Accepted Manuscripts

Tsuyoshi Sasagawa, Taiji Chida and Yuichi Niibori
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037163
Cementitious materials for the construction of a geological repository of radioactive waste alter the pH of groundwater to a highly alkaline condition (pH?13). While this alkaline groundwater dissolves silicate minerals, the soluble silicic acid polymerizes or deposits on the surface of rock with the decrease in pH by mixing with the surrounding groundwater (pH=8). In particular, the deposition of silicic acid leads to a clogging effect in flow-paths, which retards the migration of radionuclides. This study estimated the clogging of silicic acid in flow-paths with the one-dimensional advection-dispersion model considering the deposition rate constants evaluated in our previous study. As some of the most important parameters, these estimations focused on the initial supersaturated concentration of silicic acid and the density of deposited minerals. As a result, the aperture of flow-paths (initial width: 0.1 mm, flow-rate: 5 m/year, initial supersaturated concentration: 0.01, 0.1 and 1.0 mM) was almost clogged within about 200 years by the deposition of silicic acid. The period for the clogging became shorter under the conditions of higher initial supersaturated concentration and lower density of deposited minerals. In other words, the use of cementitious materials for constructing the repository might produce a retardation effect of radionuclide migration by the deposition/clogging processes of the supersaturated silicic acid.
TOPICS: Density, Flow (Dynamics), Radioactive wastes, Radioisotopes, Cements (Building materials), Construction, Rocks, Groundwater, Minerals
G.A. Ferrier, David Kerr, Joseph Metzler, Evan Veryard, M. Farahani, P.K. Chan, Mark R. Daymond and E.C. Corcoran
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037115
During normal operation in CANDU® reactors, the Stress Corrosion Cracking (SCC) of fuel sheathing is mitigated effectively, in part, using a thin graphite-based coating known as CANLUB. Mechanisms typically proposed for the demonstrated SCC mitigation offered by CANLUB include lubrication and/or chemical interactions. An additional possibility, that was recently suggested, involves the sequestering of iodine through its interaction with alkali metal and/or alkaline earth metal impurities in the CANLUB coating. This possibility is supported by the systematic analysis and testing in this paper wherein three prevalent impurities (Na, Ca, and Mg) found in CANLUB were incorporated into SCC slotted ring experiments as metal oxides. When the molar concentration of metal oxide (Na2O, CaO, or MgO) matched or exceeded the molar concentration of iodine (6 mmol = 16 mg/cm3), Na2O and CaO protected the rings from corrosion whereas MgO enhanced their corrosion. When Zircaloy-4 sheathing is subjected to mechanical stress, high temperature, and high concentrations of iodine vapour, it is better protected by siloxane coatings than by graphite-CANLUB coatings. Consequently, since metal impurities (Na, Ca, and Mg) are found more abundantly in siloxane coatings than in graphite-CANLUB coatings, Zircaloy-4 slotted rings were coated with graphite-CANLUB containing Na, Ca, and/or Mg at those more abundant concentrations. Since these concentrations remain below 6 mmol, SCC test results suggest that the siloxane's superior adhesion is an essential first step in preventing corrosion induced by 6 mmol of iodine.
TOPICS: Fuels, Stress corrosion cracking, Metal impurities, Alkali metals, Coatings, Graphite, Corrosion, Metals, Siloxanes, High temperature, Lubrication, Adhesion, Testing, Stress
Technology Review  
Lembit Sihver and Nakahiro Yasuda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037116
In this paper the causes and the radiological consequences of the Chernobyl and Fukushima Daiichi nuclear accidents is discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating Nuclear Power Plants (NPPs). In addition to that, improved radiation release barriers, including low corrosive fuel and cladding, should be developed. Stress tests should on a regular basis be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, nuclear fuel performance, decontamination and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will also be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well prepared and well established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will be presented together with training and education program, in order to ensure that professional rescue workers, including medical staff, fire fighters, police, etc., as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.
