Showing papers in "Nuclear Engineering and Technology in 2008"
TL;DR: In this paper, the optimal means of treating spent metal fuels from metal fast fuel reactors is proposed as a potential option for GNEP in order to meet the requirements of the next generation fuel cycle.
Abstract: Pyroprocessing is the optimal means of treating spent metal fuels from metal fast fuel reactors and is proposed as a potential option for GNEP in order to meet the requirements of the next generation fuel cycle. Currently, efforts for research and development are being made not only in the U.S., but also in Asian countries. Electrorefining, cathode processing by distillation, injection casting for fuel fabrication, and waste treatment must be verified by the use of genuine materials, and the engineering scale model of each device must be developed for commercial deployment. Pyroprocessing can be effectively extended to treat oxide fuels by applying an electrochemical reduction, for which various kinds of oxides are examined. A typical morphology change was observed following the electrochemical reduction, while the product composition was estimated through the process flow diagram. The products include much stronger radiation emitter than pure typical LWR Pu or weapon-grade Pu. Nevertheless, institutional measures are unavoidable to ensure proliferation-proof plant operations. The safeguard concept of a pyroprocessing plant was compared with that of a PUREX plant. The pyroprocessing is better adapted for a collocation system positioned with some reactors and a single processing facility rather than for a centralized reprocessing unit with a large scale throughput.
143 citations
TL;DR: In this article, two fuel channel designs were proposed for the CANDU-SCWR, a pressure-tube type supercritical water cooled reactor, where each pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator.
Abstract: This paper presents two of the fuel channel designs being considered for the CANDU-SCWR, a pressure-tube type supercritical water cooled reactor. The first is an insulated pressure tube design. The pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator. Each pressure tube is in direct contact with the moderator, which operates at an average temperature of about . The low temperature allows zirconium alloys to be used. A perforated metal liner protects the insulator from being damaged by the fuel bundles and erosion by the coolant. The coolant pressure is transmitted through the perforated metal liner and insulator and applied directly to the pressure tube. The second is a re-entrant design. The fuel channel consists of two concentric tubes, and a calandria tube that separates them from the moderator. The coolant enters between the annulus of the two concentric fuel channel tubes, then exits the fuel channel through the inner tube, where the fuel bundles reside. The outer tube bears the coolant pressure and its temperature will be the same as the coolant inlet temperature, . Advantages and disadvantages of these designs and the material requirements are discussed.
101 citations
TL;DR: A conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed in this article, where the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process was evaluated.
Abstract: In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of -ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.
77 citations
TL;DR: In this paper, the general corrosion behavior of austenitic and ferritic steels (316L, 304, N controlled 304L, and 410) in supercritical water is investigated.
Abstract: The general corrosion behavior of austenitic and ferritic steels(316L, 304, N controlled 304L, and 410) in supercritical water is investigated in this paper. After exposure to deaerated supercritical water at /25 MPa for up to 500 h, the four steels studied were characterized using gravimetry, scanning electron microscopy/energy dispersive X-ray spectroscopy(SEM/EDS), X-ray photoelectron spectroscopy(XPS), and X-ray diffraction(XRD). The results show that the 316L steel with a higher Cr and Ni content has the best corrosion-resistance performance among the steels tested. In addition to the oxide layer mixed with and that formed on all the samples, a loose outer layer was observed on the 410 steel. The corrosion mechanism of stainless steels in supercritical water is discussed based on the above results.
65 citations
TL;DR: In this paper, a new nuclear core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone, which is similar to that of a conventional thermal supercritical water-cooled reactor (SCWR) core with two fuel pin rows between the moderator channels.
Abstract: In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor (SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.
60 citations
TL;DR: In this article, the authors describe the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of equipment, and recent developments at the cathode processor.
Abstract: As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.
57 citations
TL;DR: In this article, the authors performed heat transfer experiments in an annulus passage using SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), which was constructed at KAERI(Korea Atomic Energy Research Institute), to investigate the heat transfer behaviors of supercritical.
