bluuebunny/crossref_metadata_2025 huggingface dataset:
A Beyond Design Basis Earthquake Study of Operating Nuclear Fuel Cycle Facilities
Authors: Liang Li , Rong Pan , Xiuyun Zhu , Guopeng Ren
Published: October 2021

After the Fukushima nuclear accident in Japan, the world attaches great importance to the study of beyond design basis earthquake. In 2016, China National Nuclear Safety Administration issued a new nuclear safety regulation, regulations for design safety of nuclear power plants (HAF102-2016), which requires that appropriate margin must be provided for new nuclear power plants in case of beyond design basis natural disasters. However, the existing nuclear fuel cycle facilities are also faced with the problem of beyond design basis earthquake. However, there is no relevant standard for seismic capacity evaluation of nuclear fuel cycle facilities beyond design basis earthquake in China. In this paper, the seismic safety evaluation standards and evaluation methods of nuclear facilities at home and abroad are investigated. Taking a nuclear fuel reprocessing facility as the research object, the dynamic response numerical analysis under beyond design basis earthquake is carried out. The seismic response is analyzed, and the weak link under the beyond design basis earthquake is found out, and reasonable and feasible seismic margin evaluation is proposed for the nuclear fuel cycle facility. At the same time, it provides technical support for weak links to take reinforcement measures.
Cause Analysis and Influence Evaluation of Cracks in Thick Slab Construction
Authors: Guixiang Yi , Liang Li
Published: October 2021

Cracks were found in the slab of a nuclear power plant when the formwork was removed. By means of ultrasonic testing, water storage test and crack width detection, the fracture distribution, depth and width characteristics are determined. On this basis, the numerical simulation analysis of hydration heat in the maintenance process is carried out to simulate the possible crack generation and distribution of the floor slab in the maintenance process. The simulation result shows that cracks under the combined action of temperature and shrinkage are consistent with the cracking characteristics of floor slab. Through numerical simulation of the cracked floor and intact floor, the change of the out of plane bearing capacity (bending and shear) of the cracked floor is compared and analyzed. In the model, considering the reduction of the bearing capacity at the crack section and the change of load transfer effect at the crack interface, the concrete model is disconnected according to the known opening situation, and the contact relationship is set on the crack interface to simulate the crack. Through comparative analysis, the mechanical properties of the cracked floor are evaluated, and the refined numerical simulation method of the working components with cracks is proposed. This paper can provide guidance for the cause analysis and influence evaluation of similar thick plate cracking phenomenon.
Mixing Process of Two Component Gases by Natural Convection and Molecular Diffusion
Authors: Ube Takeaki , Takeda Tetsuaki
Published: October 2021

A depressurization accident involving the rupture of the primary cooling pipe of the Gas Turbine High Temperature Reactor 300 cogeneration (GTHTR300C), which is a very-high-temperature reactor, is a design-based accident. When the primary pipe connected horizontally to the side of the reactor pressure vessel of GTHTR300C ruptures, molecular diffusion and local natural convection facilitate gas mixing, in addition to air ingress by counter flow. Furthermore, it is expected that a natural circulation flow around the furnace will suddenly occur. To improve the safety of GTHTR300C, an experiment was conducted using an experimental apparatus simulating the flow path configuration of GTHTR300C to investigate the mixing process of a two-component gas of helium and air. The experimental apparatus consisted of a coaxial double cylinder and a coaxial horizontal double pipe. Ball valves were connected to a horizontal inner pipe and outer pipe, and the valves were opened to simulate damage to the main pipe. As a result, it was confirmed that a stable air and helium density stratification formed in the experimental apparatus, and then a natural circulation flow was generated around the inside of the reactor.
Research on Earthquake Acceleration Alarm of Nuclear Power Plant
Authors: Liang Li , Rong Pan , Guopeng Ren , Xiuyun Zhu
Published: October 2021

Almost all nuclear power plants in the world are equipped with seismic instrument system, especially the third generation nuclear power plants in China. When the ground motion measured by four time history accelerometers of containment foundation exceeds the preset threshold, the automatic shutdown trigger signal will be generated. However, from the seismic acceleration characteristics, isolated and prominent single high frequency will be generated the acceleration peak, which has no decisive effect on the seismic response, may cause false alarm, which has a certain impact on the smooth operation of nuclear power plant. According to the principle of three elements of ground motion, this paper puts forward a method that first selects the filtering frequency band which accords with the structural characteristics of nuclear power plants, then synthesizes the three axial acceleration time history, and finally selects the appropriate acceleration peak value for threshold alarm. The results show that the seismic acceleration results obtained by this method can well represent the actual magnitude of acceleration, and can solve the problem of false alarm due to the randomness of single seismic wave, and can be used for automatic reactor shutdown trigger signal of seismic acceleration.
Effects of Non-Condensable Gas on Characteristics of Natural Circulation Flow of Isolation Condenser
Authors: Takada Tetsuya , Yamamoto Yasunori , Ono Kosuke
Published: October 2021

