Accepted Manuscripts

Rosario Delgado-Tardáguila, Marisol Corisco, Antonio Espejo, Daniel Navarro and Javier Riverola
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042358
One of the limiting conditions during operation of a Pressurized Water Reactor (PWR) is cladding integrity in class I (normal operations) or class II (most frequent). Cladding integrity is limited typically by the Departure from the Nucleate Boiling (DNB) which criterion ensures an appropriate core refrigeration. Adequate heat transfer between the fuel cladding and reactor coolant is achieved by preventing DNB that is avoided if the local heat flux is lower than the critical heat flux. The DNB is estimated though correlations based on several parameters, the Thermal Diffusion Coefficient (TDC) among others. Nevertheless and although the TDC is a variable, the thermal-hydraulic design codes specifically developed for the DNB prediction consider the TDC as a constant. This investigation is founded on a new numerical program developed to explore the effect of the TDC on the DNB. In addition to this, variables as the effect of the Turbulent Momentum Factor (FTM) and the correlation effect has been explored too. The most direct outcome of this research is the substantial extension of the existing studies of VIPRE-W TH code. The results show that TDC has an effect on the DNBR dominated by the radial power distribution. The DNBR increases up to 1.2% when TDC is variable under normal operation radial shapes.
TOPICS: Thermal diffusion, Nucleate boiling, Sensitivity analysis, Cladding systems (Building), Pressurized water reactors, Shapes, Heat flux, Critical heat flux, Design, Performance, Refrigeration, Nuclear reactor coolants, Momentum, Heat transfer, Fuels, Turbulence
Vojtech Galek, Jaroslav Stoklasa, Jan Hadrava, Jan Hrbek, Petr Bezdicka and Ivona Sedlárová
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042359
The technology of Molten Salt Oxidation (MSO) is a thermal treatment process mainly used for reprocessing of hazardous organic waste. This technology is considered as an alternative to the conventional incineration processes. The principle of the whole technology is based on flameless oxidation of materials in the molten salt with the consequent capture of gaseous products in molten alkaline salts. The melts with low melting point and high viscosity, such as a ternary mixture of carbonates Na2CO3, K2CO3 a Li2CO3, are the most used in this technology. However, the molten salts create a very corrosive environment for metal and ceramic materials, so the main aim of this experimental work was to determine the resistance of corundum samples, which were prepared by plasma spraying, and to find out its potential use as the protection of the reactor metal surface.
TOPICS: Ceramics, oxidation, Metals, Viscosity, Plasma spraying, Organic wastes, Metal surfaces, Melting point
Andriolo Lena, Mériot Clément and Bakouta Nikolai
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042360
The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EdF) Research and Development (R&D) Center, in the PERICLES Department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMRs) with the MAAP5 severe accident analysis code in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a step-wise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of lower head geometry, vessel steel ablation and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. Results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identifies modeling improvements.
TOPICS: Small modular reactors, Vessels, Heat, Pressurized water reactors, Radiation (Physics), Modeling, Approximation, Geometry, Heat losses, Industrial research, Flux (Metallurgy), Ablation (Vaporization technology), Energy dissipation, Accidents, Heat flux, Steel, Surveillance
Jorgelina Lupiano Contreras, Alicia Doval and Patricio Alberto
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042363
Difficulties are experienced during the thermal-hydraulic design of a nuclear reactor operating in the transition flow regime and are the result of the inability to accurately predict the heat transfer coefficient. Experimental values for the heat transfer coefficient in rectangular channels are compared with the calculated by correlations usually used for the design of MTR Reactors. The values predicted by Gnielinski and Kreith correlations at Re<5000 are not necesarily conservative. The Al-Arabi-Churchill correlation with the correction proposed by Jones has proved to be conservative for Reynolds between 2100 and 5000. Two alternative design approaches are proposed to solve a specific thermal-hydraulic design problem for an MTR reactor operating at Reynolds 2500. The conservative approach comprises two alternatives: the use of Al-Arabi correlation with no uncertainty factors, as it has proved to be conservative, or the use of Kreith correlation with a maximum uncertainty. In this conservative approach, maximum deviations in other input parameters are also taken into account. The best estimate plus uncertainty approach (BEPU) considers an uncertainty distribution in input parameters to generate a random sample of 59 inputs. An uncertainty distribution based on the ratio between the experimental and the calculated heat transfer coefficient, when using Kreith correlation, is considered. Results are given in terms of maximum and minimum bounds for the figure of merit used as design criterion with 95% probability and 95% confidence level. The BEPU approach offers a less penalizing design and its use depends on regulator's acceptance.
