Accepted Manuscripts

Wei Gao, Guofeng Tang, Jingyu Zhang and Qinfang Zhang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039967
Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has been studying seismic risk analysis for nuclear power plant for a long time, and completed Seismic Margin Analysis (SMA) for several plants. After Fukushima accident, seismic risk has drawn an increasing attention worldwide, and the regulatory body in China has also required the utilities to conduct a detailed analysis for seismic risk. So we turned our focus on a more intensive study of Seismic Probabilistic Safety Assessment (PSA/PRA) for nuclear power plant in recent years. Since quantification of seismic risk is a key part in Seismic PSA, lots of efforts have been devoted to its research by SNERDI. The quantification tool is the main product of this research, and will be discussed in detail in this paper. First, a brief introduction to Seismic PSA quantification methodology is presented in this paper, including fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves, and uncertainty analysis as well. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Finally, this paper introduced the development of the seismic PSA quantification tool based on the algorithms discussed in this paper. Tests and application has been made for this software based on a specific nuclear power plant seismic PSA model.
TOPICS: Industrial research, Probabilistic safety assessment, Earthquake risk, Nuclear power stations, Algorithms, Design, China, Computer software, Probability, Nuclear engineering, Public utilities, Fukushima nuclear disaster, Japan, 2011, Uncertainty analysis
Muhammad/Q Awan, Liangzhi Cao, Hongchun Wu and Chuanqi Zhao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039971
Use of FCM fuel in LWRs is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. Present work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and keeping the MTC feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configuration including QUADRISO, BISO and Matrix Mix forms. Results obtained from the implicit random treatment of the double heterogeneity of TRISO, QUADRISO and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5 %. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at EOC; however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At BOC, erbia loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.
TOPICS: Fuels, Manufacturing, Pressurized water reactors, Cycles, Feedback, Rods, Light water reactors, Accidents, Resonance, Particulate matter, Containment
Kei Sugihara, Hirotaka Sakai, Kanako Hattori, Genki Tanaka, Mitsunobu Hayashi, Toshiaki Ito and Naotaka Oda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039968
In this study, the applicability of Monte-Carlo code PHITS(1) to the equipment design of sampler and detector in the radiation monitoring system was evaluated by comparing calculation results with experimental results obtained by actual measurements of radioactive materials. In modeling a simulation configuration, reproducing the energy distribution of beta-ray emitted from specific nuclide by means of Fermi Function was performed as well as geometric arrangement of the detector in the sampler volume. The reproducing and geometric arrangement proved that the calculation results are in excellent matching with actual experimental results. Moreover, reproducing the Gaussian energy distribution to the radiation energy deposition was performed according to experimental results obtained by the multi-channel analyzer. Through the modeling and the Monte-Carlo simulation, key parameters for equipment design were identified and evaluated. Based on the results, it was confirmed that the Monte-Carlo simulation is capable of supporting the evaluation of the equipment design.
TOPICS: Sensors, Radiation measurement, Simulation, Design, Modeling, Nuclides, Radiation (Physics), Radioactive substances, Beta particles
Yankun Dou, xinfu He, Dongjie Wang, Wu Shi, Lixia Jia and Wen Yang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039969
In order to study the contribution of Mn atoms in Cu precipitates to hardening in bcc Fe matrix, the interactions of a (111){110} edge dislocations with nanosized Cu and Cu-Mn precipitates in bcc Fe have been investigated by using of molecular dynamics. The results indicate that the critical resolved shear stresses of the Cu-Mn precipitates are larger than that of Cu precipitates. Meanwhile, the critical resolved shear stresses of the Cu-Mn precipitates show a much more significant dependence on temperature and size, compared to Cu precipitates. Mn atoms exhibit strong attractive interaction with <111> crowdion and improve the fraction of transformed atoms from body centered cubic (bcc) phase to face centered cubic (fcc) phase for big size precipitates. Those all lead to the higher resistance to the dislocation glide. The increasing temperature can assist the Cu atoms rearrange back towards a bcc structure, resulting in the rapid decline of the critical resolved shear stresses. Eventually, these features are confirmed that the appearance of Mn atoms in Cu precipitates greatly facilitates the hardening in bcc Fe matrix.