TOPICS: Accidents, Fukushima nuclear disaster, Japan, 2011, Chernobyl Nuclear Accident, Chornobyl, Ukraine, 1986, Safety, Nuclear power stations, Emergency response, Emergency management, Accident management, Radiation (Physics), Evacuations, Cladding systems (Building), Fire, Ionizing radiation, Fuels, Stress, Natural disasters, Failure, Nuclear power, Education, Nuclear fuels, Biomedicine, Decontamination
Matthew Weathered, Jordan Rein, Mark Anderson, Paul Brooks and Bryan Coddington
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037118
This study characterizes the magnitude, spatial profile, and frequency spectrum of thermal striping at a junction using a novel sodium deployable optical fiber temperature sensor. Additionally, this study reveals for the first time the capability of performing cross correlation velocimetry with an optical fiber to acquire fluid flow rates in a pipe. Optical fibers were encapsulated in stainless steel capillary tubes with an inert cover gas for high temperature sodium deployment. Plots of temperature oscillation range as a function of two dimensional space highlighted locations prone to mechanical failure for particular flow momentum ratios. The power spectral density revealed that these temperature oscillations occurred at frequencies ranging from 0.1-6 Hz. Finally, the bulk flow rate of liquid sodium was calculated from thermal striping's periodic temperature oscillations using cross correlation velocimetry for flow rates of 0.25-5.76 L/min.
TOPICS: Fiber optic sensors, Sodium, Optical fiber, Oscillations, Flow (Dynamics), Temperature, Spectral energy distribution, Pipes, Failure, Momentum, Fluid dynamics, Junctions, Stainless steel, Temperature sensors, High temperature
Xu Cheng and Xiaojing Liu
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037117
Supercritical fluids (SCFs) become more and more important in various engineering applications. In nuclear power systems, SCFs are considered as coolant of the reactor core such as the supercritical water cooled reactor (SCWR), superconducting magnets and blankets in the fusion reactors or as fluid in the energy conversion systems of the next generation nuclear reactors. Accurate determination of heat transfer and the temperature of the structural material (e.g. fuel rod cladding) is of crucial importance for the system design. Thus, extensive studies on heat transfer to SCFs have been carried out in the past five decades and are still ongoing worldwide. However, no breakthrough is recognized or expected in the near future. In this paper, the status, main challenges and future R&D needs are briefly reviewed. Three aspects are taken into consideration, i.e. experimental studies, numerical analysis and model development for the prediction of heat transfer coefficient. Several key challenges and also the important subjects of the future R&D needs are identified. They are (a) data base for turbulence quantities, (b) multi solution of wall temperature, (c) extensive RANS method and (d) new prediction method for heat transfer coefficient (HTC).
TOPICS: Supercritical fluids, Heat transfer, Heat transfer coefficients, Supercritical water reactors, Fuel rods, Temperature, Fluids, Fusion reactors, Turbulence, Superconducting magnets, Coolants, Energy conversion, Cladding systems (Building), Design, Engineering systems and industry applications, Numerical analysis, Databases, Nuclear reactors, Model development, Nuclear power, Reynolds-averaged Navier–Stokes equations, Wall temperature, Water
Chuanqi Zhao, Kunpeng Wang, Liangzhi Cao and Youqi Zheng
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037119
Burnable poison (BP) is used to control excess reactivity in Supercritical Water Cooled Reactor (SCWR). It helps reducing the number of control rods. Overall BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP containments. Core performance with and without BP is compares. The results have shown that the core radial power peaking factor decreases by introducing BP. It is also shown that the core axial power peaking factor increases and the power peak moves towards the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum, and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.
