bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Validation of CFD Models Capturing the Thermal-Hydraulics in Liquid Metal Cooled Reactor Plena
Authors: Ward Brendan , Hopkins Thomas , Bindra Hitesh 
Published: October 2020

Abstract
High fidelity velocity field experimental data in a liquid metal plenum is presented and compared with numerical simulations. While work has already been established for fluids like air and water, research on low Pr fluids (Pr ≪ 1) (e.g. liquid metals) has fewer experimental data sets with validation-quality data. Work in advanced reactors using liquid metal coolant requires validated numerical simulations for safety analyses. The Gallium Thermal-hydraulic Experiment (GaTE) facility is outfitted with acoustic backscattering measurement techniques to generate the high fidelity distributed flow field data in a liquid metal plenum (a 1/20th scale of the Department of Energy’s sodium cooled Advanced Burner Test Reactor design). The high spatial and temporal resolution of the sensors are required to capture the fluctuations of velocity to allow a more direct comparison to the numerical simulations. For these simulations the coupled mass and momentum equations under the large eddy simulation (LES) framework were solved with the wall-adapting local eddy-viscosity (WALE) model for sub-grid scale formulations. Since the temperature transients of interest for reactor safety have a period of about a minute in the GaTE system, there may not be enough time to allow statistical tools to check one-to-one correspondence. So the data collection period for both data sets was extended to allow convergence of the mean and a larger sample size for other statistics during system steady-state, isothermal tests. Two characteristic velocities of the plenum inlet barrel were investigated (U = 40, 60 mm/s; Re = 7,000, 11,000). Probability distributions show good agreement between experiment and simulation with the difference only in the low-probability tails that LES is not expected to simulate. The time averaged mean axial distribution of the vertical velocity also shows good agreement between the two setups.
Numerical Simulation of Structural Optimization for Inlet Pipe of Centrifugal Pump in CAP1400 Power Plant Auxiliary System
Authors: Qu Ting , Zhang Liang 
Published: October 2020

Abstract
The centrifugal pump is the key equipment for power conversion in the CAP1400 power plant. According to the equipment layout requirements in the nuclear power plant process system design manual, the pipe section at the entrance should be a straight pipe with a length at least five times the pipe diameter. However, in actual layout, due to factors such as plant space or modularity, the layout of the inlet pipes of some centrifugal pumps cannot meet this requirement, resulting in non-uniform incoming flow at the pump inlet, which affects the safe and stable operation of the pump and system. In this paper, two different structures of rectifiers are designed and numerically simulated for the elbow part of the inlet pipe section of the SFS system centrifugal pump in the CAP1400 power plant. Firstly, the variation of flow uniformity of the pump inlet pipe at different inlet speeds is studied. Then, the influence of different structure rectifiers on the uniformity of flow at the pump inlet under the same working conditions is studied. The results show that the addition of a rectifier in the elbow can effectively improve the flow uniformity of the pump inlet section and reduce the effect of uneven inlet flow on the pump performance.
A Review of Degradation Modeling of Key Components of Sensor Circuits Based on Physical Analysis
Authors: Zhu Yonlong , Huang Xiaojin , Zhou Shuqiao 
Published: October 2020

Abstract
Nuclear power plant (NPP) accidents can cause severe effects. In order to ensure the normal operation of NPP, instrument and control system (I&C) composed of multiple sensors plays an important role in it. The sensor consists of sensitive components and functional circuits. In this paper, temperature sensor commonly used in NPP, PT100, is used as an example, and a bottom-up based physical analysis method is used to review the degradation mechanism and physical model of key components in resistance temperature detector (RTD) circuits. Resistors, capacitors, and MOSFETs are the key components of sensor circuits. Thin film resistors are the most widely used resistors due to their good performance. The transformed Arrhenius’ equation is used to describe its degradation characteristics. For the most commonly used aluminum electrolytic capacitors, the model reviewed in this paper can accurately describe the changes in equivalent series resistance (ESR), and use this as a criterion for determining capacitor failure. MOS technology is widely used in analog circuits. We have summarized three physical degradation processes, channel hot carrier (CHC), negative bias temperature instability (NBTI), and time-dependent dielectric breakdown (TDDB), that affect MOSFET performance. The physical model of the key component degradation of the sensor circuit summarized in this paper provides a basis for the subsequent establishment of a circuit-level degradation model.
Investigation on Post Accident Heat Removal From Partial Core Relocation in Lower Plenum After CDA in SFRs: 3-D CFD Analysis
Authors: Jhade Vidhyasagar , Sharma Anil Kumar , Kasinathan N. 
Published: October 2020

