-Leg Piping in Japan Sodium-Cooled Fast Reactor

The primary hot-leg piping system of the advanced sodium-cooled fast reactor under conceptual study in Japan (named Japan sodium-cooled fast reactor: JSFR) utilizes large-diameter and thin-walled pipes to ensure high coolant velocity, which inevitably leads to the occurrence of flow-induced vibration. Usually, the structural integrity of a piping system under flow-induced vibration is defined to be the maximum stress amplitude below the design fatigue limit. The present study tries to establish a reasonable methodology to estimate the high-cycle fatigue damage due to flow-induced vibration depending on its frequencies and the corresponding stress levels. An analytical procedure for probabilistic fatigue evaluation is developed and applied to the hot-leg piping system. The reasonability of the newly proposed methodology is confirmed from a test simulation.

Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

There have been incidents recently where stress corrosion cracking (SCC) observed in the dissimilar metal weld (DMW) joints connecting the reactor pressure vessel (RPV) nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR). Tensile residual stress (RS), in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM), an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

302 ナトリウム冷却高速炉の高温側1次主冷却系統合解析モデルの整備(OS3-(1),OS3 オーガナイズドセッション《熱・流体・エネルギー変換工学の進展》)

Numerical simulation method to analyze unsteady thermal-hydraulic phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) was developed in the previous studies. Since flow-induced vibration (FIV) in the hot-leg piping with the short elbow (curvature radius corresponds to diameter) is one of the targeted issues in this study, numerical estimation method for dynamic analysis of mechanical stress on piping has been developed successively. As the preliminary step in development of the fluid-structure mechanical interaction simulation method, time history response analysis of piping by using the time history data of fluid pressure obtained by the unsteady hydraulics simulation as its boundary condition was attempted. Through the numerical results, potential capability of the dynamic analysis of piping were confirmed and unsteady behavior of pipe due to the unsteady flow phenomena was analyzed.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

320 ナトリウム冷却高速炉のショートエルボ配管を有する炉上部プレナム部の非定常解析モデルの整備(OS9 オーガナイズドセッション《熱・流体・エネルギー変換工学の進展》)

Numerical simulation method to analyze unsteady thermal-hydraulic phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) has been developed. Flow-induced vibration (FIV) in the hot-leg piping with the short elbow (curvature radius corresponds to diameter) is mainly targeted in this study. Numerical simulation model of the hot-leg piping systems with the short elbow integrated into upper plenum full domain were constructed. Through the numerical simulation with the lull sector model, applicability of numerical models and mesh arrangement for transient calculation were confirmed and unsteady flow phenomena in the plenum and piping system were successfully analyzed.

310 ショートエルボ配管を有する高速炉の上部プレナム非定常流動解析手法の整備(OS3-3 オーガナイズドセッション《熱・流体・エネルギー変換工学の進展》)

Numerical simulation method to analyze unsteady flow phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) has been developed. As one of the most significant issues related to unsteady phenomena, a flow-induced vibration (FIV) in the hot-leg piping with the short elbow (curvature radius corresponds to diameter) is mainly targeted in this study. At the first step of the simulation model development, models of the hot-leg piping with the short elbow integrated into a symmetric half domain of the upper plenum were constructed. Through the numerical simulations in the half sector model, applicable numerical schemes and turbulence model were confirmed and requirements of mesh arrangement for the transient calculation for unsteady phenomena analysis were revealed.