Recovering metal values hydrometallurgically from spent dry battery cells

Science.gov (United States)

Rabah, M. A.; Barakat, M. A.; Mahrous, Y. Sh.

1999-12-01

A hydro-pyrometallurgical method was used to recover metal values from spent dry battery cells. Water-soluble ingredients were filtered, and solid residue was sorted by magnetic separation and water flotation. Parameters affecting the recovery efficiency were also studied. Results revealed that metallic parts, carbon rods, and paper were safely recovered; pure NH4Cl, MnO2, and ZnCl2 salts were obtained. Maximum recovery efficiencies reached 93 percent for manganese and 99.5 percent for zinc and NH4.

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

Bioleaching of valuable metals from spent lithium-ion mobile phone batteries using Aspergillus niger

Science.gov (United States)

Horeh, N. Bahaloo; Mousavi, S. M.; Shojaosadati, S. A.

2016-07-01

In this paper, a bio-hydrometallurgical route based on fungal activity of Aspergillus niger was evaluated for the detoxification and recovery of Cu, Li, Mn, Al, Co and Ni metals from spent lithium-ion phone mobile batteries under various conditions (one-step, two-step and spent medium bioleaching). The maximum recovery efficiency of 100% for Cu, 95% for Li, 70% for Mn, 65% for Al, 45% for Co, and 38% for Ni was obtained at a pulp density of 1% in spent medium bioleaching. The HPLC results indicated that citric acid in comparison with other detected organic acids (gluconic, oxalic and malic acid) had an important role in the effectiveness of bioleaching using A. niger. The results of FTIR, XRD and FE-SEM analysis of battery powder before and after bioleaching process confirmed that the fungal activities were quite effective. In addition, bioleaching achieved higher removal efficiency for heavy metals than the chemical leaching. This research demonstrated the great potential of bio-hydrometallurgical route to recover heavy metals from spent lithium-ion mobile phone batteries.

Technique for recovering rare-earth metals from spent sintered Nd-Fe-B magnets without external heating

Directory of Open Access Journals (Sweden)

Ryo Sasai

2016-06-01

Full Text Available To selectively recover rare-earth metals with higher purity from spent sintered Nd-Fe-B magnets without external heating, we investigated the mechano-chemical treatment of spent sintered Nd-Fe-B magnet powder with a reaction solution of HCl and (COOH2 at room temperature. The results of various experiments showed that the mechano-chemical treatment with HCl and (COOH2 is very effective for recovering the rare-earth metals contained in spent sintered Nd-Fe-B magnet powder; the recovery rate and purity of the rare-earth metals were 95.3 and 95.0 mass%, respectively, under optimal conditions ([HCl] = 0.2 mol/dm3 and [(COOH2] = 0.25 mol/dm3.

Bioleaching of metals from spent refinery petroleum catalyst using moderately thermophilic bacteria: effect of particle size.

Science.gov (United States)

Srichandan, Haragobinda; Singh, Sradhanjali; Pathak, Ashish; Kim, Dong-Jin; Lee, Seoung-Won; Heyes, Graeme

2014-01-01

The present work investigated the leaching potential of moderately thermophilic bacteria in the recovery of metals from spent petroleum catalyst of varying particle sizes. The batch bioleaching experiments were conducted by employing a mixed consortium of moderate thermophilic bacteria at 45°C and by using five different particle sizes (from 45 to >2000 μm) of acetone-washed spent catalyst. The elemental mapping by FESEM confirmed the presence of Al, Ni, V and Mo along with sulfur in the spent catalyst. During bioleaching, Ni (92-97%) and V (81-91%) were leached in higher concentrations, whereas leaching yields of Al (23-38%) were found to be lowest in all particle sizes investigated. Decreasing the particle size from >2000 μm to 45-106 μm caused an increase in leaching yields of metals during initial hours. However, the final metals leaching yields were almost independent of particle sizes of catalyst. Leaching kinetics was observed to follow the diffusion-controlled model showing the linearity more close than the chemical control. The results of the present study suggested that bioleaching using moderate thermophilic bacteria was highly effective in removing the metals from spent catalyst. Moreover, bioleaching can be conducted using spent catalyst of higher particle size (>2000 μm), thus saving the grinding cost and making process attractive for larger scale application.

Recycling of spent noble metal catalysts with emphasis on pyrometallurgical processing

Energy Technology Data Exchange (ETDEWEB)

Hagelueken, C. [Degussa Huels AG, Hanau (Germany)

1999-09-01

Precious metal catalysts for catalytic Naphta Reforming, Isomerization, Hydrogenation and other chemical and petrochemical processes are valuable assets for oil refineries and chemical companies. At the end of the service life of a reactor load of catalyst, the efficient and reliable recovery of the precious metals contained in the catalyst is of paramount importance. More than 150 years of technological advances at Degussa-Huels have resulted in refining methods for all kinds of precious metal containing materials which guarantee an optimum technical yield of the precious metals included. The refining of catalysts today is one of the important activities in the precious metals business unit. In the state-of-the-art precious metal refinery at Hanau in the centre of Germany, a wide variety of processes for the recovery of all precious metals is offered. These processes include accurate preparation, sampling and analysis as well as both wet-chemical and pyrometallurgical recovery techniques. Special emphasis in this presentation is laid on the advantages of pyrometallurgical processes for certain kinds of catalysts. To avoid any risks during transport, sampling and treatment of the spent catalyst, all parties involved in the recycling chain strictly have to follow the relevant safety regulations. Under its commitment to 'Responsible Care' standard procedures have been developed which include pre-shipment samples, safety data sheets/questionnaires and inspection of spent catalysts. These measures not only support a safe and environmentally sound catalyst recycling but also enable to determine the most suitable and economic recovery process - for the benefit of the customer. (orig.)

Recovery Of Electrodic Powder From Spent Nickel-Metal Hydride Batteries (NiMH

Directory of Open Access Journals (Sweden)

Shin S.M.

2015-06-01

Full Text Available This study was focused on recycling process newly proposed to recover electrodic powder enriched in nickel (Ni and rare earth elements (La and Ce from spent nickel-metal hydride batteries (NiMH. In addition, this new process was designed to prevent explosion of batteries during thermal treatment under inert atmosphere. Spent nickel metal hydride batteries were heated over range of 300°C to 600°C for 2 hours and each component was completely separated inside reactor after experiment. Electrodic powder was successfully recovered from bulk components containing several pieces of metals through sieving operation. The electrodic powder obtained was examined by X-ray diffraction (XRD and energy dispersive X-ray spectroscopy (EDX and image of the powder was taken by scanning electron microscopy (SEM. It was finally found that nickel and rare earth elements were mainly recovered to about 45 wt.% and 12 wt.% in electrodic powder, respectively.

Process optimization and kinetics for leaching of rare earth metals from the spent Ni-metal hydride batteries.

Science.gov (United States)

Meshram, Pratima; Pandey, B D; Mankhand, T R

2016-05-01

Nickel-metal hydride batteries (Ni-MH) contain not only the base metals, but valuable rare earth metals (REMs) viz. La, Sm, Nd, Pr and Ce as well. In view of the importance of resource recycling and assured supply of the contained metals in such wastes, the present study has focussed on the leaching of the rare earth metals from the spent Ni-MH batteries. The conditions for the leaching of REMs from the spent batteries were optimized as: 2M H2SO4, 348K temperature and 120min of time at a pulp density (PD) of 100g/L. Under this condition, the leaching of 98.1% Nd, 98.4% Sm, 95.5% Pr and 89.4% Ce was achieved. Besides the rare earth metals, more than 90% of base metals (Ni, Co, Mn and Zn) were also leached out in this condition. Kinetic data for the dissolution of all the rare earth metals showed the best fit to the chemical control shrinking core model. The leaching of metals followed the mechanism involving the chemical reaction proceeding on the surface of particles by the lixiviant, which was corroborated by the XRD phase analysis and SEM-EDS studies. The activation energy of 7.6, 6.3, 11.3 and 13.5kJ/mol was acquired for the leaching of neodymium, samarium, praseodymium and cerium, respectively in the temperature range 305-348K. From the leach liquor, the mixed rare earth metals were precipitated at pH∼1.8 and the precipitated REMs was analyzed by XRD and SEM studies to determine the phases and the morphological features. Copyright © 2015. Published by Elsevier Ltd.

Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

Science.gov (United States)

Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

2013-07-01

A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

Recovery and Separation of Valuable Metals from Spent Nickel-Metal Hydride Batteries using some Organophosphorus Extractants

International Nuclear Information System (INIS)

Aly, M.I.; Daoud, J.A.; ALy, H.F.

2012-01-01

The separation of cobalt, nickel, and rare earth elements from NiMH battery residues is evaluated in this paper. A hydrometallurgical process is developed for the recovery of metals from spent batteries and a selective separation of RE by precipitation of sodium RE double sulfate is performed. The methodology used benefits the solubility of the battery electrode materials in sulfuric or hydrochloric acids. The results obtained show that sulfuric acid is slightly less powerful in leaching (NiMH) compared to HCl acid. However, sulfuric acid was used on economic basis. Leaching solution was obtained by using 3 M H 2 SO 4 at 70 +1 degree C + 3% wt. H 2 O 2 for 5 hours. It has been shown that it is possible to recover about 98 % of the RE contained in spent NiMH batteries. The maximum recovery of nickel and cobalt metals was 99.9% and 99.4%, respectively. The effects of the main operating variables of both leaching and solvent extraction steps of nickel (II) and cobalt (II) from the leach solution using HDEHP (di-2-ethylhexyl phosphoric acid) and CYANEX 272 (di-(2,4,4 trimethyl pentyl) phosphinic acid) in kerosene were investigated aiming to maximize metal separation for recycling purposes. The developed process for the recovery and separation of nickel (II) , cobalt (II), and rare earth from spent NiMH batteries is tested and the obtained sulfate salts CoSO 4 and NiSO 4 have a high purity, suggesting that these recovered products could be used as chemical materials without further purification

Literature survey on metal waste form for metallic waste from electrorefiners for the electrometallurgical treatment of spent metallic fuels

International Nuclear Information System (INIS)

Nishimura, Tomohiro

2003-01-01

This report summarizes the recent results of the metal waste form development activities at the Argonne National Laboratory in the USA for high-level radioactive metallic waste (stainless-steel (SS) cladding hulls, zirconium (Zr), noble-metal fission products (NMFPs), etc.) from electrorefiners for the electrometallurgical treatment of spent metallic fuels. Their main results are as follows: (1) SS- 15 wt.% Zr- ∼4 wt.% NMFPs alloy was selected as the metal waste form, (2) metallurgical data, properties, long-term corrosion data, etc. of the alloy have been collected, (3) 10-kg ingots have been produced in hot tests and a 60-kg production machine is under development. The following research should be made to show the feasibility of the metal waste form in Japan: (1) degradation assessment of the metal waste form in Japanese geological repository environments, and (2) clarification of the maximum allowable contents of NMFPs. (author)

Metals recovery of spent household batteries using a hydrometallurgical process

International Nuclear Information System (INIS)

Souza, K.P.; Tenorio, J.A.S.

2010-01-01

The objective of the work is to study a method for metals recovery from a sample composed by a mixture of the main types of spent household batteries. Segregation of the main metals is investigated using a treatment route consisting of the following steps: manual identified and dismantling, grinding, electric furnace reduction, acid leaching and selective precipitation with sodium hydroxide with and without hydrogen peroxide. Before and after precipitations the solutions had been analyzed by Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP/OES) and the precipitated analyzed by Scanning Electron Microscopy (SEM) with Spectrometry of Energy Dispersion Spectroscopy (EDS). The results had indicated that the great majority of metals had been precipitated in pHs studied, also had co-precipitation or simultaneous precipitation of metals in some pHs. (author)

The role of fission products (noble metal particles) in spent fuel corrosion process in a failed container

Energy Technology Data Exchange (ETDEWEB)

Wu, L., E-mail: lwu59@uwo.ca [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Univ. of Western Ontario, Surface Science Western, London, Ontario (Canada)

2013-07-01

The corrosion/dissolution of simulated spent fuel has been studied electrochemically. Fission products within the UO{sub 2} matrix are found to have significant effect on the anodic dissolution behaviour of the fuel. It is observed that H{sub 2}O{sub 2}oxidation is accelerated on the surfaces of doped noble metal (ε) particles existing in the fuel matrix. It is concluded that H{sub 2}O{sub 2} decomposition rather than UO{sub 2} corrosion should be the dominant reaction under high H{sub 2}O{sub 2} concentrations. (author)

Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

International Nuclear Information System (INIS)

Abraham, D.P.; McDeavitt, S.M.; Park, J.

1996-01-01

Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

International Nuclear Information System (INIS)

Sabourin, G.

1998-01-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

International Nuclear Information System (INIS)

McKinnon, M.A.; Creer, J.M.

1988-10-01

REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

International Nuclear Information System (INIS)

Herrmann, S.D.; Li, S.X.

2010-01-01

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl - 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

Recovery of Platinum Group Metals from Spent Catalysts Using Iron Chloride Vapor Treatment

Science.gov (United States)

Taninouchi, Yu-ki; Okabe, Toru H.

2018-05-01

The recovery of platinum group metals (PGMs) from spent automobile catalysts is a difficult process because of their relatively low contents in the scrap. In this study, to improve the efficiency of the existing recycling techniques, a novel physical concentration method involving treatment with FeCl2 vapor has been examined. The reactions occurring between typical catalyst components and FeCl2 vapor are discussed from the thermodynamic point of view, and the validity of the proposed technique was experimentally verified. The obtained results indicate that the vapor treatment at around 1200 K (927 °C) can effectively alloy PGMs (Pt, Pd, and Rh) with Fe, resulting in the formation of a ferromagnetic alloy. It was also confirmed that cordierite and alumina (the major catalyst components) remained unreacted after the vapor treatment, while ceria species were converted into oxychlorides. The samples simulating the automobile catalyst were also subjected to magnetic separation after the treatment with FeCl2 vapor; as a result, PGMs were successfully extracted and concentrated in the form of a magnetic powder. Thus, the FeCl2 vapor treatment followed by magnetic separation can be utilized for recovering PGMs directly from spent catalysts as an effective pretreatment for the currently used recycling methods.

The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

Energy Technology Data Exchange (ETDEWEB)

Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

International Nuclear Information System (INIS)

Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

2016-01-01

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

International Nuclear Information System (INIS)

Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

2011-01-01

Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

Review and evaluation of long-term integrity on metal casks and spent fuels stored in overseas countries

International Nuclear Information System (INIS)

Sasahara, Akihiro; Saegusa, Toshiari

2009-01-01

Inspection and experimental results on the metal cask and PWR-UO 2 spent fuels practically stored for fifteen years in Idaho National Laboratory (INL) are reviewed. Experimental results on PWR-UO 2 and BWR-MOX spent fuels stored for twenty years under wet or dry condition obtained by Central Research Institute of Electric Power Industry (CRIEPI) are also reviewed. These results show that the integrity of the metal cask and PWR-spent fuels are maintained at least during dry storage for fifteen years and that Japanese electric utilities may start their self-inspection on casks and spent fuels after fifteen-year storage. The gas sampling carrying out in INL can be applied to licensing for interim dry storage facilities in Japan. New program for the fuel integrity for high burn-up fuels (>45 GWd/MTU) at transportation after dry storage has been launched by Nuclear Regulation Commission (NRC), Department of Energy (DOE) and Electric Power Research Institute (EPRI) in USA. (author)

Spent lithium-ion battery recycling - Reductive ammonia leaching of metals from cathode scrap by sodium sulphite.

Science.gov (United States)

Zheng, Xiaohong; Gao, Wenfang; Zhang, Xihua; He, Mingming; Lin, Xiao; Cao, Hongbin; Zhang, Yi; Sun, Zhi

2017-02-01

Recycling of spent lithium-ion batteries has attracted wide attention because of their high content of valuable and hazardous metals. One of the difficulties for effective metal recovery is the separation of different metals from the solution after leaching. In this research, a full hydrometallurgical process is developed to selectively recover valuable metals (Ni, Co and Li) from cathode scrap of spent lithium ion batteries. By introducing ammonia-ammonium sulphate as the leaching solution and sodium sulphite as the reductant, the total selectivity of Ni, Co and Li in the first-step leaching solution is more than 98.6% while it for Mn is only 1.36%. In detail understanding of the selective leaching process is carried out by investigating the effects of parameters such as leaching reagent composition, leaching time (0-480min), agitation speed (200-700rpm), pulp density (10-50g/L) and temperature (323-353K). It was found that Mn is primarily reduced from Mn 4+ into Mn 2+ into the solution as [Formula: see text] while it subsequently precipitates out into the residue in the form of (NH 4 ) 2 Mn(SO 3 ) 2 ·H 2 O. Ni, Co and Li are leached and remain in the solution either as metallic ion or amine complexes. The optimised leaching conditions can be further obtained and the leaching kinetics is found to be chemical reaction control under current leaching conditions. As a result, this research is potentially beneficial for further optimisation of the spent lithium ion battery recycling process after incorporating with metal extraction from the leaching solution. Copyright © 2016 Elsevier Ltd. All rights reserved.

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

Energy Technology Data Exchange (ETDEWEB)

Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

1998-07-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Recovery of metals from simulant spent lithium-ion battery as organophosphonate coordination polymers in aqueous media

Energy Technology Data Exchange (ETDEWEB)

Perez, Emilie; Andre, Marie-Laure; Navarro Amador, Ricardo [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Hyvrard, François; Borrini, Julien [SARPI VEOLIA, Direction Technique et Innovations, Zone portuaire de Limay-Porcheville, 427 route du Hazay, 78520 Limay (France); Carboni, Michaël, E-mail: michael.carboni@cea.fr [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Meyer, Daniel [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

2016-11-05

Highlights: • Original waste disposal strategies for battery. • Precipitation of metals as coordination polymers. • Organo-phosphonate coordination polymers. • Selective extraction of manganese or co-precipitation of manganese/cobalt. • The recycling process give a promising application on any waste solution. - Abstract: An innovative approach is proposed for the recycling of metals from a simulant lithium-ion battery (LIBs) waste aqueous solution. Phosphonate organic linkers are introduced as precipitating agents to selectively react with the metals to form coordination polymers from an aqueous solution containing Ni, Mn and Co in a hydrothermal process. The supernatant is analyzed by ICP-AES to quantify the efficiency and the selectivity of the precipitation and the materials are characterized by Scanning Electron Microscopy (SEM), Powder X-Ray Diffraction (PXRD), Thermogravimetric Analyses (TGA) and nitrogen gas sorption (BET). Conditions have been achieved to selectively precipitate Manganese or Manganese/Cobalt from this solution with a high efficiency. This work describes a novel method to obtain potentially valuable coordination polymers from a waste metal solution that can be generalized on any waste solution.

Electrolytic reduction of a simulated oxide spent fuel and the fates of representative elements in a Li{sub 2}O-LiCl molten salt

Energy Technology Data Exchange (ETDEWEB)

Park, Wooshin, E-mail: wooshin@kaeri.re.kr; Choi, Eun-Young; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Young-Hwan; Hur, Jin-Mok

2016-08-15

A series of electrolytic reduction experiments were carried out using a simulated oxide spent fuel to investigate the reduction behavior of elements in a mixed oxide condition and the fates of elements in the reduction process with 1.0 wt% Li{sub 2}O-LiCl. It was found out that 155% of the theoretical charge was enough to reduce the simulated. Te and Eu were expected to possibly exist in the precipitate and on the anode surface, whereas Ba and Sr showed apparent dissolution behaviors. Rare earths showed relatively low metal fractions from 28.2 to 34.0% except for Y. And the solubility of rare earths was observed to be low due to the low concentration of Li{sub 2}O. The reduction of U was successful as expected showing 99.8% of a metal fraction. Also it was shown that the reduction of ZrO{sub 2} would be effective when a relatively small amount was included in a metal oxide mixture.

Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Freedman, J.M.; Humphreys, J.R.

1975-10-01

A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

Rejuvenation of residual oil hydrotreating catalysts by leaching of foulant metals. Modelling of the metal leaching process

Energy Technology Data Exchange (ETDEWEB)

Marafi, M.; Kam, E.K.T.; Stanislaus, A.; Absi-Halabi, M. [Petroleum Technology Department, Petroleum, Petrochemicals and Materials Division, Kuwait Institute for Scientific Research, Safat (Kuwait)

1996-11-19

Increasing emphasis has been paid in recent years on the development of processes for the rejuvenation of spent residual oil hydroprocessing catalysts, which are deactivated by deposition of metals (e.g. vanadium) and coke. As part of a research program on this subject, we have investigated selective removal of the major metal foulant from the spent catalyst by chemical leaching. In the present paper, we report the development of a model for foulant metals leaching from the spent catalyst. The leaching process is considered to involve two consecutive operations: (1) removal of metal foulants along the main mass transfer channels connected to the narrow pores until the pore structure begins to develop and (2) removal of metal foulants from the pore structure. Both kinetic and mass transfer aspects were considered in the model development, and a good agreement was noticed between experimental and simulated results

Recycling of spent adsorbents for oxyanions and heavy metal ions in the production of ceramics.

Science.gov (United States)

Verbinnen, Bram; Block, Chantal; Van Caneghem, Jo; Vandecasteele, Carlo

2015-11-01

Spent adsorbents for oxyanion forming elements and heavy metals are classified as hazardous materials and they are typically treated by stabilization/solidification before landfilling. The use of lime or cement for stabilization/solidification entails a high environmental impact and landfilling costs are high. This paper shows that mixing spent adsorbents in the raw material for the production of ceramic materials is a valuable alternative to stabilize oxyanion forming elements and heavy metals. The produced ceramics can be used as construction material, avoiding the high economic and environmental impact of stabilization/solidification followed by landfilling. To study the stabilization of oxyanion forming elements and heavy metals during the production process, two series of experiments were performed. In the first series of experiments, the main pollutant, Mo was adsorbed onto iron-based adsorbents, which were then mixed with industrial sludge (3 w/w%) and heated at 1100°C for 30 min. Mo was chosen, as this element is easily adsorbed onto iron-based adsorbents and it is the element that is the most difficult to stabilize (i.e. the highest temperatures need to be reached before the concentrations in the leachate are reduced). Leaching concentration from the 97/3 sludge/adsorbent mixture before heating ranged between 85 and 154 mg/kg; after the heating process they were reduced to 0.42-1.48 mg/kg. Mo was actually stabilized, as the total Mo concentration after addition was not affected by the heat treatment. In the second series of experiments, the sludge was spiked with other heavy metals and oxyanion forming elements (Cr, Ni, Cu, Zn, As, Cd and Pb) in concentrations 5 times higher than the initial concentrations; after heat treatment the leachate concentrations were below the regulatory limit values. The incorporation of spent adsorbents in ceramic materials is a valuable and sustainable alternative to the existing treatment methods, saving raw materials in the

Potential for preparation of hot gas cleanup sorbents from spent hydroprocessing catalysts

Energy Technology Data Exchange (ETDEWEB)

Furimsky, E.; Biagini, M. [Canada Centre for Mineral and Energy Technology, Ottawa, ON (Canada). Energy Research Labs.

1996-01-01

Three spent-decoked hydroprocessing catalysts and two corresponding fresh catalysts were tested as hot gas clean-up sorbents and compared with the zinc ferrite using a simulated coal gasification gas mixture. The catalysts deposited only by coke exhibited relatively good cleaning efficiency. The catalyst deposited by coke and metals such as vanadium and nickel was less efficient. The useful life of the spent hydroprocessing catalysts may be extended if utilized as hot gas clean-up sorbents. 12 refs., 3 figs., 4 tabs.

Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

Dry refabrication technology development of spent nuclear fuel

International Nuclear Information System (INIS)

Park, Geun Il; Lee, J. W.; Song, K. C.

2012-04-01

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

Dry refabrication technology development of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Geun Il; Lee, J. W.; Song, K. C.; and others

2012-04-15

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

International Nuclear Information System (INIS)

In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

2014-01-01

A dry storage system is used for the interim storage of spent fuel prior to permanent depository and/or recycling. The spent fuel is initially stored in a water pool for more than 5 years at least after dispatch from the reactor core and is transported to dry storage. The dry cask contains a multiple number of spent fuel assemblies, which are cooled down in the spent fuel pool. The dry cask is usually filled up with helium gas for increasing the heat transfer to the environment outside the cask. The dry storage system has been used for more than a decade in United States of America (USA) and the European Union (EU). Korea is also developing a dry storage system since its spent fuel pool is anticipated to be full within 10 years. The spent fuel will be stored in a dry cask for more than 40 years. The integrity and safety of spent fuel are important for long-term dry storage. The long-term storage will experience the degradation of spent fuel such as the embrittlement of fuel cladding, thermal creep and hydride reorientation. High burn-up fuel may expedite the material degradation. It is known that the cladding temperature has a strong influence on the material degradation. Hence, it is necessary to accurately predict the local distribution of the cladding temperature using the Computational Fluid Dynamics (CFD) approach. The objective of this study is to apply the CFD method for predicting the three-dimensional distribution of fuel temperature in a dry cask. This CFD study simulated the dry cask for containing the 21 fuel assemblies under development in Korea. This paper presents the fluid velocity and temperature distribution as well as the fuel temperature. A two-step CFD approach was applied to simulate the heat and fluid flow in a dry storage of 21 spent fuel assemblies. The first CFD analysis predicted the helium flow and temperature in a dry cask by a assuming porous body of the spent fuel. The second CFD analysis was to simulate a spent fuel assembly in the

Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

International Nuclear Information System (INIS)

Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

2003-01-01

Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

Science.gov (United States)

Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

2012-01-01

An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

Residual salt separation from simulated spent nuclear fuel reduced in a LiCl-Li2O salt

International Nuclear Information System (INIS)

Hur, Jin-Mok; Hong, Sun-Seok; Seo, Chung-Seok

2006-01-01

The electrochemical reduction of spent nuclear fuel in LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel Atomic Energy Society of Japan, Tokyo, Japan, All rights reservedopyriprocess. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of the LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below the temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950degC were measured as follows: LiCl-8 wt% Li 2 O>LiCl>LiCl-8 wt% SrCl 2 >SrCl 2 . (author)

Hydrometallurgical Recovery of Metal Values From Spent Dry Battery Cells

International Nuclear Information System (INIS)

Rabah, M.A.; Barakat, M.A.; Mahrous, Y.Sh.

1999-01-01

This study focuses on the recovery of metal values from spent dry battery cells (DBC) applying a hydro-pyrometallurgical method. A process flow sheet was followed up starting with cutting the DBC with toothed cutter disc followed by water soaking and rinsing. Water soluble ingredients were filtered. Solid residue was assorted with the help of magnetic separation and water flotation.The method utilizes hydrogen peroxide to enhance dissolution of these metals in acidic or alkaline leachants. Parameters affecting the recovery efficiency such as stoichiometric ratio, solid: liquid ratio, temperature, time and ph of the system were investigated. In this concern, experiments were executed with a battery sample weighing up to 15 kg. Atomic absorption analysis showed that the input DBC contain appreciable amounts of metal zinc, zinc chloride and manganese that are recoverable.Results obtained revealed that metallic parts, carbon rods and paper were safely separated for recycling. From the water-soluble salts, pure NH 4 CI, MnO 2 and ZnCI 2 salts are obtained meeting the standard specifications. Temperature up to 55 degree enhances the recovery process. Under the optimum conditions, maximum recovery efficiency obtained amounts to 93% for Mn, and 99.5% for Zn and NH 4 CI. A model for explaining the obtained results was also given. Dissolution of metals concerned increases in the order nitric> hydrochloric acid. Results were explained in the premise of the kinetic and thermodynamic properties of the reactions involved. Cost estimate of the products shows that the prices of the products are competitive to those of the market prices

Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries

International Nuclear Information System (INIS)

Chen, Xiangping; Chen, Yongbin; Zhou, Tao; Liu, Depei; Hu, Hang; Fan, Shaoyun

2015-01-01

Highlights: • Selective precipitation and solvent extraction were adopted. • Nickel, cobalt and lithium were selectively precipitated. • Co-D2EHPA was employed as high-efficiency extraction reagent for manganese. • High recovery percentages could be achieved for all metal values. - Abstract: Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe–Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC 2 O 4 ⋅2H 2 O and Li 2 CO 3 using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor

Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiangping; Chen, Yongbin; Zhou, Tao, E-mail: zhoutao@csu.edu.cn; Liu, Depei; Hu, Hang; Fan, Shaoyun

2015-04-15

Highlights: • Selective precipitation and solvent extraction were adopted. • Nickel, cobalt and lithium were selectively precipitated. • Co-D2EHPA was employed as high-efficiency extraction reagent for manganese. • High recovery percentages could be achieved for all metal values. - Abstract: Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe–Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC{sub 2}O{sub 4}⋅2H{sub 2}O and Li{sub 2}CO{sub 3} using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor.

Simulation of the mechanical behavior of a spent fuel shipping cask in a rail accident environment

International Nuclear Information System (INIS)

Fields, S.R.

1977-02-01

A preliminary mathematical model has been developed to simulate the dynamic mechanical response of a large spent fuel shipping cask to the impact experienced in a hypothetical rail accident. The report was written to record the status of the development of the mechanical response model and to supplement an earlier report on spent fuel shipping cask accident evaluation

Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries.

Science.gov (United States)

Chen, Xiangping; Chen, Yongbin; Zhou, Tao; Liu, Depei; Hu, Hang; Fan, Shaoyun

2015-04-01

Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe-Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC2O4 ⋅ 2H2O and Li2CO3 using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor. Copyright © 2015 Elsevier Ltd. All rights reserved.

Spent fuel pyroprocessing demonstration

International Nuclear Information System (INIS)

McFarlane, L.F.; Lineberry, M.J.

1995-01-01

A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

Spent fuel reprocessing method

International Nuclear Information System (INIS)

Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

1996-01-01

Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

Residual salts separation from metal reduced electrolytically in a LiCl-Li2O molten salt

International Nuclear Information System (INIS)

Hur, Jin Mok; Oh, Seung Chul; Hong, Sun Seok; Seo, Chung Seok; Park, Seong Won

2005-01-01

The PWR spent oxide fuel can be reduced electrolytically in a hot molten salt for the conditioning and the preparation of a metallic fuel. Then the metal product is smelted into an ingot to be treated in the post process. Incidentally, the residual salt which originated from the molten salt and spent fuel elements should be separated from the metal product during the smelting. In this work, we constructed a surrogate material system to simulate the salt separation from the reduced spent fuel and studied the vaporization behaviors of the salts

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Oil removal of spent hydrotreating catalyst CoMo/Al{sub 2}O{sub 3} via a facile method with enhanced metal recovery

Energy Technology Data Exchange (ETDEWEB)

Yang, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu, Shengming, E-mail: smxu@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Beijing Key Lab of Radioactive Wastes Treatment, Tsinghua University, Beijing 100084 (China); Li, Zhen [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wang, Jianlong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Beijing Key Lab of Radioactive Wastes Treatment, Tsinghua University, Beijing 100084 (China); Zhao, Zhongwei [School of Metallurgy and Environment, Central South University, Changsha 410083, Hunan (China); Xu, Zhenghe, E-mail: zhenghe.xu@ualberta.ca [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Department of Chemical and Material Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada)

2016-11-15

Highlights: • A novel approach for oil removal from spent hydrotreating catalysts has been developed. • Oil removal possibility is analyzed through surface characteristics. • Oil is successfully removed from spent catalysts via aqueous surfactant solution. • Over 98% Mo can be leached after oil removal and thermal treatment. • The proposed deoiling method helps to avoid detrimental impurity generation (CoMoO{sub 4}) and enhance metal recovery. - Abstract: Deoiling process is a key issue for recovering metal values from spent hydrotreating catalysts. The oils can be removed with organic solvents, but the industrialized application of this method is greatly hampered by the high cost and complex processes. Despite the roasting method is simple and low-cost, it generates hardest-to-recycle impurities (CoMoO{sub 4} or NiMoO{sub 4}) and enormous toxic gases. In this study, a novel and facile approach to remove oils from the spent hydrotreating catalysts is developed. Firstly, surface properties of spent catalysts are characterized to reveal the possibility of oil removal. And then, oils are removed with water solution under the conditions of 90 °C, 0.1 wt% SDS, 2.0 wt% NaOH and 10 ml/g L/S ratio for 4 h. Finally, thermal treatment and leaching tests are carried out to further explore the advantages of oil removal. The results show that no hardest-to-recycle impurity CoMoO{sub 4} is found in XPS spectra of thermally treated samples after deoiling and molybdenum is leached completely with sodium carbonate solution. It means that the proposed deoiling method can not only remove oils simply and without enormous harmful gases generating, but also avoid the generation of detrimental impurity and promote recycling of valuable metals from spent hydrotreating catalysts.

Spent fuel storage and isolation

International Nuclear Information System (INIS)

Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

1979-02-01

The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

Using contraband simulators for portal metal detector testing

Energy Technology Data Exchange (ETDEWEB)

Murray, D.W.

1992-08-01

Because contraband materials or items are either too dangerous or too expensive, contraband simulators have been widely used to test contraband detection equipment. Very realistic bomb simulators have been used to test x-ray scanners, and common radioactive sources have been used successfully to test the operation of special nuclear material (SNM) radiation detectors. The simulators used to test early metal detectors were also reasonably successful; however, these simulators were rapidly outdated by the introduction of modern active field metal detectors. This paper describes some of the earlier attempts to develop metal detector test simulators. A successful highly enriched uranium (HEU) simulator for metal detector testing is described that has duplicated all the characteristics modern equipment is capable of detecting. The paper also describes the development needed to produce handgun simulators that could be used effectively for metal detector performance testing.

Using contraband simulators for portal metal detector testing

Energy Technology Data Exchange (ETDEWEB)

Murray, D.W.

1992-01-01

Because contraband materials or items are either too dangerous or too expensive, contraband simulators have been widely used to test contraband detection equipment. Very realistic bomb simulators have been used to test x-ray scanners, and common radioactive sources have been used successfully to test the operation of special nuclear material (SNM) radiation detectors. The simulators used to test early metal detectors were also reasonably successful; however, these simulators were rapidly outdated by the introduction of modern active field metal detectors. This paper describes some of the earlier attempts to develop metal detector test simulators. A successful highly enriched uranium (HEU) simulator for metal detector testing is described that has duplicated all the characteristics modern equipment is capable of detecting. The paper also describes the development needed to produce handgun simulators that could be used effectively for metal detector performance testing.

Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

International Nuclear Information System (INIS)

Shimizu, Masashi; Hayashi, Makoto; Kashiwakura, Jun

2003-01-01

Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

Metallic ions catalysis for improving bioleaching yield of Zn and Mn from spent Zn-Mn batteries at high pulp density of 10.

Science.gov (United States)

Niu, Zhirui; Huang, Qifei; Wang, Jia; Yang, Yiran; Xin, Baoping; Chen, Shi

2015-11-15

Bioleaching of spent batteries was often conducted at pulp density of 1.0% or lower. In this work, metallic ions catalytic bioleaching was used for release Zn and Mn from spent ZMBs at 10% of pulp density. The results showed only Cu(2+) improved mobilization of Zn and Mn from the spent batteries among tested four metallic ions. When Cu(2+) content increased from 0 to 0.8 g/L, the maximum release efficiency elevated from 47.7% to 62.5% for Zn and from 30.9% to 62.4% for Mn, respectively. The Cu(2+) catalysis boosted bioleaching of resistant hetaerolite through forming a possible intermediate CuMn2O4 which was subject to be attacked by Fe(3+) based on a cycle of Fe(3+)/Fe(2+). However, poor growth of cells, formation of KFe3(SO4)2(OH)6 and its possible blockage between cells and energy matters destroyed the cycle of Fe(3+)/Fe(2+), stopping bioleaching of hetaerolite. The chemical reaction controlled model fitted best for describing Cu(2+) catalytic bioleaching of spent ZMBs. Copyright © 2015 Elsevier B.V. All rights reserved.

Oil removal of spent hydrotreating catalyst CoMo/Al2O3 via a facile method with enhanced metal recovery.

Science.gov (United States)

Yang, Yue; Xu, Shengming; Li, Zhen; Wang, Jianlong; Zhao, Zhongwei; Xu, Zhenghe

2016-11-15

Deoiling process is a key issue for recovering metal values from spent hydrotreating catalysts. The oils can be removed with organic solvents, but the industrialized application of this method is greatly hampered by the high cost and complex processes. Despite the roasting method is simple and low-cost, it generates hardest-to-recycle impurities (CoMoO4 or NiMoO4) and enormous toxic gases. In this study, a novel and facile approach to remove oils from the spent hydrotreating catalysts is developed. Firstly, surface properties of spent catalysts are characterized to reveal the possibility of oil removal. And then, oils are removed with water solution under the conditions of 90°C, 0.1wt% SDS, 2.0wt% NaOH and 10ml/gL/S ratio for 4h. Finally, thermal treatment and leaching tests are carried out to further explore the advantages of oil removal. The results show that no hardest-to-recycle impurity CoMoO4 is found in XPS spectra of thermally treated samples after deoiling and molybdenum is leached completely with sodium carbonate solution. It means that the proposed deoiling method can not only remove oils simply and without enormous harmful gases generating, but also avoid the generation of detrimental impurity and promote recycling of valuable metals from spent hydrotreating catalysts. Copyright © 2016 Elsevier B.V. All rights reserved.