TOPICS: Flow (Dynamics), Design, Uncertainty, Heat transfer coefficients, Nuclear reactors, Probability
Cristina Guibaldo, Georgina De Micco and Ana Bohé
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042362
Uranium tetrafluoride was synthesized using a novel method which consists of a combination of carbochlorination reaction between a mixture of U3O8 and sucrose carbon with chlorine, and a solid state halogen exchange reaction between the products of the carbochlorination reaction and sodium fluoride. The thermodynamic feasibility to produce the halogen exchange reaction between UCl4 and NaF was analyzed. Reactions are favorable in standard conditions, even at low temperature. We have prepared a mixture of UCl4 and UCl2O2 by U3O8 treatment in Cl2 atmosphere with presence of sucrose carbon at 1173 K. UCl4 and UCl2O2 were obtained as a condensed product, which was collected in a quartz capsule containing NaF. The capsules were sealed after several repeated stages of argon purges and mechanic vacuum. Subsequently, they were treated at 573-623 K for 2 hours. We obtained that when NaF is the limiting reagent, the solid product of the thermal treatment of the capsules consists in a mixture of UF4 and NaCl (this last one is water soluble). Solid products were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), fluorescence spectroscopy and energy dispersive X-ray spectroscopy (EDX). Gaseous products were identified by Fourier-transform infrared spectroscopy (FTIR).
TOPICS: X-ray diffraction, Vacuum, Infrared spectroscopy, Fluorescence spectroscopy, Fourier transform infrared spectroscopy, X-ray spectroscopy, Carbon, Low temperature, Scanning electron microscopy, Fourier transforms, Quartz, Sodium, Uranium, Water
Jie Cheng, Yingwei Wu, Guanghui Su, Suizheng Qiu and Wenxi Tian
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042364
China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on PWR (pressurized water reactor) and SCWR (super-critical water reactor) water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher Tritium Breeding Ratio (TBR) and uniform heat distribution. This blanket uses the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. At first, the neutronic analysis was performed and based on the typical outboard equatorial blanket. Then, thermal and fluid dynamic analysis of the 3-D model was performed by CFX with fluid-solid coupling approach. It was found that the temperature on the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water condition. It indicated that SCWR case had smaller safety margin than PWR case, but SCWR case would lead higher outlet temperature, thermal conductivity and heat exchange efficiency also. In addition, it was found that beryllium was the dominant factor leading a higher TBR.
TOPICS: Water, Pressurized water reactors, Supercritical water reactors, Fluids, China, Heat, Temperature, Dynamic analysis, Power stations, Neutrons, Lithium, Three-dimensional models, Bridges (Structures), Ceramics, Safety, Computer simulation, Tokamaks, Coolants, Thermal conductivity, Design
Technical Brief  
Nicolas Alejandro Malinovsky
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042361
The present work shows the introduction of the ETAP software as a calculation and analysis tool for power electrical systems of the nuclear power plants under the orbit of Nucleoeléctrica Argentina S.A. Through the use of the software, the model of the electrical power system of the Atucha II Nuclear Power Plant was developed. To test the functionality of the model, studies of load flow and short-circuit analysis were conducted, yielding satisfactory results, which were contrasted with the plant design values. Once the model has been validated, this will be the basis for carrying out different studies in the plant through simulation. Furthermore, with the incorporation of ETAP as a fundamental calculation and analysis tool for power electrical systems at NASA's Engineering departments, it is expected to improve the safety, operation, quality, reliability and maintenance of both the Atucha II NPP electrical power system as well as the other nuclear power plants operated by NASA.
