Showing papers in "Nuclear Technology in 2008"
TL;DR: Nanofluids are engineered colloidal suspensions of nanoparticles in water and exhibit a very significant enhancement (up to 200%) of the boiling critical heat flux (CHF) at modest nanoparticle conc...
Abstract: Nanofluids are engineered colloidal suspensions of nanoparticles in water and exhibit a very significant enhancement (up to 200%) of the boiling critical heat flux (CHF) at modest nanoparticle conc...
207 citations
TL;DR: The U.S. Department of Energy is sponsoring the Generation IV Initiative in the United States for the purposes of developing future-generation nuclear energy systems as discussed by the authors, and six systems have been selected.
Abstract: The U.S. Department of Energy is sponsoring the Generation IV Initiative in the United States for the purposes of developing future-generation nuclear energy systems. Six systems have been selected...
107 citations
TL;DR: In-vessel retention of core melt is a key severe-accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (LWRs).
Abstract: In-vessel retention (IVR) of core melt is a key severe-accident-management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (LWRs). If the...
74 citations
TL;DR: In the DECOVALEX project as discussed by the authors, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical processes near underground waste emplacement drifts.
Abstract: As part of the ongoing international code comparison project DECOVALEX, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near underground waste emplacement drifts. The simulations were conducted for two generic repository types with open or back-filled repository drifts under higher and lower post-closure temperature, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses are currently being resolved. Good agreement in the basic thermal-mechanical responses was achieved for both repository types, even with some teams using relatively simplified thermal-elastic heat-conduction models that neglect complex near-field thermal-hydrological processes. The good agreement between the complex and simplified (and well-known) process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level. The research teams have now moved on to the second phase of the project, the analysis of THM-induced permanent (irreversible) changes and the impact of those changes on the fluid flow field near an emplacement drift.
65 citations
TL;DR: In this article, the authors investigated the recovery of uranium and transuranic (TRU) actinides from spent nuclear fuel by an electrorefining process as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative.
Abstract: Recovery of uranium and transuranic (TRU) actinides from spent nuclear fuel by an electrorefining process was investigated as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative. Experiments were performed in a shielded hot cell at the Materials and Fuels Complex at Idaho National Laboratory. The goal of these experiments was to collect, by an electrochemical process, kilogram quantities of uranium and plutonium into what is called a liquid cadmium cathode (LCC). For each experiment, a steel basket loaded with chopped spent nuclear fuel from the Experimental Breeder Reactor II acted as the anode in the electrorefiner. The cathode was a beryllium oxide crucible containing ∼26 kg of cadmium metal (the LCC). In the three experiments performed to date, between 1 and 2 kg of heavy metal was collected in the LCC after passing an integrated current between 1.80 and 2.16 MC (500 and 600 A h) from the anode to the cathode. Sample analysis of the processed LCC ingots measured detectable amounts of TRUs and rare earth elements.
60 citations
TL;DR: In this paper, the electrochemical reduction process has been developed for converting oxide nuclear fuels to metals, and the reduction mechanism has been characterized and investigated in order to investigate appropriate condi cation.
Abstract: The electrochemical reduction process has been recently developed for converting oxide nuclear fuels to metals. In order to characterize the reduction mechanism and to investigate appropriate condi...
53 citations
TL;DR: In this paper, the design concept of a subcritical advanced burner (SABR) was described, which is fueled with transuranics (TRUs) discharged from thermal reactors cast into a TRU-Zr metal fuel pin and is cooled with sodium.
Abstract: The design concept of a subcritical advanced burner reactor (SABR) is described. SABR is fueled with transuranics (TRUs) discharged from thermal reactors cast into a TRU-Zr metal fuel pin and is cooled with sodium. The reactor operates subcritical to achieve a deep-burn four-batch fuel cycle that fissions 25% of the TRU in an 8.2-yr residence time, limited by radiation damage accumulation (200 displacements per atom) in the oxygen dispersion strengthened clad and structure. The annual TRU fission rate in SABR [3000 MW(thermal)] is comparable to the annual TRU discharge of three to five 1000-MW(electric) light water reactors, depending on the plant capacity factor of SABR. A tokamak D-T fusion neutron source based on physics and technology that will be demonstrated in ITER supports the subcritical operation.