Abstract
In the present article, authors have carried out a three-dimensional (3D) Computational Fluid Dynamics (CFD) analysis of turbulent natural convection heat transfer from relocated core debris, in typical Sodium cooled Fast Reactors (SFRs), following a Core Disruptive Accident (CDA). Full-Scale analysis of complete sodium pool, i.e., hot and cold pool including immersed decay heat exchanger, has been carried out. k–ω SST model is used for turbulence closure. The model is selected based on the validation exercise. Core catcher (CC) with multiple passive jets over the Heat Shield Plate (HSP) is considered for analysis. Earlier CFD analysis with the assumption of whole core relocation on CC gave a CC temperature higher than the allowable limit. Hence, in this study, the analysis of partial relocation of the core debris on the CC has been carried out. From this, the maximum extent of relocatable core debris on the CC, which conforms with the allowable criteria, has been observed. Therefore, we have investigated the cases where the percentage of core debris relocation varies from 30–100% on HSP and remaining in the original position. This configuration may influence the decay heat removal via the hot pool. Time-dependent decay heat sources are used. Isotherms and streamlines have been presented to understand heat transfer characteristics. It has been found that with the implementation of multi jets CC, debris settled on HSP does not cross the threshold sodium boiling (∼1200 K) temperature up to 70% debris relocated to HSP with single tray configuration. Heat source surface, which remains at the core and in direct contact with coolant (liquid sodium), reaches a maximum value ∼1031 K for the case where the two-third core is intact at the core region. For HSP, it has been found that the thermal design limit exceeds (∼923 K) when 50% of debris relocates to the lower plenum. The transient study shows that time to attain maximum temperature by debris and HSP is inversely proportional to the percentage of intact core.
Experimental Study on Transient Heat Transfer for Helium Gas Flowing in a Minichannel
Authors: Xu Feng , Liu Qiusheng , Kawaguchi Satoshi , Shibahara Makoto 
Published: October 2020

Abstract
The blanket modules of first wall need bear tremendous heat flux due to the very high temperature of plasma in the nuclear fusion reactor. Therefore, it is significant to clarify the knowledge of transient heat transfer process for helium gas flowing in the tubes installed in the blanket modules.
In this research, the transient heat transfer process of turbulent forced convection for helium gas flowing in a horizontal minichannel was experimentally investigated. The test tube made of platinum with the inner diameter of 1.8 mm, the wall thickness of 0.1 mm and the effective length of 90 mm was heated by a direct current from power source. The heat generation rate of the test tube, Q̇, was raised with an exponential function, Q̇ = Q0 exp(t/τ), where Q0 is the initial heat generation rate, t is time, and τ is e-folding time of heat generation rate. The heat generation rates of the test tube were controlled and measured by a heat input control system. The flow rates were adjusted by the bypass of gas loop and measured by the turbine flow meter. The experiment was conducted under the e-folding time of heat generation rate ranged from 40 ms to 15 s. Based on experimental data, it is obvious that the heat flux and temperature difference between surface temperature of test tube and bulk temperature of helium gas increased with the exponentially increasing of heat generation rate. At the same flow velocity, the heat transfer coefficients approached constant values when the e-folding time is longer than about 1 s (quasi-steady state), but increased with a decrease of e-folding time when the e-folding time is smaller than about 1 s (transient state). The heat transfer coefficients increased with the increase in flow velocities but showed less dependent on flow velocities at shorter e-folding time. Furthermore, the Nusselt number under quasi-steady and transient condition was affected by the Reynolds number and the Fourier number.
Deep Gouging Technique for Fuel Debris Retrieval in Fukushima Daiichi
Authors: Doyen Ioana , Chagnot Christophe , Journeau Christophe , Porcheron Emmanuel , Gelain Thomas , Roulet Damien 
Published: October 2020