Abstract: Heat transfer experiments in an annulus passage were performed using SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation), which was constructed at KAERI(Korea Atomic Energy Research Institute), to investigate the heat transfer behaviors of supercritical . was selected as the working fluid to utilize its low critical pressure and temperature when compared with water. The mass flux was in the range of 400 to 1200 and the heat flux was chosen at rates up to 150 . The selected pressures were 7.75 and 8.12 MPa. At lower mass fluxes, heat transfer deterioration occurs if the heat flux increases beyond a certain value. Comparison with the tube test results showed that the degree of heat transfer deterioration in the heat flux was smaller than that in the tube. In addition, the Nusselt number correlation for a normal heat transfer mode is presented.
45 citations
TL;DR: In this article, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling, and fine oxide particles were observed in the matrix, and their chemical formulations were determined to be YSi2O7 and TiO.
Abstract: Based on a composition of 99.4 wt% AISI 316L stainless steel, 0.3 wt% Ti and 0.3 wt% Y2O3, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling. Fine oxide particles were observed in the matrix, and their chemical formulations were determined to be Y2Si2O7 and TiO. Heat treatment of the cold-rolled sample at 1200°C induced an isotropic tensile behavior at room temperature and at 700°C This result would be mainly attributed to the equiaxed grains that form as a result of the heat treatment for recrystallization.
39 citations
TL;DR: In this paper, the authors extensively review the condition monitoring techniques using empirical models in an effort to reduce or eliminate unexpected downtimes in general industry, and to illustrate the feasibility of applying them to the nuclear industry.
Abstract: The purpose of this paper is to extensively review the condition monitoring (CM) techniques using empirical models in an effort to reduce or eliminate unexpected downtimes in general industry, and to illustrate the feasibility of applying them to the nuclear industry. CM provides on-time warnings of system states to enable the optimal scheduling of maintenance and, ultimately, plant uptime is maximized. Currently, most maintenance processes tend to be either reactive, or part of scheduled, or preventive maintenance. Such maintenance is being increasingly reported as a poor practice for two reasons: first, the component does not necessarily require maintenance, thus the maintenance cost is wasted, and secondly, failure catalysts are introduced into properly working components, which is worse. This paper first summarizes the technical aspects of CM including state estimation and state monitoring. The mathematical background of CM is mature enough even for commercial use in the nuclear industry. Considering the current computational capabilities of CM, its application is not limited by technical difficulties, but by a lack of desire on the part of industry to implement it. For practical applications in the nuclear industry, it may be more important to clarify and quantify the negative impact of unexpected outcomes or failures in CM than it is to investigate its advantages. In other words, while issues regarding accuracy have been targeted to date, the concerns regarding robustness should now be concentrated on. Standardizing the anticipated failures and the possibly harsh operating conditions, and then evaluating the impact of the proposed CM under those conditions may be necessary. In order to make the CM techniques practical for the nuclear industry in the future, it is recommended that a prototype CM system be applied to a secondary system in which most of the components are non-safety grade. Recently, many activities to enhance the safety and efficiency of the secondary system have been encouraged. With the application of CM to nuclear power plants, it is expected to increase profit while addressing safety and economic issues.
39 citations
TL;DR: In this article, the authors discuss current and potential failure mechanisms of these failure mechanisms, as well as the potential failure mechanism of the failure mechanisms. But, they do not consider the effect of hydrogen (H) pickup and redistribution on the ductility of the fuel rod.
Abstract: One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.
36 citations
TL;DR: This approach uses the Architectural Analysis and Design Language to model the structural, behavioral and failure aspects of the system in a composite architecture model and then automatically generates dynamic fault trees (DFT) for comprehensive, tool-supported reliability analysis.
Abstract: Dependability-critical systems, such as digital instrumentation and control systems in nuclear power plants, necessitate engineering techniques and tools to provide assurances of their safety and reliability. Determining system reliability at the architectural design phase is important since it may guide design decisions and provide crucial information for trade-off analysis and estimating system cost. Despite this, reliability and system engineering remain separate disciplines and engineering processes by which the dependability analysis results may not represent the designed system. In this article we provide an overview and application of our approach to build architecture-based, dynamic system models for dependability-critical systems and then automatically generate dynamic fault trees (DFT) for comprehensive, tool-supported reliability analysis. Specifically, we use the Architectural Analysis and Design Language (AADL) to model the structural, behavioral and failure aspects of the system in a composite architecture model. From the AADL model, we seek to derive the DFT(s) and use Galileo's automated reliability analyses to estimate system reliability. This approach alleviates the dependability engineering - systems engineering knowledge expertise gap, integrates the dependability and system engineering design and development processes and enables a more formal, automated and consistent DFT construction. We illustrate this work using an example based on a dynamic digital feed-water control system for a nuclear reactor.