TOPICS: Hardening, Atoms, Shear stress, Dislocations, Temperature, Molecular dynamics
Karl Waedt, Yongjian Ding, Antonio Ciriello and Xinxin Lou
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039970
The generic concept of Security Controls, as initially deployed in the information security domain, is gradually used in other business domains, including industrial security for critical infrastructure and cybersecurity of nuclear safety I&C. A Security Control, or less formally, a security countermeasure can be any organizational, technical or administrative measure that helps in reducing the risk imposed by a cybersecurity threat. In order to facilitate and formalize the process of developing, precisely describing, distributing and maintaining more complex security controls, the Application Security Controls (ASC) concept is introduced by the new ISO/IEC 27034 multipart standard. An ASC is an extensible semi-formal representation of a security control (e.g. XML or JSON-based), which contains a set of mandatory and optional parts as well as possible links to other ASCs.
TOPICS: Maintenance, Security, Computer security, Safety, Risk, Countermeasures
Marcel Straetz, Joerg Starflinger, Rainer Mertz and Dieter Brillert
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039884
In case of an accident in a nuclear power plant with combined initiating events, (loss of ultimate heat sink and station blackout) an additional heat removal system could transfer the decay heat from the core to an ultimate heat sink. One specific additional heat removal system, based upon a Brayton cycle with supercritical CO2 as working fluid, is currently investigated within the EU-funded project “sCO2-HeRo” (supercritical carbon dioxide heat removal system). It serves as a self-launching, self-propelling and self-sustaining decay heat removal system used in severe accident scenarios. Since this Brayton cycle produces more electric power than it consumes, the excess electric power can be used inside the power plant, e.g. for recharging batteries. A small-scale demonstrator is attached to the PWR glass model at Gesellschaft für Simulatorschulung (GfS), Essen, Germany. In order to design and build this small-scale model, cycle calculations are performed to determine the design parameters from which a layout can be derived.
TOPICS: Heat, Fluids, Cycles, Supercritical carbon dioxide, Brayton cycle, Accidents, Design, Heat sinks, Electricity (Physics), Glass, Nuclear power stations, Pressurized water reactors, Power stations
Jayangani I Ranasinghe, Ericmoore Jossou, Linu Malakkal, Barbara Szpunar and Jerzy Szpunar
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039886
The understanding of the radial distribution of temperature in a fuel pellet, under normal operation and accident conditions, is important for a safe operation of a nuclear reactor. Therefore, in this study, we have solved the steady state heat conduction equation, to analyze the temperature profiles of a 12 mm diameter cylindrical dispersed nuclear fuels of U3O8-Al, U3Si2-Al, and UN-Al operating at 870 K. Moreover, we have also derived the thermal conductivity correlations as a function of temperature for U3Si2, UN, and Al. To evaluate the thermal conductivity correlations of U3Si2, UN, and Al we have used density functional theory (DFT) as incorporated in the Quantum ESPRESSO (QE) along with other codes such as Phonopy, ShengBTE, EPW, and BoltzTraP. However, for U3O8, we utilized the thermal conductivity correlation proposed by Pillai et al. Furthermore, the effective thermal conductivity of dispersed fuels with 5, 10, 15, 30 and 50 vol% respectively of dispersed fuel particle densities over the temperature range of 300 to 900 K was evaluated by Bruggman model. Additionally, the temperature profiles and temperature gradient profiles of the dispersed fuels were evaluated by solving the steady state heat conduction equation by using Maple code. This study not only predicts a reduction in the centerline temperature and temperature gradient in dispersed fuels but also reveals the maximum concentration of fissile material (U3O8, U3Si2, and UN) that can be incorporated in the Al matrix without the centerline melting.