TOPICS: Electricity (Physics), Modeling, Nuclear power stations, Computer software, Electronic systems, Circuits, Maintenance, Safety, Reliability, Simulation, Stress, Plant design, NASA, Flow (Dynamics)
Koichi Hata, Katsuya Fukuda and Tohru Mizuuchi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042356
Natural convection heat transfer from vertical 7x7 rod bundle in liquid sodium was numerically analyzed to optimize the thermal-hydraulic design for the bundle geometry with equilateral square array, ESA. The 7x7 test rods for diameter (D=7.6 mm), heated length (L=200 mm) and L/d (=26.32) were used in this work. The surface heat fluxes for each cylinder were equally given for a modified Rayleigh number, (R_f,L)_ij and (R_f,L)_NxxNy,S/D, ranging from 3.08x10^4 to 4.28x10^7 (q=1x10^4~7x10^6 W/m^2) in liquid temperature (TL=673.15 K). The values of S/D for vertical 7x7 rod bundle were ranged from 1.8 to 6 on the bundle geometry with equilateral square array. The general correlations for natural convection heat transfer from a vertical NxxNy rod bundle with the equilateral square and triangle arrays including the effects of array size, (R_f,L)_NxxNy,S/D and S/D were derived. The correlations for vertical NxxNy rod bundles can describe the theoretical values of (Nu_av,B)_NxxNy,S/D for each bundle geometry in the wide analytical range of S/D (=1.8 to 6) and the modified Rayleigh number ((R_f,L)_NxxNy,S/D=3.08x10^4 to 4.28x10^7) within -9.49 to 10.6 % differences.
TOPICS: Heat transfer, Natural convection, Sodium, Fuel rods, Geometry, Rayleigh number, Design, Cylinders, Flux (Metallurgy), Heat, Temperature, Rods
Technical Brief  
He Lixia, Xiao-yong Hao and Gaokui He
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042357
Thallium bromide (TlBr) is a kind of semiconductor material. Due to its promising physical properties and can be used at room temperature, it is continually studied as X and gamma ray detectors' candidate material. Both of its atomic number and density are high. It also has large band-gap (B=2.68eV) and low ionization energy. TlBr device exhibits high detection efficiency and excellent energy resolution. It can be easily fabricated or compacted in small housing. So it is a reasonable selection in the fields of nuclear material inspection and safeguards property, national security, spatial and high energy physics researches. The paper investigates the TlBr radioactive detector development and fabrication procedures. The processing detail information and signals collection are emphasized in different section. The prototype detectors were irradiated by Am-241 and corresponding spectrum was obtained. The photoelectric peak at 59.5keV is distinguished visible and the best resolution at 59.5keV is 4.15keV (7%).
TOPICS: Sensors, Resolution (Optics), Engineering prototypes, Ionization energy, Energy gap, Signals, National defense, Density, Energy (Physics), Temperature, Gamma rays, Atomic number, Semiconductors (Materials), Inspection, Manufacturing
Wu Chongzhi, Zhang Jindong, Wang Zongkui, Zhang Zengkui and Fu Shenqing
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042221
In order to avoid the misuse of metal material in nuclear projects, typical cases happened in AP1000 nuclear power projects in China are analyzed. The analysis finding indicates that some cases were caused by defective procedures or undemanding processes performance, and every case is relevant with human error. It is considered that procedural management cannot completely avoid the misuse of metal material when it is caused by human error, and spectrometry analysis is suggested to reexamine the material of key components.