TL;DR: In this paper, the authors investigated the undesirable feedback effect between an interaction layer growth and a centerline temperature increase for dispersion fuel and found that the U-Mo particle size was a dominant factor that determined the fuel temperature during irradiation.
Abstract: Because the interaction layers that form between U-Mo particles and the Al matrix degrade the thermal properties of U-Mo/Al dispersion fuel, an investigation was undertaken of the undesirable feedback effect between an interaction layer growth and a centerline temperature increase for dispersion fuel. The radial temperature distribution due to interaction layer growth during irradiation was calculated iteratively in relation to changes in the volume fractions, the thermal conductivities of the constituents, and the oxide thickness with the burnup. The interaction layer growth, which is estimated on the basis of the temperature calculations, showed a reasonable agreement with the post-irradiation examination results of the U-Mo/Al dispersion fuel rods irradiated at the HANARO reactor. The U-Mo particle size was found to be a dominant factor that determined the fuel temperature during irradiation. Dispersion fuel with larger U-Mo particles revealed lower levels of both the interaction layer formation and the fuel temperature increase. The results confirm that the use of large U-Mo particles appears to be an effective way of mitigating the thermal degradation of U-Mo/Al dispersion fuel.
TL;DR: In this paper, the authors developed a system analysis code for the conceptual development of a DEMO-Reactor and used it to identify the necessary R&D areas, both in the physics and technology areas, for the realization of the concept.
Abstract: Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.
TL;DR: In this article, a transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain.
Abstract: The objective of the paper is to analyze the thermally induced density wave oscillations in water cooled boiling water reactors A transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain A two-phase flow was simulated with a one-dimensional homogeneous equilibrium model The model treats the boundary conditions naturally and takes into account the compressibility effect of the two-phase flow The axial variation of the heat flux profile can also be handled with the model Unlike the method of characteristics analysis, the present numerical model is computationally inexpensive in terms of time and works in a Eulerian coordinate system without the loss of accuracy The model was validated against available benchmarks The model was extended for the purpose of studying the flow-induced density wave oscillations in forced circulation and natural circulation boiling water reactors Various parametric studies were undertaken to evaluate the model's performance under different operating conditions Marginal stability boundaries were drawn for type-I and type-II instabilities in a dimensionless parameter space The significance of adiabatic riser sections in different boiling reactors was analyzed in detail The effect of the axial heat flux profile was also investigated for different boiling reactors
TL;DR: The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented in this article, which is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a hightemperature v-blender.
Abstract: The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented. This blending is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a high-temperature v-blender. A labscale system was used with non-radioactive surrogate salts to determine optimal particle size distributions and time at temperature. An engineering-scale system was then installed in the Hot Fuel Examination Facility hot cell and used to demonstrate blending of actual electrorefiner salt with zeolite. In those tests, it was shown that the results are still favorable with actinide-loaded salt and that batch size of this v-blender could be increased to a level consistent with efficient production operations for EBR-II spent fuel treatment. One technical challenge that remains for this technology is to mitigate the problem of material retention in the v-blender due to formation of caked patches of salt/zeolite on the inner v-blender walls.
TL;DR: In this paper, a 1:4 scale model of a prestressed concrete containment vessel (PCCV) using an axisymmetric model and a three-dimensional model is presented.
Abstract: This paper describes the nonlinear analyses of a 1:4 scale model of a prestressed concrete containment vessel (PCCV) using an axisymmetric model and a three-dimensional model. These two models are refined by comparison of the analysis results and with testing results. This paper is especially focused on the analysis of behavior under pressure and the temperature effects revealed using an axisymmetric model. The temperature-dependent degradation properties of concrete and steel are considered. Both geometric and material nonlinearities, including thermal effects, are also addressed in the analyses. The Menetrey and Willam (1995) concrete constitutive model with non-associated flow potential is adopted for this study. This study includes the results of the predicted thermal and mechanical behaviors of the PCCV subject to high temperature loading and internal pressure at the same time. To find the effect of high temperature accident conditions on the ultimate capacity of the liner plate, reinforcement, prestressing tendon and concrete, two kinds of analyses are performed: one for pressure only and the other for pressure with temperature. The results from the test on pressurization, analysis for pressure only, and analyses considering pressure with temperatures are compared with one another. The analysis results show that the temperature directly affects the behavior of the liner plate, but has little impact on the ultimate pressure capacity of the PCCV.