TOPICS: Aluminum, Temperature distribution, Nuclear fuels, Fuels, Thermal conductivity, Temperature, Steady state, Temperature profiles, Temperature gradient, Heat conduction, Density functional theory, Melting, Accidents, Nuclear reactors, Particulate matter
Technical Brief  
Qiang Zhao, Zheng Zhang and Xiaoping Ouyang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039885
Uranium dioxide (UO2) is the typical fuel that is used in the current nuclear power plant, fission gas atoms are produced during and after the nuclear reactor operation, and the fis- sion gas atoms have a significant effect on the performance of UO2 fuel in nuclear reactor. In this paper, we investigated the diffusion of the fission gas atoms in the UO2 fuel by using the first-principles calculation method based on the density functional theory (DFT). The results indicate that the volume of the UO2 cell increased when there is a fission gas atom enter in the UO2 supercell; the elastic properties of UO2 are in good agreement with other simulation results and exper- imental data, and the fission gas atoms make the ductility of UO2 decreased; fission gas atoms prefer to occupy the octahedral interstitial site (OIS) over the uranium vacancy site and the oxygen vacancy site, and the oxygen vacancy site is the most difficult occupied site due to the formation of an oxygen vacancy is more difficult than that of the ura- nium vacancy; the diffusion barrier of a He atom in the UO2 supercell is smaller than that of a oxygen atom, that means that the diffusion of the He atom in UO2 fuel is stronger than that of the oxygen atom. Our work may shed some light on the formation mechanism of the bubbles caused by the fission gas atoms in the UO2 fuel.
TOPICS: Diffusion (Physics), Nuclear fission, Atoms, Density functional theory, Uranium, Oxygen, Fuels, Bubbles, Ductility, Nuclear reactors, Nuclear power stations, Simulation results, Nuclear power plant operations, Computational methods, Elasticity
Toru Kitagaki, Takanori Hoshino, Kimihiko Yano, Nobuo Okamura, Hiroshi Ohara, Tetsuo Fukasawa and Kenji Koizumi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039847
Evaluation of fuel debris properties in the Fukushima Daiichi Nuclear Power Plant (1F) is required to develop fuel debris removal tools. In the removal of debris resulting from the TMI-2 accident, a core-boring system played an important role. Considering the working principle of core boring, hardness, elastic modulus, and fracture toughness were found to be important fuel debris properties that profoundly influenced the performance of the boring machine. It is speculated that uranium and zirconium oxide solid solution (U,Zr)O2 is one of the major materials in the fuel debris from 1F. In addition, the Zr content of the fuel debris from 1F is expected to be higher than that of the debris from TMI-2 because the 1F reactors were boiling-water reactor (BWR). In this research, the mechanical properties of cubic (U,Zr)O2 samples containing 10%¬-65% ZrO2 are evaluated. The hardness, elastic modulus, and fracture toughness are measured by the Vickers test, ultrasonic pulse echo method, and indentation fracture method, respectively. In case of the (U,Zr)O2 samples containing less than 50% ZrO2, Vickers hardness and fracture toughness increased, and the elastic modulus decreased slightly with increasing ZrO2 content. Moreover, all of those values of the (U,Zr)O2 samples containing 65% ZrO2 increased slightly compared to (U,Zr)O2 samples containing 55% ZrO2. ZrO2 content affects fracture toughness significantly in the case of samples containing less than 10% ZrO2. Higher Zr content (exceeding 50%) has little effect on the mechanical properties.
TOPICS: Mechanical properties, Zirconium, Fuels, Fracture toughness, Elastic moduli, Boiling water reactors, Vickers hardness testing, Fukushima nuclear disaster, Japan, 2011, Nuclear power stations, Solid solutions, Uranium, Accidents, Fracture (Process), Boring machines, Echoes, Fracture (Materials)
Wang Zhu, Zhang Chunyu and Yuan Cenxi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039848
Nuclear fuel rods operate under complex radioactive, thermal and mechanical conditions. Nowadays fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermos-mechanical behavior of fuel rods. The swelling of fission products cause a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.