52 citations
TL;DR: Molten fluorides were initially developed for use in the nuclear industry as the high-temperature fluid fuel for the Molten Salt Reactor (MSR) as discussed by the authors.
Abstract: Molten fluorides were initially developed for use in the nuclear industry as the high-temperature fluid fuel for the Molten Salt Reactor (MSR). The U.S. Department of Energy Office of Nuclear Energ...
44 citations
TL;DR: In this paper, the important incidents related to fast reactor sodium components and systems are summarized, based on experience, analysis, experimental tests, and research and development for past and current SFRs.
Abstract: Eighteen sodium-cooled fast reactors (SFRs), a number that includes reactors in operation or shut down, have provided 388 reactor-years of operating experience to date. This paper summarizes the important incidents related to fast reactor sodium components and systems. The solutions incorporated, based on experience, analysis, experimental tests, and research and development for past and current SFRs, are described. The paper also describes lessons learned for future SFRs.
37 citations
TL;DR: In this paper, the fission gas release characteristics during the voloxidation and oxidation and reduction of oxide fuel (OREOX) processes of spent PWR fuel were carried out by a hot cell of the DUPIC Fuel Development Facility.
Abstract: Quantitative analysis of the fission gas release characteristics during the voloxidation and oxidation and reduction of oxide fuel (OREOX) processes of spent pressurized water reactor (PWR) fuel was carried out by spent PWR fuel in a hot cell of the DUPIC Fuel Development Facility. The release characteristics of 85Kr and 14C fission gases during voloxidation process at 500°C are closely linked to the degree of conversion efficiency of UO2 to U3O8 powder, and it can be interpreted that the release from grain boundary would be dominated during this step. Volatile fission gases of 14C and 85Kr were released to near completion during the OREOX process. Both the 14C and 85Kr have similar release characteristics under the voloxidation and OREOX process conditions. A higher burnup spent fuel showed a higher release fraction than that of a low burnup fuel during the voloxidation step. It was also observed that the release fraction of semivolatile Cs was ~16% during a reduction at 1000°C of the oxidized po...
36 citations
TL;DR: In this article, the dynamics of the molten salt reactor (MSR) was studied and the graphite-moderated channel-type MSR was selected for numerical simulation.
Abstract: The dynamics of the molten salt reactor (MSR), one of the Generation IV International Forum concepts, was studied. The graphite-moderated channel-type MSR was selected for numerical simulation. MSR...
TL;DR: In this article, the authors present a working fluid for high-temperature (500 to 1000°C) heat transport in fission and fusion applications using liquid fluoride salts and helium.
Abstract: Liquid fluoride salts and helium have desirable properties for use as working fluids for high-temperature (500 to 1000°C) heat transport in fission and fusion applications. This paper presents rece...
TL;DR: The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form as mentioned in this paper.
Abstract: The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form. Reactive metal ...
TL;DR: The Korea Atomic Energy Research Institute is currently developing the Advanced Spent-Fuel Conditioning Process (ACP) based on a pyrochemical process as mentioned in this paper, and an electrochemical reduction process has been used to reduce fuel consumption.
Abstract: The Korea Atomic Energy Research Institute is currently developing the Advanced Spent-Fuel Conditioning Process (ACP) based on a pyrochemical process. An electrochemical reduction process has been ...
TL;DR: The Global Nuclear Energy Partnership (GNEP) as discussed by the authors is an initiative to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclea...
Abstract: The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclea...
TL;DR: A range of probabilistic approaches to forecast the future spatial distribution of volcanism are discussed, including kernel, adaptive kernel, and Cox process methods.
Abstract: Because of the difficulty of describing the complex spatial and temporal patterns inherent to volcanism, the use of solely deterministic models is not sufficient for long-term estimation of volcanic hazards. In order to account for the intrinsic uncertainty of volcanism that occurs in space and time and with respect to event types and their intensity, the use of probabilistic models becomes quite natural for long-term hazard assessment. Here, we discuss a range of probabilistic approaches to forecast the future spatial distribution of volcanism, including kernel, adaptive kernel, and Cox process methods. An application to the volcanic arc of Tohoku illustrates the proposed methodology.