TL;DR: In this paper, an aide system for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique was developed to reduce human-induced or human-related unplanned reactor trips.
Abstract: periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1%), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human- induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.
TL;DR: The present method is the first successful application of a B DD truncation and is an efficient method to maintain a small BDD size by a BDD truncation during a BDS calculation.
Abstract: A Binary Decision Diagram (BDD) is a graph-based data structure that calculates an exact top event probability (TEP). It has been a very difficult task to develop an efficient BDD algorithm that can solve a large problem since it is highly memory consuming. In order to solve a large reliability problem within limited computational resources, many attempts have been made, such as static and dynamic variable ordering schemes, to minimize BDD size. Additional effort was the development of a ZBDD (Zero-suppressed BDD) algorithm to calculate an approximate TEP. The present method is the first successful application of a BDD truncation. The new method is an efficient method to maintain a small BDD size by a BDD truncation during a BDD calculation. The benchmark tests demonstrate the efficiency of the developed method. The TEP rapidly converges to an exact value according to a lowered truncation limit.
TL;DR: The KSTAR Magnet Power Supply (MPS) as discussed by the authors was dedicated to the SC coil commissioning and plasma experiment as a part of the system commissioning, and it showed that all of the hardware and controlling software operated well, and this result finally led to the success of SC coil and the KSTAR plasma experiment.
Abstract: The KSTAR Magnet Power Supply (MPS) was dedicated to the SC coil commissioning and plasma experiment as a part of the system commissioning. Although many efforts to develop large-current power supplies that are useful for high power electronic devices have been made in various application fields, such as for large metal-plating devices, there were clear discrepancies between conventional power supply technologies and that for the SC coils due to the special SC coil load conditions. Therefore, most of the power supply technologies for the SC coils were a challenge in the domestic research area due to their limited application. However, the MPS commissioning result showed that all of the hardware and controlling software operated well, and this result finally led to the success of SC coil commissioning and the KSTAR plasma experiment. This paper will describe key features of KSTAR MPS for the plasma experiment, and will also report the commissioning results of the magnet power supplies.
TL;DR: This paper describes the application of a software fault tree analysis (FTA) as one of the analysis techniques for a software safety analysis (SSA) at the design phase and its analysis results for the safety-critical software of a digital reactor protection system, which is called the KNICS RPS.
Abstract: This paper describes the application of a software fault tree analysis (FTA) as one of the analysis techniques for a software safety analysis (SSA) at the design phase and its analysis results for the safety-critical software of a digital reactor protection system, which is called the KNICS RPS, being developed in the KNICS (Korea Nuclear Instrumentation & Control Systems) project. The software modules in the design description were represented by function blocks (FBs), and the software FTA was performed based on the well-defined fault tree templates for the FBs. The SSA, which is part of the verification and validation (V&V) activities, was activated at each phase of the software lifecycle for the KNICS RPS. At the design phase, the software HAZOP (Hazard and Operability) and the software FTA were employed in the SSA in such a way that the software HAZOP was performed first and then the software FTA was applied. The software FTA was applied to some critical modules selected from the software HAZOP analysis.
TL;DR: In this article, a high burn-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, UO2 pellets, spacer grids, performance code, and fuel assembly tests.
Abstract: High burn-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, UO2 pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain UO2 pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to UO2 powder, were prepared by oxidizing and heat-treating scrap UO2 pellets. A UO2 pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rod- supporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea
TL;DR: In this article, the effect of thermal stratification on the structural integrity of the surge line of a nuclear power plant pipe was investigated, and the response characteristics for various types of top-to-bottom temperature differentials were determined.