TOPICS: Fuels, Design, Water, Manufacturing, Supercritical water reactors, Temperature, Rods, Cladding systems (Building), Safety
Josef Hasslberger, Peter Katzy, Lorenz R. Boeck and Thomas Sattelmayer
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037094
For the purpose of nuclear safety analysis, a reactive flow solver has been developed to determine the hazardous potential of large-scale hydrogen explosions. Without using empirical transition criteria, the whole combustion process including Deflagration-to-Detonation Transition (DDT) is computed within a single solver framework. In this paper we present massively parallelized three-dimensional explosion simulations in a full-scale pressurized water reactor of the Konvoi type. Several generic DDT scenarios in globally lean hydrogen-air mixtures are examined to assess the importance of different input parameters. It is demonstrated that the explosion process is highly sensitive to mixture composition, ignition location and thermodynamic initial conditions. Pressure loads on the confining structure show a profoundly dynamic behavior depending on the position in the containment. Computational cost can effectively be reduced through adaptive mesh refinement.
TOPICS: Explosions, Simulation, Computational fluid dynamics, Pressurized water reactors, Hydrogen, Ignition, Containment, Chemically reactive flow, Stress, Safety, Pressure, Combustion
Grant L. Hawkes, James W. Sterbentz, John Maki and Binh Pham
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037095
A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with Post Irradiation Examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program which supports the development of the Very-High-Temperature gas-cooled Reactor under the Next-Generation Nuclear Plant project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) TRISO fueled compacts were inserted into twelve separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple data. PIE-measured experimental data was used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all of the fuel compacts, graphite holders, and all of the graphite rings. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each non-fuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule.
TOPICS: Physics, Heat, Temperature, Nuclear fission, Neutrons, Fuels, Irradiation (Radiation exposure), Shrinkage (Materials), Fluence (Radiation measurement), Finite element model, Graphite, Nuclear power stations, Steady state, Thermal analysis, Thermocouples, Very high temperature reactors
Technical Brief  
Sri Budi Utami and Susetyo Hario Putero
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037078
The decommissioning of nuclear facilities in Indonesia are mainly based on Act No. 10/1997 on Nuclear Energy, Nuclear Energy Regulatory Agency Chairman Regulation (BCR) No. 4 Year 2009, and Gov. Reg. no 2 Year 2014 that cover General provisions and licensing requirements of decommissioning and the detailed requirements and guidelines for preparing the decommissioning plan and licensee applications. BCR No. 4 Year 2009 was developed based on the adoption and adaption from the IAEA Safety Guide, WS-G-2.1. Currently, one of three research reactors, TRIGA research reactor 2MW (the oldest which went critical at 250 kW in 1964, and was operated at maximum in 1971 upgrade to 2 MW in 2000), has operated for 45 years, but there is no decision for decommissioning this reactor yet. Indonesia has experience in decommissioning of the phosphoric acid purification facility of the Gresik petrochemical plant. Some aspects of decommissioning, which have been successfully addressed to date, are: regulation, communication; and decommissioning team. Development of human resources, technological capability, and Information flow from more advanced countries are important factor for the future of the nuclear facility decommissioning plan in Indonesia. Some regulations have still not anticipated all the regulatory challenges that might be encountered in the near future. More regulations and guidance are needed to be established by Nuclear Energy Regulatory Agency in order to complete the current regulations so that problems can all be anticipated.
TOPICS: Nuclear power stations, Nuclear decommissioning, Regulations, Nuclear power, Teams, Petrochemicals, Licensing, Personnel management, Flow (Dynamics), Safety
Mukesh Kumar Dhiman, Purnendra K. Verma, Arun K. Nayak and A. Rama Rao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037031
Fukushima accident has raised a strong concern and apprehension about the safety of a nuclear reactor failing to remove the decay heat following an extreme event. After Fukushima accident, the reactor designers worldwide analyzed the safety margin of the existing and new generation nuclear power plants for such an event. AHWR, designed in India, was also analyzed for even more severe conditions than occurred at Fukushima. AHWR equipped with several passive systems showed its robustness against this type of scenarios. However, few new passive systems were incorporated in AHWR design for maintaining the integrity of the reactor at least for 7 days as a grace period. Passive moderator cooling system and passive endshield cooling system were among the newly introduced safety system in AHWR. An experimental test facility simulating the prolonged SBO case in AHWR has been designed and built. Experiments have been performed in the test facility for simulated conditions of prolonged SBO. The current study shows the performance behavior of AHWR during prolonged SBO case through simulation in its integral test facility. The results indicate that AHWR design is capable of removing decay heat for prolonged period without operator interference.