Abstract
Laser cutting, a contactless remote controlled process, was assessed for nuclear decommissioning applications, in full through-thickness-cutting (emerging cutting) and deep gouging (non-emerging) conditions.
This work aims at presenting specific tools developed and implemented at CEA Saclay, to evaluate and to optimize new cutting strategies, in terms of robustness and reliability, in order to best adapt the laser cutting process to fuel debris retrieval specificities of Fukushima Daiichi reactors.
Several configurations for deep gouging have been tested in DELIA facility to reach more than 40 mm of cut-thickness in one pass with an 8kW laser power. First, the laser cutting process was assessed on fused cast zirconia blocks and then on 316L stainless steel and fuel debris simulant blocks. This paper presents the main results of the tests performed in representative conditions of “in site conditions” such as: edge or bulk start/stop cutting, complex surface (i.e. surface with a varying topology) cutting, stand-off variation, angle of cut, trajectories misalignment, and material heterogeneity. Moreover, the excellent performances achieved on 316L stainless steel and fuel debris simulants indicate that deep gouging applied to nuclear decommissioning is a promising technology.
Development and Application of Dose Assessment Methodology During IS-LOCA in BWR Plant
Authors: Iwanami Hironobu , Hashimoto Tomoharu , Tanaka Motoi , Tayama Ryuichi , Mizuno Satoshi , Uemura Takashi 
Published: October 2020

Abstract
A dose assessment methodology was developed for IS-LOCA (Interface System – Loss of Coolant Accident) event, which is one of the design basis accidents in BWR plants, and a dose assessment at site boundary was implemented for the accident scenario by applying the evaluation method. The main objective of the assessment is to review a conventional conservative dose assessment model in the accident scenario.
The conventional conservative model is a method due to considering all fission products (FP) in liquid migrate to airborne (i.e. 100%) when reactor water is discharged from piping break area, while the developed new methodology is a method to evaluate transition behavior of fission products precisely by using flash fraction (FF) from liquid to airborne and consider it.
As the result of dose assessment, it was confirmed that the calculated dose during IS-LOCA event was about one-fifteith reduced by adopting the new model from the conventional model. The content of this paper includes:
• Background and purpose that led to the dose assessment methodology development during IS-LOCA.
• An evaluation method by using discharged amount of reactor water at piping break area and flash fraction (FF) from liquid to airborne based on Regulatory guide 1.183 published by U.S. NRC.
• Another evaluation method that takes into account the fluctuation due to depressurization of reactor pressure per time step from the start of accident until completion of isolation at piping break area in addition to the above method.
• Comparison between dose results at site boundary based on these methods and conventional conservative model. This results will give a precise information for public and it will be useful for making emergency evacuation plan.
High-Energy X-Ray CT Measurement of Void Fraction Distribution Around Part Length Rods in a Rod Bundle at High Pressure and Temperatures
Authors: Arai Takahiro , Ui Atsushi , Furuya Masahiro , Okawa Riichiro , Iiyama Tsugumasa , Ueda Shota , Shirakawa Kenetsu , Kito Kazuaki 
Published: October 2020