Leaching of copper and zinc from spent antifouling paint particles

International Nuclear Information System (INIS)

Singh, Nimisha; Turner, Andrew

2009-01-01

Leaching of Cu and Zn from a composite of spent antifouling paint particles, containing about 300 mg g -1 and 110 mg g -1 of the respective metals, was studied in batch experiments. For a given set of simulated environmental conditions, release of Cu was independent of paint particle concentration due to attainment of pseudo-saturation, but Zn was less constrained by solubility effects and release increased with increasing particle concentration. Leaching of Cu increased but Zn decreased with increasing salinity, consistent with mechanisms governing the dissolution of Cu 2 O in the presence of chloride and Zn acrylates in the presence of seawater cations. Because of complex reaction kinetics and the presence of calcium carbonate in the paint matrix, metal leaching appeared to be greater at 4 deg. C than 19 deg. C under many conditions. These findings have important environmental and biological implications regarding the deliberate or inadvertent disposal of antifouling paint residues. - Copper and zinc are readily leached from particles of spent antifouling paint under a range of environmental conditions

Leaching of copper and zinc from spent antifouling paint particles

Energy Technology Data Exchange (ETDEWEB)

Singh, Nimisha [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Turner, Andrew [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)], E-mail: aturner@plymouth.ac.uk

2009-02-15

Leaching of Cu and Zn from a composite of spent antifouling paint particles, containing about 300 mg g{sup -1} and 110 mg g{sup -1} of the respective metals, was studied in batch experiments. For a given set of simulated environmental conditions, release of Cu was independent of paint particle concentration due to attainment of pseudo-saturation, but Zn was less constrained by solubility effects and release increased with increasing particle concentration. Leaching of Cu increased but Zn decreased with increasing salinity, consistent with mechanisms governing the dissolution of Cu{sub 2}O in the presence of chloride and Zn acrylates in the presence of seawater cations. Because of complex reaction kinetics and the presence of calcium carbonate in the paint matrix, metal leaching appeared to be greater at 4 deg. C than 19 deg. C under many conditions. These findings have important environmental and biological implications regarding the deliberate or inadvertent disposal of antifouling paint residues. - Copper and zinc are readily leached from particles of spent antifouling paint under a range of environmental conditions.

Reuse of Hydrotreating Spent Catalyst

International Nuclear Information System (INIS)

Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

2004-01-01

All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

International Nuclear Information System (INIS)

Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

2005-01-01

The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

Seal performance of thermal aged metal gasket of dual purpose metal cask for interim spent fuel storage after external impact load

International Nuclear Information System (INIS)

Takeshi Yokoyama; Masami Kato; Satoshi Itooka

2005-01-01

As for interim storage for spent nuclear fuels using dual purpose dry metal cask in Japan, we recognize one of the important technical issues that there is a possibility for the cask with degraded metal gasket during storage to apply to transportation. In our study until 2003 focused on degradation of important components for the cask safety performance during storage and application to transportation, for metal gasket, we conducted property tests for degradation and influence of lid movement on seal performance, and furthermore verification tests. From the results, we developed the method to evaluate leak rate from lid with degraded metal gasket at accidents during transportation and in addition, we found as follows: Lid would hardly move and leak rate would not increase seriously during fire event. Leak rate from lid with degraded metal gasket could be evaluated by using results of leak rate trend depending on horizontal displacement of lid by external impact load. So, with regard to influence of lid movement on seal performance, we conducted additional test for extending horizontal displacement in lid moving in 2004. In addition, seal performance was discussed from the results, both previous and latest test. (authors)

Development of the graphic design and control system based on a graphic simulator for the spent fuel dismantling equipment

Energy Technology Data Exchange (ETDEWEB)

Lee, J. Y.; Kim, S. H.; Song, T. G.; Yoon, J. S

2000-06-01

In this study, the graphic design system is developed for designing the spent fuel rod consolidation and the dismantling processes. This system is used throughout the design stages from the conceptual design to the motion analysis. Also, the real-time control system of the rod extracting equipment is developed. This system utilizes the graphic simulator which simulates the motion of the equipment in real time by synchronously connecting the control PC with the graphic server through the TCP/IP network. The developed system is expected to be used as an effective tool in designing the process equipment for the spent fuel management. And the real-time graphic control system can be effectively used to enhance the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process.

Development of the graphic design and control system based on a graphic simulator for the spent fuel dismantling equipment

International Nuclear Information System (INIS)

Lee, J. Y.; Kim, S. H.; Song, T. G.; Yoon, J. S.

2000-06-01

In this study, the graphic design system is developed for designing the spent fuel rod consolidation and the dismantling processes. This system is used throughout the design stages from the conceptual design to the motion analysis. Also, the real-time control system of the rod extracting equipment is developed. This system utilizes the graphic simulator which simulates the motion of the equipment in real time by synchronously connecting the control PC with the graphic server through the TCP/IP network. The developed system is expected to be used as an effective tool in designing the process equipment for the spent fuel management. And the real-time graphic control system can be effectively used to enhance the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process

Simulation of interim spent fuel storage system with discrete event model

International Nuclear Information System (INIS)

Yoon, Wan Ki; Song, Ki Chan; Lee, Jae Sol; Park, Hyun Soo

1989-01-01

This paper describes dynamic simulation of the spent fuel storage system which is described by statistical discrete event models. It visualizes flow and queue of system over time, assesses the operational performance of the system activities and establishes the system components and streams. It gives information on system organization and operation policy with reference to the design. System was tested and analyzed over a number of critical parameters to establish the optimal system. Workforce schedule and resources with long processing time dominate process. A combination of two workforce shifts a day and two cooling pits gives the optimal solution of storage system. Discrete system simulation is an useful tool to get information on optimal design and operation of the storage system. (Author)

Fabrication of simulated DUPIC fuel

Science.gov (United States)

Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

2000-12-01

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

The importance of pre-treatment of spent hydrotreating catalysts on metals recovery

Directory of Open Access Journals (Sweden)

Alexandre Luiz de Souza Pereira

2011-01-01

Full Text Available This work describes a three-step pre-treatment route for processing spent commercial NiMo/Al2O3 catalysts. Extraction of soluble coke with n-hexane and/or leaching of foulant elements with oxalic acid were performed before burning insoluble coke under air. Oxidized catalysts were leached with 9 mol L-1 sulfuric acid. Iron was the only foulant element partially leached by oxalic acid. The amount of insoluble matter in sulfuric acid was drastically reduced when iron and/or soluble coke were previously removed. Losses of active phase metals (Ni, Mo during leaching with oxalic acid were compensated by the increase of their recovery in the sulfuric acid leachate.

The importance of pre-treatment of spent hydrotreating catalysts on metals recovery

Energy Technology Data Exchange (ETDEWEB)

Pereira, Alexandre Luiz de Souza; Silva, Cristiano Nunes da; Afonso, Julio Carlos, E-mail: julio@iq.ufrj.b [Universidade Federal do Rio de Janeiro (IQ/UFRJ), RJ (Brazil). Inst. de Quimica. Dept. de Quimica Analitica; Mantovano, Jose Luiz [Instituto de Engenharia Nuclear (CNEN/IEN-RJ), Rio de Janeiro, RJ (Brazil). Dept. de Quimica e Materiais Nucleares

2011-07-01

This work describes a three-step pre-treatment route for processing spent commercial Ni Mo/Al{sub 2}O{sub 3} catalysts. Extraction of soluble coke with n-hexane and/or leaching of foulant elements with oxalic acid were performed before burning insoluble coke under air. Oxidized catalysts were leached with 9 mol L{sup -1} sulfuric acid. Iron was the only foulant element partially leached by oxalic acid. The amount of insoluble matter in sulfuric acid was drastically reduced when iron and/or soluble coke were previously removed. Losses of active phase metals (Ni, Mo) during leaching with oxalic acid were compensated by the increase of their recovery in the sulfuric acid leachate. (author)

Hydroprocessing using regenerated spent heavy hydrocarbon catalyst

International Nuclear Information System (INIS)

Clark, F.T.; Hensley, A.L. Jr.

1992-01-01

This patent describes a process for hydroprocessing a hydrocarbon feedstock. It comprises: contacting the feedstock with hydrogen under hydroprocessing conditions with a hydroprocessing catalyst wherein the hydroprocessing catalyst contains a total contaminant metals build-up of greater than about 4 wt. % nickel plus vanadium, a hydrogenation component selected from the group consisting of Group VIB metals and Group VIII metals and is regenerated spent hydroprocessing catalyst regenerated by a process comprising the steps: partially decoking the spent catalyst in an initial coke-burning step; impregnating the partially decoked catalyst with a Group IIA metal-containing impregnation solution; and decoking the impregnated catalyst in a final coke-burning step wherein the impregnated catalyst is contacted with an oxygen-containing gas at a temperature of about 600 degrees F to about 1400 degrees F

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

Energy Technology Data Exchange (ETDEWEB)

Margraf, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-08-02

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO₂) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO₂ results to get a proper measure for spent fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO₂). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO₂ and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.

Modelling Methods of Magnetohydrodynamic Phenomena Occurring in a Channel of the Device Used to Wash Out the Spent Automotive Catalyst by a Liquid Metal

Directory of Open Access Journals (Sweden)

Fornalczyk A.

2016-06-01

Full Text Available The recovery of precious metals is necessary for environmental and economic reasons. Spent catalysts from automotive industry containing precious metals are very attractive recyclable material as the devices have to be periodically renovated and eventually replaced. This paper presents the method of removing platinum from the spent catalytic converters applying lead as a collector metal in a device used to wash out by using mangetohydrodynamic stirrer. The article includes the description of the methods used for modeling of magnetohydrodynamic phenomena (coupled analysis of the electromagnetic, temperature and flow fields occurring in this particular device. The paper describes the general phenomena and ways of coupling the various physical fields for this type of calculation. The basic computational techniques with a discussion of their advantages and disadvantages are presented.

Simulations of Recrystallization in Metals

DEFF Research Database (Denmark)

Godiksen, Rasmus Brauner

2007-01-01

structures in the deformed metal due to local effects: Inhomogeneous boundary morphologies and dislocation-structure-dependent migration rates are observed. The effects that the dislocation structures have must be taken into account in order to create realistic recrystallization models, and through......The growth of new near-perfect grains during recrystallization of deformed metals is governed by the migration of the grain boundaries surrounding the new grains. The grain boundaries migrate through the deformed metal driven by the excess energy of the dislocation structures created during...... deformation. Recently, it has been found that recrystallization is far more inhomogeneous than previously thought. The purpose of this PhD-project is to study recrystallization by computer simulations with special focus on inhomogeneous growth. Two types of simulations have been employed: geometric...

A Qualitative Analysis of the Neutron Population in Fresh and Spent Fuel Assemblies during Simulated Interrogation using the Differential Die-Away Technique

International Nuclear Information System (INIS)

Lundkvista, Niklas; Goodsell, Alison V.; Grapea, Sophie; Hendricksb, John S.; Henzlb, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

2015-01-01

Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is being considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.

Spent fuel management

International Nuclear Information System (INIS)

2005-01-01

The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

Spent fuel interim management: 1995 update

International Nuclear Information System (INIS)

Anderson, C.K.

1995-01-01

The problems of interim away-from-reactor spent fuel storage and storage in spent fuel pools at the reactor site are discussed. An overview of the state-of-the-art in the USA, Europe, and Japan is presented. The technical facilities for away-from-reactor storage are briefly described, including wet storage pools, interactive concrete systems, metallic containers, and passive concrete systems. Reprocessing technologies are mostly at the design stage only. It is predicted that during the 20 years to come, about 50 000 tonnes of spent fuel will be stored at reactor sites regardless of the advance of spent fuel reprocessing or interim storage projects. (J.B.). 4 tabs., 2 figs

Defense by-products production and utilization program: noble metal recovery screening experiments

International Nuclear Information System (INIS)

Hazelton, R.F.; Jensen, G.A.; Raney, P.J.

1986-03-01

Isotopes of the platinum metals (rutheium, rhodium, and palladium) are produced during uranium fuel fission in nuclear reactors. The strategic values of these noble metals warrant considering their recovery from spent fuel should the spent fuel be processed after reactor discharge. A program to evaluate methods for ruthenium, rhodium, and palladium recovery from spent fuel reprocessing liquids was conducted at Pacific Northwest Laboratory (PNL). The purpose of the work reported in this docuent was to evaluate several recovery processes revealed in the patent and technical literature. Beaker-scale screening tests were initiated for three potential recovery processes: precipitation during sugar denitration of nitric acid reprocessing solutions after plutonium-uranium solvent extraction, adsorption using nobe metal selective chelates on active carbon, and reduction forming solid noble metal deposits on an amine-borane reductive resin. Simulated reprocessing plant solutions representing typical nitric acid liquids from defense (PUREX) or commercial fuel reprocessing facilities were formulated and used for evaluation of the three processes. 9 refs., 3 figs., 9 tabs

A review of metal recovery from spent petroleum catalysts and ash.

Science.gov (United States)

Akcil, Ata; Vegliò, Francesco; Ferella, Francesco; Okudan, Mediha Demet; Tuncuk, Aysenur

2015-11-01

With the increase in environmental awareness, the disposal of any form of hazardous waste has become a great concern for the industrial sector. Spent catalysts contribute to a significant amount of the solid waste generated by the petrochemical and petroleum refining industry. Hydro-cracking and hydrodesulfurization (HDS) catalysts are extensively used in the petroleum refining and petrochemical industries. The catalysts used in the refining processes lose their effectiveness over time. When the activity of catalysts decline below the acceptable level, they are usually regenerated and reused but regeneration is not possible every time. Recycling of some industrial waste containing base metals (such as V, Ni, Co, Mo) is estimated as an economical opportunity in the exploitation of these wastes. Alkali roasted catalysts can be leached in water to get the Mo and V in solution (in which temperature plays an important role during leaching). Several techniques are possible to separate the different metals, among those selective precipitation and solvent extraction are the most used. Pyrometallurgical treatment and bio-hydrometallurgical leaching were also proposed in the scientific literature but up to now they did not have any industrial application. An overview on patented and commercial processes was also presented. Copyright © 2015 Elsevier Ltd. All rights reserved.

Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility is one of main objectives of JNES activities. For the thermal and structural analyses, the radiative heat transfer analysis code S-FOKS has been developed to reduce computing time and to avoid using large memory area. In order to simulate the specular reflection, a new model (called 'model-2') is planned to install to S-FOKS code. The theoretical values with the specular reflection in simple geometry were lead to verify S-FOKS model-2. (author)

Phase 1 study of metallic cask systems for spent fuel management from reactor to repository. Volume I. Phase 1 study summary

International Nuclear Information System (INIS)

1986-02-01

It was proposed to perform a systems evaluation of metallic cask systems in order to define and examine the use of various metallic cask concepts or combination of concepts for the overall inventory management of spent fuel starting with its discharge from reactors to its emplacement in geologic repositories. This systems evaluation occurs in three phases. This three phase systems evaluation leads to a definition and recommendation of a sound and practical metallic cask system to accomplish efficient and effective management of spent fuel in the back end of the nuclear fuel cycle. Phase 1 Study objectives: establish system-wide functional criteria and assumptions; perform the systems engineering needed to define the metallic cask concepts and their feasibility; perform a screening evaluation of the technical and economic merits of the concepts; and recommend those to be included for a more detailed systems evaluation in Phase 2. Phase 2 Study objectives: refine the system-wide functional criteria and assumptions; perform the design engineering needed to enhance the validity and workability of those concepts recommended in Phase 1; and perform a more detailed systems evaluation. Phase 3 Study objectives: conclude the systems evaluation and develop an implementation plan. Volume I presents an overview of the detailed systems evaluation presented in Volume II

EBR-II spent fuel treatment demonstration project

International Nuclear Information System (INIS)

Benedict, R.W.; Henslee, S.P.

1997-01-01

For approximately 10 years, Argonne National Laboratory was developed a fast reactor fuel cycle based on dry processing. When the US fast reactor program was canceled in 1994, the fuel processing technology, called the electrometallurgical technique, was adapted for treating unstable spent nuclear fuel for disposal. While this technique, which involves electrorefining fuel in a molten salt bath, is being developed for several different fuel categories, its initial application is for sodium-bonded metallic spent fuel. In June 1996, the Department of Energy (DOE) approved a radiation demonstration program in which 100 spent driver assemblies and 25 spent blanket assemblies from the Experimental Breeder Reactor-II (EBR-II) will be treated over a three-year period. This demonstrated will provide data that address issues in the National Research Council's evaluation of the technology. The planned operations will neutralize the reactive component (elemental sodium) in the fuel and produce a low enriched uranium product, a ceramic waste and a metal waste. The fission products and transuranium elements, which accumulate in the electrorefining salt, will be stabilized in the glass-bonded ceramic waste form. The stainless steel cladding hulls, noble metal fission products, and insoluble residues from the process will be stabilized in a stainless steel/zirconium alloy. Upon completion of a successful demonstration and additional environmental evaluation, the current plans are to process the remainder of the DOE sodium bonded fuel

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

1997-12-01

Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

Separation and Recovery of Precious Metals from Leach Liquors of Spent Electronic Wastes by Solvent Extraction

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Thi Hong; Wang, Lingyun; Lee, Man Seung [Mokpo National University, Mokpo (Korea, Republic of)

2017-04-15

Solvent extraction was employed to recover precious metals (Au (III), Pd (II) and Pt (IV)) from the leach solution of spent electronic wastes containing Cu (II), Cr (III) and Fe (III). First, pure Fe (III) and Au (III) were recovered by simultaneous extraction with Cyanex 923 followed by selective stripping with HCl and Na{sub 2}S{sub 2}O{sub 3}. Second, Pt (IV), Pd (II) and Cu (II) were extracted by Alamine 336 from the raffinate. After the removal of Cu (II) by stripping with weak HCl, Pd (II) and Pt (IV) were separately stripped by controlling the concentration of thiourea in the mixture with HCl. A process flow sheet for the separation of precious metals was proposed.

Development of spent fuel remote handling technology

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

1997-12-01

Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Shin, Young Joon; Cho, S. H.; You, G. S.

2001-04-01

Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

European workshop on spent catalysts. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

In 1999 and 2002 two well attended workshops on recycling, regeneration, reuse and disposal of spent catalysts took place in Frankfurt. This series has been continued in Berlin. The workshop was organized in collaboration with DGMK, the German Society for Petroleum and Coal Science and Technology. Contributions were in the following areas of catalyst deactivation: recycling of spent catalysts in chemical and petrochemical industry, recycling of precious metal catalysts and heterogenous base metal catalysts, legal aspects of transboundary movements, catalyst regeneration, quality control, slurry catalysts, commercial reactivation of hydrotreating catalysts. (uke)

ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

Energy Technology Data Exchange (ETDEWEB)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2015-10-15

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.

ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

International Nuclear Information System (INIS)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

2015-01-01

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management

Safe transport of spent fuels after long-term storage

International Nuclear Information System (INIS)

Aritomi, M.; Takeda, T.; Ozaki, S.

2004-01-01

Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly

Process optimization and leaching kinetics of zinc and manganese metals from zinc-carbon and alkaline spent batteries using citric acid reagent

Science.gov (United States)

Yuliusman; Amiliana, R. A.; Wulandari, P. T.; Huda, M.; Kusumadewi, F. A.

2018-03-01

Zn-Carbon and Alkaline spent batteries contains heavy metals, such as zinc and manganese, which can causes environmental problem if not handled properly. Usually the recovery of these metals were done by leaching method using strong acid, but the use of strong acids as leaching reagents can be harmful to the environment. This paper concerns the recovery of Zn and Mn metals from Zn-C and alkaline spent batteries with leaching method using citric acid as the environmental friendly leaching reagent. The leaching conditions using citric acid were optimized and the leaching kinetics of Zn and Mn in citric acid solution was investigated. The leaching of 89.62% Zn and 63.26% Mn was achieved with 1.5 M citric acid, 90°C temperature, and 90 minutes stirring time. Kinetics data for the dissolution of Zn showed the best fit to chemical control shrinking core model, while the diffusion controlled model was suitable for the dissolution of Mn kinetics data. The activation energy of 6.12 and 1.73 kcal/mol was acquired for the leaching of Zn and Mn in the temperature range 60°C-90°C.

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

International Nuclear Information System (INIS)

Poineau, Frederic; Tamalis, Dimitri

2016-01-01

The isotope 99 Tc is an important fission product generated from nuclear power production. Because of its long half-life (t 1/2 = 2.13 ∙ 105 years) and beta-radiotoxicity (β - = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99 Tc ( 99 Tc → 99 Ru + β - ). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the nature of Tc in metallic spent fuel. Computational modeling

Disposal of spent fuel

International Nuclear Information System (INIS)

Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

Extension technology of store ability of spent fuel

International Nuclear Information System (INIS)

1991-05-01

It is the introduction of the extension technology of store ability of spent fuel including metal store cask, transport and store cask, concrete cask, NUHOMS and MVDS. It explains of technology of recombination of spent fuel including the purpose and real application, demonstration, presumption of expense, major interesting issue and the present condition of relevant licences permit and approvals.

Apparatus and method for reprocessing and separating spent nuclear fuels

International Nuclear Information System (INIS)

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H.; Coops, M.S.

1983-01-01

A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A non-oxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel. (author)

Leaching of the simulated borosilicate waste glasses and spent nuclear fuel under a repository condition

International Nuclear Information System (INIS)

Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung; Suh, Hang Suk

2002-12-01

Leaching behaviors of simulated waste glass and spent fuel, contacted on bentonite blocks, in synthetic granitic groundwater were investigated in this study. The leach rate of boron from borosilicate waste glass between the compacted bentonite blocks reached about 0.03 gm-2day-1 at 1500 days, like as that of molybdenum. However, the concentration of uranium in leachate pass through bentonite blocks was less than their detection limits of 2 μg/L and whose yellow amorphous compound was found on the surface of glass contacted with the bentonite blocks. The leaching mechanism of waste glasses differed with their composition. The release rate of cesium from PWR spent fuel in the simulated granitic water without bentonite was leas than $1.0x10 -5 fraction/day after 300 days. The retardation factor of cesium by a 10 -mm thickness of bentonite block was more than 100 for 4-years leaching time. The cumulative release fraction of uranium for 954 days was 0.016% (1.7x10 -7 fraction/day) in granitic water without bentonite. The gap inventory of cesium for spent fuel G23-J11 was 0.15∼0.2%. However, the release of cesium from C15-I08 was 0.9% until 60 days and has being continued after that. Gap inventories of strontium and iodine in G23-J11 were 0.033% and below 0.2%, respectively. The sum of fraction of cesium in gap and grain boundary of G23-J11 was suggested below 3% and less

Spent fuel shipping cask accident evaluation

International Nuclear Information System (INIS)

Fields, S.R.

1975-12-01

Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

Organic reductants based leaching: A sustainable process for the recovery of valuable metals from spent lithium ion batteries.

Science.gov (United States)

Chen, Xiangping; Guo, Chunxiu; Ma, Hongrui; Li, Jiazhu; Zhou, Tao; Cao, Ling; Kang, Duozhi

2018-05-01

It is significant to recover metal values from spent lithium ion batteries (LIBs) for the alleviation or prevention of potential risks towards environmental pollution and public health, as well as for the conservation of valuable metals. Herein a hydrometallurgical process was proposed to explore the possibility for the leaching of different metals from waste cathodic materials (LiCoO 2 ) of spent LIBs using organics as reductant in sulfuric acid medium. According to the leaching results, about 98% Co and 96% Li can be leached under the optimal experimental conditions of reaction temperature - 95 °C, reaction time - 120 min, reductive agent dosage - 0.4 g/g, slurry density - 25 g/L, concentration of sulfuric acid-3 mol/L in H 2 SO 4  + glucose leaching system. Similar results (96% Co and 100% Li) can be obtained in H 2 SO 4  + sucrose leaching system under optimized leaching conditions. Despite a complete leaching of Li (∼100%), only 54% Co can be dissolved in the H 2 SO 4  + cellulose leaching system under optimized leaching conditions. Finally, different characterization methods, including UV-Vis, FT-IR, SEM and XRD, were employed for the tentative exploration of reductive leaching reactions using organic as reductant in sulfuric acid medium. All the leaching and characterization results confirm that both glucose and sucrose are effective reductants during leaching, while cellulose should be further degraded to organics with low molecular weights to achieve a satisfactory leaching performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

Simulation of the ductile damage under the metal forming

International Nuclear Information System (INIS)

Bogatov, A. A.

2003-01-01

Potentiality of metal forming is limited by ductile damage. The damage degree is estimated by the scalar value ω, that is equal to 0(ω=0) before plastic strain and is equal to 1(ω=1) at the macro cracks moment. There are two criteria that describe micro damage. The value ω=ω * corresponds to the generation of micro voids that couldn't be recovered by recrystallization but do not reduce the metal strength. The value ω=ω ** corresponds to the generation of micro voids that reduce the metal strength and material long life. The models of metal damage accumulation under pure and alternate strain also the model of metal damage recovery under the recrystallization are developed. The specimen testing at high loading parameters gives the basic equations of the ductile damage mechanics. All of that gives the method to study ductile damage under the metal forming. The methodology damage nucleation and growing is shown on various examples: the void and crack development in the areas ductile damage and unlimited ductility; mathematical simulation of the metal damage under the sheet and wire drawing and others. The problems of physical simulating at the ductile damage under metal forming are shown too in this paper. The method and equipment of metal damage physical simulation are proposed. (Original)

Leaching behavior of lanthanum, nickel and iron from spent catalyst using inorganic acids

Science.gov (United States)

Astuti, W.; Prilitasari, N. M.; Iskandar, Y.; Bratakusuma, D.; Petrus, H. T. B. M.

2018-01-01

Highly technological applications of rare earth metals (REs) and scarcity of supply have become an incentive torecover the REs from various resources, which include high grade and low grade ores, as well as recycledwaste materials. Spent hydrocracking catalyst contain lanthanum and a variety of valuable metals such as nickel and iron. This study investigated the recovery of lanthanum, nickel and iron from spent hydrocracking catalyst by leaching using various inorganic acid (sulfuric acid, hydrochloric acid, and nitric acid). The effect of acid concentration, type of acid and leaching temperature was conducted to study the leaching behavior of each valuable metal from spent-catalyst. It has been shown that it is possible to recover more than 90% of lanthanum, however the leaching efficiency of nickel and iron in this process was very low. It can be concluded that the leaching process is selective for lanthanum recovery from hydrocracking spent-catalyst.

Performance Evaluation of the Neutron Coincidence Counter for the Advanced Spent Fuel Conditioning Process

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Menlove, Howard O.; Kim, H.D.; Ko, W.I.; Park, S.W.

2005-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

Computer simulations of the mechanical properties of metals

DEFF Research Database (Denmark)

Schiøtz, Jakob; Vegge, Tejs

1999-01-01

Atomic-scale computer simulations can be used to gain a better understanding of the mechanical properties of materials. In this paper we demonstrate how this can be done in the case of nanocrystalline copper, and give a brief overview of how simulations may be extended to larger length scales....... Nanocrystline metals are metals with grain sizes in the nanometre range, they have a number of technologically interesting properties such as much increased hardness and yield strength. Our simulations show that the deformation mechanisms are different in these materials than in coarse-grained materials...

Removal of uranium from spent salt from the moltensalt oxidation process

International Nuclear Information System (INIS)

Summers, L.; Hsu, P.C.; Holtz, E.V.; Hipple, D.; Wang, F.; Adamson, M.

1997-03-01

Molten salt oxidation (MSO) is a thermal process that has the capability of destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials. In this process, combustible waste and air are introduced into the molten sodium carbonate salt. The organic constituents of the waste materials are oxidized to carbon dioxide and water, while most of the inorganic constituents, including toxic metals, minerals, and radioisotopes, are retained in the molten salt bath. As these impurities accumulate in the salt, the process efficiency drops and the salt must be replaced. An efficient process is needed to separate these toxic metals, minerals, and radioisotopes from the spent carbonate to avoid generating a large volume of secondary waste. Toxic metals such as cadmium, chromium, lead, and zinc etc. are removed by a method described elsewhere. This paper describes a separation strategy developed for radioisotope removal from the mixed spent salt, as well as experimental results, as part of the spent salt cleanup. As the MSO system operates, inorganic products resulting from the reaction of halides, sulfides, phosphates, metals and radionuclides with carbonate accumulate in the salt bath. These must be removed to prevent complete conversion of the sodium carbonate, which would result in eventual losses of destruction efficiency and acid scrubbing capability. There are two operational modes for salt removal: (1) during reactor operation a slip-stream of molten salt is continuously withdrawn with continuous replacement by carbonate, or (2) the spent salt melt is discharged completely and the reactor then refilled with carbonate in batch mode. Because many of the metals and/or radionuclides captured in the salt are hazardous and/or radioactive, spent salt removed from the reactor would create a large secondary waste stream without further treatment. A spent salt clean up/recovery system is necessary to segregate these materials and minimize the amount of

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, S. H.

2004-02-01

In this research, the remote handling technology is developed for the advanced spent fuel conditioning process which gives a possible solution to deal with the rapidly increasing spent fuels. In detail, a fuel rod slitting device is developed for the decladding of the spent fuel. A series of experiments has been performed to find out the optimal condition of the spent fuel voloxidation which converts the UO 2 pellet into U 3 O 8 powder. The design requirements of the ACP equipment for hot test is established by analysing the modular requirement, radiation hardening and thermal protection of the process equipment, etc. The prototype of the servo manipulator is developed. The manipulator has an excellent performance in terms of the payload to weight ratio that is 30 % higher than that of existing manipulators. To provide reliability and safety of the ACP, the 3 dimensional graphic simulator is developed. Using the simulator the remote handling operation is simulated and as a result, the optimal layout of ACP is obtained. The supervisory control system is designed to control and monitor the several different unit processes. Also the failure monitoring system is developed to detect the possible accidents of the reduction reactor

Pyrochemical processing of DOE spent nuclear fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1995-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

Characteristics of spent nuclear fuel

International Nuclear Information System (INIS)

Notz, K.J.

1988-04-01

The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

Energy Technology Data Exchange (ETDEWEB)

Poineau, Frederic [Univ. of Nevada, Las Vegas, NV (United States); Tamalis, Dimitri [Florida Memorial Univ., Miami Gardens, FL (United States)

2016-08-01

nature of Tc in metallic spent fuel. Computational modeling and simulations were performed to shed light on experimental results and explain structural and kinetics trends.

The effectiveness of spent coffee grounds and its biochar on the amelioration of heavy metals-contaminated water and soil using chemical and biological assessments.

Science.gov (United States)

Kim, Min-Suk; Min, Hyun-Gi; Koo, Namin; Park, Jeongsik; Lee, Sang-Hwan; Bak, Gwan-In; Kim, Jeong-Gyu

2014-12-15

Spent coffee grounds (SCG) and charred spent coffee grounds (SCG-char) have been widely used to adsorb or to amend heavy metals that contaminate water or soil and their success is usually assessed by chemical analysis. In this work, the effects of SCG and SCG-char on metal-contaminated water and soil were evaluated using chemical and biological assessments; a phytotoxicity test using bok choy (Brassica campestris L. ssp. chinensis Jusl.) was conducted for the biological assessment. When SCG and SCG-char were applied to acid mine drainage, the heavy metal concentrations were decreased and the pH was increased. However, for SCG, the phytotoxicity increased because a massive amount of dissolved organic carbon was released from SCG. In contrast, SCG-char did not exhibit this phenomenon because any easily released organic matter was removed during pyrolysis. While the bioavailable heavy metal content decreased in soils treated with SCG or SCG-char, the phytotoxicity only rose after SCG treatment. According to our statistical methodology, bioavailable Pb, Cu and As, as well as the electrical conductivity representing an increase in organic content, affected the phytotoxicity of soil. Therefore, applying SCG during environment remediation requires careful biological assessments and evaluations of the efficiency of this remediation technology. Copyright © 2014 Elsevier Ltd. All rights reserved.

Organic oxalate as leachant and precipitant for the recovery of valuable metals from spent lithium-ion batteries

International Nuclear Information System (INIS)

Sun Liang; Qiu Keqiang

2012-01-01

Graphical abstract: Display Omitted Highlights: ► Vacuum pyrolysis as a pretreatment was used to separate cathode material from aluminum foils. ► Cobalt and lithium can be leached using oxalate while cobalt can be directly precipitated as cobalt oxalate. ► Cobalt and lithium can be separated efficiently from each other only in the oxalate leaching process. ► High reaction efficiency of LiCoO 2 was obtained with oxalate. - Abstract: Spent lithium-ion batteries containing lots of strategic resources such as cobalt and lithium are considered as an attractive secondary resource. In this work, an environmentally compatible process based on vacuum pyrolysis, oxalate leaching and precipitation is applied to recover cobalt and lithium from spent lithium-ion batteries. Oxalate is introduced as leaching reagent meanwhile as precipitant which leaches and precipitates cobalt from LiCoO 2 and CoO directly as CoC 2 O 4 ·2H 2 O with 1.0 M oxalate solution at 80 °C and solid/liquid ratio of 50 g L −1 for 120 min. The reaction efficiency of more than 98% of LiCoO 2 can be achieved and cobalt and lithium can also be separated efficiently during the hydrometallurgical process. The combined process is simple and adequate for the recovery of valuable metals from spent lithium-ion batteries.

Conditioning of spent mercury by amalgamation

International Nuclear Information System (INIS)

Yim, S. P.; Shon, J. S.; An, B. G.; Lee, H. J.; Lee, J. W.; Ji, C. G.; Kim, S. H.; Yoon, J. H.; Yang, M. S.

2002-01-01

Solidification by amalgamation was performed to immobilize and stabilize the liquid spent mercury. First, the appropriate metal and alloy which can convert liquid mercury into a solid form of amalgam were selected through initial tests. The amalgam form, formulated in optimum composition, was characterized and subjected to performance tests including compressive strength, water immersion, leachability and initial vaporization rate to evaluate mechanical integrity, durability and leaching properties. Finally, bench scale amalgamation trial was conducted with about 1 kg of spent mercury to verify the feasibility of amalgamation method

Simulation of the injection casting of metallic fuels

International Nuclear Information System (INIS)

Nakagawa, Tomokazu; Ogata, Takanari; Tokiwai, Moriyasu.

1989-01-01

For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

Investigation of the condition of spent-fuel pool components

International Nuclear Information System (INIS)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts

Investigation of the condition of spent-fuel pool components

Energy Technology Data Exchange (ETDEWEB)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

Determining heavy metals in spent compact fluorescent lamps (CFLs) and their waste management challenges: Some strategies for improving current conditions

International Nuclear Information System (INIS)

Taghipour, Hassan; Amjad, Zahra; Jafarabadi, Mohamad Asghari; Gholampour, Akbar; Norouz, Prviz

2014-01-01

Highlights: • Heavy metals in spent compact fluorescent lamps (CFLs) determined. • Current waste management condition of CFLs in Iran assessed. • Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. • We propose extended producer responsibility (EPR) for CFLs waste management. - Abstract: From environmental viewpoint, the most important advantage of compact fluorescent lamps (CFLs) is reduction of green house gas emissions. But their significant disadvantage is disposal of spent lamps because of containing a few milligrams of toxic metals, especially mercury and lead. For a successful implementation of any waste management plan, availability of sufficient and accurate information on quantities and compositions of the generated waste and current management conditions is a fundamental prerequisite. In this study, CFLs were selected among 20 different brands in Iran. Content of heavy metals including mercury, lead, nickel, arsenic and chromium was determined by inductive coupled plasma (ICP). Two cities, Tehran and Tabriz, were selected for assessing the current waste management condition of CFLs. The study found that waste generation amount of CFLs in the country was about 159.80, 183.82 and 153.75 million per year in 2010, 2011 and 2012, respectively. Waste generation rate of CFLs in Iran was determined to be 2.05 per person in 2012. The average amount of mercury, lead, nickel, arsenic and chromium was 0.417, 2.33, 0.064, 0.056 and 0.012 mg per lamp, respectively. Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. For improving the current conditions, we propose by considering the successful experience of extended producer responsibility (EPR) in other electronic waste management. The EPR program with advanced recycling fee (ARF) is implemented for collecting and then recycling CFLs. For encouraging consumers to take the spent CFLs back at the end of the products’ useful life, a proportion of

Determining heavy metals in spent compact fluorescent lamps (CFLs) and their waste management challenges: Some strategies for improving current conditions

Energy Technology Data Exchange (ETDEWEB)

Taghipour, Hassan, E-mail: hteir@yahoo.com [Department of Environmental Health Engineering, Tabriz University of Medical Sciences, Tabriz (Iran, Islamic Republic of); Amjad, Zahra [Student Research Committee, Department of Environmental Health Engineering, Tabriz University of Medical Sciences, Tabriz (Iran, Islamic Republic of); Jafarabadi, Mohamad Asghari [Medical Education Research Center, Department of Statistics and Epidemiology, Tabriz University of Medical Sciences, Tabriz (Iran, Islamic Republic of); Gholampour, Akbar [Department of Environmental Health Engineering, Tabriz University of Medical Sciences, Tabriz (Iran, Islamic Republic of); Norouz, Prviz [Environmental Health Engineering, Shahid Beheshti University of Medical Sciences, Tehran (Iran, Islamic Republic of)

2014-07-15

Highlights: • Heavy metals in spent compact fluorescent lamps (CFLs) determined. • Current waste management condition of CFLs in Iran assessed. • Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. • We propose extended producer responsibility (EPR) for CFLs waste management. - Abstract: From environmental viewpoint, the most important advantage of compact fluorescent lamps (CFLs) is reduction of green house gas emissions. But their significant disadvantage is disposal of spent lamps because of containing a few milligrams of toxic metals, especially mercury and lead. For a successful implementation of any waste management plan, availability of sufficient and accurate information on quantities and compositions of the generated waste and current management conditions is a fundamental prerequisite. In this study, CFLs were selected among 20 different brands in Iran. Content of heavy metals including mercury, lead, nickel, arsenic and chromium was determined by inductive coupled plasma (ICP). Two cities, Tehran and Tabriz, were selected for assessing the current waste management condition of CFLs. The study found that waste generation amount of CFLs in the country was about 159.80, 183.82 and 153.75 million per year in 2010, 2011 and 2012, respectively. Waste generation rate of CFLs in Iran was determined to be 2.05 per person in 2012. The average amount of mercury, lead, nickel, arsenic and chromium was 0.417, 2.33, 0.064, 0.056 and 0.012 mg per lamp, respectively. Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. For improving the current conditions, we propose by considering the successful experience of extended producer responsibility (EPR) in other electronic waste management. The EPR program with advanced recycling fee (ARF) is implemented for collecting and then recycling CFLs. For encouraging consumers to take the spent CFLs back at the end of the products’ useful life, a proportion of

Spent fuel's behavior under dynamic drip tests

International Nuclear Information System (INIS)

Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

1995-01-01

In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

Choosing a spent fuel interim storage system

International Nuclear Information System (INIS)

Roland, V.; Hunter, I.