An isolation condenser (IC) is a passive core cooling system in boiling water reactors. The cooling performance of IC is deteriorated when hydrogen generated in the core flows into the IC pipes. In this study, we conducted high pressure experiments using natural circulation loop with non-condensable gas injection, where helium was used to simulate hydrogen effect on the IC. The reaching distance of steam in the heat transfer tube was estimated by observing the region where nucleate boiling occurred on the outer surface of the heat transfer tube, and the heat transfer coefficient was estimated. The heat transfer coefficient hardly changed when helium was injected to the loop that indicates injected helium was not accumulated in the heat transfer tube. The system pressure at quasi-steady state increased with increasing amount of the injected helium. Since the differential pressure at the down comer section increased by helium injection, the injected helium may be accumulated in the section, leading to increment of the system pressure.
Recent Activities and New Challenges for the EUR Organization
Authors: Sorel Vincent
Published: October 2021

Since 1991, the European Utility Requirements (EUR) organization has been developing and promoting harmonized technical specifications for the new mid- and large-size LWR designs to be proposed by the vendors in Europe. The EUR Organization issues and, from time to time, modifies and updates a report entitled “the EUR Document”. It consists of a comprehensive set of requirements covering the whole Nuclear Power Plant (NPP) and encompasses all aspects (safety, performance, competitiveness) and all parts of a NPP (nuclear island and conventional island). The main EUR Organization’s results obtained during the last years are in the three following fields: (1) The latest revision, Revision E, of the EUR Document was issued in December 2016. It includes revised safety requirements taking into account the most recent European and international standards issued by WENRA, IAEA and IEC after the Fukushima-Daishi accident. (2) The assessment of new designs. The presentation briefly recalls the EUR design assessment objectives and process and the status of the different EUR assessment projects. (3) Preparation of evolution of the EUR Document towards integration of requirements for Small Modular Light Water Reactors (SMLWR). The paper concludes by an overview of the future activities of the EUR Organization.
Improved Wet Scavenging Schemes for Air Dispersion Modeling of Cs-137 in the Fukushima Accident
Authors: Zhuang Shuhan , Fang Sheng , Dong Xinwen
Published: October 2021

Wet scavenging process is critical for air dispersion modeling of Cs-137 in the Fukushima Daiichi Nuclear power plant (FDNPP) accident. Although intensively investigated, wet scavenging simulation is still subject to uncertainties caused by the biases in wet scavenging modeling and meteorological input. To reduce these uncertainties, the on-line coupled modeling feature of the Weather Research and Forecasting-Chemistry (WRF-Chem) model was utilized and both the in-cloud and below-cloud scavenging processes are considered. In this study, the in-cloud scheme Environ and below-cloud scheme Baklanov are combined with each other to form Environ-Bakla to simulate the wet deposition of Cs-137. The model is systematically compared with a previous WRF-Chem model with a single below-cloud scheme Baklanov, based on both the cumulative deposition and ambient concentration of Cs-137 based on the FDNPP accident observation. The results demonstrate that the in-cloud scavenging scheme substantially improves the cumulative deposition simulation in regions with light rain like Tochigi, Nakadori etc. With respect to the atmospheric concentration, the inclusion of in-cloud scavenging doesn’t necessarily improve the performances and the Environ-Bakla only shows fair performance under plume events with no rain or light rain.
Effect Analysis of Power Supply Topology on the Reliability of the Reactor Protection System
Authors: Zhang Haojing , Xiong Huasheng , Guo Chao , Li Duo , Huang Xiaojin
Published: October 2021

As a clean energy source, Low Temperature Reactor can be used for heating, desalination, industrial steam production, summer cooling, etc. It still has good application prospects at present. The reactor protection system is one of the important safety systems in Low Temperature Reactor, thus improving the reliability of the protection system can enhance the safety level and economic benefits of the reactors. The reliability of the protection system is directly related to its system structure, the quality of its components, and the power supply topology. The effect of the third factor is analyzed in this article.

In the aspect of system reliability analysis, the RBD method is widely used and has many public research results. Based on the block diagram of protection system, this article establishes the RBD that considers the operation failure. According to the foregoing RBD, the protection system reliability is calculated, the minimum cut sets that can cause the protection system failure are analyzed, the different importance measures of each device in the protection system is determined, and the occurrence probability of the minimum cut sets and the importance measures are sorted as a basis for subsequent optimization of power supply topology.