Abstract
A boiling experiment was conducted to acquire a three-dimensional void fraction distribution in a rod bundle with part length rods. The test section was a 5 × 5 heated rod bundle that partially simulated a BWR rod bundle, and three PLRs were arranged together in the corner. The heated length of the full-length rod was 3.71 m, which is equal in length to an actual fuel rod in BWRs, whereas the heated length of the PLR was 1.85 m. The rod bundle exhibited an axially and radially uniform heat flux. Further, the local void fraction was quantified by normalizing the intensities in the CT images of the boiling two-phase flow with those obtained under the liquid-phase and gas-phase conditions. The cross-sectional void fraction distribution was acquired at six height levels. The experimental results exhibited the void distribution around PLRs with respect to a wide range of flow conditions, inlet temperatures, inlet flow rates, bundle thermal powers, and system pressures.
Theoretical Research on Two-Phase Flow Instability in Parallel Rectangular Channels Under Periodic Perturbation
Authors: Qian Libo , Deng Jian , Huang Tao , Cai Rong 
Published: October 2020

Abstract
A theoretical model for Density Wave Oscillations (DWOs) flow instability in parallel rectangular channels under periodic heaving motion is established with a lumped mathematical model based on homogenous hypothesis. The parallel rectangular channels comprise of the entrance section, the heating section, the riser section and the upper- and lower plenums, which guarantee the isobaric pressure drop condition between channels and the model consists of boiling channel model, pressure drop model, parallel channel model, additional pressure drop model generated by heaving motions, the constitutive and numerical models. The effect of periodic perturbation is introduced through additional pressure drop in the momentum equation. The model is validated with experimental data of a twin-rectangular-channel flow instability experiment under static condition. Then the flow instability in parallel-rectangular-channel system is studied under periodic perturbation and the margin of flow instability and the threshold power of the system under static condition is calculated as basis condition for comparison. The effect of the amplitude and period of perturbation is analyzed analytically and the results show that the amplitude and period of perturbation shows little effect on flow instability. While when the additional pressure difference introduced by heaving motion is comparable with that under static condition, the effect of amplitude becomes stronger. And the period of perturbation strongly effects the threshold power when it is identical to that of natural period of the system, which can be explained by resonance between the perturbation and the system. And this effect is even stronger when the asymmetric heating condition is introduced.
Preliminary Development on Thermal-Hydraulic Analysis Code for the Spent Fuel Rod Under the Condition of Spray Cooling
Authors: Chao Guo , Jian Deng , Rong Cai , Yugao Ma , Lili Liu , Yuhao Zhang , Siyu Lv 
Published: October 2020

Abstract
According to spray cooling characteristics for failure of the cooling system in the spent fuel pool, we establish a model for analyzing spray cooling of a single spent fuel rod and develop a code named SFSC which is used to calculate the thermal-hydraulic characteristics of film on the spent fuel rod under the condition of spray cooling. The results obtained from SFSC consist of film thickness, film temperature, film evaporation, fuel pellet temperature of spent fuel rod and cladding temperature of spent fuel rod at different height of spent fuel rod are shown in this paper. At the same time, we establish same model by RELAP and compare results calculated by RELAP with that calculated by SFSC. The good agreement of two results demonstrates that SFSC is effective. SFSC fills in the blanks of thermal-hydraulic calculation for the spent fuel rod under the condition of spray cooling.
Challenge to Investigation of Fuel Debris in RPV by an Advanced Super Dragon Articulated Robot Arm: Design and Prototyping of a Lightweight Super Long Reach Articulated Manipulator
Authors: Endo Gen , Takahashi Hideharu , Kikura Hiroshige 
Published: October 2020

Abstract
A long-reach articulated manipulator that can deploy various sensors in a large but confined workspace is urgently needed in decommissioning tasks at Fukushima Daiichi Nuclear Power Stations. The most critical problem involves managing the large gravitational torques acting on the manipulator’s joints. In previous studies, several prototype models were developed; however, they were extremely heavy and bulky. This paper presents fundamental ideas for creating a lightweight super long-reach articulated manipulator. First, a mechanical structure could be utilized to support gravity. Second, a vertical force could be generated to compensate for gravity. Third, a tendon-driven mechanism could be utilized because it permits the installation of heavy actuators on the base. The tendons can transmit the actuator power to each joint. Thus, the weight of the arm is significantly reduced. We have discussed the advantages and drawbacks of each recommendation based on the hardware prototyping.
Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact by Projectiles: Outline of Impact Test
Authors: Nishida Akemi , Kang Zuoyi , Okuda Yukihiko , Tsubota Haruji , Li Yinsheng 
Published: October 2020