2001-01-01

The Transnucleaire Group has developed different modular solutions to address spent fuel interim storage needs of NPP. These solutions, that are present in Europe, USA and Asia are metal casks (dual purpose or storage only) of the TN 24 family and the NUHOMS canister based system. It is not always simple for an operator to sort out relevant choice criteria. After explaining the basic designs involved on the examples of the TN 120 WWER dual purpose cask and the NUHOMS 56 WWER for WWER 440 spent fuel, we shall discuss the criteria that govern the choice of a given spent fuel interim storage system from the stand point of the operator. In conclusion, choosing and implementing an interim storage system is a complex process, whose implications can be far reaching for the long-term success of a spent fuel management policy. (author)

A study on the alkalimetric titration with gran plot in noncomplexing media for the determination of free acid in spent fuel solutions

International Nuclear Information System (INIS)

Suh, Moo Yul; Lee, Chang Heon; Sohn, Se Chul; Kim, Jung Suk; Kim, Won Ho; Eom, Tae Yoon

1999-01-01

Based on the study of hydrolysis behaviour of U(VI) ion and major fission product metal ions such as Cs(I), Ce(III), Nd(III), Mo(VI), Ru(II), and Zr(VI) in the titration media, the performance of noncomplexing-alkalimetric titration method for the determination of free acid in the presence of these metal ions was investigated and its results were compared to those from the complexing methods. The free acidities could be determined as low as 0.05 meq in uranium solutions in which the molar ratio of U(VI)/H + was less than 5, when the end-point of titration was estimated by Gran plot. The biases in the determinations were less than ±1% and about +3% respectively for 0.4 meq and 0.05 meq of free acid at the U(VI)/H + molar ratio of up to 5. Applicability of this method to the determination of free acid in spent fuel solutions was confirmed by the analysis of nitric acid content in simulated spent fuel solutions and in a real spent fuel solution

German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

Directory of Open Access Journals (Sweden)

A. Schwenk-Ferrero

2013-01-01

Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

Pyroprocessing oxide spent nuclear fuels for efficient disposal

International Nuclear Information System (INIS)

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P.

1994-01-01

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

International Nuclear Information System (INIS)

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Svärd, Staffan Jacobsson; Jansson, Peter; Swinhoe, Martyn T.; Tobin, Stephen J.

2015-01-01

Previous simulation studies of Differential Die‐Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Martinik, Tomas, E-mail: tomas.martinik@physics.uu.se [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Henzl, Vladimir [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Grape, Sophie; Svärd, Staffan Jacobsson; Jansson, Peter [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Swinhoe, Martyn T. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Tobin, Stephen J. [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Swedish Nuclear Fuel and Waste Management Company, Blekholmstorget 30, Box 250, SE-101 24 Stockholm (Sweden)

2015-07-11

Previous simulation studies of Differential Die‐Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or

Standardized, utility-DOE compatible, spent fuel storage-transport systems

International Nuclear Information System (INIS)

Smith, M.L.

1991-01-01

Virginia Power has developed and licensed a facility for dry storage of spent nuclear fuel in metal spent fuel storage casks. The modifications to the design of these casks necessary for licensing for both storage and transport of spent fuel are discussed along with the operational advantages of dual purpose storage-transport casks. Dual purpose casks can be used for storage at utility and DOE sites (MRS or repository) and for shipment between these sites with minimal spent fuel handling. The cost for a standardized system of casks that are compatible for use at both DOE and utility sites is discussed along with possible arrangements for sharing both the cost and benefits of dual purpose storage-transport casks

Pyrolysis of marine biomass to produce bio-oil and its upgrading using a novel multi-metal catalyst prepared from the spent car catalytic converter.

Science.gov (United States)

Sabegh, Mahzad Yaghmaei; Norouzi, Omid; Jafarian, Sajedeh; Khosh, Akram Ghanbari; Tavasoli, Ahmad

2018-02-01

In order to reduce the economic and environmental consequences caused by spent car catalyst, we herein report for the first time a novel promising multi-metal catalyst prepared from spent car catalytic converters to upgrade the pyrolysis bio-oils. The physico-chemical properties of prepared catalyst were characterized by XRD, EDS, FESEM, and FT-IR analyses. The thermal stability of the multi-metal catalyst was studied with TGA. To investigate the activity of the catalyst, Conversion of Cladophora glomerata (C. glomerata) into bio-products was carried out via a fixed bed reactor with and without catalyst at the temperature of 500°C. Although the catalyst didn't catalyze the gasification reaction, bio-oil was upgraded over the catalyst. The main effect of the catalyst on the bio-oil components is deoxygenating of nitrogen compounds and promotion the ketonization reaction, which converts acid to ketone and declines the corrosive nature of bio-oil. Copyright © 2017. Published by Elsevier Ltd.

Conditioning of spent nuclear fuel for permanent disposal

International Nuclear Information System (INIS)

Laidler, J.J.

1994-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

Conditioning of spent nuclear fuel for permanent disposal

International Nuclear Information System (INIS)

Laidler, J.J.

1994-01-01

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the US Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (> 99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

Process and system to encapsulate spent nuclear fuel

International Nuclear Information System (INIS)

Gunasekaran, Muthian; Fleischer, L.R.

1980-01-01

System for encapsulating spent nuclear fuel containing active fission matter and comprised in a metal casing, where concrete covers this casing in a contiguous, uniform and complete manner. It is characterized in that this concrete contains metal fibres to raise the thermal conductivity and polymers for increasing impermeability and that convection facilities are provided for cooling the outer surface of the concrete [fr

Spent catalyst waste management. A review. Part 1. Developments in hydroprocessing catalyst waste reduction and use

Energy Technology Data Exchange (ETDEWEB)

Marafi, M.; Stanislaus, A. [Petroleum Refining Department, Petroleum Research and Studies Center, Kuwait Institute for Scientific Research, P.O. Box 24885, 13109-Safat (Kuwait)

2008-04-15

Solid catalysts containing metals, metal oxides or sulfides, which play a key role in the refining of petroleum to clean fuels and many other valuable products, become solid wastes after use. In many refineries, the spent catalysts discarded from hydroprocessing units form a major part of these solid wastes. Disposal of spent hydroprocessing catalysts requires compliance with stringent environmental regulations because of their hazardous nature and toxic chemicals content. Various options such as minimizing spent catalyst waste generation by regeneration and reuse, metals recovery, utilization to produce useful materials and treatment for safe disposal, could be considered to deal with the spent catalyst environmental problem. In this paper, information available in the literature on spent hydroprocessing catalyst waste reduction at source by using improved more active and more stable catalysts, regeneration, rejuvenation and reuse of deactivated catalysts in many cycles, and reusing in other processes are reviewed in detail with focus on recent developments. Available methods for recycling of spent hydroprocessing catalysts by using them as raw materials for the preparation of active new catalysts and many other valuable products are also reviewed. (author)

Recovery of valuable metals from waste cathode materials of spent lithium-ion batteries using mild phosphoric acid

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiangping, E-mail: chenxiangping101@163.com [School of Environmental Science and Engineering, Shaanxi University of Science & Technology, Xi’an 710021 (China); College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Ma, Hongrui, E-mail: mahr@sust.edu.cn [School of Environmental Science and Engineering, Shaanxi University of Science & Technology, Xi’an 710021 (China); Luo, Chuanbao; Zhou, Tao [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China)

2017-03-15

Graphical abstract: Cobalt can be directly recovered as Co{sub 3}(PO{sub 4}){sub 2} from waste LiCoO{sub 2} using H{sub 3}PO{sub 4} as leaching and precipitating agent. - Highlights: • Phosphoric acid was innovatively used as leaching and precipitating agent. • Over 99% Co and Li can be separated and recovered in a single leaching step. • Co and Li can be separated under mild conditions of 40 °C and 0.7 M H{sub 3}PO{sub 4}. • Activation energy values for Co and Li are 7.3 and 10.168 kJ/mol. • Cobalt phosphate (97.1% in purity) can be obtained as the leaching product. - Abstract: Sustainable recycling of valuable metals from spent lithium-ion batteries (LIBs) may be necessary to alleviate the depletion of strategic metal resources and potential risk of environmental pollution. Herein a hydrometallurgical process was proposed to explore the possibility for the recovery of valuable metals from the cathode materials (LiCoO{sub 2}) of spent LIBs using phosphoric acid as both leaching and precipitating agent under mild leaching conditions. According to the leaching results, over 99% Co can be separated and recovered as Co{sub 3}(PO{sub 4}){sub 2} in a short-cut process involved merely with leaching and filtrating, under the optimized leaching conditions of 40 °C (T), 60 min (t), 4 vol.% H{sub 2}O{sub 2}, 20 mL g{sup −1} (L/S) and 0.7 mol/L H{sub 3}PO{sub 4}. Then leaching kinetics was investigated based on the logarithmic rate kinetics model and the obtained results indicate that the leaching of Co and Li fits well with this model and the activation energies (Ea) for Co and Li are 7.3 and 10.2 kJ/mol, respectively. Finally, it can be discovered from characterization results that the obtained product is 97.1% pure cobalt phosphate (Co{sub 3}(PO{sub 4}){sub 2}).

Hydrometallurgical process for the recovery of metal values from spent lithium-ion batteries in citric acid media.

Science.gov (United States)

Chen, Xiangping; Zhou, Tao

2014-11-01

In this paper, a hydrometallurgical process has been proposed to recover valuable metals from spent lithium-ion batteries in citric acid media. Leaching efficiencies as high as 97%, 95%, 94%, and 99% of Ni, Co, Mn, and Li were achieved under the optimal leaching experimental conditions of citric acid concentration of 2 mol L(-1), leaching temperature of 80 °C, leaching time of 90 min, liquid-solid ratio of 30 ml g(-1), and 2 vol. % H2O2. For the metals recovery process, nickel and cobalt were selectively precipitated by dimethylglyoxime reagent and ammonium oxalate sequentially. Then manganese was extracted by Na-D2EHPA and the manganese-loaded D2EHPA was stripped with sulfuric acid. The manganese was recovered as MnSO4 in aqueous phase and D2EHPA could be reused after saponification. Finally, lithium was precipitated by 0.5 mol L(-1) sodium phosphate. Under their optimal conditions, the recovery percentages of Ni, Co, Mn, and Li can reach 98%, 97%, 98%, and 89%, respectively. This is a relatively simple route in which all metal values could be effectively leached and recovered in citric acid media. © The Author(s) 2014.

Safeguardability of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

2004-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

Development of spent solvent treatment process by a submerged combustion technique

International Nuclear Information System (INIS)

Uchiyama, Gunzo; Maeda, Mitsuru; Fujine, Sachio; Amakawa, Masayuki; Uchida, Katsuhide; Chida, Mitsuhisa

1994-01-01

An experimental study using a bench-scale equipment of 1 kg-simulated spent solvents per hour has been conducted in order to evaluate the applicability of a submerged combustion technique to the treatment of spent solvents contaminated with TRU elements. This report describes the experimental results on the combustion characteristics of the simulated spent solvents of tri-n-butyl phosphate and/or n-dodecane, and on the distribution behaviors of combustion products such as phosphoric acid, Ru, I, Zr and lanthanides as TRU simulants in the submerged combustion process. Also the experimental results of TRU separation from phosphoric acid solution by co-precipitation using bismuth phosphate are reported. It was shown that the submerged combustion technique was applicable to the treatment of spent solvents including the distillation residues of the solvent. Based on the experimental data, a new treatment process of spent solvent was proposed which consisted of submerged combustion, co-precipitation using bismuth phosphate, ceramic membrane filtration, cementation of TRU lean phosphate, and vitrification of TRU rich waste. (author)

Extraction Of Cobalt From Spent CMB Catalyst Using Supercritical CO2

Directory of Open Access Journals (Sweden)

Joo S.-H.

2015-06-01

Full Text Available The metal extraction from spent CMB catalyst using supercritical CO2(scCO2 was investigated with single organic system, binary organic system and ternary organic system to extract metal ions. Leaching solution of spent CMB catalyst containing 389 mg L−1 Co2+, 187 mg L−1 Mn2+, 133 mg L−1 Na+, 14.97 mg L−1 Ca2+ and 13.2 mg L−1 Mg2+. The method consists of scCO2/ligands complexation process and metal extraction process at 60°C and 200bar. The result showed the Co and Mn was selectively extracted from Mg, Ca and Na in the ternary system of mixture of Cyanex272, DEA and Alamine304-I.

Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

International Nuclear Information System (INIS)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

2012-01-01

The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

Energy Technology Data Exchange (ETDEWEB)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

Simulating cosmic metal enrichment by the first galaxies

NARCIS (Netherlands)

Pallottini, A.; Ferrara, A.; Gallerani, S.; Salvadori, S.; D'Odorico, V.

We study cosmic metal enrichment via adaptive mesh refinement hydrodynamical simulations in a (10 Mpc h-1)3 volume following the Population III (PopIII)-PopII transition and for different PopIII initial mass function (IMFs). We have analysed the joint evolution of metal enrichment on galactic and

Bioremediation of 60Co from simulated spent decontamination solutions

International Nuclear Information System (INIS)

Rashmi, K.; Naga Sowjanya, T.; Maruthi Mohan, P.; Balaji, V.; Venkateswaran, G.

2004-01-01

Bioremediation of 60 Co from simulated spent decontamination solutions by utilizing different biomass of (Neurospora crassa, Trichoderma viridae, Mucor recemosus, Rhizopus chinensis, Penicillium citrinum, Aspergillus niger and, Aspergillus flavus) fungi is reported. Various fungal species were screened to evaluate their potential for removing cobalt from very low concentrations (0.03-0.16 μM) in presence of a high background of iron (9.33 mM) and nickel (0.93 mM) complexed with EDTA (10.3 mM). The different fungal isolates employed in this study showed a pickup of cobalt in the range 8-500 ng/g of dry biomass. The [Fe]/[Co] and [Ni]/[Co] ratios in the solutions before and after exposure to the fungi were also determined. At micromolar level the cobalt pickup by many fungi especially the mutants of N. crassa is seen to be proportional to the initial cobalt concentration taken in the solution. However, R. chinensis exhibits a low but iron concentration dependent cobalt pickup. Prior saturating the fungi with excess of iron during their growth showed the presence of selective cobalt pickup sites. The existence of cobalt specific sorption sites is shown by a model experiment with R. chinensis wherein at a constant cobalt concentration (0.034 μM) and varying iron concentrations so as to yield [Fe/Co] initial ratios in solution of 10, 100, 1000 and 287 000 have all yielded a definite Co pickup capacity in the range 8-47 ng/g. The presence of Cr(III)EDTA (3 mM) in solution along with complexed Fe and Ni has not influenced the cobalt removal. The significant feature of this study is that even when cobalt is present in trace level (sub-micromolar) in a matrix of high concentration (millimolar levels) of iron, nickel and chromium, a situation typically encountered in spent decontamination solutions arising from stainless steel based primary systems of nuclear reactors, a number of fungi studied in this work showed a good sensitivity for cobalt pickup

Interim spent-fuel storage options at commercial nuclear power plants

International Nuclear Information System (INIS)

Thakkar, A.R.; Hylko, J.M.

1991-01-01

Although spent fuel can be stored safely in waterfilled pools at reactor sites, some utilities may not possess sufficient space for life-of-plant storage capability. In-pool storage capability may be increased by reracking assemblies, rod consolidation, double tiering spent-fuel racks, and by shipping spent fuel to other utility-owned facilities. Long-term on-site storage capability for spent fuel may be provided by installing (dry-type) metal casks, storage and transportation casks, concrete casks, horizontal concrete modules, modular concrete vaults, or by constructing additional (pool-type) storage installations. Experience to date has provided valuable information regarding dry-type or pool-type installations, cask handling and staffing requirements, security features, decommissioning activities, and radiological issues

Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

International Nuclear Information System (INIS)

Grosche Filho, C.E.; Chandra, U.

1986-01-01

The results of the study of simulated radwaste, spent ion-exchange resins, borates/evaporator-concentrates and animal ashes, in bituminized form, are presented and discussed. Distilled and oxidized bitumen were used for characterizing the crude material and simulated wastes-bitumen mixtures of varying weight composition 30, 40, 50, 60% by weight the dry waste material. The asphaltine and parafin contents in the bitumens were determined. Some additives and clays were used aiming best characteristics of solidified wastes. For leaching studies, granular ion-exchange resins were loaded with Cs 134 and mixtures of resins-bitumens were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. It was used a conventional screw-extruder, used in plastic industry, to determine operational conditions and process difficulties. Mixtures resins-bitumen and concentrate-bitumen in differents operational condition were prepared and analysed. (Author) [pt

Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1998-01-01

An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

Encapsulating spent nuclear fuel

International Nuclear Information System (INIS)

Fleischer, L.R.; Gunasekaran, M.

1979-01-01

A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

Recovery of Ni Metal from Spent Catalyst with Emulsion Liquid Membrane Using Cyanex 272 as Extractant

Science.gov (United States)

Yuliusman; Huda, M.; Ramadhan, I. T.; Farry, A. R.; Wulandari, P. T.; Alfia, R.

2018-03-01

In this study was conducted to recover nickel metal from spent nickel catalyst resulting from hydrotreating process in petroleum industry. The nickel extraction study with the emulsion liquid membrane using Cyanex 272 as an extractant to extract and separate nickel from the feed phase solution. Feed phase solution was preapred from spent catalyst using sulphuric acid. Liquid membrane consists of a kerosene as diluent, a Span 80 as surfactant, a Cyanex 272 as carrier and sulphuric acid solutions have been used as the stripping solution. The important parameters governing the permeation of nickel and their effect on the separation process have been studied. These parameters are surfactant concentration, extractant concentration feed phase pH. The optimum conditions of the emulsion membrane making process is using 0.06 M Cyanex 272, 8% w/v SPAN 80, 0.05 M H2SO4, internal phase extractant / phase volume ratio: 1/1, and stirring speed 1150 rpm for 60 Minute that can produce emulsion membrane with stability level above 90% after 4 hours. In the extraction process with optimum condition pH 6 for feed phase, ratio of phase emulsion/phase of feed: 1/2, and stirring speed 175 rpm for 15 minutes with result 81.51% nickel was extracted.

Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a laser system and have used it with a simulated BWR assembly. The reflected signal from the zircaloy rods depends on the position of the assembly, but in all cases is easily discernable from the reference scan of background with no assembly

Digital mock-up for the spent fuel disassembly processes

International Nuclear Information System (INIS)

Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

2000-12-01

In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

Metals recovery of spent household batteries using a hydrometallurgical process; Recuperacao de metais de sucatas de pilhas e baterias pos-consumo utilizando processamento hidrometalurgico

Energy Technology Data Exchange (ETDEWEB)

Souza, K.P.; Tenorio, J.A.S., E-mail: kprovazi@gmail.co [Universidade de Sao Paulo (USP), SP (Brazil). Dept. de Engenharia Metalurgica e de Materiais

2010-07-01

The objective of the work is to study a method for metals recovery from a sample composed by a mixture of the main types of spent household batteries. Segregation of the main metals is investigated using a treatment route consisting of the following steps: manual identified and dismantling, grinding, electric furnace reduction, acid leaching and selective precipitation with sodium hydroxide with and without hydrogen peroxide. Before and after precipitations the solutions had been analyzed by Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP/OES) and the precipitated analyzed by Scanning Electron Microscopy (SEM) with Spectrometry of Energy Dispersion Spectroscopy (EDS). The results had indicated that the great majority of metals had been precipitated in pHs studied, also had co-precipitation or simultaneous precipitation of metals in some pHs. (author)

Options and processes for spent catalyst handling and utilization.

Science.gov (United States)

Marafi, M; Stanislaus, A

2003-07-18

The quantity of spent hydroprocessing catalysts discarded as solid wastes in the petroleum refining industries has increased remarkably in recent years due to a rapid growth in the hydroprocessing capacity to meet the rising demand for low-sulfur fuels. Due to their toxic nature, spent hydroprocessing catalysts have been branded as hazardous wastes, and the refiners are experiencing pressure from environmental authorities to handle them safely. Several alternative methods such as reclamation of metals, rejuvenation and reuse, disposal in landfills and preparation of useful materials using spent catalysts as raw materials are available to deal with the spent catalyst problem. The technical feasibility as well as the environmental and economic aspects of these options are reviewed. In addition, details of two bench-scale processes, one for rejuvenation of spent hydroprocessing catalysts, and the other for producing non-leachable synthetic aggregate materials that were developed in this laboratory, are presented in this paper.

Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

Energy Technology Data Exchange (ETDEWEB)

Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

1979-09-01

A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed. (DLC)

Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

International Nuclear Information System (INIS)

Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

1979-09-01

A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed

Probing the limits of metal plasticity with molecular dynamics simulations

Science.gov (United States)

Zepeda-Ruiz, Luis A.; Stukowski, Alexander; Oppelstrup, Tomas; Bulatov, Vasily V.

2017-10-01

Ordinarily, the strength and plasticity properties of a metal are defined by dislocations--line defects in the crystal lattice whose motion results in material slippage along lattice planes. Dislocation dynamics models are usually used as mesoscale proxies for true atomistic dynamics, which are computationally expensive to perform routinely. However, atomistic simulations accurately capture every possible mechanism of material response, resolving every ``jiggle and wiggle'' of atomic motion, whereas dislocation dynamics models do not. Here we present fully dynamic atomistic simulations of bulk single-crystal plasticity in the body-centred-cubic metal tantalum. Our goal is to quantify the conditions under which the limits of dislocation-mediated plasticity are reached and to understand what happens to the metal beyond any such limit. In our simulations, the metal is compressed at ultrahigh strain rates along its [001] crystal axis under conditions of constant pressure, temperature and strain rate. To address the complexity of crystal plasticity processes on the length scales (85-340 nm) and timescales (1 ns-1μs) that we examine, we use recently developed methods of in situ computational microscopy to recast the enormous amount of transient trajectory data generated in our simulations into a form that can be analysed by a human. Our simulations predict that, on reaching certain limiting conditions of strain, dislocations alone can no longer relieve mechanical loads; instead, another mechanism, known as deformation twinning (the sudden re-orientation of the crystal lattice), takes over as the dominant mode of dynamic response. Below this limit, the metal assumes a strain-path-independent steady state of plastic flow in which the flow stress and the dislocation density remain constant as long as the conditions of straining thereafter remain unchanged. In this distinct state, tantalum flows like a viscous fluid while retaining its crystal lattice and remaining a strong

Recovery of Tin and Nitric Acid from Spent Solder Stripping Solutions

International Nuclear Information System (INIS)

Ahn, Jae-Woo; Ryu, Seong-Hyung; Kim, Tae-young

2015-01-01

Spent solder-stripping solutions containing tin, copper, iron, and lead in nitric acid solution, are by-products of the manufacture of printed-circuit boards. The recovery of these metals and the nitric acid, for re-use has economic and environmental benefits. In the spent solder-stripping solution, a systematic method to determine a suitable process for recovery of valuable metals and nitric acid was developed. Initially, more than 90% of the tin was successfully recovered as high-purity SnO 2 by thermal precipitation at 80 ℃ for 3 hours. About 94% of the nitric acid was regenerated effectively from the spent solutions by diffusion dialysis, after which there remained copper, iron, and lead in solution. Leakage of tin through the anion-exchange membrane was the lowest (0.026%), whereas Pb-leakage was highest (4.26%). The concentration of the regenerated nitric acid was about 5.1 N.

The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

International Nuclear Information System (INIS)

Ahluwalia, R.K.; Geyer, H.K.

1996-01-01

The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Park, Seong Won; Shin, Y. J.; Cho, S. H.

2004-03-01

The research on spent fuel management focuses on the maximization of the disposal efficiency by a volume reduction, the improvement of the environmental friendliness by the partitioning and transmutation of the long lived nuclides, and the recycling of the spent fuel for an efficient utilization of the uranium source. In the second phase which started in 2001, the performance test of the advanced spent fuel management process consisting of voloxidation, reduction of spent fuel and the lithium recovery process has been completed successfully on a laboratory scale. The world-premier spent fuel reduction hot test of a 5 kgHM/batch has been performed successfully by joint research with Russia and the valuable data on the actinides and FPs material balance and the characteristics of the metal product were obtained with experience to help design an engineering scale reduction system. The electrolytic reduction technology which integrates uranium oxide reduction in a molten LiCl-Li 2 O system and Li 2 O electrolysis is developed and a unique reaction system is also devised. Design data such as the treatment capacity, current density and mass transfer behavior obtained from the performance test of a 5 kgU/batch electrolytic reduction system pave the way for the third phase of the hot cell demonstration of the advanced spent fuel management technology

Numerical simulations of the metallicity distribution in dwarf spheroidal galaxies

NARCIS (Netherlands)

Ripamonti, E.; Tolstoy, E.; Helmi, A.; Battaglia, G.; Abel, T.

2006-01-01

Abstract: Recent observations show that the number of stars with very low metallicities in the dwarf spheroidal satellites of the Milky Way is low, despite the low average metallicities of stars in these systems. We undertake numerical simulations of star formation and metal enrichment of dwarf

Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

2006-01-01

Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

Two-step process of regeneration of acid(s) from ZrF{sub 4} containing spent pickle liquor and recovery of zirconium metal

Energy Technology Data Exchange (ETDEWEB)

Nersisyan, Hayk [Graduate School of Department of Materials Science & Engineering, Chungnam National University, 99 Daehakro, Yuseong-gu, Daejeon (Korea, Republic of); RASOM, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Han, Seul Ki; Choi, Jeong Hun [Graduate School of Department of Materials Science & Engineering, Chungnam National University, 99 Daehakro, Yuseong-gu, Daejeon (Korea, Republic of); Graduate School of Energy Science & Technology, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Young-Jun; Yoo, Bung Uk [Graduate School of Energy Science & Technology, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Ri, Vladislav E. [Graduate School of Department of Materials Science & Engineering, Chungnam National University, 99 Daehakro, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Jong Hyeon, E-mail: jonglee@cnu.ac.kr [Graduate School of Department of Materials Science & Engineering, Chungnam National University, 99 Daehakro, Yuseong-gu, Daejeon (Korea, Republic of); Graduate School of Energy Science & Technology, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon (Korea, Republic of); RASOM, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2017-04-01

In this paper we describe a progressive two-step process that allows zirconium fluoride (ZrF{sub 4}) contained in spent baths for etched zirconium alloys to be effectively recycled on a pilot scale and produce a high purity regenerated pickling acid. In the first step, a spent pickling liquor is treated by a BaF{sub 2} suspension to produce water insoluble Ba{sub 2}ZrF{sub 8}. After filtration of Ba{sub 2}ZrF{sub 8} more than 99.9 wt % purity pickling acid is regenerated. The precipitation mechanism of Ba{sub 2}ZrF{sub 8} is discussed and the role of BaF{sub 2} particles size on the precipitation process is demonstrated. In the second step the as-precipitated Ba{sub 2}ZrF{sub 8} is mixed with Mg and Cu metal powders and heat-treated at 1200 °C (or above) to produce CuZr alloy ingot. The characteristics of the ingot are discussed in regard to Cu concentration and the heating temperature. - Highlights: •Two-step process for recycling ZrF{sub 4} containing pickling acid on a pilot scale is developed. •Water insoluble Ba{sub 2}ZrF{sub 8} is precipitated by mixing spent pickling liquor with BaF{sub 2}. •The recycled pickling acid demonstrates more than 99.9 wt % purity. •The processing of Ba{sub 2}ZrF{sub 8} with Cu and Mg metals at 1200 °C yielded CuZr alloy. •The recovery depth of Zr was more than 95 wt%.

Modeling Adsorption Based Filters (Bio-remediation of Heavy Metal Contaminated Water)

Science.gov (United States)

McCarthy, Chris

I will discuss kinetic models of adsorption, as well as models of filters based on those mechanisms. These mathematical models have been developed in support of our interdisciplinary lab group, which is centered at BMCC/CUNY (City University of New York). Our group conducts research into bio-remediation of heavy metal contaminated water via filtration. The filters are constructed out of biomass, such as spent tea leaves. The spent tea leaves are available in large quantities as a result of the industrial production of tea beverages. The heavy metals bond with the surfaces of the tea leaves (adsorption). The models involve differential equations, stochastic methods, and recursive functions. I will compare the models' predictions to data obtained from computer simulations and experimentally by our lab group. Funding: CUNY Collaborative Incentive Research Grant (Round 12); CUNY Research Scholars Program.

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, Y. H.

2001-03-01

Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

Interatomic potentials and the simulation of lattice defects in metals

International Nuclear Information System (INIS)

Heugten, W.F.W.M. van.

1979-01-01

The computer simulation technique is applied to investigate the properties of point defects and line defects in metals. For that purpose crystallites are constructed in which these defects are simulated. In the case of line defects (dislocations) the initial positions of the atoms, surrounding the dislocations, are determined using the elastic theory of anisotropic media. Hereafter the atoms in such crystallites are allowed to relax to there minimum potential energy positions under the influence of the interatomic forces. These forces are derived from interatomic interaction potentials. These potentials are together with the boundary conditions of the simulated crystallite the main input data in these computer simulation models. The metals considered include molybdenum, tungsten and tantalum. (Auth.)

Simulated BRDF based on measured surface topography of metal

Science.gov (United States)

Yang, Haiyue; Haist, Tobias; Gronle, Marc; Osten, Wolfgang

2017-06-01

The radiative reflective properties of a calibration standard rough surface were simulated by ray tracing and the Finite-difference time-domain (FDTD) method. The simulation results have been used to compute the reflectance distribution functions (BRDF) of metal surfaces and have been compared with experimental measurements. The experimental and simulated results are in good agreement.

Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

Directory of Open Access Journals (Sweden)

Eun-Young Choi

2015-12-01

Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

Electrometallurgical treatment of oxide spent fuels

International Nuclear Information System (INIS)

Karell, E. J.

1999-01-01

The Department of Energy (DOE) inventory of spent nuclear fuel contains a wide variety of oxide fuel types that may be unsuitable for direct repository disposal in their current form. The molten-salt electrometallurgical treatment technique developed by Argonne National Laboratory (ANL) has the potential to simplify preparing and qualifying these fuels for disposal by converting them into three uniform product streams: uranium metal, a metal waste form, and a ceramic waste form. This paper describes the major steps in the electrometallurgical treatment process for oxide fuels and provides the results of recent experiments performed to develop and scale up the process

Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

International Nuclear Information System (INIS)

Yoo, Seung Hun

2011-02-01

The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

Simulation codes of chemical separation process of spent fuel reprocessing. Tool for process development and safety research

International Nuclear Information System (INIS)

Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji

2005-01-01

This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)

Status of work at PNL supporting dry storage of spent fuel

International Nuclear Information System (INIS)

Cunningham, M.E.; McKinnon, M.A.; Michener, T.E.; Thomas, L.E.; Thornhill, C.K.

1992-01-01

Three projects related to dry storage of light-water reactor spent fuel are being conducted at Pacific Northwest Laboratory. Performance testing of six dry storage systems (four metal casks and two concrete storage systems) has been completed and results compiled. Two computer codes for predicting spent fuel and storage system thermal performance, COBRA-SFS and HYDRA-II, have been developed and have been reviewed by the US Nuclear Regulatory Commission. Air oxidation testing of spent fuel was conducted from 1984 through 1990 to obtain data to support recommendations of temperature-time limits for air dry storage for periods up to 40 years

Survey of wet and dry spent fuel storage

International Nuclear Information System (INIS)

1999-07-01

Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel

Bioremediation of {sup 60}Co from simulated spent decontamination solutions

Energy Technology Data Exchange (ETDEWEB)

Rashmi, K.; Naga Sowjanya, T.; Maruthi Mohan, P.; Balaji, V.; Venkateswaran, G

2004-07-26

Bioremediation of {sup 60}Co from simulated spent decontamination solutions by utilizing different biomass of (Neurospora crassa, Trichoderma viridae, Mucor recemosus, Rhizopus chinensis, Penicillium citrinum, Aspergillus niger and, Aspergillus flavus) fungi is reported. Various fungal species were screened to evaluate their potential for removing cobalt from very low concentrations (0.03-0.16 {mu}M) in presence of a high background of iron (9.33 mM) and nickel (0.93 mM) complexed with EDTA (10.3 mM). The different fungal isolates employed in this study showed a pickup of cobalt in the range 8-500 ng/g of dry biomass. The [Fe]/[Co] and [Ni]/[Co] ratios in the solutions before and after exposure to the fungi were also determined. At micromolar level the cobalt pickup by many fungi especially the mutants of N. crassa is seen to be proportional to the initial cobalt concentration taken in the solution. However, R. chinensis exhibits a low but iron concentration dependent cobalt pickup. Prior saturating the fungi with excess of iron during their growth showed the presence of selective cobalt pickup sites. The existence of cobalt specific sorption sites is shown by a model experiment with R. chinensis wherein at a constant cobalt concentration (0.034 {mu}M) and varying iron concentrations so as to yield [Fe/Co]{sub initial} ratios in solution of 10, 100, 1000 and 287 000 have all yielded a definite Co pickup capacity in the range 8-47 ng/g. The presence of Cr(III)EDTA (3 mM) in solution along with complexed Fe and Ni has not influenced the cobalt removal. The significant feature of this study is that even when cobalt is present in trace level (sub-micromolar) in a matrix of high concentration (millimolar levels) of iron, nickel and chromium, a situation typically encountered in spent decontamination solutions arising from stainless steel based primary systems of nuclear reactors, a number of fungi studied in this work showed a good sensitivity for cobalt pickup.

Waste form development and characterization in pyrometallurgical treatment of spent nuclear fuel

International Nuclear Information System (INIS)

Ackerman, J.

1998-01-01

Electrometallurgical treatment is a compact, inexpensive method that is being developed at Argonne National Laboratory to deal with spent nuclear fuel, primarily metallic and oxide fuels. In this method, metallic nuclear fuel constituents are electrorefined in a molten salt to separate uranium from the rest of the spent fuel. Oxide and other fuels are subjected to appropriate head end steps to convert them to metallic form prior to electrorefining. The treatment process generates two kinds of high-level waste--a metallic and a ceramic waste. Isolation of these wastes has been developed as an integral part of the process. The wastes arise directly from the electrorefiner, and waste streams do not contain large quantities of solvent or other process fluids. Consequently, waste volumes are small and waste isolation processes can be compact and rapid. This paper briefly summarizes waste isolation processes then describes development and characterization of the two waste forms in more detail

Atomistic simulations of Mg-Cu metallic glasses: Mechanical properties

DEFF Research Database (Denmark)

Bailey, Nicholas; Schiøtz, Jakob; Jacobsen, Karsten Wedel

2004-01-01

The atomistic mechanisms of plastic deformation in amorphous metals are far from being understood. We have derived potential parameters for molecular dynamics simulations of Mg-Cu amorphous alloys using the Effective Medium Theory. We have simulated the formation of alloys by cooling from the melt...

Treatment of spent catalyst from the nitrogenous fertilizer industry-A review of the available methods of regeneration, recovery and disposal

International Nuclear Information System (INIS)

Singh, Bina

2009-01-01

Disposal of spent catalyst is a problem as it falls under the category of hazardous industrial waste. The recovery of metals from these catalysts is an important economic aspect as most of these catalysts are supported, usually on alumina/silica with varying percent of metal; metal concentration could vary from 2.5 to 20%. Metals like Ni, Mo, Co, Rh, Pt, Pd, etc., are widely used as a catalyst in chemical and petrochemical industries and fertilizer industries. They are generally supported on porous materials like alumina and silica through precipitation or impregnation processes. Many workers have adapted pyrometallurgy and Hydrometallurgy process for recovery of precious metals. Many workers have studied the recovery of nickel from a spent catalyst in an ammonia plant by leaching it in sulphuric acid solution (Hydrometallurgy). Ninety-nine percent of the nickel was recovered as nickel sulphate when the catalyst, having a particle size of 0.09 mm was dissolved in an 80% sulphuric acid solution for 50 min in at 70 deg. C. Many researcher have studied the extraction of metals from spent catalyst by roasting-extraction method (Pyrometallurgy). Chelating agents are the most effective extractants, which can be introduced in the soil washing fluid to enhance heavy metal extraction from contaminated soils. The advantages of chelating agents in soil cleanup include high efficiency of metal extraction, high thermodynamic stabilities of the metal complexes formed, good solubilities of the metal complexes, and low adsorption of the chelating agents on soils, But very few workers have attempted chelating agent to extract metals from spent catalyst.