The Low Temperature Reactor protection system in this article is mainly composed of transducer, set-point comparison unit, logic coincidence unit, and safety trigger unit. The power supply of these units have many possible topological structures, which have different complexity and different reliability levels. Three improvement approaches are proposed for the power supply topological structure of transducer, set-point comparison unit and logic coincidence unit, the reliability of the optimized protection system is calculated and analyzed. The results show that by improving the power supply topology, the reliability of the protection system is improved without increasing the components’ quality and redundancy.
Experimental Study on Performance Improvement of HTR-10 Helium Purification System
Authors: Wang Fangfang , Wei Liqiang , Kang Tianyu , Hu Chuangguo , Xie Feng , Chen Xiaoming , Shi Lei
Published: October 2021

Based on the historical operation data analysis and experience feedback of the HTR-10 helium purification system, the performance improvement experiment scheme of the helium purification system equipment has been designed, and the simulation experiment bench with copper oxide bed and molecular sieve bed has been built. By sorting and analyzing the purification and regeneration experiment data of copper oxide particles (CuO-ZnO-Al2O3) and 5A molecular sieve, the following conclusions have been drawn: (1) The activity of copper oxide particles after initial activation is high and their performance decreases obviously after purification and regeneration. Electron microscope scanning shows that sintering deactivation phenomenon exists in the application of CuO, and its regeneration times and service life should be concerned; (2) The operation mode of molecular sieve bed in series is more favorable to the purification of impurity gas, and moderate high temperature and large flow sweep are beneficial to the regeneration and desorption of molecular sieve. The optimized process scheme can further improve the performance of HTR-10 helium purification equipment, and reduce the content of helium impurities in the primary circuit in future, which can provide a certain reference significance to achieve the ultra-high temperature operation of HTR-10.
Research on DTS Analysis Method for 1000MWe PWR NPP
Authors: Zhang Yubin
Published: October 2021

The design transients (DTS) are intended to be used to evaluate component stress analysis, which bound the plant operation over the design plant life, and will be as a reference document regarding plant usage factor surveillance. The conditions of DTS represent bounding cases for plant events that are expected to occur, or that may occur, during the plant lifetime. For different operating conditions, DTS analysis methods would be different.

This report describes the DTS analysis method to be used when designing the major reactor coolant system components of the 1000MWe PWR plant. Some of components will be considered: Reactor Pressure Vessel, Main Coolant Lines, Reactor Coolant Pumps, Pressurizer and Steam Generators. Generally, variations in fluid pressure, fluid temperature and flowrate are used to represent the parameters for DTS. This report describes the method of DTS analysis and shows the evolution process of relevant thermal hydraulic parameters.
High Flux Reactor Review and Reactivity Control Analysis
Authors: Lin Wang , Wei Xu , Fei Xie
Published: October 2021

For over 60 years, research reactors have provided the world with a versatile tool to test materials and promote irradiation research, as well as to produce radioisotopes for medical treatments. The High Flux Reactor (HFR), as a water moderated and cooled, beryllium-reflected reactor has awarded more attention in recent years. There is a wide range of designs and applications for HFRs that based on their own situation to meet research requirements. For the purpose of reducing the volume and mass of the reactor, as well as ensuring the safety operation, it is necessary to determine the most effective reactivity control scheme, and analyze the corresponding reactivity insertion accidents. This paper is going to investigate typical high flux reactors within the same type with HFETR, summarize general description and characteristics, the uses of the high flux reactor, and reactivity control mechanisms. In addition, the associated reactivity insertion accidents were presented and analyzed. The analysis result will provide some references to further design and construction of high flux reactor.
Experimental Study on Flow Pattern of 10 mm Vertical Pipe
Authors: Xu Wangtao , He Qingche , Yan Meiyue , Zhu Wen , Zhang Luteng , Wu Dan , Huang Tao , Ma Zaiyong , Sun Wan , Pan Liangming
Published: October 2021

As a common heat exchange structure, vertical pipe is widely used in neaclear reactors and other systems. Flow pattern and flow regime map are the basis of thermal-hydraulic calculation. In this paper, the experiment of air-water two-phase flow pattern was carried out based on the visual test section of 10 mm vertical pipe. Three impedance meters were used to measure the void fraction at three axial positions (L/D = 66, 148, and 229) along the vertical pipe. Compared with the flow pattern criterion proposed by Mishima and Ishii, six superficial liquid velocities are selected for design conditions, which are 0.5 m/s, 1.0 m/s, 1.5 m/s, 2.0 m/s, 2.5 m/s and 3.0 m/s, respectively. And the superficial gas velocities are ranging from 0 to 30 m/s. Considering the inlet and export effects of pipe flow, the void fraction parameters from L/D = 148 (#2 impedance meter) were applied for analysis. Based on the experimental data and random forest algorithm, the flow pattern was identified for each condition, and the flow regime map of 10 mm vertical pipe was established. It is concluded that the flow pattern transition point of air-water flow in the 10 mm vertical pipe was different from that of the Mishima-Ishii’s model, and the inner diameter of the pipe will affect the flow pattern formation and flow pattern transition.
Research on Dimensionless Analysis Method of Scale Effects for Molten Pool Experiments
Authors: Quan Fengyang , Li Wei , Zhao Zikun , Zeng Xiao , Guo Yong , Yuan Yidan
Published: October 2021