Abstract: If there is a water flow with a range of temperature inside a pipe, the wanner water tends to float on top of the cooler water because it is lighter, resulting in the upper portion of the pipe being hotter than the lower portion. Under these conditions, such thermal stratification can play an important role in the aging of nuclear power plant piping because of the stress caused by the temperature difference and the cyclic temperature changes. This stress can limit the lifetime of the piping, even leading to penetrating cracks. Investigated in this study is the effect of thermal stratification on the structural integrity of the pressurizer surge line, which is reported to be one of the pipes most severely affected. Finite element models of the surge line are developed using several element types available in a general purpose structural analysis program and stress analyses are performed to determine the response characteristics for the various types of top-to-bottom temperature differentials due to thermal stratification. Fatigue analyses are also performed and an allowable environmental correction factor is suggested.
TL;DR: This paper introduces dynamic nodes to the existing reliability graph with general gates (RGGG) as an intuitive modeling method to model dynamic systems and uses a discrete-time method to convert an RGGG to an equivalent Bayesian network.
Abstract: Conventional static reliability analysis methods are inadequate for modeling dynamic interactions between components of a system. Various techniques such as dynamic fault tree, dynamic Bayesian networks, and dynamic reliability block diagrams have been proposed for modeling dynamic systems based on improvement of the conventional modeling methods. In this paper, we review these methods briefly and introduce dynamic nodes to the existing reliability graph with general gates (RGGG) as an intuitive modeling method to model dynamic systems. For a quantitative analysis, we use a discrete-time method to convert an RGGG to an equivalent Bayesian network and develop a software tool for generation of probability tables.
TL;DR: In this article, a deployment strategy for the sodium-cooled fast-reactor (SFR) is proposed to reduce the amount of spent fuel in PWR spent fuel disposal.
Abstract: The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.
TL;DR: In this article, the authors used a homogeneous equilibrium model when a second phase appears in the SCF blowdown or depressurization process and showed the complexity of behavior that can occur.
Abstract: The use of Supercritical Fluids(SCF) has been proposed for numerous power cycle designs as part of the Generation IV advanced reactor designs, and can provide for higher thermal efficiency. One particular area of interest involves the behavior of SCF during a blowdown or depressurization process. Currently, no data are available in the open literature at supercritical conditions to characterize this phenomenon. A preliminary computational analysis, using a homogeneous equilibrium model when a second phase appears in the process, has shown the complexity of behavior that can occur. Depending on the initial thermodynamic state of the SCF, critical flow phenomena can be characterized in three different ways; the flow can remain in single phase(high temperature), a second phase can appear through vaporization(high pressure low temperature) or condensation(high pressure, intermediate temperature). An experimental facility has been built at the University of Wisconsin to study SCF depressurization through several diameter breaks. The preliminary results obtained show that the experimental data can be predicted with good agreement by the model for all the different initial conditions.
TL;DR: The status of the database on radionuclide migration and retardation that is being developed in Korea is investigated and summarized in this paper, where the solubilities of major actinides such as D, Th, Am, Np, and Pu both in Korean bentonite porewater and in deep Korean groundwater are calculated by using the geochemical code PHREEQC (Ver. 2.0) based on the KAERI-TDB(Korea Atomic Energy Research Institute-Thermochemical Database), which is under development.
Abstract: In this study, data on radionuclide migration and retardation processes in the engineered and natural barriers of High-Level Radioactive Waste (HLW) repository have been reviewed and compiled for use in the performance assessment of a HLW disposal system in Korea. The status of the database on radionuclide migration and retardation that is being developed in Korea is investigated and summarized in this study. The solubilities of major actinides such as D, Th, Am, Np, and Pu both in Korean bentonite porewater and in deep Korean groundwater are calculated by using the geochemical code PHREEQC (Ver. 2.0) based on the KAERI-TDB(Korea Atomic Energy Research Institute-Thermochemical Database), which is under development. Databases for the diffusion coefficients ( values) and distribution coefficients ( values) of some radionuclides in the compacted Korean Ca-bentonite are developed based upon domestic experimental results. Databases for the rock matrix diffusion coefficients ( values) and distribution coefficients ( values) of some radionuclides for Korean granite rock and deep groundwater are also developed based upon domestic experimental results. Finally, data related to colloids such as the characteristics of natural groundwater colloids and the pseudo-colloid formation constants ( values) are provided for the consideration of colloid effects in the performance assessment.
TL;DR: In this paper, Ferrocyanide-anion exchange resins were tested on the ability to uptake ion and the ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation.