Acquisition of Co metal from spent lithium-ion battery using emulsion liquid membrane technology and emulsion stability test

Science.gov (United States)

Yuliusman; Wulandari, P. T.; Amiliana, R. A.; Huda, M.; Kusumadewi, F. A.

2018-03-01

Lithium-ion batteries are the most common type to be used as energy source in mobile phone. The amount of lithium-ion battery wastes is approximated by 200 – 500 ton/year. In one lithium-ion battery, there are 5 – 20% of cobalt metal, depend on the manufacturer. One of the way to recover a valuable metal from waste is leaching process then continued with extraction, which is the aim of this study. Spent lithium-ion batteries will be characterized with EDX and AAS, the result will show the amount of cobalt metal with form of LiCoO2 in the cathode. Hydrochloric acid concentration used is 4 M, temperature 80°C, and reaction time 1 hour. This study will discuss the emulsion stability test on emulsion liquid membrane. The purpose of emulsion stability test in this study was to determine optimum concentration of surfactant and extractant to produce a stable emulsion. Surfactant and extractant used were SPAN 80 and Cyanex 272 respectively with both concentrations varied. Membrane and feed phase ratios used in this experiment was 1 : 2. The optimum results of this study were SPAN 80 concentrations of 10% w/v and Cyanex 272 0.7 M.

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

Preparation and Heat-Treatment of DWPF Simulants With and Without Co-Precipitated Noble Metals

International Nuclear Information System (INIS)

Koopman, David C.:Eibling, Russel E

2005-01-01

The Savannah River National Laboratory is in the process of investigating factors suspected of impacting catalytic hydrogen generation in the Chemical Process Cell of the Defense Waste Processing Facility, DWPF. Noble metal catalyzed hydrogen generation in simulation work constrains the allowable acid addition operating window in DWPF. This constraint potentially impacts washing strategies during sludge batch preparation. It can also influence decisions related to the addition of secondary waste streams to a sludge batch. Noble metals have historically been added as trim chemicals to process simulations. The present study investigated the potential conservatism that might be present from adding the catalytic species as trim chemicals to the final sludge simulant versus co-precipitating the noble metals into the insoluble sludge solids matrix. Parallel preparations of two sludge simulants targeting the composition of Sludge Batch 3 were performed in order to evaluate the impact of the form of noble metals. Identical steps were used except that one simulant had dissolved palladium, rhodium, and ruthenium present during the precipitation of the insoluble solids. Noble metals were trimmed into the other stimulant prior to process tests. Portions of both sludge simulants were held at 97 C for about eight hours to qualitatively simulate the effects of long term storage on particle morphology and speciation. The simulants were used as feeds for Sludge Receipt and Adjustment Tank, SRAT, process simulations. The following conclusions were drawn from the simulant preparation work: (1) The first preparation of a waste slurry simulant with co-precipitated noble metals was successful, based on the data obtained. It appears that 99+% of the noble metals were retained in the simulant. (2) Better control of carbonate, hydroxide, and post-wash trim chemical additions is needed before the new method of simulant preparation will be as reproducible as the old method. (3) The two new

Transmutation of DUPIC spent fuel in the hyper system

International Nuclear Information System (INIS)

Kim, Y.H.; Song, T.Y.

2005-01-01

In this paper, the transmutation of TRUs of the DUPIC (Direct Use of Spent PWR Fuel in CANDU) spent fuel has been studied with the HYPER system, which is an LBE-cooled ADS. The DUPIC concept is a synergistic combination of PWRs and CANDUs, in which PWR spent fuels are directly re-utilized in CANDU reactors after a very simple re-fabrication process. In the DUPIC-HYPER fuel cycle, TRUs are recovered by using a pyro-technology and they are incinerated in a metallic fuel form of U-TRU-Zr. The objective of this study is to investigate the TRU transmutation potential of the HYPER core for the DUPIC-HYPER fuel cycle. All the previously-developed HYPER core design concepts were retained except that fuel is composed of TRU from the DUPIC spent fuel. In order to reduce the burnup reactivity swing, a B 4 C burnable absorber is used. The HYPER core characteristics have been analyzed with the REBUS-3/DIF3D code system. (authors)

Absorption column working study for iodine formed in spent fuel reprocessing plant gaseous effluents: hydrodynamic and mass transfer

International Nuclear Information System (INIS)

Vignau, B.

1986-09-01

The hydrodynamic and matter transfer parameters has been studied on absorption columns destined to trap iodine issued of spent fuel reprocessing plants. These columns have different packing - Raschig rings (glass, ceramic, PVC, steel) - Berl saddles (ceramic) - Weaved metallic thread (steel). The effect of dimension and of packing structure on gas pressure drop and on liquid holdup has been evaluated. The partial transfer coefficients of I 2 -Air-NaOH system has been the object of an experimental study. This system can be simulated by CO 2 -Air-NaOH system [fr

Microcomputer simulation model for facility performance assessment: a case study of nuclear spent fuel handling facility operations

International Nuclear Information System (INIS)

Chockie, A.D.; Hostick, C.J.; Otis, P.T.

1985-10-01

A microcomputer based simulation model was recently developed at the Pacific Northwest Laboratory (PNL) to assist in the evaluation of design alternatives for a proposed facility to receive, consolidate and store nuclear spent fuel from US commercial power plants. Previous performance assessments were limited to deterministic calculations and Gantt chart representations of the facility operations. To insure that the design of the facility will be adequate to meet the specified throughput requirements, the simulation model was used to analyze such factors as material flow, equipment capability and the interface between the MRS facility and the nuclear waste transportation system. The simulation analysis model was based on commercially available software and application programs designed to represent the MRS waste handling facility operations. The results of the evaluation were used by the design review team at PNL to identify areas where design modifications should be considered. 4 figs

Molecular dynamics simulation of self-diffusion coefficients for liquid metals

International Nuclear Information System (INIS)

Ju Yuan-Yuan; Zhang Qing-Ming; Gong Zi-Zheng; Ji Guang-Fu

2013-01-01

The temperature-dependent coefficients of self-diffusion for liquid metals are simulated by molecular dynamics methods based on the embedded-atom-method (EAM) potential function. The simulated results show that a good inverse linear relation exists between the natural logarithm of self-diffusion coefficients and temperature, though the results in the literature vary somewhat, due to the employment of different potential functions. The estimated activation energy of liquid metals obtained by fitting the Arrhenius formula is close to the experimental data. The temperature-dependent shear-viscosities obtained from the Stokes—Einstein relation in conjunction with the results of molecular dynamics simulation are generally consistent with other values in the literature. (atomic and molecular physics)

Determination of plutonium content in TRR spent fuel by nondestructive neutron counting

International Nuclear Information System (INIS)

Chen, Y.-F.; Sheu, R.-J.; Chiao, L.-H.; Yuan, M.-C.; Jiang, S.-H.

2010-01-01

For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240 Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240 Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

Leaching of spent fuel in simulated disposal condition and separation of plutonium species as a function of oxidation state

International Nuclear Information System (INIS)

Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

2000-11-01

The influences of compacted bentonite on the leaching of spent fuel in bentonite-saturated ground water at room temperature were investigated by analyzing the components of leachates as well as the alterated surface of them. And the plutonium species was separated by ion exchangers. The amounts of Cs, Sb, Sr, Am, Ru, Pu and U released from spent fuel by bentonite-saturated solution for the initial 165 days were 2.0, 0.2, 0.2, 0.02, 0.005, 5x10 - 4, 0.05 % of inventory, respectively. These values correspond to several ∼ several tens times as much as those through bentonite block which were compacted to 1.4 g/cm 3 . The comparison of the cesium released in groundwater and bentonite-saturated solution through bentonite block is simular values, whose lower concentration in leachant indicates that most of radionuclides are retained by compacted bentonite, even though alkali metal such as Cs. The separation of plutonium species as a function of oxidation state by ion exchanger was succeed by two columns' method with packing materials as SiO - , SiO-SO 3 -

Electrodialytic decontamination of spent ion exchange resins

International Nuclear Information System (INIS)

Nott, B.R.

1982-01-01

Development of a novel electrodialytic decontamination process for the selective removal of radioactive Cs from spent ion exchange resins containing large amounts of Li is described. The process involves passage of a dc electric current through a bed of the spent ion exchange resin in a specially designed electrodialytic cell. The radiocesium so removed from a volume of the spent resin is concentrated onto a much smaller volume of a Cs selective sorbent to achieve a significant radioactive waste volume reduction. Technical feasibility of the electrodialytic resin decontamination process has been demonstrated on a bench scale with a batch of simulated spent ion exchange resin and using potassium cobalt ferrocyanide as the Cs selective sorbent. A volume reduction factor between 10 and 17 has been estimated. The process appears to be economically attractive. Improvements in process economics can be expected from optimization of the process. Other possible applications of the EDRD process have been identified

Studies on recycling and utilization of spent catalysts. Preparation of active hydrodemetallization catalyst compositions from spent residue hydroprocessing catalysts

Energy Technology Data Exchange (ETDEWEB)

Marafi, Meena; Stanislaus, Antony [Petroleum Refining Department, Petroleum Research and Studies Center, Kuwait Institute for Scientific Research, P.O. Box 24885, Safat (Kuwait)

2007-02-15

Spent catalysts form a major source of solid wastes in the petroleum refining industries. Due to environmental concerns, increasing emphasis has been placed on the development of recycling processes for the waste catalyst materials as much as possible. In the present study the potential reuse of spent catalysts in the preparation of active new catalysts for residual oil hydrotreating was examined. A series of catalysts were prepared by mixing and extruding spent residue hydroprocessing catalysts that contained C, V, Mo, Ni and Al{sub 2}O{sub 3} with boehmite in different proportions. All prepared catalysts were characterized by chemical analysis and by surface area, pore volume, pore size and crushing strength measurements. The hydrodesulfurization (HDS) and hydrodemetallization (HDM) activities of the catalysts were evaluated by testing in a high pressure fixed-bed microreactor unit using Kuwait atmospheric residue as feed. A commercial HDM catalyst was also tested under similar operating conditions and their HDS and HDM activities were compared with that of the prepared catalysts. The results revealed that catalyst prepared with addition of up to 40 wt% spent catalyst to boehmite had fairly high surface area and pore volume together with large pores. The catalyst prepared by mixing and extruding about 40 wt% spent catalyst with boehmite was relatively more active for promoting HDM and HDS reactions than a reference commercial HDM catalyst. The formation of some kind of new active sites from the metals (V, Mo and Ni) present in the spent catalyst is suggested to be responsible for the high HDM activity of the prepared catalyst. (author)

Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Romanato, Luiz Sergio, E-mail: romanato@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade

2011-07-01

Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2011-01-01

Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

Study of plasticity in metals by numerical simulations

International Nuclear Information System (INIS)

Clouet, E.

2013-01-01

We present a study of the plastic behaviour in metals based on the modelling of dislocation properties. Different simulation tools have been used and developed to study plasticity in structural materials, in particular metals used in the nuclear industry. In iron or zirconium alloys, plasticity is controlled at low temperature by the glide of screw dislocations. Atomistic simulations can be used to model dislocation core properties and thus to obtain a better knowledge of the mechanisms controlling dislocation glide. Such atomistic simulations need nevertheless some special care because of the long range elastic field induced by the dislocations. We have therefore developed a modelling approach relying both on atomistic simulations, using either empirical interatomic potentials or ab initio calculations, and on elasticity theory. Such an approach has been used to obtain dislocation intrinsic core properties. These simulations allowed us to describe, in iron, the variations of these core properties with the dislocation character. In zirconium, we could identity the origin of the high lattice friction and obtain a better understanding of the competition between the different glide systems. At high temperature, dislocations do not only glide but can also cross-slip or climb. This leads to a motion of the dislocations out of their glide plane which needs to be considered when modelling the plastic flow. We performed a study of dislocation climb at different scales, leading to the implementation of a dislocation climb model in dislocation dynamics simulations. (author) [fr

Simulated electron affinity tuning in metal-insulator-metal (MIM) diodes

Science.gov (United States)

Mistry, Kissan; Yavuz, Mustafa; Musselman, Kevin P.

2017-05-01

Metal-insulator-metal diodes for rectification applications must exhibit high asymmetry, nonlinearity, and responsivity. Traditional methods of improving these figures of merit have consisted of increasing insulator thickness, adding multiple insulator layers, and utilizing a variety of metal contact combinations. However, these methods have come with the price of increasing the diode resistance and ultimately limiting the operating frequency to well below the terahertz regime. In this work, an Airy Function Transfer Matrix simulation method was used to observe the effect of tuning the electron affinity of the insulator as a technique to decrease the diode resistance. It was shown that a small increase in electron affinity can result in a resistance decrease in upwards of five orders of magnitude, corresponding to an increase in operating frequency on the same order. Electron affinity tuning has a minimal effect on the diode figures of merit, where asymmetry improves or remains unaffected and slight decreases in nonlinearity and responsivity are likely to be greatly outweighed by the improved operating frequency of the diode.

Advanced waste forms from spent nuclear fuel

International Nuclear Information System (INIS)

Ackerman, J.P.; McPheeters, C.C.

1995-01-01

More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed

Atomic-scale simulations of the mechanical deformation of nanocrystalline metals

DEFF Research Database (Denmark)

Schiøtz, Jakob; Vegge, Tejs; Di Tolla, Francesco

1999-01-01

that the main deformation mode is sliding in the grain boundaries through a large number of uncorrelated events, where a few atoms (or a few tens of atoms) slide with respect to each other. Little dislocation activity is seen in the grain interiors. The localization of the deformation to the grain boundaries......Nanocrystalline metals, i.e., metals in which the grain size is in the nanometer range, have a range of technologically interesting properties including increased hardness and yield strength. We present atomic-scale simulations of the plastic behavior of nanocrystalline copper. The simulations show...

Processes and Technologies for the Recycling of Spent Fluorescent Lamps

Directory of Open Access Journals (Sweden)

Kujawski Wojciech

2014-09-01

Full Text Available The growing industrial application of rare earth metals led to great interest in the new technologies for the recycling and recovery of REEs from diverse sources. This work reviews the various methods for the recycling of spent fluorescent lamps. The spent fluorescent lamps are potential source of important rare earth elements (REEs such as: yttrium, terbium, europium, lanthanum and cerium. The characteristics of REEs properties and construction of typical fl uorescent lamps is described. The work compares also current technologies which can be utilized for an efficient recovery of REEs from phosphors powders coming from spent fluorescent lamps. The work is especially focused on the hydrometallurgical and pyrometallurgical processes. It was concluded that hydrometallurgical processes are especially useful for the recovery of REEs from spent fluorescent lamps. Moreover, the methods used for recycling of REEs are identical or very similar to those utilized for the raw ores processing.

Initial results for electrochemical dissolution of spent EBR-II fuel

International Nuclear Information System (INIS)

Li, S. X.

1998-01-01

Initial results are reported for the anode behavior of spent metallic nuclear fuel in an electrorefining process. The anode behavior has been characterized in terms of the initial spent fuel composition and the final composition of the residual cladding hulls. A variety of results have been obtained depending on the experimental conditions. Some of the process variables considered are average and maximum cell voltage, average and maximum anode voltage, amount of electrical charge passed (coulombs or amp-hours) during the experiment, and cell resistance. The main goal of the experiments has been the nearly complete dissolution of uranium with the retention of zirconium and noble metal fission products in the cladding hulls. Analysis has shown that the most indicative parameters for determining an endpoint to the process, recognizing the stated goal, are the maximum anode voltage and the amount of electrical charge passed. For the initial experiments reported here, the best result obtained is greater than 98% uranium dissolution with approximately 50% zirconium retention. Noble metal fission product retention appears to be correlated with zirconium retention

Control system design specification of advanced spent fuel management process units

Energy Technology Data Exchange (ETDEWEB)

Ahn, S. H.; Kim, S. H.; Yoon, J. S

2003-06-01

In this study, the design specifications of instrumentation and control system for advanced spent fuel management process units are presented. The advanced spent fuel management process consists of several process units such as slitting device, dry pulverizing/mixing device, metallizer, etc. In this study, the control and operation characteristics of the advanced spent fuel management mockup process devices and the process devices developed in 2001 and 2002 are analysed. Also, a integral processing system of the unit process control signals is proposed, which the operation efficiency is improved. And a redundant PLC control system is constructed which the reliability is improved. A control scheme is proposed for the time delayed systems compensating the control performance degradation caused by time delay. The control system design specification is presented for the advanced spent fuel management process units. This design specifications can be effectively used for the detail design of the advanced spent fuel management process.

Finite element simulation and Experimental verification of Incremental Sheet metal Forming

Science.gov (United States)

Kaushik Yanamundra, Krishna; Karthikeyan, R., Dr.; Naranje, Vishal, Dr

2018-04-01

Incremental sheet metal forming is now a proven manufacturing technique that can be employed to obtain application specific, customized, symmetric or asymmetric shapes that are required by automobile or biomedical industries for specific purposes like car body parts, dental implants or knee implants. Finite element simulation of metal forming process is being performed successfully using explicit dynamics analysis of commercial FE software. The simulation is mainly useful in optimization of the process as well design of the final product. This paper focuses on simulating the incremental sheet metal forming process in ABAQUS, and validating the results using experimental methods. The shapes generated for testing are of trapezoid, dome and elliptical shapes whose G codes are written and fed into the CNC milling machine with an attached forming tool with a hemispherical bottom. The same pre-generated coordinates are used to simulate a similar machining conditions in ABAQUS and the tool forces, stresses and strains in the workpiece while machining are obtained as the output data. The forces experimentally were recorded using a dynamometer. The experimental and simulated results were then compared and thus conclusions were drawn.

Partition wall structure in spent fuel storage pool and construction method for the partition wall

International Nuclear Information System (INIS)

Izawa, Masaaki

1998-01-01

A partitioning wall for forming cask pits as radiation shielding regions by partitioning inside of a spent fuel storage pool is prepared by covering both surface of a concrete body by shielding metal plates. The metal plate comprises opposed plate units integrated by welding while sandwiching a metal frame as a reinforcing material for the concrete body, the lower end of the units is connected to a floor of a pool by fastening members, and concrete is set while using the metal plate of the units as a frame to form the concrete body. The shielding metal plate has a double walled structure formed by welding a lining plate disposed on the outer surface of the partition wall and a shield plate disposed to the inner side. Then the term for construction can be shortened, and the capacity for storing spent fuels can be increased. (N.H.)

Sulfuric acid baking and leaching of spent Co-Mo/Al2O3 catalyst.

Science.gov (United States)

Kim, Hong-In; Park, Kyung-Ho; Mishra, Devabrata

2009-07-30

Dissolution of metals from a pre-oxidized refinery plant spent Co-Mo/Al(2)O(3) catalyst have been tried through low temperature (200-450 degrees C) sulfuric acid baking followed by mild leaching process. Direct sulfuric acid leaching of the same sample, resulted poor Al and Mo recoveries, whereas leaching after sulfuric acid baking significantly improved the recoveries of above two metals. The pre-oxidized spent catalyst, obtained from a Korean refinery plant found to contain 40% Al, 9.92% Mo, 2.28% Co, 2.5% C and trace amount of other elements such as Fe, Ni, S and P. XRD results indicated the host matrix to be poorly crystalline gamma- Al(2)O(3). The effect of various baking parameters such as catalyst-to-acid ratio, baking temperature and baking time on percentage dissolutions of metals has been studied. It was observed that, metals dissolution increases with increase in the baking temperature up to 300 degrees C, then decreases with further increase in the baking temperature. Under optimum baking condition more than 90% Co and Mo, and 93% Al could be dissolved from the spent catalyst with the following leaching condition: H(2)SO(4)=2% (v/v), temperature=95 degrees C, time=60 min and Pulp density=5%.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

Science.gov (United States)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

Reprocessing method for spent fuel

International Nuclear Information System (INIS)

Fujie, Makoto; Shoji, Yuichi; Kobayashi, Tsuguyuki.

1997-01-01

After reducing oxides of uranium (U), plutonium (Pu) and miner actinides in spent fuels by magnesium (Mg) in a molten salt, rear earth element oxides and salts of alkali metals and alkaline earth metals contained in the molten salt phase are separated and removed. Further, the Mg phase containing the reduced metals is evaporated to separate and remove Mg, thereby recovering U, Pu and minor actinides. In a lithium (Li) process, Li 2 O also generated in the reduction step is regenerated to Li simultaneously, and the reduction is conducted while suppressing the Li 2 O concentration in the molten salt low. This can improve the reduction rate of oxides of U, Pu and minor actinides compared with conventional cases. Since Li 2 O is regenerated into Li in the reduction step of the Li process, deposited Li 2 O is not carried to an electrolysis purification step, and recovering rate of U, Pu and minor actinides is not lowered. (T.M.)

Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

Energy Technology Data Exchange (ETDEWEB)

Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools (SFSP's) will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a full size laser system operating in air and have used an array of 6 zircaloy BWR tubes to simulate an assembly. The reflective signal from the zircaloy rods is a strong function of position of the assembly, but in all cases is easily discernable from the reference scan of the background with no assembly. A design for a SFSP laser surveillance system incorporating laser ranging is discussed. 10 figures

Computer simulations of nanoindentation in Mg-Cu and Cu-Zr metallic glasses

DEFF Research Database (Denmark)

Paduraru, Anca; Andersen, Ulrik Grønbjerg; Thyssen, Anders

2010-01-01

The formation of shear bands during plastic deformation of Cu0.50Zr0.50 and Mg0.85Cu0.15 metallic glasses is studied using atomic-scale computer simulations. The atomic interactions are described using realistic many-body potentials within the effective medium theory, and are compared with similar...... simulations using a Lennard-Jones description of the material. The metallic glasses are deformed both in simple shear and in a simulated nanoindentation experiment. Plastic shear localizes into shear bands with a width of approximately 5 nm in CuZr and 8 nm in MgCu. In simple shear, the shear band formation...... is very clear, whereas only incipient shear bands are seen in nanoindentation. The shear band formation during nanoindentation is sensitive to the indentation velocity, indenter radius and the cooling rate during the formation of the metallic glass. For comparison, a similar nanoindentation simulation...

MSO spent salt clean-up recovery process; TOPICAL

International Nuclear Information System (INIS)

Adamson, M G; Brummond, W A; Hipple, D L; Hsu, P C; Summers, L J; Von Holtz, E H; Wang, F T

1997-01-01

An effective process has been developed to separate metals, mineral residues, and radionuclides from spent salt, a secondary waste generated by Molten Salt Oxidation (MSO). This process includes salt dissolution, pH adjustment, chemical reduction and/or sulfiding, filtration, ion exchange, and drying. The process uses dithionite to reduce soluble chromate and/or sulfiding agent to suppress solubilities of metal compounds in water. This process is capable of reducing the secondary waste to less than 5% of its original weight. It is a low temperature, aqueous process and has been demonstrated in the laboratory[1

Simulating the long-term chemistry of an upland UK catchment: Heavy metals

Energy Technology Data Exchange (ETDEWEB)

Tipping, E. [Centre for Ecology and Hydrology (Lancaster), Library Avenue, Bailrigg, Lancaster LA1 4AP (United Kingdom)]. E-mail: et@ceh.ac.uk; Lawlor, A.J. [Centre for Ecology and Hydrology (Lancaster), Library Avenue, Bailrigg, Lancaster LA1 4AP (United Kingdom); Lofts, S. [Centre for Ecology and Hydrology (Lancaster), Library Avenue, Bailrigg, Lancaster LA1 4AP (United Kingdom); Shotbolt, L. [Environment Department, University of York, Heslington, York YO10 5DD (United Kingdom)

2006-05-15

CHUM-AM was used to investigate the behaviours of atmospherically-deposited heavy metals (Ni, Cu, Zn, Cd and Pb) in three moorland sub-catchments in Cumbria UK. The principal processes controlling cationic metals are competitive partitioning to soil organic matter, chemical interactions in solution, and chemical weathering. Metal deposition histories were generated by combining measured data for the last 30 years with local lake sediment records. For Ni, Cu, Zn and Cd, default parameters for the interactions with organic matter provided reasonable agreement between simulated and observed present-day soil metal pools and average streamwater concentrations. However, for Pb, the soil binding affinity in the model had to be increased to match the observations. Simulations suggest that weakly-sorbing metals (Ni, Zn, Cd) will respond on timescales of decades to centuries to changes in metal inputs or acidification status. More strongly-sorbing metals (Cu, Pb) will respond over centuries to millennia. - Catchment turnover times for the strongly-retained metals Cu and Pb are of the order of centuries, whereas those for the more mobile Ni, Zn and Cd are appreciably shorter.

Separation of radionuclides from spent decontamination solutions and/or evaporator concentrates

International Nuclear Information System (INIS)

Sebesta, F.; John, J.; Rosikova, K.; Motl, A.

1999-01-01

Separation of radionuclides from spent alkaline decontamination solutions has been tested in model experiments with strontium separation from simulant solution. The composite absorbers tested included TiO-PAN and NaTiO-PAN materials (titanium dioxide or sodium titanate incorporated into a matrix of polyacrylonitrile binder). As an alkaline simulant, solution of 1 M NaOH + 1 M NaNO 3 + 10 -4 M Ca(NO 3 ) 2 + 10 -5 M Sr(NO 3 ) 2 spiked with a carrier-free 85 Sr tracer, was used. The experiments were performed at a flow rate of 12.5 BV/hr. Some experiments with real and simulant spent decontamination solutions are described

Simulation studies of current transport in metal-insulator-semiconductor Schottky barrier diodes

International Nuclear Information System (INIS)

Chand, Subhash; Bala, Saroj

2007-01-01

The current-voltage characteristics of Schottky diodes with an interfacial insulator layer are analysed by numerical simulation. The current-voltage data of the metal-insulator-semiconductor Schottky diode are simulated using thermionic emission diffusion (TED) equation taking into account an interfacial layer parameter. The calculated current-voltage data are fitted into ideal TED equation to see the apparent effect of interfacial layer parameters on current transport. Results obtained from the simulation studies shows that with mere presence of an interfacial layer at the metal-semiconductor interface the Schottky contact behave as an ideal diode of apparently high barrier height (BH), but with same ideality factor and series resistance as considered for a pure Schottky contact without an interfacial layer. This apparent BH decreases linearly with decreasing temperature. The effects giving rise to high ideality factor in metal-insulator-semiconductor diode are analysed. Reasons for observed temperature dependence of ideality factor in experimentally fabricated metal-insulator-semiconductor diodes are analysed and possible mechanisms are discussed

Operational analysis and improvement of a spent nuclear fuel handling and treatment facility using discrete event simulation

International Nuclear Information System (INIS)

Garcia, H.E.

2000-01-01

Spent nuclear fuel handling and treatment often require facilities with a high level of operational complexity. Simulation models can reveal undesirable characteristics and production problems before they become readily apparent during system operations. The value of this approach is illustrated here through an operational study, using discrete event modeling techniques, to analyze the Fuel Conditioning Facility at Argonne National Laboratory and to identify enhanced nuclear waste treatment configurations. The modeling approach and results of what-if studies are discussed. An example on how to improve productivity is presented.

Thermal simulation of the magnesium thermal of metallic uranium reduction

International Nuclear Information System (INIS)

Borges, W.A.; Saliba-Silva, A.M.

2008-01-01

Metallic uranium production is vital to fabricate fuel elements for nuclear research reactors and to produce radioisotopes and radiopharmaceuticals. Metallic uranium is got via magnesiothermal reduction of UF 4 . This reaction is carried out inside a closed graphite crucible inserted in a metallic reactor adequately sealed without any outside contact. The assembled set is gradually heated up inside a pit furnace up to reach the reaction ignition temperature (between 600-650 deg C). The optimization of the reactive system depends on the mathematical modeling using simulation by finite elements and computational calculation with specialized programs. In this way, the reactants' thermal behavior is forecast until they reach the ignition temperature. The optimization of the uranium production reaction is based on minimization of thermal losses using better the exo thermal reaction heat. As lower the thermal losses, as higher would be the heat amount to raise the temperature of reaction products. This promotes the adequate melting of uranium and slag, so allowing better metal/slag separation with higher metallic yield. This work shows how the mathematical simulation is made and supplies some preliminary results. (author)

Loss of cooling accident simulation of nuclear power station spent-fuel pool

Energy Technology Data Exchange (ETDEWEB)

Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

2014-07-01

The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

NOBLE METAL CHEMISTRY AND HYDROGEN GENERATION DURING SIMULATED DWPF MELTER FEED PREPARATION

Energy Technology Data Exchange (ETDEWEB)

Koopman, D

2008-06-25

Simulations of the Defense Waste Processing Facility (DWPF) Chemical Processing Cell vessels were performed with the primary purpose of producing melter feeds for the beaded frit program plus obtaining samples of simulated slurries containing high concentrations of noble metals for off-site analytical studies for the hydrogen program. Eight pairs of 22-L simulations were performed of the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. These sixteen simulations did not contain mercury. Six pairs were trimmed with a single noble metal (Ag, Pd, Rh, or Ru). One pair had all four noble metals, and one pair had no noble metals. One supporting 4-L simulation was completed with Ru and Hg. Several other 4-L supporting tests with mercury have not yet been performed. This report covers the calculations performed on SRNL analytical and process data related to the noble metals and hydrogen generation. It was originally envisioned as a supporting document for the off-site analytical studies. Significant new findings were made, and many previous hypotheses and findings were given additional support as summarized below. The timing of hydrogen generation events was reproduced very well within each of the eight pairs of runs, e.g. the onset of hydrogen, peak in hydrogen, etc. occurred at nearly identical times. Peak generation rates and total SRAT masses of CO{sub 2} and oxides of nitrogen were reproduced well. Comparable measures for hydrogen were reproduced with more variability, but still reasonably well. The extent of the reproducibility of the results validates the conclusions that were drawn from the data.

Multi-purpose container technologies for spent fuel management

International Nuclear Information System (INIS)

2000-12-01

The management of spent nuclear fuel is an integral part of the nuclear fuel cycle. Spent fuel management resides in the back end of the fuel cycle, and is not revenue producing as electric power generation is. It instead results in a cost associated power generation. It is a major consideration in the nuclear power industry today. Because technologies, needs and circumstances vary from country to country, there is no single, standardized approach to spent fuel management. The projected cumulative amount of spent fuel generated worldwide by 2010 will be 330 000 t HM. When reprocessing is accounted for, that amount is likely to be reduced to 215 000 t HM, which is still more than twice as much as the amount now in storage. Considering the limited capacity of at-reactor (AR) storage, various technologies are being developed for increasing storage capacities. At present, many countries are developing away-from-reactor (AFR) storage in the form of pool storage or as dry storage. Further these AFR storage systems may be at-reactor sites or away-from-reactor sites (e.g. centrally located interim storage facilities, serving several reactors). The dry storage technologies being developed are varied and include vaults, horizontal concrete modules, concrete casks, and metal casks. The review of the interim storage plans of several countries indicates that the newest approaches being pursued for spent fuel management use dual-purpose and multi-purpose containers. These containers are envisaged to hold several spent fuel assemblies, and be part of the transport, storage, and possibly geological disposal systems of an integrated spent fuel management system

Collision simulations of an exclusive ship of spent nuclear fuels

International Nuclear Information System (INIS)

Kitamura, Ou; Endo, Hisayoshi

2000-01-01

Exclusive ships for sea transport of irradiated nuclear fuels operating in Japanese territorial waters are required to be built with the special hull structure against collision. To comply with the official notice 'KAISA No. 520' issued by the Ministry of Transport, the side structure of any such exclusive ship must be designed to secure the specified energy absorption capability based on Minorsky's ship collision model. The Shipbuilding Research Association of Japan (JSRA) has studied the safety in sea transport of nuclear fuels intermittently for these several decades. Recently, the adoption of finite element method has made detailed collision analyses practicable. Since 1998, the regulation research panel No. 46 of JSRA has carried out a series of finite element collision simulations in order to estimate the realistic damage to a typical exclusive ship of spent nuclear fuels. The expected structural responses, global motions and energy absorption capabilities of both colliding and struck ships during collision were investigated. The results of the investigations have shown that the ship is very likely to withstand the collision even with one of the world's largest ship. This is due mainly to her hull structure specially strengthened beyond the crushing strength of the colliding bow structures. (author)

Physical modeling of spent-nuclear-fuel container

Directory of Open Access Journals (Sweden)

Wang Liping

2012-11-01

Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

Prototype spent-fuel canister design, analysis, and test

International Nuclear Information System (INIS)

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

Development of parallel benchmark code by sheet metal forming simulator 'ITAS'

International Nuclear Information System (INIS)

Watanabe, Hiroshi; Suzuki, Shintaro; Minami, Kazuo

1999-03-01

This report describes the development of parallel benchmark code by sheet metal forming simulator 'ITAS'. ITAS is a nonlinear elasto-plastic analysis program by the finite element method for the purpose of the simulation of sheet metal forming. ITAS adopts the dynamic analysis method that computes displacement of sheet metal at every time unit and utilizes the implicit method with the direct linear equation solver. Therefore the simulator is very robust. However, it requires a lot of computational time and memory capacity. In the development of the parallel benchmark code, we designed the code by MPI programming to reduce the computational time. In numerical experiments on the five kinds of parallel super computers at CCSE JAERI, i.e., SP2, SR2201, SX-4, T94 and VPP300, good performances are observed. The result will be shown to the public through WWW so that the benchmark results may become a guideline of research and development of the parallel program. (author)

Process Simulation of Aluminium Sheet Metal Deep Drawing at Elevated Temperatures

International Nuclear Information System (INIS)

Winklhofer, Johannes; Trattnig, Gernot; Lind, Christoph; Sommitsch, Christof; Feuerhuber, Hannes

2010-01-01

Lightweight design is essential for an economic and environmentally friendly vehicle. Aluminium sheet metal is well known for its ability to improve the strength to weight ratio of lightweight structures. One disadvantage of aluminium is that it is less formable than steel. Therefore complex part geometries can only be realized by expensive multi-step production processes. One method for overcoming this disadvantage is deep drawing at elevated temperatures. In this way the formability of aluminium sheet metal can be improved significantly, and the number of necessary production steps can thereby be reduced. This paper introduces deep drawing of aluminium sheet metal at elevated temperatures, a corresponding simulation method, a characteristic process and its optimization. The temperature and strain rate dependent material properties of a 5xxx series alloy and their modelling are discussed. A three dimensional thermomechanically coupled finite element deep drawing simulation model and its validation are presented. Based on the validated simulation model an optimised process strategy regarding formability, time and cost is introduced.

Quantum-based Atomistic Simulation of Transition Metals

International Nuclear Information System (INIS)

Moriarty, J A; Benedict, L X; Glosli, J N; Hood, R Q; Orlikowski, D A; Patel, M V; Soderlind, P; Streitz, F H; Tang, M; Yang, L H

2005-01-01

First-principles generalized pseudopotential theory (GPT) provides a fundamental basis for transferable multi-ion interatomic potentials in d-electron transition metals within density-functional quantum mechanics. In mid-period bcc metals, where multi-ion angular forces are important to structural properties, simplified model GPT or MGPT potentials have been developed based on canonical d bands to allow analytic forms and large-scale atomistic simulations. Robust, advanced-generation MGPT potentials have now been obtained for Ta and Mo and successfully applied to a wide range of structural, thermodynamic, defect and mechanical properties at both ambient and extreme conditions of pressure and temperature. Recent algorithm improvements have also led to a more general matrix representation of MGPT beyond canonical bands allowing increased accuracy and extension to f-electron actinide metals, an order of magnitude increase in computational speed, and the current development of temperature-dependent potentials

Mobility of U, Np, Pu, Am and Cm from spent nuclear fuel into bentonite clay

International Nuclear Information System (INIS)

Ramebaeck, H.; Skaalberg, M.; Eklund, U.B.; Kjellberg, L.; Werme, L.

1998-01-01

The mobility of uranium, neptunium, plutonium, americium and curium from spent nuclear fuel (UO 2 ) into compacted bentonite was studied. Pieces of spent BWR UO 2 fuel was embedded in a compacted bentonite clay/low saline synthetic groundwater system. After a contact time of six years the bentonite was sliced into 0.1 mm thick slices and analysed for its content of actinides. Radiometric as well as inductively coupled plasma mass spectrometry (ICP-MS) were used for the analysis. The influence on the mobility by the addition of metallic iron, metallic copper and vivianite (Fe(II)-mineral) to the bentonite clay was investigated. The results show a low mobility of actinides in bentonite clay. Except for uranium the mobility of the other actinides could, after six years of diffusion time, only be detected less than 1 mm from the spent fuel. (orig.)

Yugoslav spent nuclear fuel management program and international perspectives

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.; Milosevic, M.