Abstract
Studies on the local damage to reinforced concrete (RC) panels subjected to projectile impact have mainly focused on collisions that occur at an angle normal to the structure; thus, research on oblique impact is scarce. Due to this, we conducted research focusing on oblique impact to enable more realistic impact assessment of projectile collisions. To date, the validity of the analytical method has been confirmed by comparing the results with those of previous tests, and the local damage to RC panels that have collided with projectiles has been analytically investigated focusing on the impact angle. Therefore, this study aims to confirm the validity of the analysis method by conducting specific impact tests under various conditions, including the impact angle, by obtaining the relevant data. This paper outlines the test for the local damage to RC panels subjected to both normal and oblique impacts.
Analytical Study on Dynamic Response of Reinforced Concrete Structure With Internal Equipment Subjected to Projectile Impact
Authors: Okuda Yukihiko , Kang Zuoyi , Nishida Akemi , Tsubota Haruji , Li Yinsheng 
Published: October 2020

Abstract
In case of a projectile impact on a reactor building of a nuclear power plant, stress waves propagate from the impacted wall to the structure’s interior. It is important to assess the effect of dynamic responses generated by the projectile’s impact on internal equipment, because stress waves are likely to excite high-frequency vibrations of internal equipment. The OECD (Organization for Economic Co-operation and Development) / NEA (Nuclear Energy Agency) launched the IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Projectiles) benchmark project in order to assess the dynamic response of a nuclear facility to projectile impact, and the third phase of IRIS (IRIS 3) [1] contributes to the investigation of the dynamic responses of reinforced concrete (RC) structures that house internal equipment. We have participated in IRIS 3 and have performed calibration analyses of projectile impact tests on a structure that models a reactor building that houses internal equipment. Specifically, we have developed and validated a numerical approach to investigation of impact responses of an RC structure that houses internal equipment through calibration correction. This paper presents partial simulation results of the dynamic responses of this structure and discusses the effects of support conditions of the internal equipment and stress wave propagation.
Experimental Study on Aerosol Collection by Spray Droplets: Application to Fission Products Removal in Containment
Authors: Sun Haomin , Leblois Yohan , Gelain Thomas , Porcheron Emmanuel 
Published: October 2020

Abstract
During a severe accident of a pressurized water reactor (PWR), fission products (FPs) may be released and transported to the containment. The containment spray can be utilized to remove the aerosols of FPs, which contributes to retaining the FPs in the sump. Therefore, it is important to develop an analytical model for predicting the aerosol removal efficiency by the containment spray for nuclear safety. The containment spray is required to have a very high spray coverage ratio where the droplets are expected to impact the containment side walls. In such condition, the gas flow induced by the droplets will behave differently from that in the other condition without the droplet impaction where a stable gas circulation is expected between the boundary of the spray and the side walls. Since the aerosol removal efficiency also depends on the gas flow behavior, to develop the aerosol removal model applicable for the containment spray from this viewpoint, several aerosol removal experiments were carried out in the TOSQAN IRSN facility. Based on the droplet mass flux measurements, it was confirmed that the spray coverage ratios were comparable to that of the containment spray and many droplets impacted the side wall. According to the aerosol measurement results, it was found the aerosol removal was more effective in higher spray water flow rates. Since larger particles were removed more quickly, the mean particle diameter was decreasing during the spray operation and approached to a value. Based on a CFD simulation for the experiment, gas velocity field was investigated. The measured and calculated particle number concentration decays agreed well.