TL;DR: In this paper, the authors address the long-term uranium supply from first principles, summarizing estimates of the abundance of uranium in the crust of the earth as a function of concentration and accessibility.
Abstract: We address the long-term uranium supply from first principles, summarizing estimates of the abundance of uranium in the crust of the earth as a function of concentration and accessibility. Defining the supply curve as a functional relationship between the cumulative quantity of uranium extracted and the cost of extracting the next unit of uranium, we note that a supply curve requires a crustal abundance model plus a correlation between ore grade and extraction cost. Surveying a number of supply curves that appear in the literature, we observe that while estimates vary widely (we observe an order of magnitude difference in forecasts of the quantity of uranium available at $100/kg U or less), they generally reflect expectations that uranium availability will be significantly greater than the Red Book numbers imply. Furthermore, by comparison with historical data for more than 40 other minerals, we show that the assumption of time invariance when formulating a supply curve is not borne out by experience. In fact, the price of most other minerals has decreased with time as well as with cumulative quantity extracted. Neither the Red Book nor the other supply curves we survey explicitly accounts for the unit-based technological learning that fosters this behavior.
TL;DR: The gas-cooled fast-reactor (GFR) has received increased attention in the past decade under the impetus provided by the Generation-IV International Forum (GIF).
Abstract: The gas-cooled fast reactor (GFR) has received increased attention in the past decade under the impetus provided by the Generation-IV International Forum. The GFR given principal attention is a ver...
TL;DR: In this paper, the authors derived a kinetic model for the reduction of oxide spent nuclear fuel in a radial flow reactor and extended it to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction.
Abstract: A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be dependent only upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.
TL;DR: The PArticle Stress Analysis (PASTA) code was written to evaluate stresses in coated particle fuel embedded in graphite of high-temperature reactors (HTRs) as discussed by the authors.
Abstract: The PArticle STress Analysis (PASTA) code was written to evaluate stresses in coated particle fuel embedded in graphite of high-temperature reactors (HTRs). Existing models for predicting stresses in coated particle fuels were extended with a treatment of stresses induced by dimensional change of the matrix graphite and stresses caused by neighboring particles. PASTA was applied to two practical cases in order to evaluate the significance of this model extension. Thermal hydraulics, neutronics, and fuel depletion calculation tools were used to calculate the fuel conditions in these cases. Stresses in the first fuel loading of the High-Temperature Engineering Test Reactor (HTTR) and in the fuel of a 400-MW(thermal) pebble bed reactor were analyzed. It is found that the presence of the matrix material plays a significant role in the determination of the stresses that apply to a single isolated TRISO particle as well as in the transmittal of the stresses between particles in actual pebble designs.
TL;DR: With the current and forthcoming need to develop new nuclear power plants, decommission existing nuclear plants, and satisfy future demands to minimize nuclear waste, it is important to examine and examine and... as discussed by the authors.
Abstract: With the current and forthcoming need to develop new nuclear power plants, decommission existing nuclear plants, and satisfy future demands to minimize nuclear waste, it is important to examine and...
TL;DR: The self-scraping phenomenon of the graphite cathode was interpreted to be due to the formation of a uranium graphite intercalation compound as discussed by the authors, where the uranium atoms elongate at the innermost layer of graphite's outermost layer, and the deposited dendrite falls off spontaneously as the gravitational force exceeds the bonding strength of the layers.
Abstract: The conventional electrorefiners to treat a metallic spent fuel equipped with a steel cathode have a sticking characteristic that hinders their overall processing efficiency. The critical question in order to enhance their throughput is how to decrease the sticking coefficient of the cathode. In order to realize this purpose, the conventional steel cathode was replaced with a graphite one. The graphite cathode exhibited self-scraping behavior in which the electrodeposited uranium dendrite falls from the cathode surface on its own without any kind of mechanical operation such as a scraping and rotation of the electrode. This self-scraping phenomenon of the graphite cathode was interpreted to be due to the formation of a uranium graphite intercalation compound. In this self-scraping mechanism, uranium atoms elongate at the graphite's outermost layer by an intercalation reaction, so the deposited uranium dendrite falls off spontaneously as the gravitational force exceeds the bonding strength of the layers. Based on our preliminary work, a self-scraping should increase the efficiency of an electrorefiner due to the elimination of a mechanical scraping as well as the electrolytic stripping steps of the cathode.