TOPICS: Fuels, Modeling, Pressurized water reactors, Fuel rods, Nuclear fission, Displacement, Tension, Physics, Thermal expansion, Stress, Transients (Dynamics), Cladding systems (Building), Accidents, Finite element analysis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039778
Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian Super-Critical Water-cooled Reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of ThO2-based fuel irradiation behaviour is significantly less advanced than that of UO2 fuel, although ThO2 possesses superior thermal conductivity, thermal expansion, higher melting temperature and oxidation resistance that may improve both fuel performance and safety. The Fuel and Sheath Modelling Tool (FAST), a fuel performance model for uranium dioxide (UO2) fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include Thoria (ThO2), Thorium Uranium Dioxide (Th,U)O2 and Thorium Plutonium Dioxide (Th,Pu)O2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O2 fuel. The development and integration of ThO2 and (Th,U)O2 models into the existing FAST model led to the Multi-Pellet Material Fuel and Sheath Modelling Tool (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO2 and (Th,U)O2 MPM-FAST model, a comparison between modelling results with post-irradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO2 and (Th,U)O2 fuel compared to UO2 . Results are encouraging when compared to PIE data.
TOPICS: Engineering prototypes, Uranium, Nuclear fuels, Fuels, Modeling, Sustainability, Irradiation (Radiation exposure), Supercritical water reactors, Nuclear industry, Melting, Thermal conductivity, Design, Safety, Thermal expansion, Temperature, Cycles, Military systems, Model development, oxidation, Collaboration, Water
Technical Brief  
Martin Schulc, Michal Košt'ál, Evžen Novák, Jan Šimon and Nicola Burianová
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039774
The presented work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6E8 n/s was employed for the irradiation of sodium iodide. The spectrum averaged cross sections were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the Net Peak Areas measured using the high purity germanium spectroscopy. The measured spectrum averaged cross section for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to 8.98 ± 0.32 µb. The resulting spectrum averaged cross section for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as 2.044 ± 0.0072 mb. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.
TOPICS: Neutron sources, Chemical kinetics, Spectroscopy, Irradiation (Radiation exposure), Cross section (Physics), Germanium, Sodium, Emissions, Nuclear fission, Neutrons
Technical Brief  
Hu Chundong, Wu Mingshan, Xie Yahong, Wei Jianglong and Yu Ling
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039694
During the process of beam extraction in positive ion source under high voltage region, a large number of electrons are produced in the gaps of grids. After back-streaming acceleration, these electrons go back to arc chamber or impinge grids and heat electron dump or grids, which are harmful for the safety of ion source. Under the situation of poor beam extraction optics, a large part of the primary beam ions bombard the surface of suppressor grid. And this process produces a large number of electrons. Due to the huge extracted voltage, the secondary electron emission coefficient of the suppressor grid surface is also great, when beam ions bombard on it. As a result, the grids' current grows. According the measurement of the current of suppressor grid and the calculation of the perveance of the corresponding shoot, we can analyze the effect of beam divergence angle on back-streaming electron. When the beam divergence angle increases, the number of back-streaming electrons increases rapidly, and grids current changes significantly, especially the current of gradient grid and suppressor grid. The results can guide the parameters operating on the ion source for EAST-NBI and find the reasonable operation interval of perveance and the best one to ensure the safety and stable running of the ion source, which has great significance on the development of long pulse, high power ion source.
TOPICS: Electrons, Silencers, Ions, Safety, Electron emission, Heat, Optics
Sumit Vishnu Prasad and Arun Nayak
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039636
The present experimental investigation in a scaled facility of an Indian PHWR is focused on the heat transfer behaviour from the calandria vessel to the calandria vault during a prolonged severe accident condition in presence of decay heat. The transient heat transfer simulates the conditions from single phase to boiling in the calandria vault water, partial uncovery of the calandria vessel due to boil off of water in the vault, refill of calandria vault. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1100°C. Decay heat in the melt pool was simulated by using high watt cartridge type's heaters. The temperature distributions inside the molten pool, across the calandria vessel wall thickness and vault water were measured for prolonged period which can be divided into various phases viz. Single phase natural convection heat transfer in calandria vault, boiling heat transfer in calandria vault, partial uncovery of calandria vessel and refilling calandria vault. Experimental results showed that the once the crust formed, the inner vessel temperature remained very low and vessel integrity maintained. Even boiling of calandria vault water and uncovery of calandria vessel had negligible effect on melt, calandria vessel and vault water temperature. The heat transfer coefficients on outer vessel surface were obtained and compared with various conditions.