2002-01-01

Spent nuclear fuel stored in the Vinca Institute of Nuclear Sciences, Yugoslavia, consists of about 2.5 tons of metal uranium (initial enrichment 2%) and about 20 kg uranium dioxide (dispersed in aluminum matrix, initial fuel uranium enrichment 80%). This spent nuclear fuel is generated in operation of the RA heavy water research reactor during 1959-1984 period. Both types of fuel are of ex-USSR origin, have the same shape and dimensions and approximately the same initial mass of 235 nuclide. They are known as the TVR-S type of fuel elements. The total of 8030 spent fuel elements are stored at the RA research reactor premises, almost all in the spent fuel pool filled by ordinary water. The last used 480 high-enriched uranium spent fuel elements are kept in the drained RA reactor core since 1984. Fuel layer of both enrichments is covered with thin aluminium cladding. Due to non-suitable chemical parameters of water in the spent fuel storage pool, the corrosion processes penetrated aluminium cladding and aluminium walls od storage containers during storage period long from 20 to 40 years. Activity of fission products ( 137 Cs) is detected in water samples during water inspection in 1996 and experts of the lAEA Russia and USA were invited to help. By end of 2001, some remediation of the water transparency of the storage pool and inspections of water samples taken from the storage containers with the spent fuel elements were carried out by the Vinca Institute staff and with the help of experts from the Russia and the IAEA. Following new initiatives on international perspective on spent fuel management, a proposal was set by the IAEA, and was supported by the governments of the USA and the Russian Federation to ship the spent fuel elements of the RA research reactor to Mayak spent fuel processing plant in Russia. This paper describes current status of the reactor RA spent fuel elements, initiative for new Yugoslav spent fuel management program speculates on some of the

Physical properties of Pd and Al transition metals and Pd-Al binary metal alloy investigated by using molecular dynamics simulation

International Nuclear Information System (INIS)

Coruh, A.; Uludogan, M.; Tomak, M.; Cagin, T.

2002-01-01

In this study, physical properties, such as Pair Distribution Function g(r), Structure Factor S(k)''1'',''4, Diffusion Coefficient D''2''.''4, Intermediate Scattering function S(k,t)''3'',''4 and Dynamical Structure Factor S(k,w)''3'',''4 of some transition metals and metal alloys are investigated by using molecular dynamics simulation method. The simulation is specified for Pd, Al transition metals and Pd-Al binary metal alloys in the liquid form for different concentrations and at various temperatures by using Quantum Sutton-Chen (Q-SC) inter atomic potential. Intermediate scattering function and dynamical structure factor are calculated for various values of wave vector k. Results are in good agreement with published data''1'',''3'',''4

Simulation of the metallic powders compaction process

International Nuclear Information System (INIS)

Prado, J.M.; Riera, M.D.

1998-01-01

The simulation by means of finite elements of the forming processes of mechanical components is a very useful tool for their design and validation. In this work, the simulation of the compaction of a metal powder is presented. The finite element software ABAQUS is used together with the modified CAM-clay plasticity model in order to represent the elastoplastic behaviour of the material. Density distributions are obtained and therefore the motion of the compaction punches which improve this distribution can be found. Stress distribution in the different parts of the mould can also be determined. (Author) 9 refs

Source Term Characteristics Analysis for Structural Components in PWR spent fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kook, Dong Hak; Choi, Heui Joo; Cho, Dong Keun [KAERI, Daejeon (Korea, Republic of)

2010-12-15

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core under different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.32x1015 Bequerels, 238 Watts, 4.32x109 m3 water, respectively, at 10 years after discharge. Those values correspond to 0.6 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 25{approx}50 % and 35{approx}40 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important

Speciation analysis and leaching behaviors of selected trace elements in spent SCR catalyst.

Science.gov (United States)

Dai, Zejun; Wang, Lele; Tang, Hao; Sun, Zhijun; Liu, Wei; Sun, Yi; Su, Sheng; Hu, Song; Wang, Yi; Xu, Kai; Liu, Liang; Ling, Peng; Xiang, Jun

2018-09-01

This study investigated heavy metal chemical speciation and leaching behavior from a board-type spent selective catalytic reduction (SCR) catalyst containing high concentrations of vanadium, chromium, nickel, copper, zinc, and lead. A three-step sequential extraction method, standard toxicity characteristic leaching procedure (TCLP), and leaching characteristic tests have been performed. It was found that the mobility of six heavy metals in the spent SCR catalyst was significantly different. The mobility of the six heavy metals exhibited the following order: Ni > Zn > V > Cr > As > Cu. Meanwhile, TCLP test results revealed relatively high Zn and Cr leaching rate of 83.20% and 10.35%, respectively. It was found that leaching rate was positively correlated with available contents (sum of acid soluble, reducible and oxidizable fractions). Leaching characteristics tests indicated that pH substantially affected the leaching of these heavy metals. In particular, the leaching of Cr, Ni, Cu, and Zn was positively influenced by strong acid, while V and As were easily released in the presence of strong acid and strong alkali (pH 11). In terms of kinetics, the leaching of Cr, Ni, Cu, Zn, and As within the spent catalyst was dominated by erosion and dissolution processes, which were rapid reaction processes. V was released in large amounts within 1 h, but its leaching amount sharply decreased with time due to readsorption. Copyright © 2018 Elsevier Ltd. All rights reserved.

Smelting Associated with the Advanced Spent Fuel Conditioning Process

International Nuclear Information System (INIS)

Hur, J-M.; Jeong, M-S.; Lee, W-K.; Cho, S-H.; Seo, C-S.; Park, S-W.

2004-01-01

The smelting process associated with the advanced spent fuel conditioning process (ACP) of Korea Atomic Energy Research Institute was studied by using surrogate materials. Considering the vaporization behaviors of input materials, the operation procedure of smelting was set up as (1) removal of residual salts, (2) melting of metal powder, and (3) removal of dross from a metal ingot. The behaviors of porous MgO crucible during smelting were tested and the chemical stability of MgO in the salt-being atmosphere was confirmed

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

1998-03-01

This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

Ab initio simulation of dislocation cores in metals

International Nuclear Information System (INIS)

Ventelon, L.

2008-01-01

In the framework of the multi scale simulation of metals and alloys plasticity, the aim of this study is to develop a methodology of ab initio dislocations study and to apply it to the [111] screw dislocation in the bc iron. (A.L.B.)

Recent developments in spent fuel management in Norway - 59260

International Nuclear Information System (INIS)

Bennett, Peter J.; Oberlaender, Barbara C.

2012-01-01

Spent Nuclear Fuel (SNF) in Norway has arisen from irradiation of fuel in the NORA, Jeep I and Jeep II reactors at Kjeller, and in the Heavy Boiling Water Reactor (HBWR) in Halden. In total there is some 16 tonnes of SNF, with 12 tons of aluminium-clad fuel, of which 10 tonnes is metallic uranium fuel and the remainder oxide (UO 2 ). The portion of this fuel that is similar to commercial fuel (UO 2 clad in Zircaloy) may be suitable for direct disposal on the Swedish model or in other repository designs. However, metallic uranium and/or fuels clad in aluminium are chemically reactive and there would be risks associated with direct disposal. Two committees were established by the Government of Norway in January 2009 to make recommendations for the interim storage and final disposal of spent fuel in Norway. The Technical Committee on Storage and Disposal of Metallic Uranium Fuel and Al-clad Fuels was formed with the mandate to recommend treatment (i.e. conditioning) options for metallic uranium fuel and aluminium-clad fuel to render them stable for long term storage and disposal. This committee, whose members were drawn from the nuclear industry, reported in January 2010, and recommended commercial reprocessing as the best option for these fuels. The Phase-2 committee, which in part based its work on the work of previous committees and on the report of the Technical Committee, had the mandate to find the most suitable technical solution and localisation for intermediate storage for spent nuclear fuel and long-lived waste. The membership of this committee was chosen to represent a broad cross section of stakeholders. The committee evaluated different solutions and their associated costs, and recommended one of the options. The committee's report published in early 2011. This paper summarises the conclusions of the two committees, and thereby illustrates the steps taken by one country to formulate a strategy for the long-term management of its SNF. (authors)

A review on methods of recovery of acid(s) from spent pickle liquor of steel industry.

Science.gov (United States)

Ghare, N Y; Wani, K S; Patil, V S

2013-04-01

Pickling is the process of removal of oxide layer and rust formed on metal surface. It also removes sand and corrosion products from the surface of metal. Acids such as sulfuric acid, hydrochloric acid are used for pickling. Hydrofluoric acid-Nitric acid mixture is used for stainless steel pickling. Pickling solutions are spent when acid concentration in pickling solutions decreases by 75-85%, which also has metal content up to 150-250 g/ dm3. Spent pickling liquor (SPL) should be dumped because the efficiency of pickling decreases with increasing content of dissolved metal in the bath. The SPL content depends on the plant of origin and the pickling method applied there. SPL from steel pickling in hot-dip galvanizing plants contains zinc(II), iron, traces of lead, chromium. and other heavy metals (max. 500 mg/dm3) and hydrochloric acid. Zinc(II) passes tothe spent solution after dissolution of this metal from zinc(II)-covered racks, chains and baskets used for transportation of galvanized elements. Unevenly covered zinc layers are usually removed in another pickling bath. Due to this, zinc(II) concentration increases even up to 110 g/dm3, while iron content may reach or exceed even 80 g/dm3 in the same solution. This review presents an overview on different aspects of generation and treatment of SPL with recourse to recovery of acid for recycling. Different processes are described in this review and higher weightage is given to membrane processes.

Real life experimental determination of platinum group metals content in automotive catalytic converters

Science.gov (United States)

Yakoumis, I.; Moschovi, A. M.; Giannopoulou, I.; Panias, D.

2018-03-01

The real life experimental protocol for the preparation of spent automobile catalyst samples for elemental analysis is thoroughly described in the following study. Collection, sorting and dismantling, homogenization and sample preparation for X-Ray fluorescence spectroscopy and Atomic Adsorption Spectroscopy combined with Inductive coupled plasma mass spectrometry are discussed in detail for both ceramic and metallic spent catalysts. The concentrations of Platinum Group Metals (PGMs) in spent catalytic converters are presented based on typical consignments of recycled converters (more than 45,000 pieces) from the Greek Market. The conclusions clearly denoted commercial metallic catalytic foil contains higher PGMs loading than ceramic honeycombs. On the other hand, the total PGMs loading in spent ceramic catalytic converters has been found higher than the corresponding value for the metallic ones.

Bioleaching of spent hydro-processing catalyst using acidophilic bacteria and its kinetics aspect

International Nuclear Information System (INIS)

Mishra, Debaraj; Kim, Dong J.; Ralph, David E.; Ahn, Jong G.; Rhee, Young H.

2008-01-01

Bioleaching of metals from hazardous spent hydro-processing catalysts was attempted in the second stage after growing the bacteria with sulfur in the first stage. The first stage involved transformation of elemental sulfur particles to sulfuric acid through an oxidation process by acidophilic bacteria. In the second stage, the acidic medium was utilized for the leaching process. Nickel, vanadium and molybdenum contained within spent catalyst were leached from the solid materials to liquid medium by the action of sulfuric acid that was produced by acidophilic leaching bacteria. Experiments were conducted varying the reaction time, amount of spent catalysts, amount of elemental sulfur and temperature. At 50 g/L spent catalyst concentration and 20 g/L elemental sulfur, 88.3% Ni, 46.3% Mo, and 94.8% V were recovered after 7 days. Chemical leaching with commercial sulfuric acid of the similar amount that produced by bacteria was compared. Thermodynamic parameters were calculated and the nature of reaction was found to be exothermic. Leaching kinetics of the metals was represented by different reaction kinetic equations, however, only diffusion controlled model showed the best correlation here. During the whole process Mo showed low dissolution because of substantiate precipitation with leach residues as MoO 3 . Bioleach residues were characterized by EDX and XRD

Bioleaching of spent hydro-processing catalyst using acidophilic bacteria and its kinetics aspect

Energy Technology Data Exchange (ETDEWEB)

Mishra, Debaraj [Mineral and Material Processing Division, Korea Institute of Geosciences and Mineral Resources, Daejeon 305-350 (Korea, Republic of); Department of Microbiology, Chungnam National University, Daejeon 305-764 (Korea, Republic of); Kim, Dong J. [Mineral and Material Processing Division, Korea Institute of Geosciences and Mineral Resources, Daejeon 305-350 (Korea, Republic of)], E-mail: djkim@kigam.re.kr; Ralph, David E. [AJ Parker CRC for Hydrometallurgy, Murdoch University, South Street Murdoch, Perth 6153 (Australia); Ahn, Jong G. [Mineral and Material Processing Division, Korea Institute of Geosciences and Mineral Resources, Daejeon 305-350 (Korea, Republic of); Rhee, Young H. [Department of Microbiology, Chungnam National University, Daejeon 305-764 (Korea, Republic of)

2008-04-15

Bioleaching of metals from hazardous spent hydro-processing catalysts was attempted in the second stage after growing the bacteria with sulfur in the first stage. The first stage involved transformation of elemental sulfur particles to sulfuric acid through an oxidation process by acidophilic bacteria. In the second stage, the acidic medium was utilized for the leaching process. Nickel, vanadium and molybdenum contained within spent catalyst were leached from the solid materials to liquid medium by the action of sulfuric acid that was produced by acidophilic leaching bacteria. Experiments were conducted varying the reaction time, amount of spent catalysts, amount of elemental sulfur and temperature. At 50 g/L spent catalyst concentration and 20 g/L elemental sulfur, 88.3% Ni, 46.3% Mo, and 94.8% V were recovered after 7 days. Chemical leaching with commercial sulfuric acid of the similar amount that produced by bacteria was compared. Thermodynamic parameters were calculated and the nature of reaction was found to be exothermic. Leaching kinetics of the metals was represented by different reaction kinetic equations, however, only diffusion controlled model showed the best correlation here. During the whole process Mo showed low dissolution because of substantiate precipitation with leach residues as MoO{sub 3}. Bioleach residues were characterized by EDX and XRD.

Alternatives for recovering metals from spent catalysts for hydrotreating of heavy hydrocarbons: a case study; Alternativas para la recuperacion de metales a partir de catalizadores gastados del hidrotratamiento de hidrocarburos pesados: un caso de estudio

Energy Technology Data Exchange (ETDEWEB)

Alonso, Fernando; Ramirez, Sergio; Ancheyta, Jorge; Mavil, Martha [Instituto Mexicano del Petroleo, Mexico, D.F. (Mexico)]. E-mail: jancheyt@imp.mx

2008-05-15

The increasing production of spent hydrotreating catalysts used for processing heavy hydrocarbons and the problems related to their disposal are described in this work. These catalysts contain important amounts of heavy metals such as molybdenum (Mo), nickel (Ni), cobalt (Co) and vanadium (V), which can be recovered and hence an economical benefit may be obtained. The results of experimental tests for alkaline leaching (NaOH) to recover V and Mo, and the effect of operating conditions on metal recovery are also presented. The results show that, in general, the highest recovery of Mo is obtained at pH 8.5 and leaching time of 12 hours, while in the case of V, the highest recovery is observed at pH 9.0 and 8 hours. In both cases, the leaching solution contained 10 wt % alkaline. Based on the experimental information and data from a commercial plant, a preliminary economy study was developed, in which the expected economical benefits of metals recovery from spent catalysts used for hydrotreating heavy hydrocarbon are estimated. [Spanish] En el presente trabajo se describe la problematica de la creciente produccion de catalizadores gastados de los procesos de hidrotratamiento de hidrocarburos pesados. Estos catalizadores contienen cantidades importantes de metales pesados como molibdeno (Mo), niquel (Ni), cobalto (Co) y vanadio (V), que son susceptibles de recuperarse y obtener con ello un beneficio economico. Tambien se presentan resultados de pruebas experimentales de lixiviacion alcalina (NaOH) para la recuperacion de V y Mo, y el efecto de las variables de operacion sobre la recuperacion de metales. En general, se encontro que las mejores recuperaciones de Mo fueron a pH de 8.5 y 12 h, mientras que para el V fueron a pH de 9.0 y 8 h, ambos a una concentracion del agente lixiviante de 10% en peso. Con base en la informacion experimental obtenida y datos de una planta industrial se presenta un estudio economico preliminar, en el que se estiman los beneficios

Numerical simulation of heat transfer in metal foams

Science.gov (United States)

Gangapatnam, Priyatham; Kurian, Renju; Venkateshan, S. P.

2018-02-01

This paper reports a numerical study of forced convection heat transfer in high porosity aluminum foams. Numerical modeling is done considering both local thermal equilibrium and non local thermal equilibrium conditions in ANSYS-Fluent. The results of the numerical model were validated with experimental results, where air was forced through aluminum foams in a vertical duct at different heat fluxes and velocities. It is observed that while the LTE model highly under predicts the heat transfer in these foams, LTNE model predicts the Nusselt number accurately. The novelty of this study is that once hydrodynamic experiments are conducted the permeability and porosity values obtained experimentally can be used to numerically simulate heat transfer in metal foams. The simulation of heat transfer in foams is further extended to find the effect of foam thickness on heat transfer in metal foams. The numerical results indicate that though larger foam thicknesses resulted in higher heat transfer coefficient, this effect weakens with thickness and is negligible in thick foams.

Full scale simulations of accidents on spent-nuclear-fuel shipping systems

International Nuclear Information System (INIS)

Yoshimura, H.R.

1978-01-01

In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

Numerical simulation on the explosive boiling phenomena on the surface of molten metal

International Nuclear Information System (INIS)

Chen Deqi; Peng Cheng; Wang Qinghua; Pan Liangming

2014-01-01

In this paper, numerical simulation was carried out to investigate the explosive boiling phenomenon on high temperature surface also the influence of vapor growth rate during explosive boiling, vapor condensation in sub-cooled water and the subsequent effect on flowing and heat transfer. The simulation result indicates that the steam on the molten metal surface grows with very high speed, and it pushes away the sub-cooled water around and causes severe flowing. The steam clusters which block the sub-cooled water to rewet the molten metal surface are appearing at the same time. During the growth, lifting off as well as condensation of the steam clusters, the sub-cooled water around is strongly disturbed, and obvious vortexes appear. Conversely, the vortex will influence the steam cluster detachment and cub-cooled water rewetting the metal surface. This simulation visually displays the complex explosive boiling phenomena on the molten metal surface with high temperature. (authors)

Determining heavy metals in spent compact fluorescent lamps (CFLs) and their waste management challenges: some strategies for improving current conditions.

Science.gov (United States)

Taghipour, Hassan; Amjad, Zahra; Jafarabadi, Mohamad Asghari; Gholampour, Akbar; Norouz, Prviz

2014-07-01

From environmental viewpoint, the most important advantage of compact fluorescent lamps (CFLs) is reduction of green house gas emissions. But their significant disadvantage is disposal of spent lamps because of containing a few milligrams of toxic metals, especially mercury and lead. For a successful implementation of any waste management plan, availability of sufficient and accurate information on quantities and compositions of the generated waste and current management conditions is a fundamental prerequisite. In this study, CFLs were selected among 20 different brands in Iran. Content of heavy metals including mercury, lead, nickel, arsenic and chromium was determined by inductive coupled plasma (ICP). Two cities, Tehran and Tabriz, were selected for assessing the current waste management condition of CFLs. The study found that waste generation amount of CFLs in the country was about 159.80, 183.82 and 153.75 million per year in 2010, 2011 and 2012, respectively. Waste generation rate of CFLs in Iran was determined to be 2.05 per person in 2012. The average amount of mercury, lead, nickel, arsenic and chromium was 0.417, 2.33, 0.064, 0.056 and 0.012 mg per lamp, respectively. Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. For improving the current conditions, we propose by considering the successful experience of extended producer responsibility (EPR) in other electronic waste management. The EPR program with advanced recycling fee (ARF) is implemented for collecting and then recycling CFLs. For encouraging consumers to take the spent CFLs back at the end of the products' useful life, a proportion of ARF (for example, 50%) can be refunded. On the other hand, the government and Environmental Protection Agency should support and encourage recycling companies of CFLs both technically and financially in the first place. Copyright © 2014 Elsevier Ltd. All rights reserved.

Direct reduction of uranium dioxide and few other metal oxides to corresponding metals by high temperature molten salt electrolysis

International Nuclear Information System (INIS)

Mohandas, K.S.

2017-01-01

Molten salt based electro-reduction processes, capable of directly converting solid metal oxides to metals with minimum intermediate steps, are being studied worldwide. Production of metals apart, the process assumes importance in nuclear technology in the context of pyrochemical reprocessing of spent oxide fuels, for it serves as an intermediate step to convert spent oxide fuel to a metal alloy, which in turn can be processed by molten salt electro-refining method to gain the actinides present in it. In the context of future metal fuel fast reactor programme, the electrochemical process was studied for conversion of solid UO_2 to U metal in LiCl-1wt.% Li_2O melt at 650 °C with platinum anode at the Metal Processing Studies Section, PMPD, IGCAR. A brief overview of the work is presented in the paper

Melting of Uranium Metal Powders with Residual Salts

International Nuclear Information System (INIS)

Jin-Mok Hur; Dae-Seung Kang; Chung-Seok Seo

2007-01-01

The Advanced Spent Fuel Conditioning Process (ACP) of the Korea Atomic Energy Research Institute focuses on the conditioning of Pressurized Water Reactor spent oxide nuclear fuel. After the oxide reduction step of the ACP, the resultant metal powders containing ∼ 30 wt% residual LiCl-Li 2 O should be melted for a consolidation of the fine metal powders. In this study, we investigated the melting behaviors of uranium metal powders considering the effects of a LiCl-Li 2 O residual salt. (authors)

Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

1992-12-01

Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal

Simulation of Thermal, Neutronic and Radiation Characteristics in Spent Nuclear Fuel and Radwaste Facilities

International Nuclear Information System (INIS)

Poskas, P.; Bartkus, G.

1999-01-01

The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)

Capabilities for processing shipping casks at spent fuel storage facilities

International Nuclear Information System (INIS)

Baker, W.H.; Arnett, L.M.

1978-01-01

Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

A simulator study of adverse wear with metal and cement debris contamination in metal-on-metal hip bearings.

Science.gov (United States)

Halim, T; Clarke, I C; Burgett-Moreno, M D; Donaldson, T K; Savisaar, C; Bowsher, J G

2014-03-01

Third-body wear is believed to be one trigger for adverse results with metal-on-metal (MOM) bearings. Impingement and subluxation may release metal particles from MOM replacements. We therefore challenged MOM bearings with relevant debris types of cobalt-chrome alloy (CoCr), titanium alloy (Ti6Al4V) and polymethylmethacrylate bone cement (PMMA). Cement flakes (PMMA), CoCr and Ti6Al4V particles (size range 5 µm to 400 µm) were run in a MOM wear simulation. Debris allotments (5 mg) were inserted at ten intervals during the five million cycle (5 Mc) test. In a clean test phase (0 Mc to 0.8 Mc), lubricants retained their yellow colour. Addition of metal particles at 0.8 Mc turned lubricants black within the first hour of the test and remained so for the duration, while PMMA particles did not change the colour of the lubricant. Rates of wear with PMMA, CoCr and Ti6Al4V debris averaged 0.3 mm(3)/Mc, 4.1 mm(3)/Mc and 6.4 mm(3)/Mc, respectively. Metal particles turned simulator lubricants black with rates of wear of MOM bearings an order of magnitude higher than with control PMMA particles. This appeared to model the findings of black, periarticular joint tissues and high CoCr wear in failed MOM replacements. The amount of wear debris produced during a 500 000-cycle interval of gait was 30 to 50 times greater than the weight of triggering particle allotment, indicating that MOM bearings were extremely sensitive to third-body wear. Cite this article: Bone Joint Res 2015;4:29-37. ©2015 The British Editorial Society of Bone & Joint Surgery.

Metallic Fuel Casting Development and Parameter Optimization Simulations

International Nuclear Information System (INIS)

Fielding, Randall S.; Kennedy, J.R.; Crapps, J.; Unal, C.

2013-01-01

Conclusions: • Gravity casting is a feasible process for casting of metallic fuels: – May not be as robust as CGIC, more parameter dependent to find right “sweet spot” for high quality castings; – Fluid flow is very important and is affected by mold design, vent size, super heat, etc.; – Pressure differential assist was found to be detrimental. • Simulation found that vent location was important to allow adequate filling of mold; • Surface tension plays an important role in determining casting quality; • Casting and simulations high light the need for better characterized fluid physical and thermal properties; • Results from simulations will be incorporated in GACS design such as vent location and physical property characterization

A study on the smelting of electrolytically reduced spent fuel by using surrogates

International Nuclear Information System (INIS)

Hur, Jin-Mok; Jeong, Myoung-Soo; Cho, Soo-Haeng; Seo, Chung-Seok; Park, Seong-Won

2005-01-01

A smelting as a part of the advanced spent fuel conditioning process (ACP) was studied by using surrogate materials. Residual salts including LiCl-Li 2 O were successfully separated from the metal components by an evaporation at 950degC. The melting of the metal was characterized, especially by considering the oxidation of the fine metal particles. The operation procedure of the smelting was set up as 1) removal of residual salts, 2) melting of the metal powder, and 3) a solidification of the melted mass to an ingot. (author)

Thermochemical treatment of spent ion exchange resins

International Nuclear Information System (INIS)

Ojovan, M.I.; Petrov, G.A.; Dmitriev, S.A.; Trusov, B.G.; Semenov, K.N.; Klimov, V.L.

2001-01-01

Spent ion exchange resins (IER) is a principal type of radioactive waste constantly generated by nuclear plants of various functions. The reduction of volume of this waste and its treatment to the forms suitable for long-term disposal is an urgent problem facing the present-day atomic energetics. Nowadays the technological process THOR (Studsvik, Sweden) based on the thermodestruction of IER is the best developed and realized on the industrial scale. Unfortunately, this process requires expensive equipment and great energy consumption for the moisture to be evaporated and thereafter IER to be destroyed by heat. Meanwhile the capability of some elements (Mg, Al, Si, Ti etc.) has long been known and practical use found for active interaction with water in combustion regime. This property of the metals has been used in the development of new technology of treatment of IERs in SIA ''Radon''. Wet IER is mixed with powder metal fuel (PMF) which represents a mixture of metal powder, a quantity of burning activator and some technological additives. On initiation, the mixture of IER with PMF burns without extra energy supply to generate enough heat for the moisture to be evaporated and products of IER decomposition to be destroyed and evaporated. To burn out the products of IER evaporation the air is used. The thermodynamic simulation data and the results of experiments using a pilot plant show that radionuclides contained in IER are chemically bound in ash residue consisting of metal oxides, spinel, silicates, etc. According to the experimental data, radionuclides in amounts of 90% or more of Cs-137 and up to 95% of Sr-90 and Co-60 are fixed in the ash residue. The residue volume is several times less than the initial volume of IER. Concentrations of hazard gases in off-gases do not exceed maximum permissible ones accepted in different countries. The technological process is easy to perform, it does not require sophisticated equipment and great energy consumption which

Development of Experimental Facilities for Advanced Spent Fuel Management Technology

Energy Technology Data Exchange (ETDEWEB)

You, G. S.; Jung, W. M.; Ku, J. H. [and others

2004-07-01

The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

Spent fuel treatment at ANL-West

International Nuclear Information System (INIS)

Goff, K.M.; Benedict, R.W.; Levinskas, D.

1994-01-01

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Cycle Facility at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will employ a pyrochemical process that also has applications for treating most of the fuel types within the Department of Energy complex. The treatment equipment is in its last stage of readiness, and operations will begin in the Fall of 1994

Development of the spent fuel disassembling process by utilizing the 3D graphic design technology

International Nuclear Information System (INIS)

Song, T. K.; Lee, J. Y.; Kim, S. H.; Yun, J. S.

2001-01-01

For developing the spent fuel disassembling process, the 3D graphic simulation has been established by utilizing the 3D graphic design technology which is widely used in the industry. The spent fuel disassembling process consists of a downender, a rod extraction device, a rod cutting device, a pellet extracting device and a skeleton compaction device. In this study, the 3D graphical design model of these devices is implemented by conceptual design and established the virtual workcell within kinematics to motion of each device. By implementing this graphic simulation, all the unit process involved in the spent fuel disassembling processes are analyzed and optimized. The 3D graphical model and the 3D graphic simulation can be effectively used for designing the process equipment, as well as the optimized process and maintenance process

Compaction simulation of nano-crystalline metals with molecular dynamics analysis

Directory of Open Access Journals (Sweden)

Khoei A.R.

2016-01-01

Full Text Available The molecular-dynamics analysis is presented for 3D compaction simulation of nano-crystalline metals under uniaxial compaction process. The nano-crystalline metals consist of nickel and aluminum nano-particles, which are mixed with specified proportions. The EAM pair-potential is employed to model the formation of nano-particles at different temperatures, number of nano-particles, and mixing ratio of Ni and Al nano-particles to form the component into the shape of a die. The die-walls are modeled using the Lennard-Jones inter-atomic potential between the atoms of nano-particles and die-walls. The forming process is model in uniaxial compression, which is simulated until the full-dense condition is attained at constant temperature. Numerical simulations are performed by presenting the densification of nano-particles at different deformations and distribution of dislocations. Finally, the evolutions of relative density with the pressure as well as the stress-strain curves are depicted during the compaction process.

Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

Thermal analysis of a one-element PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Fields, S.R.

1979-06-01

The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

Simulation of spent sulfite liquor fermentation using the object oriented knowledge based shell G2

Energy Technology Data Exchange (ETDEWEB)

Polakovic, M; Hoernsten, E G; Mandenius, C F

1992-10-01

This report demonstrates that simulation is a valuable tool, which can provide useful information for industrial fermentor operation and design. The key to good simulation is reliable fermentation kinetics. Starting point is the kinetics found in literature or obtained in laboratory experiments. This need not necessarily give a correct description of full-scale plant behaviour for several reasons, like population distribution (different characteristics of recycled biomass), natural selection of microorganism and metabolic behaviour modification during long-term operation, etc. Therefore, it is highly recommended to verify the kinetics on real plant data obtained either from permanent monitoring or especially designed plant measurements. We wanted to use the unique design of the MoDo ethanol plant in order to obtain sufficient information concerning the fermentation kinetics formulation based on normal steady-state operation. Unfortunately, this was not possible from the data obtained, because we could only estimate the fermentation rates in the first fermentor. The rest of the cascades was only flown through by the mash. A solution worth to try is to increase the flow rate of spent sulfite liquor, or to decrease the fermentor medium volume and then make new measurements. If this would help to formulate the process kinetics, simulation could then be used more efficiently for improving the current process or in design of the new one. (21 refs., 2 figs., 5 tabs.).

Critical and subcritical parameters of the system simulating plutonium metal dissolution

International Nuclear Information System (INIS)

Vasilev, Yury Yu.; Ryazanov, Boris G.; Sviridov, Victor I.; Mozhayeva, Lubov I.

2003-01-01

Dissolution of plutonium metal was simulated using the Monte Carlo computer code to calculate criticality safety limits for the process. Calculations were made for the constant masses of plutonium charged to the dissolving vessel considering distribution of plutonium in metal and solution phases. Critical parameters and limits were calculated as a function of dissolving vessel volume and plutonium metal mass. 240 Pu content was assumed to be from 0% to 10% (mass). Critical parameters were evaluated for the system with a water reflector. Results of this paper may be used in the designing process equipment for plutonium metal dissolution. (author)

Determination of reduction yield of lithium metal reduction process

International Nuclear Information System (INIS)

Choi, In Kyu; Cho, Young Hwan; Kim, Taek Jin; Jee, Kwang Young

2004-01-01

Metal reduction of spent oxide fuel is the first step for the effective storage of spent fuel in Korea as well as transmutation purpose of long-lived radio-nuclides. During the reduction of uranium oxide by lithium metal to uranium metal, lithium oxide is stoichiometrically produced. By determining the concentration of lithium oxide in lithium chloride, we can estimate that how much uranium oxide is converted to uranium metal. Previous method to determine the lithium oxide concentration in lithium chloride is tedious and timing consuming. This paper describe the on-line monitoring method of lithium oxide during the reduction process

Dry Refabrication Technology Development of Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lee, Jung Won; Park, G. I.; Park, C. J.

2010-04-01

Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

Development of code SFINEL (Spent fuel integrity evaluator)

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

1999-01-01

SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

Compact approach to long-term monitored retrievable storage of spent fuel

International Nuclear Information System (INIS)

Muir, D.W.

1986-01-01

We examine a new approach to monitored retrievable storage (MRS) that is extremely compact in terms of total land use and may offer increased security and reduced environmental impact, relative to current designs. This approach involves embedding the spent fuel assemblies in monolithic blocks of metallic aluminum. While this would clearly require increased effort in the spent-fuel packaging phase, it would offer in return the above-mentioned environmental advantages, plus the option of easily extending the surface-storage time scale from several years to several decades if a need for longer storage times should arise in the future

CFD simulations on the dynamics of liquid sloshing and its control in a storage tank for spent fuel applications

International Nuclear Information System (INIS)

Sanapala, V.S.; Velusamy, K.; Patnaik, B.S.V.

2016-01-01

Highlights: • Dynamics of sloshing in partially filled spent fuel storage tanks is numerically simulated. • Two type of baffle plates were examined towards the control of slosh suppression. • An optimum baffles configuration was obtained, after carrying out systematic investigations. • This vertical baffle design was effective, when tested for a seismic excitation (El centro). - Abstract: Spent nuclear liquid waste is often kept in partially filled storage tanks. When such storage tanks are subjected to wind and/or earthquake induced excitations, this could lead to detrimental conditions. Therefore, storage tank designers should ensure safe design margins and develop methodologies to overcome a wide range of possible scenarios. In the present study, systematic numerical simulations are carried out to investigate the sloshing dynamics of liquid in a storage tank, subjected to seismic excitation. As a precursor, the influence of resonant harmonic excitation on the free surface displacement, pressure distribution, slosh forces etc. is studied. To suppress the free surface fluctuations and the associated slosh force, two types of baffles viz., ring and vertical baffle are examined. Based on the response to an imposed harmonic excitation, the vertical baffle plate in the middle of the tank, was found to be effective and its dimensions are systematically optimized. This baffle geometry was tested for a well known seismic excitation (El Centro) and it was observed to effectively suppress free surface fluctuations and the slosh forces.

Determining Bond Sodium Remaining in Plenum Region of Spent Nuclear Driver Fuel

International Nuclear Information System (INIS)

Vaden, D.; Li, S.X.

2008-01-01

The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electro-chemical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials (REF 1). Upon immersion into the ER electrolyte, the sodium used to thermally bond the fuel to the clad jacket chemically reacts with the UCl3 in the electrolyte producing NaCl and uranium metal. The uranium in the spent fuel is separated from the cladding and fission products by taking advantage of the electro-chemical potential differences between uranium and the other fuel components. Assuming all the sodium in the thermal bond is converted to NaCl in the ER, the difference between the cumulative bond sodium mass in the fuel elements and the cumulative sodium mass found in the driver ER electrolyte inventory provides an upper mass limit for the sodium that migrated to the upper gas region, or plenum section, of the fuel element during irradiation in the reactor. The plenums are to be processed as metal waste via melting and metal consolidation operations. However, depending on the amount of sodium in the plenums, additional processing may be required to remove the sodium before metal waste processing

Biohydrometallurgical methods for metals recovery from waste materials

OpenAIRE

J. Willner; J. Kadukova; A. Fornalczyk; M. Saternus

2015-01-01

The article draws attention to recently conducted research of bacterial leaching of metals from various polymetallic waste. These wastes are the carriers of valuable metals: base metals, precious and platinum group metals (e.g. electronic waste, spent catalysts) or rare earth elements.

High Antioxidant Action and Prebiotic Activity of Hydrolyzed Spent Coffee Grounds (HSCG) in a Simulated Digestion-Fermentation Model: Toward the Development of a Novel Food Supplement.

Science.gov (United States)

Panzella, Lucia; Pérez-Burillo, Sergio; Pastoriza, Silvia; Martín, María Ángeles; Cerruti, Pierfrancesco; Goya, Luis; Ramos, Sonia; Rufián-Henares, José Ángel; Napolitano, Alessandra; d'Ischia, Marco

2017-08-09

Spent coffee grounds are a byproduct with a large production all over the world. The aim of this study was to explore the effects of a simulated digestion-fermentation treatment on hydrolyzed spent coffee grounds (HSCG) and to investigate the antioxidant properties of the digestion and fermentation products in the human hepatocellular carcinoma HepG2 cell line. The potentially bioaccessible (soluble) fractions exhibited high chemoprotective activity in HepG2 cells against oxidative stress. Structural analysis of both the indigestible (insoluble) and soluble material revealed partial hydrolysis and release of the lignin components in the potentially bioaccessible fraction following simulated digestion-fermentation. A high prebiotic activity as determined from the increase in Lactobacillus spp. and Bifidobacterium spp. and the production of short-chain fatty acids (SCFAs) following microbial fermentation of HSCG was also observed. These results pave the way toward the use of HSCG as a food supplement.

Development of spent fuel remote handling technology

Energy Technology Data Exchange (ETDEWEB)

Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

2007-02-15

In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

Gold biorecovery from e-waste: An improved strategy through spent medium leaching with pH modification.

Science.gov (United States)

Natarajan, Gayathri; Ting, Yen-Peng

2015-10-01

Rapid technological advancement and relatively short life time of electronic goods have resulted in an alarming growth rate of electronic waste which often contains significant quantities of toxic and precious metals. Compared to conventional recovery methods, bioleaching is an environmentally friendly process for metal extraction. Gold was bioleached from electronic scrap materials (ESM) via gold-cyanide complexation using cyanide produced from pure and mixed cultures of cyanogenic bacteria Chromobacterium violaceum, Pseudomonas aeruginosa and Pseudomonas fluorescens. As ESM was toxic to the bacteria, a two-step bioleaching approach was adopted where the solid waste was added to the bacterial culture after it has reached maximum growth and cyanide production during early stationary phase. Pure culture of C. violaceum showed the highest cyanide production, yielding maximum gold recovery of 11.3% at 0.5% w/v pulp density of ESM in two-step bioleaching. At the same pulp density of ESM, spent medium bioleaching using bacterial cell-free metabolites achieved gold recovery of 18%. Recovery increased to 30% when the pH of the spent medium was increased to shift the equilibrium in favor of cyanide ions production. It is demonstrated for the first time that pH modification of spent medium further improved metal solubilization and yielded higher metal recovery (compared to two-step bioleaching). Copyright © 2015 Elsevier Ltd. All rights reserved.