In a severe accident of light water reactors, the loss of coolant and heat sink can cause core degradation. The molten fuel pellet (UO2), molten cladding material (Zr-ZrO2), and molten structural materials (stainless steel) form molten corium, and finally relocate in the lower head of the reactor pressure vessel (RPV). In case of coolant loss, the corium molten pool form in the lower plenum which will threaten the integrity of RPV. To prevent the RPV failure, the ex-reactor vessel cooling (ERVC) system, which floods the RPV with cooling water, is applied to ensure the survivability of RPV and eventually achieve the goal of fission product retention. Therefore, the molten pool’s behavior is well studied by a large number of experiments, such as BALI, LIVE-2D, COPRA. Natural heat convection, which is related to the latent heat of fission products and the cooling of ex-vessel, will induce heat flux alone the RPV lower plenum. The peak value of heat flux is estimated at around 70°. However, there is still uncertainty to predict the Nuup/Nudn ratio especially when the condition of the prototype molten pool (1017) to be extrapolated. This research aims at analyzing the scale effects based on Buckinhum’s PI theory. The molten pool phenomenon related to dimensionless parameters is firstly calculated. After rigorous calculation, the influence of each dimensionless parameter is evaluated according to the behavior of the corium pool. Finally, the experimental results of BALI and COPRA are obtained, in order to evaluate the uncertainty between the chosen simulant from those two experiments.
A Preliminary Evaluation of the Computational Fluid Dynamics Capabilities in MOOSE
Authors: Weiss Abdullah G. , Abdoelatef M. Gomaa , Bani Ahmad Mohammad T. H. , Ahmed Karim , Kimber Mark L.
Published: October 2021

The Multiphysics Object Oriented Simulation Environment (MOOSE) is a code package that couples a variety of physics modules, allowing for highly accessible multiphysics simulations. The physics modules include a finite-element Navier-Stokes (N-S) module that is designed to solve laminar fluid dynamics problems. The usage of this module in multiple recent studies coupled with the growing interest in MOOSE for usage in non-LWR safety studies by the Nuclear Regulatory Commission (NRC) prompted the authors to investigate the computational fluid dynamics (CFD) capabilities of MOOSE. A 2D flow past a round cylinder scenario is simulated in the MOOSE environment to investigate the effectiveness of the N-S module. The simulation assumed an unsteady laminar flow with a Reynolds number of 200, where the unsteadiness in the vortex street is captured using a Euler-implicit time-integration scheme. To verify the results from MOOSE, a similar simulation is conducted using the well-utilized OpenFOAM finite volume code, and the velocity profiles of both transient simulations are compared. Grid and input sensitivity studies are conducted in MOOSE to quantify the numerical and input errors, respectively. The numerical and input errors are also visualized with contour plots in order to qualitatively understand the evolution of the errors across time and space. Results show that MOOSE struggles with capturing the instabilities in the flow, as it takes more than 20 seconds before a vortex street is observed. Excellent agreement in the velocities is seen between MOOSE and OpenFOAM upstream of the cylinder, but huge discrepancies can be seen where instabilities are expected around the cylinder and upstream of it. The discrepancies become more severe as the transient simulations progress in time. The numerical error is most prominent in the wake of the cylinder and around the boundaries of the channel. The input errors are also prominent in the wake but are most prominent around the cylinder walls. The results demonstrate the difficulties faced by the N-S module in MOOSE when it comes to simulating unsteady flow and prompts questions about the effectiveness of finite element CFD solvers.
Numerical Calculations of the Effective Thermal Conductivity of the Dispersion Fuel Sphere With the Internal Heat Sources
Authors: Ziping Liu , Yu Ji , Han Zhang , Jun Sun
Published: October 2021

The effective thermal conductivity of dispersion fuel sphere is an important parameter for its’ temperature prediction. Whether the existing effective thermal conductivity models which does not consider the impact of inner heat source can predict the average temperature of heat generating dispersion fuel sphere is unknown. This study established the three-dimensional heat conduction model to calculate the temperature distribution of dispersion fuel sphere with inner heat source, then the effective thermal conductivity for average temperature prediction was calculated by conduction differential equation. The fuel particles are set to be uniformed with uniform inner heat source, fuel particles are defined uniformly distributed in probability in the fuel sphere. Through the analyses, the existing models overestimate the effective thermal conductivity of dispersion fuel sphere with inner heat source, and then underpredict the average temperature of dispersion fuel sphere. The distribution of inner heat source will affect the effective thermal conductivity. The values of inner heat source will not affect the effective thermal conductivity. This paper revealed the importance of inner heat source on effective thermal conductivity, and lay the foundation for further study.
Research on Optimization and Verification Method of Sensor Arrangement in the Chemical and Volume Control System
Authors: Zhou Gui , Wang Hang , Peng Minjun
Published: October 2021

In order to avoid the nuclear accidents during the operation of nuclear power plants, it is necessary to always monitor the status of relevant facilities and equipment. The premise of condition monitoring is that the sensor can provide sufficient and accurate operating parameters. Therefore, the sensor arrangement must be rationalized.