TOPICS: Safety, Test facilities, Fukushima nuclear disaster, Japan, 2011, Heat, Cooling systems, Design, Nuclear reactors, Nuclear power stations, Robustness, Simulation
Peter Katzy, Josef Hasslberger, Lorenz R. Boeck and Thomas Sattelmayer
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036984
The presented work aims to improve CFD explosion modeling for lean hydrogen-air mixtures on under-resolved grids. Validation data is obtained from an entirely closed laboratory scale explosion channel (GraVent facility). Investigated hydrogen-air concentrations range from 6 to 19 vol.-%. Initial conditions are p = 1 atm and T = 293 K. Two highly time-resolved optical measurement techniques are applied simultaneously: (1) 10 kHz shadowgraphy captures line-of-sight integrated macroscopic flame propagation; and (2) 20 kHz OH-PLIF (planar laser-induced fluorescence of the OH radical) resolves microscopic flame topology without line-of-sight integration. This paper presents the experiment, measurement techniques, data evaluation methods and simulation results. The evaluation methods encompass the determination of flame tip velocity over distance and a detailed time-resolved quantification of flame topology based on OH-PLIF images. One parameter is the length of wrinkled flame fronts in the OH-PLIF plane obtained through automated post-processing. It reveals the expected enlargement of flame surface area by instabilities on microscopic level. A strong effect of mixture composition is observed. Simulations based on the new model formulation, incorporating the microscopic enlargement of the flame front, show a promising behavior, where the impact of the augmented flame front on the observed flame front velocities can be detected.
TOPICS: Combustion, Simulation, Flames, Hydrogen, Topology, Evaluation methods, Explosions, Lasers, Optical measurement, Fluorescence, Computational fluid dynamics, Modeling, Simulation results
Yusuke Fujiwara, Takahiro Nemoto, Daisuke Tochio, Masanori Shinohara, Masato Ono and Shoji Takada
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036985
In the High Temperature Engineering Test Reactor (HTTR), the Vessel Cooling System (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV and the biological shielding concrete were quite slow during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C. The temperature increase of water cooling tube panel of VCS was calculated about 15 °C which is sufficiently small in the view point of property protection. As the results, it was confirmed that the cooling capacity of VCS can be kept sufficiently even in case that one of two water cooling lines of VCS loses its function.
TOPICS: Cooling, Cooling systems, Vessels, High temperature, Temperature, Heat, Water, Concretes, Control systems, Temperature control, Thermal energy, Coolants, Feedback, Helium, Reactor vessels
Alex Matev
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036986
Many researchers are investigating the potential of lead-bismuth cooled fast reactors for producing electricity, as well as for the safe transmutation of minor actinides and the nuclear incineration of long-lived fission products. The paper presents the results from simulating with the RELAP5-3D code of natural circulation in a generic design of a pool-type nuclear reactor with lead-bismuth eutectic alloy (LBEA) as a primary, and water/steam as a secondary coolant. The simulation results provide valuable insights in the evolution of key reactor safety-relevant phenomena and support also the qualified use of system analysis codes as RELAP5-3D for the simulation of transients in pool-type reactor systems.