Ion bombardment induced smoothing of amorphous metallic surfaces: Experiments versus computer simulations

International Nuclear Information System (INIS)

Vauth, Sebastian; Mayr, S. G.

2008-01-01

Smoothing of rough amorphous metallic surfaces by bombardment with heavy ions in the low keV regime is investigated by a combined experimental-simulational study. Vapor deposited rough amorphous Zr 65 Al 7.5 Cu 27.5 films are the basis for systematic in situ scanning tunneling microscopy measurements on the smoothing reaction due to 3 keV Kr + ion bombardment. The experimental results are directly compared to the predictions of a multiscale simulation approach, which incorporates stochastic rate equations of the Langevin type in combination with previously reported classical molecular dynamics simulations [Phys. Rev. B 75, 224107 (2007)] to model surface smoothing across length and time scales. The combined approach of experiments and simulations clearly corroborates a key role of ion induced viscous flow and ballistic effects in low keV heavy ion induced smoothing of amorphous metallic surfaces at ambient temperatures

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

Biohydrometallurgical methods for metals recovery from waste materials

Directory of Open Access Journals (Sweden)

J. Willner

2015-01-01

Full Text Available The article draws attention to recently conducted research of bacterial leaching of metals from various polymetallic waste. These wastes are the carriers of valuable metals: base metals, precious and platinum group metals (e.g. electronic waste, spent catalysts or rare earth elements.

Molecular Models for DSMC Simulations of Metal Vapor Deposition

OpenAIRE

Venkattraman, A; Alexeenko, Alina A

2010-01-01

The direct simulation Monte Carlo (DSMC) method is applied here to model the electron‐beam (e‐beam) physical vapor deposition of copper thin films. A suitable molecular model for copper‐copper interactions have been determined based on comparisons with experiments for a 2D slit source. The model for atomic copper vapor is then used in axi‐symmetric DSMC simulations for analysis of a typical e‐beam metal deposition system with a cup crucible. The dimensional and non‐dimensional mass fluxes obt...

Source Term Characterization for Structural Components in 17 x 17 KOFA Spent Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Cho, Dong Keun; Kook, Dong Hak; Choi, Heui Joo; Choi, Jong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-12-15

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.40 x 10{sup 15} Bequerels, 236 Watts, 4.34 x 10{sup 9} m{sup 3}-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20 {approx} 45 % and 30 {approx} 45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

Paper summary inventory assessment of DOE spent nuclear fuels

International Nuclear Information System (INIS)

Abbott, D.G.; Bringhurst, A.R.; Fillmore, D.L.

1994-01-01

The U.S. Department of Energy (DOE) has determined that it will not longer reprocess its spent nuclear fuel. This decision made it necessary to manage this fuel for long-term interim storage and ultimate disposal. DOE is developing a computerized database of its spent nuclear fuel inventory. This database contains information about the fuels and the fuel storage locations. There is approximately 2,618 metric tons initial heavy metal of fuel, stored at 12 locations. For analysis in an environmental impact statement, the fuel has been divided into six categories: naval, aluminum-based, Hanford defense, graphite, commercial-type, and test and experimental. This paper provides a discussion of the development of the database, and includes summary inventory information and a brief description of the fuels

Spent NiMH batteries-The role of selective precipitation in the recovery of valuable metals

Science.gov (United States)

Bertuol, Daniel Assumpção; Bernardes, Andréa Moura; Tenório, Jorge Alberto Soares

The production of electronic equipment, such as computers and cell phones, and, consequently, batteries, has increased dramatically. One of the types of batteries whose production and consumption has increased in recent times is the nickel metal hydride (NiMH) battery. This study evaluated a hydrometallurgical method of recovery of rare earths and a simple method to obtain a solution rich in Ni-Co from spent NiMH batteries. The active materials from both electrodes were manually removed from the accumulators and leached. Several acid and basic solutions for the recovery of rare earths were evaluated. Results showed that more than 98 wt.% of the rare earths were recovered as sulfate salts by dissolution with sulfuric acid, followed by selective precipitation at pH 1.2 using sodium hydroxide. The complete process, precipitation at pH 1.2 followed by precipitation at pH 7, removed about 100 wt.% of iron and 70 wt.% of zinc from the leaching solution. Results were similar to those found in studies that used solvent extraction. This method is easy, economic, and does not pose environmental threats of solvent extraction.

Development of a water boil-off spent-fuel calorimeter system

International Nuclear Information System (INIS)

Creer, J.M.; Shupe, J.W. Jr.

1981-05-01

A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW

Platinum-group metals from nuclear reactions as a possible resource

International Nuclear Information System (INIS)

Jensen, G.A.

1985-03-01

Spent nuclear fuels contain significant quantities of three of the platinum-group metals (ruthenium, rhodium, and palladium), and a related element technetium, which is nearly absent in nature. Applications for ruthenium, rhodium, and palladium are well established. Since the supply of these and other platinum metals is largely from foreign sources, they are considered strategic materials. Existing and future spent nuclear fuels contain quantities of these platinum metals that exceed the United States reserve base. Technetium has properties similar to platinum metals and has unique, useful properties of its own. The technical feasibility of recovering and using fission product platinum metals (and technetium) extensively in industry depends on: thoroughly decontaminating platinum-group metals from all other radioactive materials in the waste stream; separating platinum-group metals from one another in very high purity; using applications where appropriate control of the residual radioactivity is possible; and whether or not the United States will recover or process spent fuel prior to repository storage. If the radioactivity must be removed, isotope separation or long term storage to allow decay of the contained radioisotopes may be possible. 7 figs., 7 tabs

Sustainable recovery of valuable metals from spent lithium-ion batteries using DL-malic acid: Leaching and kinetics aspect.

Science.gov (United States)

Sun, Conghao; Xu, Liping; Chen, Xiangping; Qiu, Tianyun; Zhou, Tao

2018-02-01

An eco-friendly and benign process has been investigated for the dissolution of Li, Co, Ni, and Mn from the cathode materials of spent LiNi 1/3 Co 1/3 Mn 1/3 O 2 batteries, using DL-malic acid as the leaching agent in this study. The leaching efficiencies of Li, Co, Ni, and Mn can reach about 98.9%, 94.3%, 95.1%, and 96.4%, respectively, under the leaching conditions of DL-malic acid concentration of 1.2 M, hydrogen peroxide content of 1.5 vol.%, solid-to-liquid ratio of 40 g l -1 , leaching temperature of 80°C, and leaching time of 30 min. In addition, the leaching kinetic was investigated based on the shrinking model and the results reveal that the leaching reaction is controlled by chemical reactions within 10 min with activation energies (Ea) of 21.3 kJ·mol -1 , 30.4 kJ·mol -1 , 27.9 kJ·mol -1 , and 26.2 kJ·mol -1 for Li, Co, Ni, and Mn, respectively. Diffusion process becomes the controlled step with a prolonged leaching time from 15 to 30 min, and the activation energies (Ea) are 20.2 kJ·mol -1 , 28.9 kJ·mol -1 , 26.3 kJ·mol -1 , and 25.0 kJ·mol -1 for Li, Co, Ni, and Mn, respectively. This hydrometallurgical route was found to be effective and environmentally friendly for leaching metals from spent lithium batteries.

Testing of Metal Cask and Concrete Cask

International Nuclear Information System (INIS)

Shirai, K.; Wataru, M.; Takeda, H.; Tani, J.; Arai, T.; Saegusa, T.

2015-01-01

In Japan, the first interim spent fuel storage facility (ISF) outside of nuclear power plant site in use of dual-purpose metal cask is being planned to start its commercial operation in 2012 in Mutsu city, Aomori prefecture. The CRIEPI (Central Research Institute of Electric Power Industry) has executed several study programs on demonstrative testing for interim storage of spent fuel, mainly related to metal cask and concrete cask storage technology to reflect in Japanese safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy and Trade Industry). On top of that, the Japan Nuclear Energy Safety Organization (JNES) has executed study programs on spent fuel integrity, etc. This paper introduces the summary of these research programs. (author)

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

International Nuclear Information System (INIS)

2015-10-01

At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact

Fiscal Year (FY) 2017 Activities for the Spent Fuel Nondestructive Assay Project

Energy Technology Data Exchange (ETDEWEB)

Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McMath, Garrett Earl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Grogan, Brandon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-11

The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify results from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.

Methods for expanding the capacity of spent fuel storage facilities

International Nuclear Information System (INIS)

1990-06-01

At the beginning of 1989 more than 55,000 metric tonnes of heavy metal (MTHM) of spent Light Water Reactor (LWR) and Heavy Water Reactor (HWR) fuel had been discharged worldwide from nuclear power plants. Only a small fraction of this fuel has been reprocessed. The majority of the spent fuel assemblies are currently held at-reactor (AR) or away-from-reactor (AFR) in storage awaiting either chemical processing or final disposal depending on the fuel concept chosen by individual countries. Studies made by NEA and IAEA have projected that annual spent fuel arising will reach about 10,000 t HM in the year 2000 and cumulative arising will be more than 200,000 t HM. Taking into account the large quantity of spent fuel discharged from NPP and that the first demonstrations of the direct disposal of spent fuel or HLW are expected only after the year 2020, long-term storage will be the primary option for management of spent fuel until well into the next century. There are several options to expand storage capacity: (1) to construct new away-from-reactor storage facilities, (2) to transport spent fuel from a full at-reactor pool to another site for storage in a pool that has sufficient space to accommodate it, (3) to expand the capacity of existing AR pools by using compact racks, double-tierce, rod consolidation and by increasing the dimensions of existing pools. The purpose of the meeting was: to exchange new information on the international level on the subject connected with the expansion of storage capacities for spent fuel; to elaborate the state-of-the-art of this problem; to define the most important areas for future activity; on the basis of the above information to give recommendations to potential users for selection and application of the most suitable methods for expanding spent fuel facilities taking into account the relevant country's conditions. Refs, figs and tabs

Recycling of spent lithium-ion battery with polyvinyl chloride by mechanochemical process.

Science.gov (United States)

Wang, Meng-Meng; Zhang, Cong-Cong; Zhang, Fu-Shen

2017-09-01

In the present study, cathode materials (C/LiCoO 2 ) of spent lithium-ion batteries (LIBs) and waste polyvinyl chloride (PVC) were co-processed via an innovative mechanochemical method, i.e. LiCoO 2 /PVC/Fe was co-grinded followed by water-leaching. This procedure generated recoverable LiCl from Li by the dechlorination of PVC and also generated magnetic CoFe 4 O 6 from Co. The effects of different additives (e.g. alkali metals, non-metal oxides, and zero-valent metals) on (i) the conversion rates of Li and Co and (ii) the dechlorination rate of PVC were investigated, and the reaction mechanisms were explored. It was found that the chlorine atoms in PVC were mechanochemically transformed into chloride ions that bound to the Li in LiCoO 2 to form LiCl. This resulted in reorganization of the Co and Fe crystals to form the magnetic material CoFe 4 O 6 . This study provides a more environmentally-friendly, economical, and straightforward approach for the recycling of spent LIBs and waste PVC compared to traditional processes. Copyright © 2017. Published by Elsevier Ltd.

Shielding Performance Measurements of Spent Fuel Transportation Container

Directory of Open Access Journals (Sweden)

SUN Hong-chao

2015-11-01

Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

1979-01-01

The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Rechard, R.P.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency's Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

Radiation heat transfer model in a spent fuel pool by TRACE code

International Nuclear Information System (INIS)

Sanchez-Saez, F.; Carlos, S.; Villanueva, J.F.; Martorell, S.

2014-01-01

Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. The work presented consists of the thermohydraulic simulation of spent fuel pool behavior after a loss of coolant throughout transfer channel with loss of cooling transient is produced. The simulation is done with the TRACE code. One of the most important variables that define the behavior of the pool is cladding temperature, which evolution depends on the heat emission. In this work convection and radiation heat transfer is considered. When both heat transfer models are considered, a clear delay in achieving the maximum peak cladding temperature (1477 K) is observed compared with the simulation in which only convection heat transfer is considered. (authors)

Burnup simulations and spent fuel characteristics of ZrO{sub 2} based inert matrix fuels

Energy Technology Data Exchange (ETDEWEB)

Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Department of Theoretical and Applied Mechanics, Cornell University, Ithaca, NY (United States)]. E-mail: mrd6@cornell.edu; Herring, S.T. [Idaho National Laboratory, Idaho Falls, ID (United States); Cady, K.B. [Department of Theoretical and Applied Mechanics, Cornell University, Ithaca, NY (United States)

2007-03-31

Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO{sub 2} based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.

A study on the expulsion of iodine from spent-fuel solutions

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1995-02-01

During dissolution of spent nuclear fuels, some radioiodine remains in spent-fuel solutions. Its expulsion to dissolver off-gas is important to minimize iodine escape to the environment. In our current work, the iodine remaining in spent-fuel solutions varied from 0 to 10% after dissolution of spent PWR-fuel specimens (approximately 3 g each). The amount remaining probably was dependent upon the dissolution time required. The cause is ascribable to the increased nitrous acid concentration that results from NOx generated during dissolution. The presence of nitrous acid was confirmed spectrophotometrically in an NO-HNO{sub 3} system at 100{degrees}C. Experiments examining NOx concentration versus the quantity of iodine in a simulated spent-fuel solution indicate that iodine (I{minus}) in spent fuels is subjected to the following three reactions: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid arising from NOx, and (3) formation of colloidal iodine (AgI, PdI{sub 2}), the major iodine species in a spent-fuel solution. Reaction (2) competes with reaction (3) to control the quantity of iodine remaining in solution. The following two-step expulsion process to remove iodine from a spent-fuel solution was derived from these experiments: Step One - Heat spent-fuel solutions without NOx sparging. When aged colloidal iodine is present, an excess amount of iodate should be added to the solution. Step Two - Sparge the fuel solution with NOx while heating. Effect of this new method was confirmed by use of a spent PWR-fuel solution.

Spent fuel test. Climax data acquisition system integration report

International Nuclear Information System (INIS)

Nyholm, R.A.; Brough, W.G.; Rector, N.L.

1982-06-01

The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

Spent Fuel Test - Climax data acquisition system operations manual

International Nuclear Information System (INIS)

Nyholm, R.A.

1983-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

Storage of water reactor spent fuel in water pools. Survey of world experience

International Nuclear Information System (INIS)

1982-01-01

Following discharge from a nuclear reactor, spent fuel has to be stored in water pools at the reactor site to allow for radioactive decay and cooling. After this initial storage period, the future treatment of spent fuel depends on the fuel cycle concept chosen. Spent fuel can either be treated by chemical processing or conditioning for final disposal at the relevant fuel cycle facilities, or be held in interim storage - at the reactor site or at a central storage facility. Recent forecasts predict that, by the year 2000, more than 150,000 tonnes of heavy metal from spent LWR fuel will have been accumulated. Because of postponed commitments regarding spent fuel treatment, a significant amount of spent fuel will still be held in storage at that time. Although very positive experience with wet storage has been gained over the past 40 years, making wet storage a proven technology, it appears desirable to summarize all available data for the benefit of designers, storage pool operators, licensing agenices and the general public. Such data will be essential for assessing the viability of extended water pool storage of spent nuclear fuel. In 1979, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD jointly issued a questionnaire dealing with all aspects of water pool storage. This report summarizes the information received from storage pool operators

Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

Energy Technology Data Exchange (ETDEWEB)

Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

2006-08-15

Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

Overview of the spent nuclear fuel project at Hanford

International Nuclear Information System (INIS)

Daily, J.L.

1995-02-01

The Spent Nuclear Fuel Project's mission at Hanford is to open-quotes Provide safe, economic and environmentally sound management of Hanford spent nuclear fuel in a manner which stages it to final disposition.close quotes The inventory of spent nuclear fuel (SNF) at the Hanford Site covers a wide variety of fuel types (production reactor to space reactor) in many facilities (reactor fuel basins to hot cells) at locations all over the Site. The 2,129 metric tons of Hanford SNF represents about 80% of the total US Department of Energy (DOE) inventory. About 98.5% of the Hanford SNF is 2,100 metric tons of metallic uranium production reactor fuel currently stored in the 1950s vintage K Basins in the 100 Area. This fuel has been slowly corroding, generating sludge and contaminating the basin water. This condition, coupled with aging facilities with seismic vulnerabilities, has been identified by several groups, including stakeholders, as being one of the most urgent safety and environmental concerns at the Hanford Site. As a direct result of these concerns, the Spent Nuclear Fuel Project was recently formed to address spent fuel issues at Hanford. The Project has developed the K Basins Path Forward to remove fuel from the basins and place it in dry interim storage. Alternatives that addressed the requirements were developed and analyzed. The result is a two-phased approach allowing the early removal of fuel from the K Basins followed by its stabilization and interim storage consistent with the national program

Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Sihm Kvenangen, Karen

2007-06-15

In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

International Nuclear Information System (INIS)

Sihm Kvenangen, Karen

2007-06-01

In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation

Spent fuel characterization for the commercial waste and spent fuel packaging program

International Nuclear Information System (INIS)

Fish, R.L.; Davis, R.B.; Pasupathi, V.; Klingensmith, R.W.

1980-03-01

This document presents the rationale for spent fuel characterization and provides a detailed description of the characterization examinations. Pretest characterization examinations provide quantitative and qualitative descriptions of spent fuel assemblies and rods in their irradiated conditions prior to disposal testing. This information is essential in evaluating any subsequent changes that occur during disposal demonstration and laboratory tests. Interim examinations and post-test characterization will be used to identify fuel rod degradation mechanisms and quantify degradation kinetics. The nature and behavior of the spent fuel degradation will be defined in terms of mathematical rate equations from these and laboratory tests and incorporated into a spent fuel performance prediction model. Thus, spent fuel characterization is an essential activity in the development of a performance model to be used in evaluating the ability of spent fuel to meet specific waste acceptance criteria and in evaluating incentives for modification of the spent fuel assemblies for long-term disposal purposes

Technical concept for a test of geologic storage of spent reactor fuel in the climax granite, Nevada Test Site

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

1979-01-01

We plan to emplace spent fuel assemblies from an operating commercial nuclear reactor in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent fuel will be emplaced with 6 electrical simulator canisters in a storage drift 420 m below in surface and their effects compared. Two adjacent drifts will contain electrical heaters, operated to simulate the temperature-stress-displacement fields of a large repository. We describe the test objectives, the technical issues, the site, the preoperational measurement program, thermal and mechanical response calculations, the characteristics of the spent fuel, the field instrumentation and data-acquisition systems, and the system for handling the spent fuel

Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

Energy Technology Data Exchange (ETDEWEB)

Lee, Sanghoon; Cho, Sangsoon; Jeon, Jeeon; Kim, Kiyoung; Seo, Kiseog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-02-15

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches, those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

International Nuclear Information System (INIS)

Lee, Sanghoon; Cho, Sangsoon; Jeon, Jeeon; Kim, Kiyoung; Seo, Kiseog

2014-01-01

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches, those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters

Separation of Metals From Spent Catalysts Waste by Bioleaching Process

OpenAIRE

Sirin Fairus, Tria Liliandini, M.Febrian, Ronny Kurniawan

2010-01-01

A kind of waste that hard to be treated is a metal containing solid waste. Leaching method is one thealternative waste treatment. But there still left an obstacle on this method, it is the difficulty to find theselective solvent for the type of certain metal that will separated. Bioleaching is one of the carry ablealternative waste treatments to overcome that obstacle. Bioleaching is a metal dissolving process orextraction from a sediment become dissolve form using microorganisms. On this met...

Catalytic oxidative pyrolysis of spent organic ion exchange resins from nuclear power plants

International Nuclear Information System (INIS)

Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.; Shirsat, A.N.; Bharadwaj, S.R.

2005-08-01

The spent IX resins from nuclear power reactors are highly active solid wastes generated during operations of nuclear reactors. Catalytic oxidative pyrolysis of these resins can lead to high volume reduction of these wastes. Low temperature pyrolysis of transition metal ion loaded IX resins in presence of nitrogen was carried out in order to optimize catalyst composition to achieve maximum weight reduction. Thermo gravimetric analysis of the pyrolysis residues was carried out in presence of air in order to compare the oxidative characteristics of transition metal oxide catalysts. Copper along with iron, chromium and nickel present in the spent IX resins gave the most efficient catalyst combination for catalytic and oxidative pyrolysis of the residues. During low temperature catalytic pyrolysis, 137 Cesium volatility was estimated to be around 0.01% from cationic resins and around 0.1% from anionic resins. During oxidative pyrolysis at 700 degC, nearly 10 to 40% of 137 Cesium was found to be released to off gases depending upon type of resin and catalyst loaded on to it. The oxidation of pyrolytic residues at 700 degC gave weight reduction of 15% for cationic resins and 93% for anionic resins. Catalytic oxidative pyrolysis is attractive for reducing weight and volume of spent cationic resins from PHWRs and VVERs. (author)

Solving optimisation problems in metal forming using Finite Element simulation and metamodelling techniques

NARCIS (Netherlands)

Bonte, M.H.A.; van den Boogaard, Antonius H.; Huetink, Han

2005-01-01

During the last decades, Finite Element (FEM) simulations of metal forming processes have become important tools for designing feasible production processes. In more recent years, several authors recognised the potential of coupling FEM simulations to mathematical optimisation algorithms to design

Vacuum pyrolysis and hydrometallurgical process for the recovery of valuable metals from spent lithium-ion batteries

International Nuclear Information System (INIS)

Sun, Liang; Qiu, Keqiang

2011-01-01

Highlights: → The cathode active materials LiCoO 2 from spent lithium-ion batteries peeled completely from aluminum foils by vacuum pyrolysis and hydrometallurgical process. → The aluminum foils were excellent without damage after vacuum pyrolysis. → The pyrolysis products organic fluorine compounds from organic electrolyte and binder were collected and enriched. → High leaching efficiencies of cobalt and lithium were obtained with H 2 SO 4 and H 2 O 2 . - Abstract: Spent lithium-ion batteries contain lots of strategic resources such as cobalt and lithium together with other hazardous materials, which are considered as an attractive secondary resource and environmental contaminant. In this work, a novel process involving vacuum pyrolysis and hydrometallurgical technique was developed for the combined recovery of cobalt and lithium from spent lithium-ion batteries. The results of vacuum pyrolysis of cathode material showed that the cathode powder composing of LiCoO 2 and CoO peeled completely from aluminum foils under the following experimental conditions: temperature of 600 o C, vacuum evaporation time of 30 min, and residual gas pressure of 1.0 kPa. Over 99% of cobalt and lithium could be recovered from peeled cobalt lithium oxides with 2 M sulfuric acid leaching solution at 80 o C and solid/liquid ratio of 50 g L -1 for 60 min. This technology offers an efficient way to recycle valuable materials from spent lithium-ion batteries, and it is feasible to scale up and help to reduce the environmental pollution of spent lithium-ion batteries.

Vacuum pyrolysis and hydrometallurgical process for the recovery of valuable metals from spent lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Sun, Liang [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Key Laboratory of Resources Chemistry of Nonferrous Metals, Central South University, Ministry of Education of the People' s Republic of China (China); Qiu, Keqiang, E-mail: qiuwhs@sohu.com [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Key Laboratory of Resources Chemistry of Nonferrous Metals, Central South University, Ministry of Education of the People' s Republic of China (China)

2011-10-30

Highlights: {yields} The cathode active materials LiCoO{sub 2} from spent lithium-ion batteries peeled completely from aluminum foils by vacuum pyrolysis and hydrometallurgical process. {yields} The aluminum foils were excellent without damage after vacuum pyrolysis. {yields} The pyrolysis products organic fluorine compounds from organic electrolyte and binder were collected and enriched. {yields} High leaching efficiencies of cobalt and lithium were obtained with H{sub 2}SO{sub 4} and H{sub 2}O{sub 2}. - Abstract: Spent lithium-ion batteries contain lots of strategic resources such as cobalt and lithium together with other hazardous materials, which are considered as an attractive secondary resource and environmental contaminant. In this work, a novel process involving vacuum pyrolysis and hydrometallurgical technique was developed for the combined recovery of cobalt and lithium from spent lithium-ion batteries. The results of vacuum pyrolysis of cathode material showed that the cathode powder composing of LiCoO{sub 2} and CoO peeled completely from aluminum foils under the following experimental conditions: temperature of 600 {sup o}C, vacuum evaporation time of 30 min, and residual gas pressure of 1.0 kPa. Over 99% of cobalt and lithium could be recovered from peeled cobalt lithium oxides with 2 M sulfuric acid leaching solution at 80 {sup o}C and solid/liquid ratio of 50 g L{sup -1} for 60 min. This technology offers an efficient way to recycle valuable materials from spent lithium-ion batteries, and it is feasible to scale up and help to reduce the environmental pollution of spent lithium-ion batteries.

Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1992-01-01

Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. The maximum rod temperature is a limiting factor in the amount of spent fuel that can be loaded in a transportation cask. The scope of this work is to demonstrate that reasonable and conservative spent fuel rod temperature predictions can be made using commercially available thermal analysis codes. The demonstration is accomplished by a comparison between numerical temperature predictions, with a commercially available thermal analysis code, and experimental temperature data for electrical rod heaters simulating a horizontally oriented spent fuel rod bundle

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

Energy Technology Data Exchange (ETDEWEB)

N/A

2000-04-14

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

Advanced Wear Simulation for Bulk Metal Forming Processes

Directory of Open Access Journals (Sweden)

Behrens Bernd-Arno

2016-01-01

Full Text Available In the recent decades the finite element method has become an essential tool for the cost-efficient virtual process design in the metal forming sector in order to counter the constantly increasing quality standards, particularly from the automotive industry as well as intensified international competition in the forging industry. An optimized process design taking precise tool wear prediction into account is a way to increase the cost-efficiency of the bulk metal forming processes. The main objective of the work presented in this paper is a modelling algorithm, which allows predicting die wear with respect to a geometry update during the forming simulation. Changes in the contact area caused by geometry update lead to the different die wear distribution. It primarily concerns the die areas, which undergo high thermal and mechanical loads.

Metal release from simulated fixed orthodontic appliances.

Science.gov (United States)

Hwang, C J; Shin, J S; Cha, J Y

2001-10-01

Most orthodontic appliances and archwires are stainless steel or nickel-titanium (NiTi) alloys that can release metal ions, with saliva as the medium. To measure metal released from the fixed orthodontic appliances currently in use, we fabricated simulated fixed orthodontic appliances that corresponded to half of the maxillary arch and soaked them in 50 mL of artificial saliva (pH 6.75 +/- 0.15, 37 degrees C) for 3 months. We used brackets, tubes, and bands made by Tomy (Tokyo, Japan). Four groups were established according to the appliance manufacturer and the type of metal in the .016 x .022-in archwires. Groups A and B were stainless steel archwires from Ormco (Glendora, Calif) and Dentaurum (Ispringen, Germany), respectively, and groups C and D were both NiTi archwires with Ormco's copper NiTi and Tomy's Bioforce sentalloy, respectively. Stainless steel archwires were heat treated in an electric furnace at 500 degrees C for 1 minute and quenched in water. We measured the amount of metal released from each group by immersion time. Our conclusions were as follows: (1) there was no increase in the amount of chromium released after 4 weeks in group A, 2 weeks in group B, 3 weeks in group C, and 8 weeks in group D; (2) there was no increase in the amount of nickel released after 2 weeks in group A, 3 days in group B, 7 days in group C, and 3 weeks in group D; and (3) there was no increase in the amount of iron released after 2 weeks in group A, 3 days in group B, and 1 day in groups C and D. In our 3-month-long investigation, we saw a decrease in metal released as immersion time increased.

Recent advances during the treatment of spent EBR-II fuel

International Nuclear Information System (INIS)

Westphal, B.R.; Mariani, R.D.; Vaden, D.E.; Sherman, S.R.; Li, S.X.; Keiser, D.D. Jr.

2000-01-01

Several recent advances have been achieved for the electrometallurgical treatment of spent nuclear fuel. In anticipation of production operations at Argonne National Laboratory-West, development of both electrorefining and metal processing has been ongoing in the post-demonstration phase in order to further optimize the process. These development activities show considerable promise. This paper discusses the results of recent experiments as well as plans for future investigations

Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

International Nuclear Information System (INIS)

Casal, V.; Graf, E.; Hartmann, W.

1976-09-01

The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

Science.gov (United States)

Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Simulation of short-term annealing of displacement cascades in FCC metals

International Nuclear Information System (INIS)

Heinisch, H.L.; Doran, D.G.; Schwartz, D.M.

1980-01-01

Computer models have been developed for the simulation of high energy displacement cascades. The objective is the generation of defect production functions for use in correlation analysis of radiation effects in fusion reactor materials. In particular, the stochastic cascade annealing simulation code SCAS has been developed and used to model the short-term annealing behavior of simulated cascades in FCC metals. The code is fast enough to make annealing of high energy cascades practical. Sets of cascades from 5 keV to 100 keV in copper were generated by the binary collision code MARLOWE

Numerical Simulations of Particle Deposition in Metal Foam Heat Exchangers

Science.gov (United States)

Sauret, Emilie; Saha, Suvash C.; Gu, Yuantong

2013-01-01

Australia is a high-potential country for geothermal power with reserves currently estimated in the tens of millions of petajoules, enough to power the nation for at least 1000 years at current usage. However, these resources are mainly located in isolated arid regions where water is scarce. Therefore, wet cooling systems for geothermal plants in Australia are the least attractive solution and thus air-cooled heat exchangers are preferred. In order to increase the efficiency of such heat exchangers, metal foams have been used. One issue raised by this solution is the fouling caused by dust deposition. In this case, the heat transfer characteristics of the metal foam heat exchanger can dramatically deteriorate. Exploring the particle deposition property in the metal foam exchanger becomes crucial. This paper is a numerical investigation aimed to address this issue. Two-dimensional (2D) numerical simulations of a standard one-row tube bundle wrapped with metal foam in cross-flow are performed and highlight preferential particle deposition areas.

Dissolution of heavy metals from electrostatic precipitator (ESP) dust ...

African Journals Online (AJOL)

SIBOO

Key words: Fungal leaching, sponge iron, electrostatic precipitator (ESP) dust, metal dissolution. INTRODUCTION ... ability of micro organisms to transform solid compounds ..... of metals from spent lithium ion secondary batteries using A.

Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

International Nuclear Information System (INIS)

2000-01-01

DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate

Novel Approach for in Situ Recovery of Lithium Carbonate from Spent Lithium Ion Batteries Using Vacuum Metallurgy.

Science.gov (United States)

Xiao, Jiefeng; Li, Jia; Xu, Zhenming

2017-10-17

Lithium is a rare metal because of geographical scarcity and technical barrier. Recycling lithium resource from spent lithium ion batteries (LIBs) is significant for lithium deficiency and environmental protection. A novel approach for recycling lithium element as Li 2 CO 3 from spent LIBs is proposed. First, the electrode materials preobtained by mechanical separation are pyrolyzed under enclosed vacuum condition. During this process the Li is released as Li 2 CO 3 from the crystal structure of lithium transition metal oxides due to the collapse of the oxygen framework. An optimal Li recovery rate of 81.90% is achieved at 973 K for 30 min with a solid-to-liquid ratio of 25 g L -1 , and the purity rate of Li 2 CO 3 is 99.7%. The collapsed mechanism is then presented to explain the release of lithium element during the vacuum pyrolysis. Three types of spent LIBs including LiMn 2 O 4 , LiCoO 2 , and LiCo x Mn y Ni z O 2 are processed to prove the validity of in situ recycling Li 2 CO 3 from spent LIBs under enclosed vacuum condition. Finally, an economic assessment is taken to prove that this recycling process is positive.

The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

International Nuclear Information System (INIS)

Saphier, D.; Madell, J.T.

1982-01-01

A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

International Nuclear Information System (INIS)

Butkovich, T.R.

1980-05-01

A generic test of the geologic storage of spent-fuel assemblies is being made at Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes, ADINA and ADINAT

Spent fuel workshop'2002

International Nuclear Information System (INIS)

Poinssot, Ch.

2002-01-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

Energy Technology Data Exchange (ETDEWEB)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function {Phi}(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

International Nuclear Information System (INIS)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

2011-01-01

Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function Φ(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

Some thermal analysis aspects of metal encapsulated waste

International Nuclear Information System (INIS)

Jardine, L.J.; Steindler, M.J.

1978-01-01

This paper is to summarize two waste management schemes: (1) packaging for extended storage of LWR spent fuel assemblies, with the capability for simple conversion either to terminal storage if a ''throwaway'' fuel cycle is ultimately adopted or to a form that can be reprocessed and (2) packaging for the terminal storage of solidified high-level wastes when the reprocessing of spent fuel is initiated. Only concepts utilizing metals or metal alloys to encapsulate either spent fuel or solidified high-level waste forms have been considered. Conceptual process flow sheets have been constructed to allow potential advantages and disadvantages of encapsulation alternatives to be identified in comparison with more conventional reference processes. Identification is also made of uncertainties of the analysis due to a lack of fundamental data required to perform evaluations. 3 tables

Postmortem metallurgical examination of a fire-exposed spent fuel shipping cask

International Nuclear Information System (INIS)

Rack, H.J.; Yoshimura, H.R.

1980-04-01

A potmortem examination of a large fire-exposed rail-transported spent fuel shipping container has revealed the presence of two macrofissures in the outer cask shell. The first, a part-thru crack located within the seam weld fusion zone of the outer cask shell, was typical of hot cracks that may be found in stainless steel weldments. The second, located within the stainless steel base metal, apparently originated at microcracks formed during the welding of a copper-stainless steel dissimilar metal joint. The latter microcrack then propagated during the fire-test, ultimately penetrating the outer shall of the cask. 18 figures, 2 tables

Simulation of heavy metal contamination of fresh water bodies: toxic ...

African Journals Online (AJOL)

Michael Horsfall

www.bioline.org.br/ja. Simulation of heavy metal contamination of fresh water bodies: toxic effects in the ... 96 hours (though sampling was done at the 48th hour). Biochemical markers of ... silver, while enhancing the bioavailability of mercury in Ceriodaphnia ..... Biochemical and molecular disorders of bilirubin metabolism.

Optical properties of metallic nanoparticles basic principles and simulation

CERN Document Server

Trügler, Andreas

2016-01-01

This book introduces the fascinating world of plasmonics and physics at the nanoscale, with a focus on simulations and the theoretical aspects of optics and nanotechnology. A research field with numerous applications, plasmonics bridges the gap between the micrometer length scale of light and the secrets of the nanoworld. This is achieved by binding light to charge density oscillations of metallic nanostructures, so-called surface plasmons, which allow electromagnetic radiation to be focussed down to spots as small as a few nanometers. The book is a snapshot of recent and ongoing research and at the same time outlines our present understanding of the optical properties of metallic nanoparticles, ranging from the tunability of plasmonic resonances to the ultrafast dynamics of light-matter interaction. Beginning with a gentle introduction that highlights the basics of plasmonic interactions and plasmon imaging, the author then presents a suitable theoretical framework for the description of metallic nanostructu...

Simulated Tip Rub Testing of Low-Density Metal Foam

Science.gov (United States)

Bowman, Cheryl L.; Jones, Michael G.

2009-01-01

Preliminary acoustic studies have indicated that low-density, open-cell, metal foams may be suitable acoustic liner material for noise suppression in high by-pass engines. Metal foam response under simulated tip rub conditions was studied to assess whether its durability would be sufficient for the foam to serve both as a rub strip above the rotor as well as an acoustic treatment. Samples represented four metal alloys, nominal cell dimensions ranging from 60 to 120 cells per inch (cpi), and relative densities ranging from 3.4 to 10 percent. The resulting rubbed surfaces were relatively smooth and the open cell structure of the foam was not adversely affected. Sample relative density appeared to have significant influence on the forces induced by the rub event. Acoustic responses of various surface preparations were measured using a normal incidence tube. The results of this study indicate that the foam s open-cell structure was retained after rubbing and that the acoustic absorption spectra variation was minimal.

Method of processing spent ion exchange resins

International Nuclear Information System (INIS)

Mori, Kazuhide; Tamada, Shin; Kikuchi, Makoto; Matsuda, Masami; Aoyama, Yoshiyuki.

1985-01-01

Purpose: To decrease the amount of radioactive spent ion exchange resins generated from nuclear power plants, etc and process them into stable inorganic compounds through heat decomposition. Method: Spent ion exchange resins are heat-decomposed in an inert atmosphere to selectively decompose only ion exchange groups in the preceeding step while high molecular skeltons are completely heat-decomposed in an oxidizing atmosphere in the succeeding step. In this way, gaseous sulfur oxides and nitrogen oxides are generated in the preceeding step, while gaseous carbon dioxide and hydrogen requiring no discharge gas procession are generated in the succeeding step. Accordingly, the amount of discharged gases requiring procession can significantly be reduced, as well as the residues can be converted into stable inorganic compounds. Further, if transition metals are ionically adsorbed as the catalyst to the ion exchange resins, the ion exchange groups are decomposed at 130 - 300 0 C, while the high molecular skeltons are thermally decomposed at 240 - 300 0 C. Thus, the temperature for the heat decomposition can be lowered to prevent the degradation of the reactor materials. (Kawakami, Y.)