As one of the nuclear auxiliary systems, the chemical and volume control system plays an important role in ensuring the safe operation of nuclear power plants. There are plenty of sensor measuring points arranged in the chemical and volume control system. These sensors are not only for detecting faults, but also for running and controlling services.

Particle swarm algorithm has many applications in solving the problem of sensor layout optimization but the disadvantage of the basic particle swarm optimization algorithm is that the parameters are fixed, the particles are single, and it is easy to fall into the local optimization. In this paper, the basic particle swarm optimization algorithm is improved by Non-linearly adjusting inertia weight factor, asynchronously changing learning factor, and variating particle. The improved particle swarm optimization algorithm is used to optimize the sensor placement.

The numerical analysis verified that a smaller number of sensors can meet the fault detection requirements of the chemical and volume control system in this paper, and Experiments have proved that the improved particle swarm algorithm can improve the basic particle swarm algorithm, which is easy to fall into the shortcomings of local optimization and single particles. This method has good applicability, and could be also used to optimize other systems with sufficient parameters and consistent objective function.
Vertical-Downward Two-Phase Flow Regime Identification by Probabilistic Neural Network (PNN) and Nonlinear Support Vector Machine (SVM)
Authors: Zhong Wenyi , Qiao Shouxu , Hao Sijia , Li Xupeng , Tan Sichao
Published: October 2021

The present study proposes a new feature extraction method based on non-stationary conductivity probe signals. Two types of discriminative network models, i.e., the probabilistic neural network (PNN) and nonlinear support vector machine (SVM), are established for flow regime identification using small sample sets. The eigenvectors are composed of 16 feature quantities obtained by wavelet packet decomposition (WPD) and 8 feature quantities in the time-domain derived from the reconstructed low-frequency signals. The 8 features include maximum, minimum, standard deviation, arithmetic mean, kurtosis, peak factor, impulse factor and margin factor. The signals are normalized based on features rather than samples before flow regime identification. In the current study, WPD results show that the conductivity probe signals in two-phase flow are mostly in low frequency. The identification accuracy of the nonlinear SVM is 90.47%, which is better than 83.33% by the PNN method. This study verifies the superiority of nonlinear SVM in solving small samples and nonlinear flow regime classification problems. However, the accuracy of flow regime identification near flow regime transitional boundaries still remains questionable and needs further improvement.
Application of Combining 3D Model and Survey on Site to Simulate Dome Lifting
Authors: He Wei-ting , Zhou Yuan-xia , Yang Jie
Published: October 2021

Dome lifting is one of the most important milestones for nuclear power project. The lifting operation must be properly planned and carried out in safety manner. The simulation and analysis are usually done before lifting activities to ensure the smooth lifting implementation. By using of 3D model and survey data on site to assist dome lifting, the risks related to lifting activities can be discovered and the risk elimination scheme can be formulated at early stage. Combining with Power Plant Design Management System (PDMS), design inspection software NAVISWORKS and survey data on site, the accuracy of simulation and verification can be increased, the deviations and non-conformities can be easily found. Through the implementation of this technical scheme, it helps to ensure the smooth and safe implementation of the milestone of dome lifting for nuclear power project. This paper gives an example for combining 3D model and survey on site to simulate dome lifting, methods and suggestions are given for guidance for other projects.
Experimental Study on the Flow Characteristics of Rod Bundle Under Rolling Motion
Authors: Li Xin , Qi Peiyao , Tan Sichao , Qi Chao , Qiao Shouxu
Published: October 2021