TOPICS: Simulation, Coolants, Eutectic alloys, Fast neutron reactors, Nuclear reactors, Simulation results, Steam, Water, Transients (Dynamics), Design, Nuclear fission, Systems analysis, Safety
Shinichiro Uesawa, Yasuo Koizumi, Mitsuhiko Shibata and Hiroyuki Yoshida
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036987
Seawater was injected into reactor cores in the accident at the Fukushima Daiichi Nuclear Power Station. Saturated pool nucleate boiling heat transfer experiments of NaCl solution, natural seawater and artificial seawater as well as distilled water were performed to examine the effect of salts on boiling heat transfer. The heat transfer surface was made of a printed coper circuit board. A boiling state was recorded with a high speed video camera. A surface temperature distribution was measured with an infrared camera. The concentration of the NaCl solution and the artificial sweater was varied in the range of 3.5 ~ 10 wt% in the experiments. Boiling curves were well predicted with the Rohsenow correlation although large coalescent bubble formation was suppressed in NaCl, natural seawater and artificial seawater experiments. Deposit of calcium sulfate CaSO4 on the heat transfer surface was observed in the experiments of artificial seawater. This deposit layer formation resulted in the initiation of slow heat transfer surface temperature excursion at a lower heat flux than a usual critical heat flux. A unique relation was confirmed between the concentration of artificial seawater in bulk fluid and the vaporization rate on the heat transfer surface at which the slow heat transfer surface temperature excursion was initiated. This relation suggested that if the bulk concentration of sea salts in seawater exceeded 11 wt%, the deposition of calcium sulfate on the heat transfer surface might occur even if a heat flux was zero.
TOPICS: Heat transfer, Seas, Nucleate boiling, Seawater, Boiling, Temperature, Heat flux, Critical heat flux, Printed circuit boards, Fukushima nuclear disaster, Japan, 2011, Nuclear power stations, Temperature distribution, Video cameras, Water, Fluids, Bubbles, Accidents
Arnold Gad-Briggs, Pericles Pilidis and Theoklis Nikolaidis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036983
The control system for Generation IV Nuclear Power Plant (NPP) design must ensure load variation when changes to critical parameters affect grid demand, plant efficiency and component integrity. The objective of this study is to assess the load following capabilities of cycles when inventory pressure control is utilised. Cycles of interest are Simple Cycle Recuperated (SCR), Intercooled Cycle Recuperated (ICR) and Intercooled Cycle without recuperation (IC). Firstly, part power performance of the IC is compared to results of the SCR and ICR. Subsequently, the load following capabilities are assessed when the cycle inlet temperature is varied. This was carried out using a tool designed for this study. Results show that the IC takes ~2.7% longer than the ICR to reduce the power output to 50% when operating in Design Point (DP) for similar valve flows, which correlates to the volumetric increase for the IC inventory storage tank. However, the ability of the IC to match the ICR’s load following capabilities is severely hindered because the IC is most susceptible to temperature variation. Furthermore, the IC takes longer than the SCR and ICR to regulate the reactor power by a factor of 51 but this is severely reduced, when regulating NPP power output. However, the IC is the only cycle that does not compromise reactor integrity and cycle efficiency when regulating the power. The analyses intend to aid the development of cycles specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.
TOPICS: Stress, Gas turbines, Cycles, Helium, Nuclear power stations, Very high temperature reactors, Design, Temperature, Control systems, Coolants, Flow (Dynamics), Valves, Fast neutron reactors, Storage tanks, Pressure control
Zhifei Yang, Yali Chen and Hu Luo
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036892
To respond the urgent needs of verification, training and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under Power Condition, Shutdown Condition, Spent Fuel Pool Condition, and Refueling Conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis program MAAP5 (Modular Accident Analysis Program 5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination and verification.
TOPICS: Drills (Tools), Industrial research, Accident management, Accidents, Spent nuclear fuels, Public utilities
Oriol Costa Garrido, Samir El Shawish and Leon Cizelj
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036736
Large sets of fluid temperature histories and a recently proposed thermal fatigue assessment procedure are employed in this paper to deliver more accurate statistics of predicted lives of pipes and their uncertainties under turbulent fluid mixing circumstances. The wide variety of synthetic fluid temperatures, generated with an improved spectral method, results in a set of estimated distributions of fatigue lives through linear one-dimensional heat diffusion, thermal stress estimates and fatigue assessment codified rules. The results of the fatigue analysis indicate that, in order to avoid the inherent uncertainties due to comparatively short fluid temperature histories to the estimated fatigue lives, a conservative safe design against thermal fatigue could be attempted with the lower bounds of the predicted life distributions, such as the 5% failure probability life (5% of samples fail). The impact of the convection heat transfer coefficient on the predictions is also studied in a sensitivity analysis. This represents a detailed attempt to correlate the uncertainties in the physical fluid mixing conditions and heat transfer to the estimated fatigue life using spectral methods.