TOPICS: Spectroscopy, Errors, Metals, Nuclear power, China
Technology Review  
Igor Pioro, Romney Duffey, Pavel Kirillov, Roman Pioro, Alexander Zvorykin and Rachid Machrafi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042194
It is well known that electrical-power generation plays the key role in advances in industry, agriculture, technology, and the standard of living. Also, strong power industry with diverse energy sources is very important for a country independence. In general, electrical energy can be mainly generated from: 1) non-renewable energy sources (75.5% of the total electricity generation) such as coal (38.3%), natural gas (23.1%), oil (3.7%), and nuclear (10.4%); and 2) renewable energy sources (24.5%) such as hydro, biomass, wind, geothermal, solar, and marine power. Today, the main sources for electrical-energy generation are: 1) thermal power (61.4%) - primarily using coal and secondarily using natural gas; 2) "large" hydro-electric plants (16.6%); and 3) nuclear power (10.4%). The balance of the energy sources (11.6%) is from using oil, biomass, wind, geothermal and solar, and have visible impact just in a few countries. This paper presents the current status of electricity generation in the world, various sources of industrial electricity generation, and role of nuclear power with a comparison of nuclear-energy systems to other energy systems. A comparison of the latest data on electricity generation with those several years old shows that world usage of coal, gas, nuclear, and oil has decreased by 1 ? 2%, but usage of renewables has increased by 1% for hydro, and 2% for other renewable sources.
TOPICS: Nuclear power, Electric power generation, Coal, Energy resources, Natural gas, Solar energy, Geothermal engineering, Biomass, Renewable energy sources, Wind, Fossil fuels, Ocean energy, Electricity (Physics), Thermal energy, Energy / power systems, Energy industry
Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka and Hiroyuki Ohshima
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042191
In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. Evaluations of thermal-hydraulics in FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, thermal-hydraulics in FAIDUS is investigated with the in-house subchannel analysis code named ASFRE, which can be applied to a wire-wrapped fuel pin bundle using a distributed resistance model (DRM) and a turbulence mixing model. Before analysis of the FAIDUS, validations of the DRM and the turbulence mixing model were performed respectively in comparisons with the pressure drop coefficients through the simulated FAs obtained by water experiments and the temperature distribution in the simulated FAs obtained by sodium experiments. After confirmation of applicability of ASFRE to FAs through these validations, thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct were conducted for comparison. The numerical results indicated that significant asymmetric temperature distribution did not occur in FAIDUS compared to the distribution in the typical FA at a high flow rate condition. In addition, it was shown that the temperature distribution in FAIDUS was similar to that in the typical FA at a low flow rate condition, because the local flow acceleration and the flow redistribution caused by the buoyancy force was much effective in FAIDUS.
TOPICS: Fuels, Manufacturing, Ducts, Thermal hydraulics, Sodium fast reactors, Flow (Dynamics), Temperature distribution, Design, Turbulence, Safety, Buoyancy, Pressure drop, Sodium, Wire, Accidents, Water
Kevin Fernández-Cosials, Alfonso Barbas Espa and Jose Garcia Laruelo
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042192
Nuclear energy has created controversy since its conception during the 50's. Arguments against it have been constant through the years until the current state, on which the majority of western societies are against it as seen in recent surveys. Additionally, confidence in science and scientists is also relatively low. It is in this context that Jóvenes Nucleares appeared: an organization created by the Spanish Nuclear Society and formed by young people interested in nuclear energy. One of its main goals was the spreading of nuclear science into society. This was made through lectures at high schools, content creation and enveloping communication campaigns. A critical approach was always taken, trying to separate from the lobby argumentation, and promoting a strong critical thought. In this paper, as an example of a communication campaign, the Basic Course of Nuclear Fusion is presented. This campaign involved creating the informative content, gathering it into a book, the development of a lecture to be delivered in universities or high schools, and a strong advertisement effort through social media and presentations in congresses. This campaign has been possible thanks to voluntary work, being the main cost of the campaign the printing of the book. The early results predict a great support to this new format included into the Jóvenes Nucleares divulgation activities as perceived in the attendance and feedback provided by the audience. With these activities, Jóvenes Nucleares aspires to put another grain of sand towards narrowing the gap between science and society.