The floating reactor system will be rolling, heaving and other movements affected by waves, wind, etc. These motions will introduce additional inertial force field into the rod bundle, thus affecting the flow characteristics in the rod bundle channel. In order to study the influence of rolling motion on the flow characteristics of rod bundle, a visualization research of the flow field of rod bundle channel with a pitch-to-diameter ratio of 1.326 was carried out under rolling motion. The results show that under a small driving force, the rolling motion has a significant effect on the flow field in the rod bundle, affecting the velocity distribution in different sub-channels, and there is transverse mixing between adjacent sub-channels. With the increase of driving force, the influence of rolling motion is gradually weakened. The flow field distribution under rolling motion is significantly different from that under pulsating flow. The experimental results show that the influence of rolling motion on the middle sub-channel of the rod bundle channel is small, and the influence on the edge sub-channel is large. The velocity field of the subchannels on both sides of the edge fluctuates periodically, and the wave phase is inverse. This study shows that the flow field variation caused by rolling motion is different from that caused by pulsating flow, and the flow field of fuel assembly in floating reactor system under ocean condition is further studied.
Investigation of Applicability of Subchannel Analysis Code ASFRE on Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure in Sodium Cooled Fast Reactor
Authors: Kikuchi Norihiro , Imai Yasutomo , Yoshikawa Ryuji , Doda Norihiro , Tanaka Masaaki
Published: October 2021

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR against a core disruptive accident. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named ASFRE has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE were validated by comparisons with experimental data. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house CFD code named SPIRAL was conducted to make further discussion on applicability of ASFRE to the thermal-hydraulics analysis of FAIDUS. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
Heat Transfer Performance for Helium Gas Flowing in a Minichannel With Different Inner Diameters
Authors: Xu Feng , Liu Qiusheng , Shibahara Makoto
Published: October 2021

The high heat load on the first wall of the helium cooled blanket is removed by tube flow of helium gas. Heat transfer augmentation is considered to be acquired by downsizing of channels. Therefore, this paper experimentally studied the influence of inner diameter on the heat transfer performance of helium gas flowing in a minichannel. The helium gas flowed in the small platinum tubes with the inner diameters of 0.8 mm and 1.8 mm, respectively. The heat generation rate of the tube was controlled by a heat input subsystem and raised with an exponential equation. The surface temperature and heat flux of the tubes were obtained under a wide range of e-folding time at different flow velocities. The heat transfer coefficients of different inner diameter tubes were compared at the same conditions. The heat transfer performance of the 0.8 mm-diameter tube was compared with a classical correlation. The experimental results showed that the heat transfer performance in the minichannel is better than a conventional large-diameter tube. The heat transfer coefficients of the 0.8 mm-diameter tube were higher than those of the 1.8 mm-diameter tube. The heat transfer process was enhanced with reducing the inner diameter of the minichannel. The heat transfer process was divided into two parts including transient and quasi-steady-state regions.
Study on Laminar Turbulent Transition in Square Arrayed Rod Bundles
Authors: Dutra Carolina S. B. , Merzari Elia
Published: October 2021

The study of coolant flow behavior in rod bundles is of relevance to the design of nuclear reactors. Although laminar and turbulent flows have been researched extensively, there are still gaps in understanding the process of laminar-turbulent transition. Such a process may involve the formation of a gap vortex street as the consequence of a related linear instability.

In the present work, a parametric study was performed to analyze the spatially developing turbulence in a simplified geometry setting. The geometry includes two square arrayed rod bundle subchannels with periodic boundary conditions in the cross-section. The pitch-to-diameter ratios range from 1.05 to 1.20, and the length of the domain was selected to be 100 diameters. No-slip condition at the wall, and inlet-outlet configuration were employed. Then, to investigate the stability of the flow, the Reynolds number was varied from 250 to 3000. The simulations were carried out using the spectral-element code Nek5000, with a Direct Numerical Simulation (DNS) approach. Data were analyzed to examine this Spatio-temporal developing instability. In particular, we evaluate the location of onset and spatial growth of the instability.
Phase-Field Model for Recrystallization of Impurities in Sodium Coolant
Authors: Kawaguchi Munemichi
Published: October 2021

In researches and developments (R&Ds) on sodium (Na) management technology, the experimental data related to a cold trap and a plugging meter have been accumulated in each country because the impurities such as sodium oxide (Na2O), sodium hydride (NaH) and/or sodium hydroxide (NaOH) in Na coolant accelerate the corrosion on the stainless-steel surface. The cold trap has stainless-steel wire mesh and wall where the saturated impurities are recrystallized to remove the impurity in the Na coolant. Similarly, the plugging meter has stainless-steel orifice where the saturated impurities are recrystallized to measure the saturation concentration (plug temperature).