Abstract: Ferrocyanide-anion exchange resin was prepared and the prepared ion exchange resins were tested on the ability to uptake ion. The prepared ion exchange resins were resin-KCoFC, resin-KNiFC, and resin-KCuFC. The three tested ion exchange resins showed ion exchange selectivity on the ion of the surrogate soil decontamination solution, and resin-KCoFC showed the best ion uptake capability among the tested ion exchange resins. The ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation. A regeneration feasibility study of the spent ion exchange resins was also performed by the successive application of hydrogen peroxide and hydrazine. The desorption of the ion from the ion exchange resin satisfied the electroneutrality condition in the oxidation step; the desorption of the ion in the reduction step could also be reduced by adding the ion.
TL;DR: In this paper, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate at Yonggwang Nuclear Power Plant Unit 3.
Abstract: Most pressurized water reactors use Venturi flow meters to measure the feedwater flow rate. However, fouling phenomena, which allow corrosion products to accumulate and increase the differential pressure across the Venturi flow meter, can result in an overestimation of the flow rate. In this study, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate. The available data was divided into two groups by fuzzy c means clustering in order to reduce the training time. The data for training the soft-sensing model was selected from each data group with the aid of a subtractive clustering scheme because informative data increases the learning effect. The proposed soft-sensing model was confirmed with the real plant data of Yonggwang Nuclear Power Plant Unit 3. The root mean square error and relative maximum error of the model were quite small. Hence, this model can be used to validate and monitor existing hardware feedwater flow meters.
TL;DR: The KSTAR Integrated Control System (KICS) as discussed by the authors has successfully fulfilled its missions of surveillance, device operation, machine protection interlock, and data acquisition and management, and achieved the goal of first plasma in July 2008 through the four month's commissioning.
Abstract: After more than 10 years construction, KSTAR (Korea Superconducting Tokamak Advanced Research) had finally completed its assembly in June 2007, and then achieved the goal of first-plasma in July 2008 through the four month's commissioning. KSTAR was constructed with fully superconducting magnets with material of and NbTi, and their operation temperatures are maintained below 4.5K by the help of Helium Refrigerator System. During the first-plasma operation, plasmas of maximum current of 133kA and maximum pulse width of 865ms were obtained. The KSTAR Integrated Control System (KICS) has successfully fulfilled its missions of surveillance, device operation, machine protection interlock, and data acquisition and management. These and more were all KSTAR commissioning requirements. For reliable and safe operation of KSTAR, 17 local control systems were developed. Those systems must be integrated into the logically single control system, and operate regardless of their platforms and location installed. In order to meet these requirements, KICS was developed as a network-based distributed system and adopted a new framework, named as EPICS (Experimental Physics and Industrial Control System). Also, KICS has some features in KSTAR operation. It performs not only 24 hour continuous plant operation, but the shot-based real-time feedback control by exchanging the initiatives of operation between a central controller and a plasma control system in accordance with the operation sequence. For the diagnosis and analysis of plasma, 11 types of diagnostic system were implemented in KSTAR, and the acquired data from them were archived using MDSpius (Model Driven System), which is widely used in data management of fusion control systems. This paper will cover the design and implementation of the KSTAR integrated control system and the data management and visualization systems. Commissioning results will be introduced in brief.
TL;DR: An experimental loop operating with water at supercritical conditions (25MPa, in the test section) is designed for operation in the research reactor LVR-15 in UJV Rez, Czech Republic as mentioned in this paper.
Abstract: An experimental loop operating with water at supercritical conditions(25MPa, in the test section) is designed for operation in the research reactor LVR-15 in UJV Rez, Czech Republic. The loop should serve as an experimental facility for corrosion tests of materials for in-core as well as out-of-core structures, for testing and optimization of suitable water chemistry for a future HPLWR and for studies of radiolysis of water at supercritical conditions, which remains the domain where very few experimental data are available. At present, final necessary calculations(thermalhydraulic, neutronic, strength) are being performed on the irradiation channel, which is the most challenging part of the loop. The concept of the primary and auxiliary circuits has been completed. The design of the loop shall be finished in the course of the year 2007 to start the construction, out-of-pile testing to verify proper functioning of all systems and as such to be ready for in-pile tests by the end of the HPLWR Phase 2 European project by the end of 2009.