TOPICS: Fluids, Turbulence, Pipes, Fatigue life, Fatigue, Temperature, Uncertainty, Statistics as topic, Probability, Sensitivity analysis, Thermal diffusion, Heat transfer, Failure, Fatigue analysis, Thermal stresses, Convection, Design
A. Gad-Briggs, P. Pilidis and T. Nikolaidis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036737
An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. Firstly, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modelling and performance simulation tool designed for this study. Results show that the ICR takes ~102% longer than the SCR to reduce the load to 50% in Design Point (DP) performance conditions for similar valve flows, which correlates to the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.
TOPICS: Control systems, Gas turbines, Cycles, Helium, Nuclear power stations, Stress, Design, Very high temperature reactors, Fast neutron reactors, Flow (Dynamics), Modeling, Valves, Coolants, Transients (Dynamics), Simulation
Hiroo Kondo, Takuji Kanemura, Tomohiro Furukawa, Yasushi Hirakawa, Eiichi Wakai and Juan Knaster
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036513
A liquid Li jet flowing at 15 m/s under a high vacuum of 10^-3 Pa is intended to serve as a beam target (Li target) in the planned International Fusion Materials Irradiation Facility (IFMIF). The engineering validation and engineering design activities (EVEDA) for the IFMIF are being implemented under the Broader Approach agreement. As a major activity of the Li target facility, the EVEDA Li test loop was constructed and a stable Li target was demonstrated. This study focuses on a cavitation-like acoustic noise detected in a downstream conduit where the Li target flowed under vacuum conditions. This noise was investigated using acoustic-emission (AE) sensors installed via acoustic wave guides. The sound intensity of the noise was examined against the cavitation number of the Li target. In addition, fast Fourier transform (FFT) and continuous wavelet transform (CWT) were performed to characterize the acoustic noise. Owing to the acoustic noise’s intermittency, high frequency, and the dependence on cavitation number, we conclude that this acoustic noise is generated when cavitation bubbles collapse. The location of the cavitation was fundamental for presuming the mechanism. In this study, the propagation of acoustic waves was used to localize the cavitation and a method to determine the location of cavitation was formulated. As a result, we found that cavitation occurred only in a narrow area where the Li target impinged on the downstream conduit; therefore, we concluded that this cavitation was induced by the impingement.
TOPICS: Irradiation (Radiation exposure), Cavitation, Lithium, Acoustics, Noise (Sound), Vacuum, Sensors, Waveguides, Wavelet transforms, Acoustic emissions, Acoustic intensity, Collapse, Fast Fourier transforms, Waves, Bubbles, Engineering design
Yan Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4032312
In modeling and simulation, model-form uncertainty arises from the lack of knowledge and simplification during modeling process and numerical treatment for ease of computation. Traditional uncertainty quantification approaches are based on assumptions of stochasticity in real, reciprocal, or functional spaces to make them computationally tractable. This makes the prediction of important quantities of interest such as rare events difficult. In this paper, a new approach to capture model-form uncertainty is proposed. It is based on fractional calculus, and its flexibility allows us to model a family of non-Gaussian processes, which provides a more generic description of the physical world. A generalized fractional Fokker-Planck equation (fFPE) is used to describe the drift-diffusion processes under long-range correlations and memory effects. A new model calibration approach based on the maximum mutual information is proposed to reduce model-form uncertainty, where an optimization procedure is taken.

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