The recrystallization is common physical chemistry phenomena, in which is dominated by the temperature and the concentration. To date, a phase-field model (PFM) has been developed extensively as a powerful tool to predict microstructure evolution of micro–meso scale. Especially, the PFM has some simulation merits: it can calculate movement of interface without explicit trace of the interface (calculate robustly the large deformation), and this simulation results are known to be consistent with the thermodynamics data. In this study, we developed the simulation models for complex recrystallization of one impurity in Na coolant using the PFM, and confirmed that the simulation results are reasonable for the experimental data with the operating experience. In the simulation results for the one impurity of Na2O or NaH, we concluded that the kinetics of recrystallization was determined by the solubility and/or diffusion behavior in liquid Na.
Research on Acoustic Characteristic of Steam Injection Under Small delta-T Below Saturation
Authors: Li Hui , Li Yong , Xiao Qi , Li Dongyang , Tan Sichao
Published: October 2021

Bubbling deaeration is widely applied in marine steam power system (MSPS), and direct contact condensation (DCC) of steam injected into water under low subcooled water is the thermodynamic foundation of this technology. Due to the small delta-T below saturation of the flow field, the steam condensation potential is small, and there is a violent steam plume evolution. The process and the growth, merging, breaking and collapse of bubbles, which produce severe pressure oscillation and vibration noise. In response to this problem, the flow pattern evolution and acoustic characteristics of underwater steam injection in the range of different subcooling degrees (10–0.2°C) and steam mass flow (18kg/h) were studied through experiments. Studies have found that the degree of subcooling has an important impact on the evolution of the steam plume and the spectral characteristics of noise. The change in the shape of the steam plume determines the amplitude and spectral characteristics of the noise sound pressure. As the ambient water temperature rises, the steam plume shape changes differently, and the phenomena such as “tail oscillation”, “congestion”, and “steam recirculation” appear in sequence. Affected by the evolution of the above-mentioned steam plume shape, the noise sound pressure amplitude changes with the subcooling. The temperature first rises and then falls, and it rises slightly when it is close to the saturation temperature. At the same time, the noise spectrum shifts from medium and low frequency to low frequency. When it is close to the saturation temperature, the spectrum tends to move to medium frequency.
Comparison of Pebble Bed Velocity Profiles Between High-Fidelity and Intermediate-Fidelity Codes
Authors: Reger David , Merzari Elia , Balestra Paolo , Schunert Sebastian , Hassan Yassin A.
Published: October 2021

Recent interest for the development of high-temperature gas reactors has increased the need for more advanced understanding of flow characteristics in randomly packed pebble beds. A proper understanding of these flow characteristics can provide a better idea of the cooling capabilities of the system in both normal operation and accident scenarios. In order to enhance the accuracy of computationally efficient, intermediate fidelity modeling, high-fidelity simulation may be used to generate correlative data. For this research, NekRS, a GPU-enabled spectral-element computational fluid dynamics code, was used in order to produce the high-fidelity flow data for beds of 1,568 and 45,000 pebbles. Idaho National Lab’s Pronghorn porous media code was used as the intermediate fidelity code. The results of the high-fidelity model were separated into multiple concentric regions in order to extract porosity and velocity averages in each region. The porosity values were input into the Pronghorn model and the resulting velocity profile was compared with that from NekRS. Both cases were run with a Reynolds number of 20,000 based on pebble diameter. The Pronghorn results were found to significantly overestimate the velocity in the outermost region indicating that changes in the porosity alone do not cause the difference in fluid velocity. We conclude that further work is necessary to develop a more effective drag coefficient correlation for the near-wall region and improve predictive capabilities of intermediate fidelity models.
U.S. Nuclear Power Plant Performance Assessment Using the Versatile Economic Risk Tool (VERT)
Authors: Miller Jaden C. , Ercanbrack Spencer C. , Pope Chad L.
Published: October 2021

This paper addresses the use of a new nuclear power plant performance risk analysis tool. The new tool is called Versatile Economic Risk Tool (VERT). VERT couples Idaho National Laboratory’s SAPHIRE and RAVEN software packages. SAPHIRE is traditionally used for performing probabilistic risk assessment and RAVEN is a multi-purpose uncertainty quantification, regression analysis, probabilistic risk assessment, data analysis and model optimization software framework. Using fault tree models, degradation models, reliability data, and economic information, VERT can assess relative system performance risks as a function of time. Risk can be quantified in megawatt hours (MWh) which can be converted to dollars. To demonstrate the value of VERT, generic pressurized water reactor and boiling water reactor fault tree models were developed along with time dependent reliability data to investigate the plant systems, structures, and components that impacted performance from the year 1980 to 2020. The results confirm the overall notion that US nuclear power plant industry operational performance has been improving since 1980. More importantly, the results identify equipment that negatively or positively impact performance. Thus, using VERT, individual plant operators can target systems, structures, and components that merit greater attention from a performance perspective.
Investigation on Thermal Stability of Sintered Magnesia in Sodium for Core Catcher Application in SFRs
Authors: Shukla Prabhat Kumar , Hemanth Rao E. , M. Muthuganesh , Vetrivendan E. , Polaki S. R. , Das Sanjay Kumar , Ningshen S. , Chaurasia Pramod Kumar , Ponraju D. , Athmalingam S. , Venkatraman B.
Published: October 2021