TL;DR: In this paper, the chemical behavior of lanthanide oxides has been studied both for the electrolytic reduction process and the electrorefining process, and the solubility of the oxides was measured under both electrolytic and electro-refining conditions.
Abstract: The chemical behavior oflanthanide oxides has been studied both for the electrolytic reduction process and the electrorefining process. At high concentration of Li 2 O in LiCl, lanthanide oxides reacted with Li 2 O to form mixed oxides, LiLnO 2 (Ln = lanthanides), which decomposed to the starting materials at relatively low Li 2 O concentration. The chemical behavior of lanthanide oxides under the condition of electrorefining process was investigated by optical fiber spectrophotometry and X-ray diffraction. Lanthanide oxides reacted with U 3+ to produce Ln 3+ and UO 2 . The solubility of lanthanide oxides was measured under the electrolytic reduction and the electrorefining condition. All of the lanthanide oxides except Eu 2 O 3 had relatively low solubility values in LiCl-KCl eutectic mixture at 450°C. Electrochemical behavior of Br - , I - , and Se 2- in LiCl was also investigated by cyclic voltammetry and by X-ray diffraction. All of the anions reacted with platinum anode and gave platinum compounds.
TL;DR: In this article, a Markov model approach is developed for the evaluation of proliferation resistance (PR) of nuclear energy systems, where extrinsic and intrinsic barriers associated with the energy systems are considered.
Abstract: A Markov model approach is developed for the evaluation of proliferation resistance (PR) of nuclear energy systems. The focus of this study is to create a high-fidelity probabilistic assessment model that better represents nuclear energy systems. Both extrinsic and intrinsic barriers associated with the energy systems are considered. Modeling uncertainty and safeguards false alarms, composite safeguards approaches, concealment, and human performance are particularly discussed in detail and incorporated in the Markov model. These features are anticipated to have significant impacts on PR assessment. The Markov model approach is adapted to a hypothetical example sodium fast reactor (ESFR) system using physically meaningful parameters that can be obtained from physical processes. Development of metrics for six PR measures is discussed. Computation of the PR measures using the Markov model of the ESFR system is illustrated. The results obtained in this study demonstrate applicability and effectiveness of the Markov model approach in the PR assessment.
TL;DR: In this article, a series of computational fluid dynamics simulations have been conducted to analyze the heat transfer enhancement in a fully heated rod bundle with mixing-vane spacers, and the predicted Nusselt n
Abstract: A series of computational fluid dynamics (CFD) simulations has been conducted to analyze the heat transfer enhancement in a fully heated rod bundle with mixing-vane spacers The predicted Nusselt n
TL;DR: In this article, the authors proposed a safe, secure, environmentally sound, and potentially economic solution for high-level radioactive wastes, including spent fuel and fissile materials, in 4-to 5-km-deep boreholes.
Abstract: Disposal in 4- to 5-km-deep boreholes offers a safe, secure, environmentally sound, and potentially economic solution for high-level radioactive wastes, including spent fuel and fissile materials. ...
TL;DR: In this paper, the eutectic MgCl 2 -NaCl-KCl ternary salt has been proposed as the alternate electrolyte for the conventional 2Cs Cl-NaCl electrolyte, which is used for oxide processing.