Computational simulation studies of the reduction process of UF4 to metallic uranium

International Nuclear Information System (INIS)

Borges, Wesden de Almeida

2011-01-01

The production of metallic uranium is essential for production of fuel elements for using in nuclear reactors manufacturing of radioisotopes and radiopharmaceuticals. In IPEN, metallic uranium is produced by magnesiothermical reduction of UF 4 . This reaction is performed in a closed graphite crucible inserted in a sealed metal reactor and no contact with the outside environment. The set is gradually heated in an oven pit, until it reaches the ignition temperature of the reaction (between 600-650 degree C). The modeling of the heating profile of the system can be made using simulation programs by finite element method. Through the thermal profiles in the load, we can have a notion of heating period required for the reaction to occur, allowing the identification of the same group in a greater or smaller yield in metallic uranium production. Thermal properties of UF 4 are estimated, obtaining thermal conductivity and heat capacity using the Flash Laser Method, and for the load UF 4 + Mg, either. The results are compared to laboratory tests to simulate the primary production process. (author)

A study on the safety of spent fuel management. Radioactive source term modelling

Energy Technology Data Exchange (ETDEWEB)

Chun, Kwan Sik; Lee, Hoo Keun; Park, Keun Il; Hwoang, Jung Ki; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1992-02-01

The types and probabilities of events which may occur during the process of reception, transfer and storage of spent fuels in an away-from-reactor (AFR) spent fuel storage facility were analyzed in order to calculate the amount of radioactive material released to operation area and atmosphere, and the basic model for predicting the radioactive source-term under normal and abnormal operations were developed. Also, oxidation and dissolution of U0{sub 2} pellet was investigated to estimate the amount of radioactive materials released from spent fuel and the release characteristics of radionuclides from defected spent fuel rods was analyzed. Basic information using FIRAC code to analyze the ventilation system during fire accident was prepared and FIRIN was detached from FIRAC modified to simulate the compartment fire by personal computer. (Author).

Redesign of the spent fuel storage racks at the Trojan Nuclear Plant

International Nuclear Information System (INIS)

Stump, K.

1987-01-01

The spent fuel pool (SFP) at the Trojan Nuclear Plant located near Prescott, Oregon, was originally designed to hold 1.33 cores worth of spent fuel assemblies. Due to the delay in the site selection and preparation process for the spent fuel repository, the SFP storage capacity was increased in 1978 from 260 assemblies to 651 assemblies and in 1983 was increased again from 651 to 1408 assemblies to allow Trojan to continue operations through the year 2003 with a full core reserve in the SFP. Now it appears unlikely that a high level waste repository will be in operation before 2010. This indicates that a further capacity increase in the SFP is required to allow commercial operation until 2010, at which time the repository should be open to receive spent fuel. To accomplish this, an increase of seven times the original SFP capacity of 260 assemblies is needed. This paper presents a spent fuel assembly rack design that enables the required capacity increase in the SFP to be met. By the use of a boron carbide - silicon polymer inside a titanium/vanadium honeycomb as a neutron absorber between the fuel assemblies and by increasing the metal to water ratio of the spent fuel pool to harden the neutron energy spectrum the capacity of the SFP is increased to 1880 assemblies for an increase of 7.23 times the original spent fuel pool capacity. The multiplication factor for the pool with every fuel assembly slot filled in the new rack system is 0.62; well below the NRC regulatory limit of keff < 0.95. The capacity increase with allow the commercial operation of the Trojan Nuclear Plant through 2010 with a full core reserve in the spent fuel pool

Numerical simulation on single bubble rising behavior in liquid metal using moving particle semi-implicit method

International Nuclear Information System (INIS)

Zuo Juanli; Tian Wenxi; Qiu Suizheng; Chen Ronghua; Su Guanghui

2011-01-01

The gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovational conceptual design to enhance the natural circulation ability of reactor core. The two-phase flow character of gas-liquid metal makes significant improvement of the natural circulation capacity and reactor safety. In present basic study, the rising behavior of a single nitrogen bubble in five kinds of liquid metals (lead bismuth alloy, liquid kalium, sodium, potassium sodium alloy and lithium lead alloy) was numerically simulated using moving particle semi-implicit (MPS) method. The whole growing process of single nitrogen bubble in liquid metal was captured. The bubble shape and rising speed of single nitrogen bubble in each liquid metal were compared. The comparison between simulation results using MPS method and Grace graphical correlation shows a good agreement. (authors)

Leaching of metals from end-of-life solar cells.

Science.gov (United States)

Chakankar, Mital; Su, Chun Hui; Hocheng, Hong

2018-04-10

The issue of recycling waste solar cells is critical with regard to the expanded use of these cells, which increases waste production. Technology establishment for this recycling process is essential with respect to the valuable and hazardous metals present therein. In the present study, the leaching potentials of Acidithiobacillus thiooxidans, Acidithiobacillus ferrooxidans, Penicillium chrysogenum, and Penicillium simplicissimum were assessed for the recovery of metals from spent solar cells, with a focus on retrieval of the valuable metal Te. Batch experiments were performed to explore and compare the metal removal efficiencies of the aforementioned microorganisms using spent media. P. chrysogenum spent medium was found to be most effective, recovering 100% of B, Mg, Si, V, Ni, Zn, and Sr along with 93% of Te at 30 °C, 150 rpm and 1% (w/v) pulp density. Further optimization of the process parameters increased the leaching efficiency, and 100% of Te was recovered at the optimum conditions of 20 °C, 200 rpm shaking speed and 1% (w/v) pulp density. In addition, the recovery of aluminum increased from 31 to 89% upon process optimization. Thus, the process has considerable potential for metal recovery and is environmentally beneficial.

Risk assessment in spent fuel storage and transportation

International Nuclear Information System (INIS)

Pandimani, S.

1989-01-01

Risk assessment in various stages of nuclear fuel cycle is still an active area of Nuclear safety studies. From the results of risk assessment available in literature, it can be determined that the risk resulting from shipments of plutonium and spent-fuel are much greater than that resulting from the transport of other materials within the nuclear fuel cycle. In India spent fuels are kept in Spent Fuel Storage Pool (SFSP) for about 240-400 days, which is relatively a longer period compared to the usual 120 days as recommended by regulatory authorities. After cooling spent fuels are transported to the reprocessing sites which are mostly situated close to the plants. India has two high level waste treatment facilities, one PREFRE (Plutonium Reprocessing and Fuel Recycling) at Tarapur and the other one, a unit of Nuclear Fuel Complex at Hyderabad. This paper presents the risk associated with spent fuel storage and transportation for the Indian conditions. All calculations are based on a typical CANDU reactor system. Simple fault tree models are evolved for SFSP and for Transportation Accident Mode (TAM) for both road and rail. Fault tree quantification and risk assessment are done to each of these models. All necessary data for SFSP are taken mostly from Reactor Safety Study, (1975). Similarly, the data for rail TAM are taken from Annual Statistical Statements, (1987-8) and that for road TAM from Special Issue on Motor Vehicle Accident Statistics in India, (1986). Simulation method is used wherever necessary. Risk is also estimated for normal/accident free transport

Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

International Nuclear Information System (INIS)

Cooper, T.D.; Pajunen, A.L.

1998-01-01

The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

Molecular Dynamics Simulations of displacement cascades in metallic systems

International Nuclear Information System (INIS)

Doan, N.V.; Tietze, H.

1995-01-01

We use Molecular Dynamics Computer Simulations to investigate defect production induced by energetic displacement cascades up to 10 keV in pure metals (Cu, Ni) and in ordered intermetallic alloys NiAl, Ni 3 Al. Various model potentials were employed to describe the many-body nature of the interactions: the RGL (Rosato-Guillope-Legrand) model was used in pure Cu and Ni simulations; the modified version of the Vitek, Ackland and Cserti potentials (due to Gao, Bacon and Ackland) in Ni 3 Al and the EAM potentials of Foiles and Daw modified by Rubini and Ballone in NiAl, Ni 3 Al were used in alloy simulations. Atomic mixing and disordering were studied into details owing to imaging techniques and determined at different phases of the cascades. Some mixing mechanisms were identified. Our results were compared with existing data and those obtained by similar Molecular Dynamics Simulations available in the literature. (orig.)

Conceptual study of dry storage method for spent fuel assemblies based on honeycomb concrete overpack (COP). Phase 1

International Nuclear Information System (INIS)

Hida, Yoshio; Hayashi, Shigeki; Katsuyama, Yoshiaki; Hashimoto, Hirohide; Murata, Takashi

2017-01-01

The amount of spent fuel assemblies currently stored in Japan is approximately 15,000 tU. Most of these are stored in storage pools, although dry storage method will be safer, as was revealed in the accident of the Fukushima Daiichi Nuclear Power Plant. In addition, Japan has established a national policy of the nuclear fuel cycle. All spent fuel assemblies are designated for reprocessing. However, the reprocessing plant in Japan is currently under regulatory review for compliance with newly established safety standards. Beyond this, shortfalls in its processing capacity mean interim storage facilities for spent fuel are required. The Tokyo Electric Power Company Holdings, Incorporated and the Japan Atomic Power Company are currently building an interim dry storage facility with a storage capacity of 5,000 tU in Aomori Prefecture, while Chubu Electric Power Company, Inc. is currently building a dry storage facility with a storage capacity of 400 tU in the Hamaoka Nuclear Power Station. These facilities consist of earthquake-resistant buildings and dry storage casks. Within the buildings, metal transportable storage casks loaded with spent fuel assemblies are placed vertically with spaces between the casks and supported by earthquake-proof measures that prevent toppling or other movement. These structures entail significant cost and construction efforts. At the Fukushima Daiichi Nuclear Power Plant, a temporary dry storage facility has been built within the premises to store spent fuel generated during decommissioning. Part of this facility is already in operation. Here, each metal cask containing spent fuel is mounted on an earthquake-resistant concrete mat, which is anchored to the ground. Each cask is enclosed in a concrete box for additional radiation shielding, and the casks are spaced at intervals. This approach requires a large plot of land. The dry storage method for spent fuel presented here does not require a building. The dry metal casks containing spent

[Using sequential indicator simulation method to define risk areas of soil heavy metals in farmland.

Science.gov (United States)

Yang, Hao; Song, Ying Qiang; Hu, Yue Ming; Chen, Fei Xiang; Zhang, Rui

2018-05-01

The heavy metals in soil have serious impacts on safety, ecological environment and human health due to their toxicity and accumulation. It is necessary to efficiently identify the risk area of heavy metals in farmland soil, which is of important significance for environment protection, pollution warning and farmland risk control. We collected 204 samples and analyzed the contents of seven kinds of heavy metals (Cu, Zn, Pb, Cd, Cr, As, Hg) in Zengcheng District of Guangzhou, China. In order to overcame the problems of the data, including the limitation of abnormal values and skewness distribution and the smooth effect with the traditional kriging methods, we used sequential indicator simulation method (SISIM) to define the spatial distribution of heavy metals, and combined Hakanson index method to identify potential ecological risk area of heavy metals in farmland. The results showed that: (1) Based on the similar accuracy of spatial prediction of soil heavy metals, the SISIM had a better expression of detail rebuild than ordinary kriging in small scale area. Compared to indicator kriging, the SISIM had less error rate (4.9%-17.1%) in uncertainty evaluation of heavy-metal risk identification. The SISIM had less smooth effect and was more applicable to simulate the spatial uncertainty assessment of soil heavy metals and risk identification. (2) There was no pollution in Zengcheng's farmland. Moderate potential ecological risk was found in the southern part of study area due to enterprise production, human activities, and river sediments. This study combined the sequential indicator simulation with Hakanson risk index method, and effectively overcame the outlier information loss and smooth effect of traditional kriging method. It provided a new way to identify the soil heavy metal risk area of farmland in uneven sampling.

Immobilization of radioactive waste sludge from spent fuel storage pool

International Nuclear Information System (INIS)

Pavlovic, R.; Plecas, I.

1998-01-01

In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

A burner for the combustion of spent tall oil soap

Energy Technology Data Exchange (ETDEWEB)

Hughes, P.M.; Wong, J.K.; Moffatt, B.; Belanger, G. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Energy Technology Centre; Soriano, D. [Brais Malouin and Associates, Montreal, PQ (Canada)

2003-07-01

Efficiency in industrial processes applies both to the form of energy involved and the many by-products resulting from the process. Tall oil soap (TOS) is a white frothy substance created during the pulping process. It contains chemicals that can be extracted for use in other industries. The processing of TOS results in a product called spent TOS. This study examined the incineration process to derive process heat from the calorific value in spent TOS. Brais Malouin and Associates (BMA) proposed that an atomizing nozzle should be used for use with this liquid in an incinerating burner. The efficiency of atomization of spent TOS with the BMA nozzle was determined by the Canada Centre for Mineral and Energy Technology (CANMET), which also characterized the combustion in a simulated boiler situation. The combustion tests were performed in the Pilot-Scale Research Boiler at the CANMET Energy Technology Centre (CETC). Pre-heating was done with a number 2 oil flame. Flame stability was determined by observing the flame through sight ports and by measuring the gas in the furnace. The experiments showed that spent TOS could successfully burn with a number 2 oil, in a proportion of 81 spent TOS to 19 oil mass ratio. As the amount of spent TOS was increased, the amount of sulphur dioxide, nitrogen oxide (NOx) and carbon monoxide decreased. The number 2 fuel oil was responsible for the sulphur dioxide in the exhaust. It is believed that the reduction in the carbon monoxide in the exhaust is attributable to the water-gas shift reaction. As the proportion of spent TOS increased, it was shown that the amount of NOx in the exhaust decreased rapidly. A bluish-green molten deposit formed in the furnace near the burner came from copper and manganese found in the ash of the spent TOS. 7 refs., 7 tabs., 16 figs.

An independent spent-fuel storage installation at Surry Station: Design and operation

International Nuclear Information System (INIS)

McKay, H.S.; Wakeman, B.H.; Pickworth, J.M.; Routh, S.D.; Hopkins, W.C.

1989-07-01

Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation (ISFSI) was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry ISFSI will accommodate 84 such casks with a total storage capacity of 811 MTU of spent PWR fuel assemblies. The ISFSI is located at the Surry Station in a wooded area approximately 1000 meters (3300 feet) east of the reactor facilities. Construction of the first of three reinforced concrete storage pads and its associated support systems was completed in March 1986. The operating license and Technical Specifications were issued by the US NRC on July 2, 1986. Initial loading operations of a General Nuclear Systems, Inc., CASTOR V/21 storage cask began in September 1986. The first two CASTOR V/21 casks were placed in storage at the ISFSI in December 1986. 16 refs., 33 figs., 16 tabs

Leaching characteristics of the metal waste form from the electrometallurgical treatment process: Product consistency testing

International Nuclear Information System (INIS)

Johnson, S. G.; Keiser, D. D.; Frank, S. M.; DiSanto, T.; Noy, M.

1999-01-01

Argonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Nb, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/1--4 wt% noble metal fission products/1--2 wt % U. Leaching results are presented from several tests and sample types: (1) 2 week monolithic immersion tests on actual metal waste forms produced from irradiated cladding hulls, (2) long term (>2 years) pulsed flow tests on samples containing technetium and uranium and (3) crushed sample immersion tests on cold simulated metal waste form samples. The test results will be compared and their relevance for waste form product consistency testing discussed

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Safety analysis of spent fuel transport and storage casks under extreme impact conditions

International Nuclear Information System (INIS)

Wolff, D.; Wieser, G.; Ballheimer, V.; Voelzke, H.; Droste, B.

2005-01-01

following impact of building structures of a nuclear storage facility. In this context we present methods to calculate the deformation of cask components and relative displacements between the metallic seals and their counterparts with respect to the preservation of the leak tightness. In this paper we also discuss examples of potential explosion impacts from dangerous goods on CASTOR spent fuel casks during transportation: experimental investigation of an LPG rail tank explosion next to a cask and numerical simulation of a detonation blast wave due to 21 Mg explosives in a distance of 25 m to a cask. The presented examples demonstrate the wide safety margins of the investigated monolithic cast iron transport and storage casks. (author)

Numerical and experimental validation of a particle Galerkin method for metal grinding simulation

Science.gov (United States)

Wu, C. T.; Bui, Tinh Quoc; Wu, Youcai; Luo, Tzui-Liang; Wang, Morris; Liao, Chien-Chih; Chen, Pei-Yin; Lai, Yu-Sheng

2018-03-01

In this paper, a numerical approach with an experimental validation is introduced for modelling high-speed metal grinding processes in 6061-T6 aluminum alloys. The derivation of the present numerical method starts with an establishment of a stabilized particle Galerkin approximation. A non-residual penalty term from strain smoothing is introduced as a means of stabilizing the particle Galerkin method. Additionally, second-order strain gradients are introduced to the penalized functional for the regularization of damage-induced strain localization problem. To handle the severe deformation in metal grinding simulation, an adaptive anisotropic Lagrangian kernel is employed. Finally, the formulation incorporates a bond-based failure criterion to bypass the prospective spurious damage growth issues in material failure and cutting debris simulation. A three-dimensional metal grinding problem is analyzed and compared with the experimental results to demonstrate the effectiveness and accuracy of the proposed numerical approach.

Mathematical simulation of the behaviour of the spent organic extractive solution near the injection well area in the case of underground disposal

International Nuclear Information System (INIS)

Istomin, A.D.; Noskov, M.D.; Balakhonov, V.G.; Zubkov, A.A.; Egorov, G.F.

2005-01-01

A mathematical model is presented of the processes in the collector seam under combined disposal of organic and radioactive wastes in porous geological strata of deep bedding. The model describes filtration, mass transfer, sorption and desorption of radionuclides, radioactive decay, decomposition of organic components and heat transfer. The computer software is developed. The results of simulating the thermal field dynamics, behaviour of the components of the spent organic extractive solution and water radioactive wastes in the collector seam of deep bedding are presented [ru

Selection of dissolution process for spent fuels and preparation of corrosion test solution simulated to dissolver (contract research)

International Nuclear Information System (INIS)

Motooka, Takafumi; Terakado, Shogo; Koya, Toshio; Hamada, Shozo; Kiuchi, Kiyoshi

2001-03-01

In order to evaluate the reliability of reprocessing equipment materials used in the Rokkasho Reprocessing Plant, we have proceeded a mock-up test and laboratory tests for getting corrosion parameters. In a dissolver made of zirconium, the simulation of test solutions to the practical solution which includes the high concentration of radioactive elements such as FP and TRU is one of the important issues with respect to the life prediction. On this experiment, the dissolution process of spent fuels and the preparation of test solution for evaluating the corrosion resistance of dissolver materials were selected. These processes were tested in the No.3 cell of WASTEF. The test solution for corrosion tests was prepared by adjusting the uranium and nitric acid concentrations. (author)

Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

Energy Technology Data Exchange (ETDEWEB)

Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

2016-08-15

Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

International Nuclear Information System (INIS)

L. Angers

2001-01-01

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR

Development of an expert system for the simulation model for casting metal substructure of a metal-ceramic crown design.

Science.gov (United States)

Matin, Ivan; Hadzistevic, Miodrag; Vukelic, Djordje; Potran, Michal; Brajlih, Tomaz

2017-07-01

Nowadays, the integrated CAD/CAE systems are favored solutions for the design of simulation models for casting metal substructures of metal-ceramic crowns. The worldwide authors have used different approaches to solve the problems using an expert system. Despite substantial research progress in the design of experts systems for the simulation model design and manufacturing have insufficiently considered the specifics of casting in dentistry, especially the need for further CAD, RE, CAE for the estimation of casting parameters and the control of the casting machine. The novel expert system performs the following: CAD modeling of the simulation model for casting, fast modeling of gate design, CAD eligibility and cast ability check of the model, estimation and running of the program code for the casting machine, as well as manufacturing time reduction of the metal substructure. The authors propose an integration method using common data model approach, blackboard architecture, rule-based reasoning and iterative redesign method. Arithmetic mean roughness values was determinated with constant Gauss low-pass filter (cut-off length of 2.5mm) according to ISO 4287 using Mahr MARSURF PS1. Dimensional deviation between the designed model and manufactured cast was determined using the coordinate measuring machine Zeiss Contura G2 and GOM Inspect software. The ES allows for obtaining the castings derived roughness grade number N7. The dimensional deviation between the simulation model of the metal substructure and the manufactured cast is 0.018mm. The arithmetic mean roughness values measured on the casting substructure are from 1.935µm to 2.778µm. The realized developed expert system with the integrated database is fully applicable for the observed hardware and software. Values of the arithmetic mean roughness and dimensional deviation indicate that casting substructures are surface quality, which is more than enough and useful for direct porcelain veneering. The

Method of dissolving metal ruthenium

International Nuclear Information System (INIS)

Tsuno, Masao; Soda, Yasuhiko; Kuroda, Sadaomi; Koga, Tadaaki.

1988-01-01

Purpose: To dissolve and clean metal ruthenium deposited to the inner surface of a dissolving vessel for spent fuel rods. Method: Metal ruthenium is dissolved in a solution of an alkali metal hydroxide to which potassium permanganate is added. As the alkali metal hydroxide used herein there can be mentioned potassium hydroxide, sodium hydroxide and lithium hydroxide can be mentioned, which is used as an aqueous solution from 5 to 20 % concentration in view of the solubility of metal ruthenium and economical merit. Further, potassium permanganate is used by adding to the solution of alkali metal hydroxide at a concentration of 1 to 5 %. (Yoshihara, H.)

Baseline metal enrichment from Population III star formation in cosmological volume simulations

Science.gov (United States)

Jaacks, Jason; Thompson, Robert; Finkelstein, Steven L.; Bromm, Volker

2018-04-01

We utilize the hydrodynamic and N-body code GIZMO coupled with our newly developed sub-grid Population III (Pop III) Legacy model, designed specifically for cosmological volume simulations, to study the baseline metal enrichment from Pop III star formation at z > 7. In this idealized numerical experiment, we only consider Pop III star formation. We find that our model Pop III star formation rate density (SFRD), which peaks at ˜ 10- 3 M⊙ yr- 1 Mpc- 1 near z ˜ 10, agrees well with previous numerical studies and is consistent with the observed estimates for Pop II SFRDs. The mean Pop III metallicity rises smoothly from z = 25 to 7, but does not reach the critical metallicity value, Zcrit = 10-4 Z⊙, required for the Pop III to Pop II transition in star formation mode until z ≃ 7. This suggests that, while individual haloes can suppress in situ Pop III star formation, the external enrichment is insufficient to globally terminate Pop III star formation. The maximum enrichment from Pop III star formation in star-forming dark matter haloes is Z ˜ 10-2 Z⊙, whereas the minimum found in externally enriched haloes is Z ≳ 10-7 Z⊙. Finally, mock observations of our simulated IGM enriched with Pop III metals produce equivalent widths similar to observations of an extremely metal-poor damped Lyman alpha system at z = 7.04, which is thought to be enriched by Pop III star formation only.

Development of JSTAMP-Works/NV and HYSTAMP for Multipurpose Multistage Sheet Metal Forming Simulation

Science.gov (United States)

Umezu, Yasuyoshi; Watanabe, Yuko; Ma, Ninshu

2005-08-01

Since 1996, Japan Research Institute Limited (JRI) has been providing a sheet metal forming simulation system called JSTAMP-Works packaged the FEM solvers of LS-DYNA and JOH/NIKE, which might be the first multistage system at that time and has been enjoying good reputation among users in Japan. To match the recent needs, "faster, more accurate and easier", of process designers and CAE engineers, a new metal forming simulation system JSTAMP-Works/NV is developed. The JSTAMP-Works/NV packaged the automatic healing function of CAD and had much more new capabilities such as prediction of 3D trimming lines for flanging or hemming, remote control of solver execution for multi-stage forming processes and shape evaluation between FEM and CAD. On the other way, a multi-stage multi-purpose inverse FEM solver HYSTAMP is developed and will be soon put into market, which is approved to be very fast, quite accurate and robust. Lastly, authors will give some application examples of user defined ductile damage subroutine in LS-DYNA for the estimation of material failure and springback in metal forming simulation.

Development of JSTAMP-Works/NV and HYSTAMP for Multipurpose Multistage Sheet Metal Forming Simulation

International Nuclear Information System (INIS)

Umezu, Yasuyoshi; Watanabe, Yuko; Ma, Ninshu

2005-01-01

Since 1996, Japan Research Institute Limited (JRI) has been providing a sheet metal forming simulation system called JSTAMP-Works packaged the FEM solvers of LS-DYNA and JOH/NIKE, which might be the first multistage system at that time and has been enjoying good reputation among users in Japan. To match the recent needs, 'faster, more accurate and easier', of process designers and CAE engineers, a new metal forming simulation system JSTAMP-Works/NV is developed. The JSTAMP-Works/NV packaged the automatic healing function of CAD and had much more new capabilities such as prediction of 3D trimming lines for flanging or hemming, remote control of solver execution for multi-stage forming processes and shape evaluation between FEM and CAD.On the other way, a multi-stage multi-purpose inverse FEM solver HYSTAMP is developed and will be soon put into market, which is approved to be very fast, quite accurate and robust.Lastly, authors will give some application examples of user defined ductile damage subroutine in LS-DYNA for the estimation of material failure and springback in metal forming simulation

Guidebook on spent fuel storage

International Nuclear Information System (INIS)

1984-01-01

The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

Pyrochemical recovery of easily reducible species from spent nuclear fuel

International Nuclear Information System (INIS)

Jouault, C.

2000-01-01

The purpose of the reprocessing of spent fuel is to separate noble metals and other easily reducible species, actinides and lanthanides. A thermodynamic and bibliographical study allowed us to elaborate a process which realises these separations in several steps. The experimental validation of the steps concerning the extraction of noble metals and easily reducible species required to imagine an apparatus which is conformed to the study of the two steps in question: the reduction by a gas of fission product oxides and the extraction of the metallic particles, obtained by reduction, by digestion in a liquid metal. Experiments on digestion, carried on molybdenum and ruthenium particles, allowed us to conclude that the transfer of metallic particles from a molten salt into a liquid metal is ruled by phenomena of complex wettability between the metallic particle, the molten salt, the liquid metal and the gas. The transfer from the salt to the metal is a chain of two steps: emersion of the particles from the salt to go into the gas, and then transfer from the gas into the metal. Kinetics are limited by the transfer through the metal surface. Kinetics study withdrew the experimental parameters and the metals properties which influence the digestion rate. A model on the transfer into a liquid metal of a particle trapped at the fluid/metal interface ratified the experimental conclusions and informed on the stirring influence. All the results allow us to think that the extraction of noble metals and easily reducible species are feasible in this way. (author) [fr

Ageing of metallic gaskets for spent fuel casks: Century-long life forecast from 25,000-h-long experiments

International Nuclear Information System (INIS)

Sassoulas, H.; Morice, L.; Caplain, P.; Rouaud, C.; Mirabel, L.; Beal, F.

2006-01-01

An experimental programme is being carried out that aims at quantifying the relaxation of four types of metallic HELICOFLEX[reg] seals during their use in spent nuclear fuel storage casks. Two types of lining are taken into account: aluminium and silver. Tests longer than 10,000 h are implemented only for silver. For each type of lining, two different section diameters are investigated. The work aims at evaluating the minimum residual linear load that can be guaranteed for a seal after a particular time of relaxation. This relaxation depends on the evolution of the seal temperature with time. Therefore, holds of seals tightened between two flanges have been performed at several constant temperatures, including 100 and 200 deg. C. Residual load and 'useful' recovery have been measured after the holds. Results are interpreted according to two methods: a time extrapolation, and a time-temperature equivalence parameter. Both methods are based on linear relationships and are assessed through a statistical analysis (calculation of scatter) which is also used to determine a minimum guaranteed residual load. Finite element simulations of the relaxation of a seal have also been performed in order to justify qualitatively that the time extrapolation method is safe. For silver lining seals, the use of a time-temperature equivalence parameter equal to T (11 + log 1 (t)) appears justified and this enables us to assess the maximum temperature at which seals can be 'safely' used 'up to a century'. Using the available ageing results (longest holds: 25,000 h), and the proposed prediction method, it can be proven that the two types of silver lining seals which are evaluated will retain a residual linear load of at least 100 N mm -1 of seal perimeter after one century of use in a cask, if the initial temperature of the seal after closing the cask is less than or equal to 100 deg. C

Spent fuels program

International Nuclear Information System (INIS)

Shappert, L.B.

1983-01-01

The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

Project management for the Virginia power spent fuel storage project

International Nuclear Information System (INIS)

Smith, M.

1992-01-01

Like Duke Power, Virginia Power has been involved in spent fuel storage expansion studies for a long time - possibly a little longer than Duke Power. Virginia Power's initial studies date back to the late 70s and into the early 80s. Large variety of storage techniques are reviewed including reracking and transshipment. Virginia Power also considered construction a new spent fuel pool. This was one of the options that was considered early on since Virginia Power started this process before any dry storage techniques had been proven. Consolidation of spent fuel is something that was also studied. Finally, construction of dry storage facility was determined to be the technology of choice. They looked a large variety of dry storage technologies and eventually selected dry storage in metal casks at Surry. There are many of reasons why a utility may choose one technology over another. In Virginia Power's situation, additional storage was needed at Surry much earlier than at other utilities. Virginia Power was confronted with selecting a storage technique and having to be a leader in that it was the first U.S. utility to implement a dry storage system

Numerical Simulation of Multiphase Magnetohydrodynamic Flow and Deformation of Electrolyte-Metal Interface in Aluminum Electrolysis Cells

Science.gov (United States)

Hua, Jinsong; Rudshaug, Magne; Droste, Christian; Jorgensen, Robert; Giskeodegard, Nils-Haavard

2018-06-01

A computational fluid dynamics based multiphase magnetohydrodynamic (MHD) flow model for simulating the melt flow and bath-metal interface deformation in realistic aluminum reduction cells is presented. The model accounts for the complex physics of the MHD problem in aluminum reduction cells by coupling two immiscible fluids, electromagnetic field, Lorentz force, flow turbulence, and complex cell geometry with large length scale. Especially, the deformation of bath-metal interface is tracked directly in the simulation, and the condition of constant anode-cathode distance (ACD) is maintained by moving anode bottom dynamically with the deforming bath-metal interface. The metal pad deformation and melt flow predicted by the current model are compared to the predictions using a simplified model where the bath-metal interface is assumed flat. The effects of the induced electric current due to fluid flow and the magnetic field due to the interior cell current on the metal pad deformation and melt flow are investigated. The presented model extends the conventional simplified box model by including detailed cell geometry such as the ledge profile and all channels (side, central, and cross-channels). The simulations show the model sensitivity to different side ledge profiles and the cross-channel width by comparing the predicted melt flow and metal pad heaving. In addition, the model dependencies upon the reduction cell operation conditions such as ACD, current distribution on cathode surface and open/closed channel top, are discussed.

Release of U(VI) from spent biosorbent immobilized in cement concrete blocks

Energy Technology Data Exchange (ETDEWEB)

Venkobachar, C.; Iyengar, L.; Mishra, U.K.; Chauhan, M.S. [Indian Inst. of Tech., Kanpur (India)

1995-12-01

This paper deals with cementation as the method for the disposal of spent biosorbent, Ganoderma lucidum (a wood rotting macrofungi) after it is used for the removal of Uranium. Results on the uranium release during the curing of cement-concrete (CC) blocks indicated that placing the spent sorbent at the center of the blocks during their casting yields better immobilization of uranium as compared to the homogeneous mixing of the spent sorbent with the cement. Short term leach tests indicated that the uranium release was negligible in simulated seawater, 1.8% in 0.2 N sodium carbonate and 6.0% in 0.2 N HCl. The latter two leachates were used to represent the extreme environmental conditions. It was observed that the presence of the spent biosorbent up to 5% by weight did not affect the compressive strength of CC blocks. Thus cementation technique is suitable for the immobilization of uranium loaded biosorbent for its ultimate disposal.

Release of U(VI) from spent biosorbent immobilized in cement concrete blocks

International Nuclear Information System (INIS)

Venkobachar, C.; Iyengar, L.; Mishra, U.K.; Chauhan, M.S.

1995-01-01

This paper deals with cementation as the method for the disposal of spent biosorbent, Ganoderma lucidum (a wood rotting macrofungi) after it is used for the removal of Uranium. Results on the uranium release during the curing of cement-concrete (CC) blocks indicated that placing the spent sorbent at the center of the blocks during their casting yields better immobilization of uranium as compared to the homogeneous mixing of the spent sorbent with the cement. Short term leach tests indicated that the uranium release was negligible in simulated seawater, 1.8% in 0.2 N sodium carbonate and 6.0% in 0.2 N HCl. The latter two leachates were used to represent the extreme environmental conditions. It was observed that the presence of the spent biosorbent up to 5% by weight did not affect the compressive strength of CC blocks. Thus cementation technique is suitable for the immobilization of uranium loaded biosorbent for its ultimate disposal

Development of ultrasonic immersion inspection technique for spent fuel canisters

International Nuclear Information System (INIS)

Schankula, J.J.

1982-07-01

This report summarizes ultrasonic nondestructive testing development for metal matrix supported spent fuel disposal canisters. The work has concentated in two areas: inspection for lack of bond at the shell/matrix interface and inspection for voids in the matrix. The capabilities and limitations of these techniques have been fully established. Unbonded areas as small as 4 mm in diameter and voids 6 mm in diameter, 25 mm deep in the matrix, can readily be detected

Hydrometallurgical method for recycling rare earth metals, cobalt, nickel, iron, and manganese from negative electrodes of spent Ni-MH mobile phone batteries

International Nuclear Information System (INIS)

Santos, Vinicius Emmanuel de Oliveira dos; Lelis, Maria de Fatima Fontes; Freitas, Marcos Benedito Jose Geraldo de

2014-01-01

A hydrometallurgical method for the recovery of rare earth metals, cobalt, nickel, iron, and manganese from the negative electrodes of spent Ni-MH mobile phone batteries was developed. The rare earth compounds were obtained by chemical precipitation at pH 1.5, with sodium cerium sulfate (NaCe(SO 4 ) 2 .H 2 O) and lanthanum sulfate (La 2 (SO 4 ) 3 .H 2 O) as the major recovered components. Iron was recovered as Fe(OH) 3 and FeO. Manganese was obtained as Mn 3 O 4 .The recovered Ni(OH) 2 and Co(OH) 2 were subsequently used to synthesize LiCoO 2 , LiNiO 2 and CoO, for use as cathodes in ion-Li batteries. The anodes and recycled materials were characterized by analytical techniques. (author)

Structure of Cu64.5Zr35.5 metallic glass by reverse Monte Carlo simulations

International Nuclear Information System (INIS)

Fang, X. W.; Huang, Li; Wang, C. Z.; Ho, K. M.; Ding, Z. J.

2014-01-01

Reverse Monte Carlo simulations (RMC) have been widely used to generate three dimensional (3D) atomistic models for glass systems. To examine the reliability of the method for metallic glass, we use RMC to predict the atomic configurations of a “known” structure from molecular dynamics (MD) simulations, and then compare the structure obtained from the RMC with the target structure from MD. We show that when the structure factors and partial pair correlation functions from the MD simulations are used as inputs for RMC simulations, the 3D atomistic structure of the glass obtained from the RMC gives the short- and medium-range order in good agreement with those from the target structure by the MD simulation. These results suggest that 3D atomistic structure model of the metallic glass alloys can be reasonably well reproduced by RMC method with a proper choice of input constraints

Final environmental statement: US Spent Fuel Policy. Storage of foreign spent power reactor fuel

International Nuclear Information System (INIS)

1980-05-01

In October 1977, the Department of Energy (DOE) announced a Spent Fuel Storage Policy for nuclear power reactors. Under this policy, as approved by the President, US utilities will be given the opportunity to deliver spent fuel to US Government custody in exchange for payment of a fee. The US Government will also be prepared to accept a limited amount of spent fuel from foreign sources when such action would contribute to meeting nonproliferation goals. Under the new policy, spent fuel transferred to the US Government will be delivered - at user expense - to a US Government-approved site. Foreign spent fuel would be stored in Interim Spent Fuel Storage (ISFS) facilities with domestic fuel. This volume of the environmental impact statement includes effects associated with implementing or not implementing the Spent Fuel Storage Policy for the foreign fuels. The analyses show that there are no substantial radiological health impacts whether the policy is implemented or not. In no case considered does the population dose commitment exceed 0.000006% of the world population dose commitment from natural radiation sources over the period analyzed. Full implementation of the US offer to accept a limited amount of foreign spent fuel for storage provides the greatest benefits for US nonproliferation policy. Acceptance of lesser quantities of foreign spent fuel in the US or less US support of foreign spent fuel storage abroad provides some nonproliferation benefits, but at a significantly lower level than full implementation of the offer. Not implementing the policy in regard to foreign spent fuel will be least productive in the context of US nonproliferation objectives. The remainder of the summary provides a brief description of the options that are evaluated, the facilities involved in these options, and the environmental impacts, including nonproliferation considerations, associated with each option

Spent nuclear fuel storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2005-01-01

When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

A guide for validation of FE-Simulations in bulk metal forming

International Nuclear Information System (INIS)

Tekkaya, A. Erman

2005-01-01

Numerical analysis of metal forming processes is an everyday practice in industry. Forming loads, material flow, forming defects such as underfills, laps and even cracks, stresses in dies and punches, as well as product properties like new hardness distribution, dimensional accuracies and residual stresses are predicted by numerical analysis and used for technology generation. Most of the numerical analysis is done by the finite element method made available for engineers and technicians by numerous by powerful commercial software packages. These software packages act as black-boxes and usually hide the complicated numerical procedures and even their crucial parameters from the applier. Therefore, the question arises during the industrial applications: how accurate is the simulation and how can the results can be assessed? The aim of this paper is to provide a guideline to assess the results of metal forming simulations. Although some ideas are valid for any metal forming process, bulk forming is the process concern. The paper will address firstly the possible sources of error in a finite element analysis of bulk forming processes. Then, some useful elementary knowledge will be summarized. Various levels of validation such as result and ability validation and assessment will be discussed. Finally, interpretation of results will be treated. In this content also some suggestions will be given. (author)

Current state and perspectives of spent fuel storage in Russia

International Nuclear Information System (INIS)

Kurnosov, V.A.; Tichonov, N.S.; Makarchuk, T.F.