TOPICS: Nuclear fusion, Nuclear power, Printing, Sands, Feedback
Hernán A. Castro, Raul A. Rodriguez, Vittorio Luca and Hugo L. Bianchi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042193
Treatment and conditioning of spent ion exchange resins from nuclear facilities is a complex process that not only should contemplate obtaining a stable product suitable for long-term storage and/or disposal but also have to take into account the treatment of secondary currents generated during the process. The combination of low temperature pyrolysis treatment and high performance plasma treatment (HPPT) of the off-gas generated could be a novel solution for organic matrix nuclear wastes with economic and safety advantages. In the present work results of lab scale studies associated to the pyrolysis off-gas characterization and the performance and operating parameters influence on the removal of model compounds in a laboratory-scale flow reactor, using inductively coupled plasma under sub-atmospheric conditions, are shown. The pyrolysis off-gas stream was largely characterized and the evolution of main compounds of interest as function of temperature process was established. The results of plasma assays with model compound demonstrate a high destruction and removal efficiency (>99.9%) and a good control over the final gas products. First results of a bench scale arrangement combining both processes are presented and bode well for the application of this combined technology.
TOPICS: Ion exchange, Plasmas (Ionized gases), Pyrolysis, Resins, Storage, Flow (Dynamics), Temperature, Low temperature, Currents, Nuclear power stations, Radioactive wastes, Safety, Assaying
Akihiro Mano, Yoshihito Yamaguchi, Jinya Katsuyama and Yinsheng Li
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042115
Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants. In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, we improved PASCAL-SP considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate, and probability of crack detection. By using the improved version of PASCAL-SP, we numerically evaluated the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC. Moreover, we evaluated the influence of leak detection and non-destructive examination on failure probabilities. Based on the obtained numerical results, we concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.
TOPICS: Fracture mechanics, Fracture (Materials), Probability, Pipes, Failure, Pressurized water reactors, Welded joints, Uncertainty, Water, Leakage, Nuclear power, Nuclear power stations, Crack detection, Nickel, Alloys, Stress corrosion cracking, Nondestructive evaluation
Arnold Gad-Briggs, Pericles Pilidis and Theoklis Nikolaidis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042116
The Simple Cycle Recuperated (SCR) and Intercooled Cycle Recuperated (ICR) are highly efficient Brayton helium gas turbine cycles, designed for the Gas-cooled Fast Reactor (GFR) and Very-High-Temperature Reactor (VHTR) Generation IV (Gen IV) Nuclear Power Plants (NPPs). This paper documents risk analyses which considers technical and economic aspects of the NPP. The sensitivity analyses are presented that interrogate the plant design, performance and operational schedule and range from component efficiencies, system pressure losses, operating at varied power output due to short term load-following or long term reduced power operations to prioritise other sources such as renewables. The sensitivities of the economic and construction schedule are also considered in terms of the discount rates, capital and operational costs and increased costs in Decontamination and Decommissioning (D&D) activity due to changes in the discount rates. This was made possible by using a tool designed for this study to demonstrate the effect on the 'non-contingency' baseline Levelised Unit Electricity Cost (LUEC) of both cycles. The SCR with a cycle efficiency of 50%, has a cheaper baseline LUEC of $58.41/MWh in comparison to the ICR (53% cycle efficiency), which has a LUEC of $58.70/MWh. However, the cost of the technical and economic risks is cheaper for the ICR resulting in a final LUEC of $70.45/MWh (ICR) in comparison to the SCR ($71.62/MWh) for the year 2020 prices.
TOPICS: Cycles, Helium, Nuclear power stations, Gas turbines, Financial risk, Very high temperature reactors, Fast neutron reactors, Risk analysis, Renewable energy sources, Sensitivity analysis, Plant design, Nuclear decommissioning, Pressure, Decontamination, Construction, Stress
Ashok Kumar and Anindya Chatterjee
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042117
The thin shell design code RCC-MR is used for fast breeder reactor components operating at high temperatures. Thin shells from such applications can be designed using linear elastic buckling analysis, following procedures given in RCC-MR. For human safety, such procedures can and should be examined by the broader scientific community. Among such procedures, RCC-MR provides three alternative methods to quantify an imperfection value; and that value is used in subsequent calculations to determine safe loads. Of these methods, the third seems nonconservative. Here we examine that third method using detailed numerical examples. These examples, found by trial and error, are the main contribution of this paper. The first example is a nonuniform cylindrical shell closed with a spherical endcap, under external pressure. The second is a cylinder with an ellipsoidal head under internal pressure. The third is an L-shaped pipe with an end load. In all three cases, the new computed imperfection quantity is found to be surprisingly small compared to the actual value used for computations (e.g., 25 times smaller), and in two cases the result is insensitive to the actual imperfection. We explain how the three examples "trick" the imperfection quantification method in three different ways. We suggest that this imperfection quantification method in RCC-MR should be reexamined. The primary value of our paper lies not in new mechanics, but in identifying unexpected ways in which a particular step in shell design using RCC_MR could be nonconservative.