Sodium cooled Fast Reactors (SFRs) are built with several engineered safety features to prevent and mitigate core melt accidents. Core catcher is a passive mitigation device, which retains core debris resulting from hypothetical core melt accident in a subcritical and coolable geometry. The core catcher of Prototype Fast Breeder Reactor (PFBR) uses Austenitic Stainless Steel (ASS) as structural material. However, in the view of enhanced safety, core catchers for future reactors need to be designed for large thermal loads resulting from whole core melting. Provision of suitable refractory material as protection layer for ASS made core catcher plate can be one of the promising options to enhance the melt retention capability. Thermal stability of protective layer during impact of molten corium or fragmented core debris at high temperature in presence of liquid sodium is very essential for its satisfactory performance. Based on preliminary studies, sintered magnesia has been chosen as candidate material and experiments were carried out to investigate its stability in liquid sodium. Accelerated sodium compatibility tests were carried out by exposing magnesia specimens in sodium for 240 h at 850 °C. Tests were also carried out to assess its stability in liquid sodium under condition envisaged during whole core relocation over core catcher. Tests were carried out by subjecting the specimens under dead weight up to the boiling point of sodium in lower plenum of a typical SFR. Sodium exposed magnesia specimens were subjected to XRD and SEM analysis to assess material degradation. The specimens were also tested for change in compressive strength due to high temperature sodium exposure. The experimental results are discussed in the paper.
Analysis of Passive Tube Condensation With Non-Condensable Gas Using Heat and Mass Analogy Model
Authors: Cotul Ugur , Revankar Shripad T.
Published: October 2021

In this study, we used the heat and mass analogy model to be able to predict the heat transfer properties of a condenser tube operating in passive mode. The most important advantage of analogy model comparing boundary layer model is simplicity and fast computation, that’s why it can be applied to various engineering problems for many cases. The heat and mass analogy model is based on the heat transfer balance between liquid film and gas mixture area. The main problem for the liquid film region is the heat transfer coefficient (HTC) which is affected negatively in the presence of non-condensable gas. Therefore, our main goal is to increase the HTC and condensation heat transfer rate by updating the analogy code. In the gas-vapor mixture region, heat transfer mainly occurred as latent condensation and sensible heat transfer. In order to maintain this balance between the mixture and liquid film, the interface temperature is iterated. After defining a specified tolerance value of the heat and mass analogy model codes, this iteration process was started to be used at the entrance of a condenser tube. The gas and vapor mixture is considered to be saturated at the liquid/gas interface in the heat and mass transfer analogy model. Via boundary layer study of species concentration and energy balance, the non-condensable gas effect on condensation is added into the equation. For the condensation heat transfer coefficient of turbulent vapor flow associated with laminar condensate, numerical predictions were made and they were satisfactory. The predictions were compared with the experimental data from the literature to be able to test the model. Non-condensable gas mass fraction and vapor-non-condensable mixture temperature were presented in the form of radial and axial profiles.
UHT Test Facility Updates and Oxidation Tests for Accident Tolerant Fuel Development
Authors: Byers William A. , Wang Guoqiang
Published: October 2021

Research and development on accident tolerant fuel (ATF) have been widely pursued by the nuclear industry around the world. Westinghouse is developing new accident tolerant fuel design, EnCore® Fuel. Chromium (Cr) coated zirconium alloy and silicon-carbide (SiC) fuel cladding are the excellent candidates for application in fuel cladding of light water reactors (LWRs)[1, 2]. The ATF fuel is being developed to deliver beyond-design-basis (BDB) and design-basisaltering (DB) safety margins, withstanding far more severe conditions than the current fuel. The products in this portfolio have enhanced temperature performance enabling survival during BDB and DB accidents. The products also minimize interactions of fuel and fuel rod materials with water. The Cr-coated zirconium alloy cladding exhibits sustained high temperature performance and inhibits the detrimental effects of the zirconium-steam reaction. Westinghouse’s silicon-carbide (SiC) fuel cladding offers safety benefits in severe accident conditions, particularly compared to the significant hydrogen-and-heat producing reactions that occur above 1200° C for zirconium fuel cladding. These advancements will increase safety margins and enable the transition to high burnup facilities as well.

To support the accident tolerant fuel (ATF) development, the Westinghouse Ultra High Temperature (UHT) test facility was built at the Westinghouse Churchill site in early 2016. The UHT test facility has been upgraded recently. The current UHT reactor was designed to promote better sealing of the reaction chamber and a smaller uncertainty on cladding temperature measurement. Simulating large break LOCA steam conditions, the UHT oxidation tests have been performed with SiC Optimized ZIRLO® and Cr-coated zirconium alloy cladding. In the EnCore Fuel program, one of the key tasks is to test the safety features and margin gain of the EnCore Fuel. This paper describes the designs, safety features and baseline commissioning tests for the upgraded Ultra High Temperature (UHT) test facility.