1999-01-01

Twenty-nine power units at nine nuclear power plants, having a total installed capacity of 22 GW(e), are now in operation in the Russian Federation. They produce approximately 12% of the generated electricity in the country. The annual spent fuel arising is approximately 790 tU. The concept of the closed fuel cycle was adopted as the basis for nuclear power development in the Russian Federation, but until now this concept is only implemented for the fuel cycles of WWER-440 and BN-600 reactors. The WWER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed. It is meant to be stored in intermediate storage facilities at the NPP sites. The status of the spent fuel (SF) stored in the storage facilities is given in the paper. The principal characteristics of the fuel cycles of the Russian NPPs in the period up to 2015 is also given in the report. The key variant of the current spent fuel management at RBMK-1000 NPPs is storage in at-reactor and in away-from-reactor wet storage facilities at the power plant site with a capacity of 2,000 W. The storage capacity at the operating RBMKs (including the increase due to denser fuel assembly arrangement) will provide SF reception from the NPPs only up to 2005. For RBMK spent fuel, intermediate dry storage is foreseen at power plant sites in metallic concrete casks and thereafter transportation to the central storage facility at the RT-2 plant for long-term storage. The SF will be reprocessing after completion of the reprocessing plant at RT-2. In the Programme of Nuclear Power Development in the Russian Federation for the period 1998 to 2005 and for the period until 2010 year, provisions are made for the construction of a central dry storage facility before 2010. The facility will have a design capacity of 30,000 tU for WWER-1000 and RBMK-1000 spent fuel and is part of the reprocessing plant RT-2. The paper considers

Structural evaluation of spent nuclear fuel storage facilities under aircraft crash impact (2). Horizontal impact test onto reduced scale metal cask due to aircraft engine missile

International Nuclear Information System (INIS)

Namba, Kosuke; Shirai, Koji; Saegusa, Toshiari

2009-01-01

In this study, to confirm the sealing performance of a metal cask subjected to impact force due to possible commercial aircraft crash against a spent fuel storage facility, the horizontal impact test was carried out. In the test, an aircraft engine missile with a speed of 57.3 m/s attacked the reduced scale metal cask containing helium gas, which stands vertically. Then the leak rate and sliding displacement of the lid were measured. The leak rate increased rapidly and reached to 4.0 x 10 -6 Pa·m 3 /sec. After that, the leak rate decreased slowly and converged to 1.0x10 -6 Pa·m 3 /sec after 20 hours from the impact test. The leak rate of a full scale cask was evaluated using that of reduced scale cask obtained by the test. Then the leak rate of the full scale cask was 3.5x10 -5 Pa·m 3 /sec. This result showed that the sealing performance of the full scale metal cask would not be affected immediately by the horizontal impact of the aircraft engine with a speed of 57.3 m/s. (author)

Experimental program to determine maximum temperatures for dry storage of spent fuel

International Nuclear Information System (INIS)

Knox, C.A.; Gilbert, E.R.; White, G.D.

1985-02-01

Although air is used as a cover gas in some dry storage facilities, other facilities use inert cover gases which must be monitored to assure inertness of the atmosphere. Thus qualifying air as a cover gas is attractive for the dry storage of spent fuels. At sufficiently high temperatures, air can react with spent fuel (UO 2 ) at the site of cladding breaches that formed during reactor irradiation or during dry storage. The reaction rate is temperature dependent; hence the rates can be maintained at acceptable levels if temperatures are low. Tests with spent fuel are being conducted at Pacific Northwest Laboratory (PNL) to determine the allowable temperatures for storage of spent fuel in air. Tests performed with nonirradiated UO 2 pellets indicated that moisture, surface condition, gamma radiation, gadolinia content of the fuel pellet, and temperature are important variables. Tests were then initiated on spent fuel to develop design data under simulated dry storage conditions. Tests have been conducted at 200 and 230 0 C on spent fuel in air and 275 0 C in moist nitrogen. The results for nonirradiated UO 2 and published data for irradiated fuel indicate that above 230 0 C, oxidation rates are unacceptably high for extended storage in air. The tests with spent fuel will be continued for approximately three years to enable reliable extrapolations to be made for extended storage in air and inert gases with oxidizing constituents. 6 refs., 6 figs., 3 tabs

DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

International Nuclear Information System (INIS)

Rudisill, T; John Mickalonis, J

2006-01-01

The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Cosmic ray muons for spent nuclear fuel monitoring

Science.gov (United States)

Chatzidakis, Stylianos

There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Analysis of radiation shields of BNPP spent fuel pool

International Nuclear Information System (INIS)

Ayoobian, N.; Hadad, K.; Nematollahi, M. R.

2007-01-01

Radioactive protection is one of the most important subjects in nuclear power plants safety. Analysis of BNPP spent fuel pool shielding , as a main source of energetic γ-rays was the main goal of this project. Firstly, we simulated the reactor core using WIMSD-4 neutronic code and the amount of fission product in the fuel assembly (FA) was calculated during the reactor operation. Then, by obtaining the results from the previous calculation and by using MCNP4C nuclear code , the intensity of γ-rays was obtained in layers of spent fuel pool shields. The results have shown that no significant γ-rays passed through these shields. Finally, an accident and resulting exposure dose above the pool was analyzed

Spent fuel storage facility, Kalpakkam

International Nuclear Information System (INIS)

Shreekumar, B.; Anthony, S.

2017-01-01

Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

Energy Technology Data Exchange (ETDEWEB)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D. [KAERI, Daejeon (Korea, Republic of)

2012-01-15

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study.

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

International Nuclear Information System (INIS)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D.

2012-01-01

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study

Modeling Adsorption Kinetics (Bio-remediation of Heavy Metal Contaminated Water)

Science.gov (United States)

McCarthy, Chris

My talk will focus on modeling the kinetics of the adsorption and filtering process using differential equations, stochastic methods, and recursive functions. The models have been developed in support of our interdisciplinary lab group which is conducting research into bio-remediation of heavy metal contaminated water via filtration through biomass such as spent tea leaves. The spent tea leaves are available in large quantities as a result of the industrial production of tea beverages. The heavy metals bond with the surfaces of the tea leaves (adsorption). Funding: CUNY Collaborative Incentive Research Grant.

Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

International Nuclear Information System (INIS)

1986-04-01

This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984

Numerical simulation of systems of shear bands in ductile metal with inclusions

Science.gov (United States)

Plohr, Jeeyeon

2017-06-01

We develop a method for numerical simulations of high strain-rate loading of mesoscale samples of ductile metal with inclusions. Because of its small-scale inhomogeneity, the composite material is prone to localized shear deformation. This method employs the Generalized Method of Cells to ensure that the micro mechanical behavior of the metal and inclusions is reflected properly in the behavior of the composite at the mesoscale. To find the effective plastic strain rate when shear bands are present, we extend and apply the analytic and numerical analysis of shear bands of Glimm, Plohr, and Sharp. Our tests of the method focus on the stress/strain response in uniaxial-strain flow, both compressive and tensile, of depleted uranium metal containing silicon carbide inclusions. In results, we verify the elevated temperature and thermal softening at shear bands in our simulations of pure DU and DU/SiC composites. We also note that in composites, due the asymmetry caused by the inclusions, shear band form at different times in different subcells. In particular, in the subcells near inclusions, shear band form much earlier than they do in pure DU.

Numerical modelling of adsorption of metallic particles on graphite substrate via molecular dynamics simulation

International Nuclear Information System (INIS)

Rafii-Tabar, H.

1998-01-01

A computer-based numerical modelling of the adsorption process of gas phase metallic particles on the surface of a graphite substrate has been performed via the application of molecular dynamics simulation method. The simulation related to an extensive STM-based experiment performed in this field, and reproduces part of the experimental results. Both two-body and many-body inter-atomic potentials have been employed. A Morse-type potential describing the metal-carbon interactions at the interface was specially formulated for this modelling. Intercalation of silver in graphite has been observed as well as the correct alignments of monomers, dimers and two-dimensional islands on the surface. (author)

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

International Nuclear Information System (INIS)

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

Energy Technology Data Exchange (ETDEWEB)

You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P

2006-09-15

The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO{sub 2} into U-metal. For demonstration of this process, {alpha}-{gamma} type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for {gamma}-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

Recovery Of Electrodic Powder From Spent Lithium Ion Batteries (LIBs

Directory of Open Access Journals (Sweden)

Shin S.M.

2015-06-01

Full Text Available This study was focused on recycling process newly proposed to recover electrodic powder enriched in cobalt (Co and lithium (Li from spent lithium ion battery. In addition, this new process was designed to prevent explosion of batteries during thermal treatment under inert atmosphere. Spent lithium ion batteries (LIBs were heated over the range of 300°C to 600°C for 2 hours and each component was completely separated inside reactor after experiment. Electrodic powder was successfully recovered from bulk components containing several pieces of metals through sieving operation. The electrodic powder obtained was examined by X-ray diffraction (XRD, energy dispersive X-ray spectroscopy (EDS, and atomic absorption spectroscopy (AA and furthermore image of the powder was taken by scanning electron microscopy (SEM. It was finally found that cobalt and lithium were mainly recovered to about 49 wt.% and 4 wt.% in electrodic powder, respectively.

Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

International Nuclear Information System (INIS)

You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

2006-09-01

The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

Spent fuel workshop'2002

Energy Technology Data Exchange (ETDEWEB)

Poinssot, Ch

2002-07-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

Bioleaching of spent Zn-Mn or Ni-Cd batteries by Aspergillus species.

Science.gov (United States)

Kim, Min-Ji; Seo, Ja-Yeon; Choi, Yong-Seok; Kim, Gyu-Hyeok

2016-05-01

This research explores the recovery of metals from spent Zn-Mn or Ni-Cd batteries by a bioleaching using six Aspergillus species. Two different nutrients, malt extract and sucrose, were used to produce different types of organic acids. Oxalic acid and citric acid were shown to be the dominant organic acid in malt extract and sucrose media, respectively. In the bioleaching, the metal removal was higher in sucrose media than malt extract. All species, except A. niger KUC5254, showed more than 90% removal of metals from Zn-Mn battery. For Ni-Cd battery, more than 95% of metals was extracted by A. niger KUC5254 and A. tubingensis KUC5037. As a result, A. tubingensis KUC5037 which is a non-ochratoxigenic fungus was considered to have the greatest potential for improving the safety and efficiency of the bioleaching. Copyright © 2015 Elsevier Ltd. All rights reserved.

Method of pyrolysis for spent ion-exchange resins

International Nuclear Information System (INIS)

Aoyama, Yoshiyuki; Matsuda, Masami; Kawamura, Fumio; Yusa, Hideo.

1985-01-01

Purpose: To prevent the generation of noxious sulfur oxide and ammonia on the pyrolysis for spent ion-exchange resins discharged from nuclear power plants. Method: In the case where the pyrolysis is made for the cationic exchange resins having sulfonic acids as the ion-exchange group, alkali metals or alkaline earth metals capable of reacting with sulfonic acid groups to form solid sulfates are previously deposited by way of ion-exchange reactions prior to the pyrolysis. In another case of the anionic exchange resins having quarternary ammonium groups as the ion-exchange groups, halogenic elements capable of reacting with the ammonium groups to form solid ammonium salts are deposited to the ion-exchange resins through ion-exchange reactions prior to the pyrolysis. As a result, the amount of the binders used can be reduced, and this method can be used in a relatively simple processing facility. (Horiuchi, T.)

Atomistic simulation of nanoformed metallic glass

Energy Technology Data Exchange (ETDEWEB)

Wu, Cheng-Da, E-mail: nanowu@cycu.edu.tw

2015-07-15

Highlights: • STZ forms at substrate surface underneath punch. • Atoms underneath punch have higher speeds at larger mold displacement. • Stick-slip phenomenon becomes more obvious with increasing imprint speed. • Great pattern transfer is obtained with unloading at low temperatures. - Abstract: The effects of forming speed and temperature on the forming mechanism and mechanics of Cu{sub 50}Zr{sub 25}Ti{sub 25} metallic glass are studied using molecular dynamics simulations based on the second-moment approximation of the many-body tight-binding potential. These effects are investigated in terms of atomic trajectories, flow field, slip vectors, internal energy, radial distribution function, and elastic recovery of nanoimprint lithography (NIL) patterns. The simulation results show that a shear transformation zone (STZ) forms at the substrate surface underneath the mold during the forming process. The STZ area increases with mold displacement (D). The movement speed of substrate atoms underneath the mold increases with increasing D value. The movement directions of substrate atoms underneath the mold are more agreeable for a larger D value. The stick-slip phenomenon becomes more obvious with increasing D value and imprint speed. The substrate energy increases with increasing imprint speed and temperature. Great NIL pattern transfer is obtained with unloading at low temperatures (e.g., room temperature)

Integrated simulation of continuous-scale and discrete-scale radiative transfer in metal foams

Science.gov (United States)

Xia, Xin-Lin; Li, Yang; Sun, Chuang; Ai, Qing; Tan, He-Ping

2018-06-01

A novel integrated simulation of radiative transfer in metal foams is presented. It integrates the continuous-scale simulation with the direct discrete-scale simulation in a single computational domain. It relies on the coupling of the real discrete-scale foam geometry with the equivalent continuous-scale medium through a specially defined scale-coupled zone. This zone holds continuous but nonhomogeneous volumetric radiative properties. The scale-coupled approach is compared to the traditional continuous-scale approach using volumetric radiative properties in the equivalent participating medium and to the direct discrete-scale approach employing the real 3D foam geometry obtained by computed tomography. All the analyses are based on geometrical optics. The Monte Carlo ray-tracing procedure is used for computations of the absorbed radiative fluxes and the apparent radiative behaviors of metal foams. The results obtained by the three approaches are in tenable agreement. The scale-coupled approach is fully validated in calculating the apparent radiative behaviors of metal foams composed of very absorbing to very reflective struts and that composed of very rough to very smooth struts. This new approach leads to a reduction in computational time by approximately one order of magnitude compared to the direct discrete-scale approach. Meanwhile, it can offer information on the local geometry-dependent feature and at the same time the equivalent feature in an integrated simulation. This new approach is promising to combine the advantages of the continuous-scale approach (rapid calculations) and direct discrete-scale approach (accurate prediction of local radiative quantities).

The Spent Fuel Management in Finland and Modifications of Spent Fuel Storages

International Nuclear Information System (INIS)

Maaranen, Paeivi

2014-01-01

The objective of this presentation is to share the Finnish regulator's (STUK) experiences on regulatory oversight of the enlargement of a spent fuel interim storage. An overview of the current situation of spent fuel management in Finland will also be given. In addition, the planned modifications and requirements set for spent fuel storages due to the Fukushima accident are discussed. In Finland, there are four operating reactors, one under construction and two reactors that have a Council of State's Decision-in-Principle to proceed with the planning and licensing of a new reactor. In Olkiluoto, the two operating ASEA-Atom BWR units and the Areva EPR under construction have a shared interim storage for the spent fuel. The storage was designed and constructed in 1980's. The option for enlarging the storage was foreseen in the original design. Considering three operating units to produce their spent fuel and the final disposal to begin in 2022, extra space in the spent fuel storage is estimated to be needed in around 2014. The operator decided to double the number of the spent fuel pools of the storage and the construction began in 2010. The capacity of the enlarged spent fuel storage is considered to be sufficient for the three Olkiluoto units. The enlargement of the interim storage was included in Olkiluoto NPP 1 and 2 operating license. The licensing of the enlargement was conducted as a major plant modification. The operator needed the approval from STUK to conduct the enlargement. Prior to the construction of this modification, the operator was required to submit the similar documentation as needed for applying for the construction license of a nuclear facility. When conducting changes in an old nuclear facility, the new safety requirements have to be followed. The major challenge in the designing the enlargement of the spent fuel storage was to modify it to withstand a large airplane crash. The operator chose to cover the pools with protecting slabs and also to

Simulations of rapid pressure-induced solidification in molten metals

International Nuclear Information System (INIS)

Patel, Mehul V.; Streitz, Frederick H.

2004-01-01

The process of interest in this study is the solidification of a molten metal subjected to rapid pressurization. Most details about solidification occurring when the liquid-solid coexistence line is suddenly transversed along the pressure axis remain unknown. We present preliminary results from an ongoing study of this process for both simple models of metals (Cu) and more sophisticated material models (MGPT potentials for Ta). Atomistic (molecular dynamics) simulations are used to extract details such as the time and length scales that govern these processes. Starting with relatively simple potential models, we demonstrate how molecular dynamics can be used to study solidification. Local and global order parameters that aid in characterizing the phase have been identified, and the dependence of the solidification time on the phase space distance between the final (P,T) state and the coexistence line has been characterized

Implementation of virtual models from sheet metal forming simulation into physical 3D colour models using 3D printing

Science.gov (United States)

Junk, S.

2016-08-01

Today the methods of numerical simulation of sheet metal forming offer a great diversity of possibilities for optimization in product development and in process design. However, the results from simulation are only available as virtual models. Because there are any forming tools available during the early stages of product development, physical models that could serve to represent the virtual results are therefore lacking. Physical 3D-models can be created using 3D-printing and serve as an illustration and present a better understanding of the simulation results. In this way, the results from the simulation can be made more “comprehensible” within a development team. This paper presents the possibilities of 3D-colour printing with particular consideration of the requirements regarding the implementation of sheet metal forming simulation. Using concrete examples of sheet metal forming, the manufacturing of 3D colour models will be expounded upon on the basis of simulation results.

INEL metal recycle annual report, FY-94

International Nuclear Information System (INIS)

Bechtold, T.E.

1994-09-01

In 1992, the mission of the Idaho Chemical Processing Plant was changed from reprocessing of spent nuclear fuels to development of technologies for conditioning of spent nuclear fuels and other high-level wastes for disposal in a geologic repository. In addition, the Department of Energy (DOE) directed Idaho National Engineering Laboratory (INEL) to develop a program plan addressing the management of radioactive contaminated scrap metal (RSM) within the DOE complex. Based on discussions with the EM-30 organization, the INEL Metal Recycle program plan was developed to address all issues of RSM management. Major options considered for RSM management were engineered interim storage, land disposal as low-level waste, and beneficial reuse/recycle. From its inception, the Metal Recycle program has emphasized avoidance of storage and disposal costs through beneficial reuse of RSM. The Metal Recycle program plan includes three major activities: Site-by-site inventory of RSM resources; validation of technologies for conversion of RSM to usable products; and identification of parties prepared to participate in development of a RSM recycle business

Numerical simulation of gas metal arc welding parametrical study

International Nuclear Information System (INIS)

Szanto, M.; Gilad, I.; Shai, I.; Quinn, T.P.

2002-01-01

The Gas Metal Arc Welding (GMAW) is a widely used welding process in the industry. The process variables are usually determined through extensive experiments. Numerical simulation, reduce the cost and extends the understanding of the process. In the present work, a versatile model for numerical simulation of GMAW is presented. The model provides the basis for fundamental understanding of the process. The model solves the magneto-hydrodynamic equations for the flow and temperature fields of the molten electrode and the plasma simultaneously, to form a fully coupled model. A commercial CFD code was extended to include the effects of radiation, Lorentz forces, Joule heating and thermoelectric effects. The geometry of the numerical model assembled to fit an experimental apparatus. To demonstrate the method, an aluminum electrode was modeled in a pure argon arc. Material properties and welding parameters are the input variables in the numerical model. In a typical process, the temperature distribution of the plasma is over 15000 K, resulting high non-linearity of the material properties. Moreover, there is high uncertainty in the available property data, at that range of temperatures. Therefore, correction factors were derived for the material properties to adjust between the numerical and the experimental results. Using the compensated properties, parametric study was performed. The effects of the welding parameters on the process, such the working voltage, electrode feed rate and shielding gas flow, were derived. The principal result of the present work is the ability to predict, by numerical simulation, the mode, size and frequency of the metal transferred from the electrode, which is the main material and energy source for the welding pool in GMAW

Simulating the production of free defects in irradiated metals

International Nuclear Information System (INIS)

Heinisch, H.L.

1995-01-01

Under cascade-producing irradiation by high energy neutrons or charged particles, only a small fraction of the initially displaced atoms contribute to the population of free defects that are available to migrate throughout the metal and cause microstructural changes. Although, in principle, computer simulations of free defect production could best be done using molecular dynamics, in practice, the wide ranges of time and distance scales involved can be done only by a combination of atomistic models that employ various levels of approximation. An atomic-scale, multi-model approach has been developed that combines molecular dynamics, binary collision models and stochastic annealing simulation. The annealing simulation is utilized in calibrating binary collision simulations to the results of molecular dynamics calculations, as well as to model the subsequent migration of the defects on more macroscopic time and size scales. The annealing simulation and the method of calibrating the multi-model approach are discussed, and the results of simulations of cascades in copper are presented. The temperature dependence of free defect production following simulated annealing of isolated cascades in copper shows a differential in the fractions of free vacancies and interstitial defects escaping from the cascade above stage V. This differential, a consequence of the direct formation of interstitial clusters in cascades and the relative thermal stability of vacancy and interstitial clusters during subsequent annealing, is the basis for the production bias mechanism of void swelling. (orig.)

2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

International Nuclear Information System (INIS)

Yaksh, M.; Wang, C.

2004-01-01

In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

Spent fuel management in Japan

International Nuclear Information System (INIS)

Mineo, H.; Nomura, Y.; Sakamoto, K.

1998-01-01

In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

Knowledge Based Cloud FE Simulation of Sheet Metal Forming Processes.

Science.gov (United States)

Zhou, Du; Yuan, Xi; Gao, Haoxiang; Wang, Ailing; Liu, Jun; El Fakir, Omer; Politis, Denis J; Wang, Liliang; Lin, Jianguo

2016-12-13

The use of Finite Element (FE) simulation software to adequately predict the outcome of sheet metal forming processes is crucial to enhancing the efficiency and lowering the development time of such processes, whilst reducing costs involved in trial-and-error prototyping. Recent focus on the substitution of steel components with aluminum alloy alternatives in the automotive and aerospace sectors has increased the need to simulate the forming behavior of such alloys for ever more complex component geometries. However these alloys, and in particular their high strength variants, exhibit limited formability at room temperature, and high temperature manufacturing technologies have been developed to form them. Consequently, advanced constitutive models are required to reflect the associated temperature and strain rate effects. Simulating such behavior is computationally very expensive using conventional FE simulation techniques. This paper presents a novel Knowledge Based Cloud FE (KBC-FE) simulation technique that combines advanced material and friction models with conventional FE simulations in an efficient manner thus enhancing the capability of commercial simulation software packages. The application of these methods is demonstrated through two example case studies, namely: the prediction of a material's forming limit under hot stamping conditions, and the tool life prediction under multi-cycle loading conditions.

Safety aspects of dry spent fuel storage and spent fuel management

International Nuclear Information System (INIS)

Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)

Mock-up facilities for the development of an advanced spent fuel management process using molten salt technology

International Nuclear Information System (INIS)

Young-Joon Shin; Ik-Soo Kim; Seung-Chul Oh; Soo-Haeng Cho; Yo-Taik Song; Hyun-Soo Park

2000-01-01

The Korea Atomic Energy Research Institute (KAERI) has investigated a new approach to spent fuel storage technology that would reduce the total storage volume and the amount of decay heat. The technology utilizes the reduction of oxide fuel to a metal to reduce the volume and preferentially removing the fission products to reduce the decay heat. The uranium oxide is reduced to uranium metal by Li metal in a molten LiCl salt bath. During the reduction process, fission products are dissolved into the LiCl bath and some of the highly radioactive elements, such as Sr and Cs, are preferentially removed from the bath. The reduced uranium metal is cast into an ingot, put into a storage capsule, and stored using conventional storage methods. The fission products are treated as high level radioactive wastes. Each process of the technology has been studied and analyzed for technical feasibility, and has come to the point for designing and constructing of the mock-up for a demonstration of the technology. This paper presents the detailed design of the mock-up of the system and operational characteristics, along with all the details of the equipment for the system. KAERI plans to use the mock-up for the demonstration using an in-active spent fuel specimen. (authors)

Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

Postulated accident scenarios for the on-site transport of spent nuclear fuel

International Nuclear Information System (INIS)

Morandin, G.; Sauve, R.

2004-01-01

Once a spent fuel container is loaded with spent fuel it typically travels on-site to a processing building for permanent lid attachment. During on-site transport a lid clamp is utilized to ensure the container lid remains in place. The safe on-site transport of spent nuclear fuel must rely on the structural integrity of the transport container and system of transport. Regard for on-site traffic and safe, efficient travel routes are important and manageable with well thought-out planning. Non-manageable incidences, such as flying debris from tornado force winds or postulated blasts in proximity to the transport container, that may result in high velocity impact and shock loading on the transport system must be considered. This paper consists of simulations that consider these types of postulated accident scenarios using detailed nonlinear finite element techniques

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

The metallicity distribution of H I systems in the EAGLE cosmological simulations

Science.gov (United States)

Rahmati, Alireza; Oppenheimer, Benjamin D.

2018-06-01

The metallicity of strong H I systems, spanning from damped Lyman α absorbers (DLAs) to Lyman-limit systems (LLSs), is explored between z = 5 → 0 using the EAGLE high-resolution cosmological hydrodynamic simulation of galaxy formation. The metallicities of LLSs and DLAs steadily increase with time in agreement with observations. DLAs are more metal rich than LLSs, although the metallicities in the LLS column density range (N_{H I }≈ 10^{17}-10^{20} cm^{-2}) are relatively flat, evolving from a median H I-weighted metallicity of {Z}≲ 10^{-2} Z_{⊙} at z = 3 to ≈10-0.5 Z⊙ by z = 0. The metal content of H I systems tracks the increasing stellar content of the Universe, holding ≈ 5 {per cent} of the integrated total metals released from stars at z = 0. We also consider partial LLS (pLLS, N_{H I}≈ 10^{16}-10^{17} cm^{-2}) metallicities, and find good agreement with Wotta et al. for the fraction of systems above (37 per cent) and below (63 per cent) 0.1 Z⊙. We also find a large dispersion of pLLS metallicities, although we do not reproduce the observed metallicity bimodality and instead we make the prediction that a larger sample will yield more pLLSs around 0.1 Z⊙. We underpredict the median metallicity of strong LLSs, and predict a population of Z 3 that are not observed, which may indicate more widespread early enrichment in the real Universe compared to EAGLE.

Simulated optical properties of noble metallic nanopolyhedra with different shapes and structures

Science.gov (United States)

Zhang, An-Qi; Qian, Dong-Jin; Chen, Meng

2013-11-01

The optical properties of nanostructured architectures are highly sensitive to their compositions, structures, dimensions, geometries and embedding mediums. Nanopolyhedra, including homogeneous metal nanoparticles and core-shell structures, have unique optical properties. In the beginning of this study, Discrete Dipole Approximation (DDA) method has been introduced. Then the simulated extinction spectra of single-component metal nanoparticles and Au@Ag polyhedra were calculated using both Mie and DDA methods. The influence of morphology and components on the optical response is discussed and well-supported by previously published experimental results. It is observed that the Localized Surface Plasmon Resonance peaks are mainly decided by sharp vertexes and symmetry of noble metallic polyhedra, as well as the structure of the Au@Ag core-shell nanoparticles.

Spent fuel test-climax: a test of geologic storage of high-level waste in granite

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

1981-01-01

A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented

An Indian perspective for transportation and storage of spent fuel

International Nuclear Information System (INIS)

Dey, P.K.

2005-01-01

with stainless steel cavity was also designed for spent PHWR fuel. Fuel transportation is subjected to highly explicit safety and security regulations, constantly updated by international and national experts. It is noted that the radioactive material transportation regulations comprise two distinct objectives. Security or physical protection, consisting in the preventive losses, disappearances, thefts or misappropriation of nuclear materials. Safety, which consists in controlling the irradiation, contamination and criticality hazards inherent in the transportation of radioactive materials, with a view to ensuring that man and the environment remain unaffected by the potential pollution involved. Certain principles underline the transport regulations setup by IAEA and the universally adopted rule is that transport safety must be based on three lines of defense. Viz. the concept of a package, the reliability of transport and the efficacy of specific resources to deal with an accident. Spent fuel transport is carried out in 'type B' packages, designed to withstand severe accident conditions, simulated by tests, validated by approval certificates and subject to inspection. (author)

Survival of encapsulated potentially probiotic Lactobacillus plantarum D6SM3 with bioemulsifier derived from spent yeast in simulated gastrointestinal conditions

Directory of Open Access Journals (Sweden)

Paweena Dikit

2015-08-01

Full Text Available The effect of encapsulation with three kinds of emulsifier (Tween 80, gum arabic and bioemulsifier extracted from spent yeast on the survival of Lactobacillus plantarum D6SM3 in simulated gastrointestinal tract during storage at 4°C and room temperature was investigated. The survival of all encapsulated cells treated in simulated gastric juice was higher than free cells at both pH 2.5 and 3.0. The viability of the free and encapsulated cells showed a gradual decline throughout the storage period at 4°C. However, the viability rapidly declined at room temperature. In addition, the droplet size distribution of encapsulated cells was compared between those with and without an emulsifier by using the laser diffraction method. The particle size and polydispersity value of encapsulated cells were controlled better in emulsion with emulsifier added. The surface of encapsulated cells with emulsifier exhibited smoother characteristics than those without emulsifier.

A concept to combine DOE waste minimization goals with commercial utility needs for a universal container system for spent nuclear fuel storage, transportation, and disposal

International Nuclear Information System (INIS)

Falci, F.P.; Smith, M.L.; Sorenson, K.B.

1993-01-01

The concept of storing, transporting, and disposing of spent fuel using a single package has obvious advantages. Coupling this concept with using contaminated scrap metal from the EM Complex will help reduce a significant portion of waste that would otherwise need to be packaged, stored, and disposed of as low level radioactive waste. Assuming a material of cost of $1 per pound for 800,000 tons of metal needed for universal containers, the potential material cost savings from manufacturing these containers from what would otherwise be a waste product is about $1.5 billion. Clearly, this concept is novel and has significant obstacles that need to be addressed and overcome; particularly in the regulatory arena. However, the potential benefits warrant the evaluation of the proposal on several fronts. DOE OCRWM should seriously consider the universal cask concept for management of spent fuel. DOE EM should pursue the development of melting contaminated scrap for the manufacture of casks. Finally, EM and OCRWM should cooperate on the evaluation of using EM contaminated scrap metal for the manufacture of universal casks for OCRWM spent fuel

Pyrometallurgical processing of Integral Fast Reactor metal fuels

International Nuclear Information System (INIS)

Battles, J.E.; Miller, W.E.; Gay, E.C.

1991-01-01

The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

Disposal of spent nuclear fuel

International Nuclear Information System (INIS)

1979-12-01

This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

Collective processing device for spent fuel

International Nuclear Information System (INIS)

Irie, Hiroaki; Taniguchi, Noboru.

1996-01-01

The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

International Nuclear Information System (INIS)

Hoovler, G.S.; Baldwin, M.N.

1981-04-01

Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

International Nuclear Information System (INIS)

Creer, J.M.; McKinnon, M.A.; Collantes, C.E.

1990-01-01

To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

Oxidation of elemental mercury by modified spent TiO2-based SCR-DeNOx catalysts in simulated coal-fired flue gas.

Science.gov (United States)

Zhao, Lingkui; Li, Caiting; Zhang, Xunan; Zeng, Guangming; Zhang, Jie; Xie, Yin'e

2016-01-01

In order to reduce the costs, the recycle of spent TiO2-based SCR-DeNOx catalysts were employed as a potential catalytic support material for elemental mercury (Hg(0)) oxidation in simulated coal-fired flue gas. The catalytic mechanism for simultaneous removal of Hg(0) and NO was also investigated. The catalysts were characterized by Brunauer-Emmett-Teller (BET), scanning electron microscope (SEM), X-ray diffraction (XRD), and X-ray photoelectron spectroscopy (XPS) method. Results indicated that spent TiO2-based SCR-DeNOx catalyst supported Ce-Mn mixed oxides catalyst (CeMn/SCR1) was highly active for Hg(0) oxidation at low temperatures. The Ce1.00Mn/SCR1 performed the best catalytic activities, and approximately 92.80% mercury oxidation efficiency was obtained at 150 °C. The inhibition effect of NH3 on Hg(0) oxidation was confirmed in that NH3 consumed the surface oxygen. Moreover, H2O inhibited Hg(0) oxidation while SO2 had a promotional effect with the aid of O2. The XPS results illustrated that the surface oxygen was responsible for Hg(0) oxidation and NO conversion. Besides, the Hg(0) oxidation and NO conversion were thought to be aided by synergistic effect between the manganese and cerium oxides.

Economical evaluation on spent fuel storage technology away from reactor

International Nuclear Information System (INIS)

Itoh, Chihiro; Nagano, Koji; Saegusa, Toshiari

2000-01-01

Concerning the spent fuel storage away from reactor, economical comparison was carried out between metal cask and water pool storage technology. The economic index was defined by levelized cost (Unit storage cost) calculated on the assumption that the storage cost is paid at the receipt of the spent fuel at the storage facility. It is found that the cask storage is economical for small and large storage capacity. Unit storage cost of pool storage, however, is getting close to that of cask storage in case of storage capacity of 10,000 ton. Then, the unit storage cost is converted to power generation cost using data of the burn up of the fuel, etc. The cost is obtained as yen 0.09/kWh and yen 0. 15/kWh for cask storage and pool storage, respectively in case of the capacity of 5,000 tonU and the cooling time of 5 years. (author)

Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

International Nuclear Information System (INIS)

Patrick, W.C.; Montan, D.N.; Ballou, L.B.

1981-01-01

The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test

ASSESSMENT OF MICROBIAL LOAD OF SAUSAGES PREPARED FROM DIFFERENT COMBINATION OF SPENT DUCK AND SPENT HEN MEAT

Directory of Open Access Journals (Sweden)

Rajesh Kumar

2016-12-01

Full Text Available Aim of the present study was to assess the microbial load of sausages prepared from different combination of spent duck and spent hen meat. The combination are 100% spent duck (T1, 75%+ 25% spent duck and spent hen (T2, 50%+50% spent duck and spent hen (T3, 25%+75% spent duck and spent hen (T4 and 100% spent hen (T5. All the samples of different combination were subjected to total plate count (TPC, total psychrophilic count (TPSC and total Coliform count (TCC. Mean of TPC for T1, T2, T3, T4 and T5 were 4.69, 4.62, 4.60, 4.49 and 4.46 log 10 CFU/gm respectively, while mean TPSC were 4.46, 4.46, 4.43, 4.36 and 4.36 log CFU/gm respectively There were no significant (p<0.05 difference between the different group of combination of sausages for TPS as well as TPSC but varies significantly (p<0.05 from 14th day of storage in both cases. The coliform group of bacteria will not be detected in any combination of sausages. It is concluded that microbial load of sausage prepared from spent duck is high and it is decreases as the percentage of duck meat decreases but, the upper limit of bacteria in each group of sausages is within limit and hence it is safe for human consumption.

Theories, Methods and Numerical Technology of Sheet Metal Cold and Hot Forming Analysis, Simulation and Engineering Applications

CERN Document Server

Hu, Ping; Liu, Li-zhong; Zhu, Yi-guo

2013-01-01

Over the last 15 years, the application of innovative steel concepts in the automotive industry has increased steadily. Numerical simulation technology of hot forming of high-strength steel allows engineers to modify the formability of hot forming steel metals and to optimize die design schemes. Theories, Methods and Numerical Technology of Sheet Metal Cold and Hot Forming focuses on hot and cold forming theories, numerical methods, relative simulation and experiment techniques for high-strength steel forming and die design in the automobile industry. Theories, Methods and Numerical Technology of Sheet Metal Cold and Hot Forming introduces the general theories of cold forming, then expands upon advanced hot forming theories and simulation methods, including: • the forming process, • constitutive equations, • hot boundary constraint treatment, and • hot forming equipment and experiments. Various calculation methods of cold and hot forming, based on the authors’ experience in commercial CAE software f...

Integrated model of Korean spent fuel and high level waste disposal options - 16091

International Nuclear Information System (INIS)

Hwang, Yongsoo; Miller, Ian

2009-01-01

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21. century. The model addresses alternative design concepts for disposal of SNF of different types (Candu, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model's results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses. (authors)

The corrosion properties of Zr-Cr-NM alloy metallic waste form for longterm disposal

Energy Technology Data Exchange (ETDEWEB)

Han, Seung Youb; Jang, Seon Ah; Eun, Hee Chul; Choi, Jung Hoon; Lee, Ki Rak; Park, Hwan Seo; Ahn, Do Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-06-15

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidifcation host matrix for the noble metal fssion product waste in anode sludge from the electro-refning process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confrm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

Probability of spent fuel transportation accidents

International Nuclear Information System (INIS)

McClure, J.D.