TOPICS: Heat, Heat transfer, Simulation, Accidents, Vessels, Water, Heavy water reactors, Boiling, Natural convection, Borosilicate glasses, Temperature distribution, Transient heat transfer, Water temperature, Heat resistant glass, Temperature, Heat transfer coefficients, Wall thickness
Sapna Singh, Garima Singal, Arun Nayak and Umasankari Kannan
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039594
In a natural circulation Boiling Water Reactor (BWR), the core power varies in both axial and radial direction inside the reactor core. The variation along the axial direction is more or less constant throughout the reactor; however, there exists variation of reactor power in the radial direction. The channels located at the periphery have low power compared to the center of the core and are equipped with orifices at their inlet. This creates non-uniformity in the radial direction in the core. This study has been performed in order to understand the effect of this radial variation of power on the stability characteristics of the reactor. Four channels of a pressure tube type natural circulation boiling water reactor have been considered. The reactor has been modeled using RELAP5/MOD3.2. Before using the model, it was first benchmarked with experimental measurements and then the characteristics of both low power and high power oscillations; respectively known as Type-I and Type-II instability, has been investigated. It was observed that the Type-I instability shows slight destabilizing effect of increase in power variation among the different channels. However, in the case of Type-II instability, it was found out that the oscillations get dampens with an increase in power variation among the channels. A similar effect was found for the presence of orifices at the inlet in different channels. However, increase in number of orificed channels showed stabilizing effect for both Type-I and Type-II instabilities.
TOPICS: Stability, Boiling water reactors, Orifices, Oscillations, Pressure
Wolfgang Flaig, Rainer Mertz and Joerg Starflinger
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039595
Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor-system with supercritical CO2 as the working fluid. In case of a severe accident this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger needs to be as compact and efficient as possible. Therefore, a diffusion welded plate heat exchanger (DWHE) with mini channels was developed and manufactured. This DWHE was tested to gain data of the transferrable heat power and the pressure loss. A multipurpose facility has been built at IKE for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa and temperatures up to 150 °C. This paper describes the development and set-up of the facility as well as the first experimental investigation.
TOPICS: Heat, Heat exchangers, Test facilities, Supercritical carbon dioxide, Nuclear power stations, Carbon dioxide, Supercritical fluids, Accidents, Pressure, Flow (Dynamics), Nuclear reactors, Low temperature, Turbines, Temperature, Diffusion (Physics), Heat transfer, Fluids, Compressors, Thermal energy, Coolants, Solar power stations, Waste heat
Xiong Wenbin, Cao Jian, Huang Chaoyun, Bie Yewang, Wang Yanqi and Li Jufeng
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039596
This study investigates the reactor core physical properties of the AP1000, which applies the MCNP4a program to model the AP1000 reactor core with the parameters and data from the Design Control Document (DCD, Rev.19) of the AP1000 Nuclear Power Plant, which has been submitted to the NRC. The model is applied to calculate and verify the physical parameters of AP1000 core design. The results match well with the design values in the DCD of the AP1000 nuclear power plant. The model will be modified according to the actual reactor core arrangement, such as AP1000 reactors at China's Sanmen and Haiyang sites, and then compared with the commissioning test results in the future.