Abstract: Reducing the cooling time of spent fast breeder reactor (FBR) fuel, thus reducing the doubling time and introducing metallic fuels into FBRs, is essential for meeting the increasing energy demand of India. Development of pyrochemical reprocessing technology for processing the spent FBR fuels is another prerequisite. Accordingly, studies on the molten salt electrorefining process for metallic fuels and the oxide electrowinning process for oxide fuels have been carried out at the Indira Gandhi Centre for Atomic Research, Kalpakkam. A laboratory-scale argon atmosphere facility for molten salt electrorefining process studies is operational. Using this facility, studies on all the unit operations of the process have been carried out on uranium alloys. A code, PRAGAMAN, based on thermochemical modeling has been developed to simulate the electrotransport behavior of elements during the electrorefining process. Based on our studies, the eutectic MgCl 2 -NaCl-KCl ternary salt has been proposed as the alternate electrolyte for the conventional 2Cs Cl-NaCl electrolyte for oxide processing. A facility to demonstrate the remotization of all the process steps of the molten salt electrorefining process flow sheet for metallic fuels at 1- to 3-kg scale is being set up. Basic electrochemical studies on the reduction behavior of the chlorides and oxychlorides of uranium and the lanthanides in molten salts have also been carried out. This paper describes the studies carried out so far and the plans for the near future.
TL;DR: The SuperCritical Water-cooled Reactor (SCWR) is one of the candidates for the fourth-generation nuclear power plant, and it uses light water as a coolant as discussed by the authors.
Abstract: The SuperCritical Water-cooled Reactor (SCWR) is one of the candidates for the fourth-generation nuclear power plant, and it uses light water as a coolant. Heat transfer between a fuel assembly and...
TL;DR: In this article, the stability performance of the Economic Simplified Boiling Water Reactor (ESBWR) was studied with the downscaled GENESIS facility, which includes an artificial void reactivity feedback system for simulating the neutronic-thermal-hydraulic coupling.
Abstract: The stability performance of the Economic Simplified Boiling Water Reactor (ESBWR) is studied with the downscaled GENESIS facility. The GENESIS design is based on fluid-to-fluid modeling and includes an artificial void reactivity feedback system for simulating the neutronic-thermal-hydraulic coupling. The experiments show that the ESBWR thermal-hydraulic oscillatory mode is very stable at nominal conditions, exhibiting a decay ratio DR = 0.12 and a remarkably low resonance frequency f res =0.11 Hz. This result indicates a static pressure head-driven phenomenon since this frequency corresponds well to typical frequencies found for density wave oscillations traveling through the core plus chimney sections. For the reactor-kinetic oscillatory mode, we found a decay ratio DR = 0.30 and a resonance frequency f res = 0.75 Hz. This corresponds well to density wave oscillations traveling through the core indicating the instability mechanism is driven by the interplay between the core friction and the neutronic response due to void changes in the core. By comparing these results with those obtained with the TRACG computational code, it was found that they agree very well. In addition, the stability performance of the thermal-hydraulic and the reactor-kinetic mode is investigated for a wide range of conditions, confirming the existence of large margins to instabilities of the ESBWR design.
TL;DR: In this paper, a new approach to remove fission products including decay heat elements was proposed to minimize the amount of waste salt for a repository, while removing the high decay heat fission product [Cs, Sr, Ba, and Y including other rare earth (RE) elements] from the waste salts generated during a chloride pyroprocessing procedure.
Abstract: In this work, a new approach to remove fission products including decay heat elements was proposed. This study aims at providing a new way to minimize the amount of waste salt for a repository, while removing the high decay heat fission products [Cs, Sr, Ba, and Y including other rare earth (RE) elements] from the waste salts generated during a chloride pyroprocessing procedure. These elements were removed in consecutive order from the pyroprocessing units. First, Cs could be released in the form of an oxide gas during voloxidation of UO 2 and captured by a fly-ash filter. Then, Sr was recovered in the form of carbonate precipitates from the LiCl waste salt generated during the course of an electoreduction process, by using Li 2 CO 3 . Finally, RE elements plus yttrium in the spent LiCl-KCl waste salt generated during electrorefining were removed in the form of oxides (or oxychlorides) by using an oxygen sparging method. It was confirmed that the removal yields of each element were ∼90% for Cs at ∼1473 K, >99% for Sr at a molar ratio of [Li 2 CO 3 /SrCl 2 = 3], and >99% for the RE elements plus yttrium. Using these successes as a basis, a reference flow sheet for removing the high decay heat elements from pyroprocessing units is presented in this work. Also, a salt regeneration system to minimize the amount of waste salt is proposed in this study.