TOPICS: Design, Buckling, Thin shells, Stress, Pipes, Pressure, Safety, Computation, Cylinders, Errors, External pressure, Shells, Breeder reactors, High temperature
Ganesh Lal Kumawat, Anuj Kumar Kansal, N. K. Maheshwari and Avaneesh Sharma
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042120
The clearance between fuel rods is maintained by spacer grid or helical wire wrap. Thermal-hydraulic characteristics inside fuel rod bundle are strongly influenced by the spacer grid geometry and the bundle pitch-to-diameter (P/D) ratio. This includes the maximum fuel temperature, critical heat flux, as well as pressure drop through the fuel bundle. An understanding of the detailed structure of flow mixing and heat transfer in a fuel rod bundle geometry is therefore an important aspect of reactor core design, both in terms of the reactor's safe and reliable operation, and with regard to optimum power extraction. In the present study, computational fluid dynamics (CFD) simulations are performed to investigate isothermal turbulent flow mixing and heat transfer behaviour in 4×4 rod bundle with twist-vane spacer grid with P/D ratio of 1.35. This work is carried out under International Atomic Energy Agency (IAEA) co-ordinated research project titled as "Application of Computational Fluid Dynamics (CFD) Codes for Nuclear Power Plant Design". CFD simulations are performed using open source CFD code OpenFOAM. Numerical results are compared with experimental data from Korea Atomic Energy Research Institute (KAERI) and found to be in good agreement.
TOPICS: Heat transfer, Flow (Dynamics), Simulation, Computational fluid dynamics, Fuel rods, Geometry, Nuclear power, Fuels, Turbulence, Wire, Clearances (Engineering), Temperature, Pressure drop, Critical heat flux, Nuclear design, Design
Felix Boldt
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042118
During operation of light water reactors, the Zircaloy fuel rod cladding is susceptible for hy-drogen uptake. When the local solubility limit of hydrogen in Zircaloy is reached additional hy-drogen precipitates as zirconium hydride, which affects the ductility of the fuel rod cladding. Especially the radially aligned hydrides enhance embrittlement, while circumferential (azimuthal) hydrides have a less detrimental effect. In this work the influence of high temperatures during the dry storage period on hydride dissolution and precipitation is demonstrated. Therefore, in a descriptive example scenario being discussed, the simulation of a limited heat removal from the cask will heat up the dry storage cask for days and causes dissolution of hydrides in the cladding. Depending on the threshold stress for reorientation the following cooldown results on different hydride precipitation behaviour. The threshold stress lead to an enhanced or delayed precipitation of radial hydrides. The GRS fuel rod code TESPA-ROD is equipped with a new model for hydrogen solubility and applied to long-term storage transients. In this article hydride refers to zirconium hydrides formed inside the fuel rod cladding.
TOPICS: Stress, Hydrogen, Precipitation, Fuel rods, Cladding systems (Building), Storage, Zirconium, Heat, Simulation, Transients (Dynamics), Ductility, High temperature, Light water reactors, Embrittlement
Ignacio Gómez and Laurent Laborde
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042119
In the event of a loss of integrity of the Main Coolant Line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the ASTEC V2.1integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: first, with the 5 equation approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; then, with the recently implemented 6 equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: • The use of 5 equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 (kg/cm2/s) and the gas titles below 5.93*10^(-4). Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. • The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas titles. However, the qualitative prediction for high gas titles still shows some deviations due to the decrease of the regular friction term.
TOPICS: Friction, Momentum, Flux (Metallurgy), Simulation, Coolants, Differential equations, Engineering simulation, Circuits, Test facilities, Containment, Algebra, Flow (Dynamics), Fluids, Pressure

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