TOPICS: Design, Nuclear power stations
Yu Ji, Jun Sun and Lei Shi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039600
Hydrogen is adopted as coolant for regenerative cooling nozzle and reactor core in nuclear thermal propul-sion (NTP), which is a promising technology for human space exploration in the near future due to its large thrust and high specific impulse. During the cooling process, the hydrogen alters its state from subcritical to supercritical, accompanying with great variations of fluid properties and heat transfer characteristics. This paper is intended to study heat transfer processes of supercritical pressure hydrogen under high extremely heat flux by using numerical approach. To begin with, the models explaining the variation of density, specific heat capacity, viscosity and thermal conductivity are introduced. Later on, the convective heat transfer to supercritical pressure hydrogen in a straight tube is investigated numerically by employing a computational model, which is simplified from experiments performed by Hendricks et al. During the simulation, the stand-ard k-e model combining the enhanced wall treatment is used to formulate the turbulent viscosity, and the results validates the approach through successful prediction of wall temperature profile and bulk tempera-ture variation. Besides, the heat transfer deterioration which may occur in the heat transport of supercritical fluids is also observed. According to the results, it is deduced that the flow acceleration or the velocity profile distortion to "M" shape due to properties variation of hydrogen contributes to the suppression of turbulence and the subsequent deteriorated heat transfer.
TOPICS: Pressure, Convection, Hydrogen, Heat transfer, Cooling, Turbulence, Viscosity, Thrust, Simulation, Coolants, Supercritical fluids, Thermal conductivity, Impulse (Physics), Fluids, Nozzles, Specific heat, Flow (Dynamics), Heat, Shapes, Wall temperature, Heat flux, Density
Songbai Cheng, Ting Zhang, Jinjiang Cui, Pengfeng Gong and Yujia Qian
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039597
Studies on debris bed formation behavior are important for improved evaluation of core relocation and debris bed coolability that might be encountered in a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). Motivated to clarify the flow-regime characteristics underlying this behavior, both experimental investigations and empirical-model development are being performed at the Sun Yat-sen University in China. As for the experimental study, several series of simulated experiments are being conducted by discharging various solid particles into water pools. To obtain a comprehensive understanding, a variety of parameters, including particle size (0.125~8mm), particle density (glass, aluminum, alumina, zirconia, steel, copper and lead), particle shape (spherical and non-spherical), water depth (0~80cm) along with the particle release pipe diameter (10~40mm) were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of flow regimes, termed respectively as the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertia dominant regime, were identifiable. As for empirical-model development, aside from a base model which is restricted to predictions of spherical particles, in this paper considerations on how to cover more realistic conditions (esp. debris of non-spherical shapes) are also discussed. It is shown that by coupling the base model with an extension scheme, respectable agreement between experiments and model predictions for regime transition can be achieved for both spherical and non-spherical particles given our current range of conditions.
TOPICS: Flow (Dynamics), Particulate matter, Water, Sodium fast reactors, Shapes, Density, Inertia (Mechanics), Accidents, Convection, Pipes, China, Particle size, Copper, Aluminum, Steel, Glass
Oksana Klok, Konstantina Lambrinou, Serguei Gavrilov, Jun Lim and Iris De Graeve
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039598
This work focuses on the effect of dissolved in liquid LBE oxygen concentration on the onset of dissolution corrosion in a solution-annealed 316L austenitic stainless steel. Specimens made of the same 316L stainless steel heat were exposed for 1000 hours at 450 °C to static liquid LBE with controlled concentrations of dissolved oxygen, i.e., 10-5, 10-6 and 10-7 mass%. The corroded 316L steel specimens were analyzed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). A complete absence of dissolution corrosion was observed in the steel specimens exposed to liquid LBE with 10-5 and 10-6 mass% oxygen. In the same specimens, isolated 'islands' of FeCr-containing oxides were also detected, indicating the localized onset of oxidation corrosion under these exposure conditions. On the other hand, dissolution corrosion with a maximum depth of 59 µm was detected in the steel specimen exposed to liquid LBE with 10-7 mass% oxygen. This suggests that the threshold oxygen concentration associated with the onset of dissolution corrosion in this 316L steel heat lies between 10-6 and 10-7 mass% oxygen for the specific exposure conditions (i.e., 1000 hours, 450 °C, static liquid LBE).
TOPICS: Corrosion, Oxygen, Stainless steel, Steel, Heat, Ferrochromium, X-ray spectroscopy, Scanning electron microscopy, oxidation

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In