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News Article | November 14, 2016
Site: www.theenergycollective.com

The development of new nuclear fuels is crucial to the success of new fast reactor designs. Examples include TRISO fuel for HTGRs and Molten Salt fuel for 21st century iterations of the work done at Oak Ridge in the 1960s. (WNN) Russia has started testing its new type of nuclear fuel, REMIX, at the MIR research reactor at the Research Institute of Atomic Reactors in Dimitrovgrad, which is in the Ulyanovsk region. Rostaom said on 11/3 that REMIX fuel rods manufactured in July had been “immersed in the active zone” of MIR. Development of REMIX (from Regenerated Mixture) fuel is part of state nuclear corporation Rosatom’s strategy to enable better use of recycled uranium and plutonium on an industrial scale in pressurized water reactors. Some of the plutonium may come from nuclear weapons decommissioned as part of international treaties. Russia is also using this inventory of surplus plutonium to make MOX fuel for its BN-800 fast reactor which was recently connected to the grid to generate electricity. A loop-type research reactor, MIR is designed mainly for testing fuel elements, fuel assemblies and other core components of different types of operating and promising nuclear power reactors. The first data from testing the fuel in MIR will include the “swelling, gassing and distribution of fission products and, of course, the isotopic composition of the used fuel rods,” the head of innovation at the Khlopin Radium Institute, Andrey Belozub, said in the Rosatom statement. Use of the MIR research reactor is an “extremely important step”, Rosatom said, towards full implementation of the project to introduce REMIX into the Russian fuel cycle. According to World Nuclear News, REMIX fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessing used fuel, with a low-enriched uranium (LEU, up to 17% U-235) make-up comprising about 20% of the mix. This gives fuel initially with about 1% Pu-239 and 4% U-235 which can sustain burn-up of 50 GWd/t over four years. REMIX fuel can be repeatedly recycled with 100% core load in current VVER-1000 reactors, and correspondingly reprocessed many times – up to five times according to Russian nuclear fuel manufacturer Tenex, so that with less than three fuel loads in circulation a reactor could run for 60 years using the same fuel, with LEU recharge and waste removal on each cycle. (WNN) Canadian reactor designer StarCore Nuclear has applied to the Canadian Nuclear Safety Commission (CNSC) to begin the vendor design review process for its Generation IV high temperature gas reactor (HTGR). The CNSC’s pre-licensing vendor review process is an optional service to provide an assessment of a nuclear power plant design based on a vendor’s reactor technology. The three-phase review is not a required part of the licensing process for a new nuclear power plant, but aims to verify the acceptability of a nuclear power plant design with respect to Canadian nuclear regulatory requirements and expectations. Earlier this year the CNSC agreed to conduct a phase 1 vendor design review for Terrestrial Energy’s integral molten salt reactor design concept. StarCore CEO David Dabney said the company’s application to the CNSC, lodged on 24 October, marked the culmination of eight years’ work. “We are confident that our plant size and technology will enable us to bring safe, clean energy to the many remote sites in Northern Canada that currently have no choice other than to use costly, unreliable and polluting carbon-based fuels,” he said. Montréal-based StarCore, founded in 2008, is focused on developing small modular reactors (SMRs) to provide power and potable water to remote communities in Canada. Its standard HTGR unit would produce 20 MWe (36 MWth), expandable to 100 MWe, from a unit small enough to be delivered by truck. The helium-cooled reactor uses Triso fuel, – spherical particles of uranium fuel coated by carbon which effectively gives each tiny particle its own primary containment system, manufactured by BWXT Technologies. Each reactor would require refueling at five-yearly intervals. StarCore describes its reactor as “inherently safe.” The use of helium, which does not become radioactive, as a coolant means that any loss of coolant would be “inconsequential”, the company says. The reactors would be embedded 50 metres underground in concrete silos sealed with ten-tonne caps. DOE Inks Deal with GE-Hitachi for Laser Enrichment Plant at Paducah The Department of Energy (DOE) has agreed to sell depleted uranium to GE-Hitachi Global Laser Enrichment, LLC (GLE) over a 40-year period which would be enriched at a proposed GLE state-of-the-art facility. DOE has agreed to sell 300,000 tonnes of depleted uranium hexafluoride (UF6) to GE Hitachi Global Laser Enrichment (GLE) for re-enrichment at a proposed plant to be built near DOE’s Paducah site in Kentucky. The agreement paves the way for commercialization of Silex laser enrichment technology. The proposed new facility would use depleted uranium to produce natural uranium which is used for production of fuel for civil nuclear reactors. The facility would be built near DOE’s Paducah Gaseous Diffusion Plant in western Kentucky. The construction and operation of the billion-dollar facility at Paducah could to bring approximately 800 to 1,200 jobs to the local community. “This agreement furthers the Energy Department’s environmental cleanup mission while reducing cleanup costs, creating good local jobs, and supporting an economical enrichment enterprise for our energy needs,” said Energy Secretary Ernest Moniz. GLE will finance, construct, own and operate the Paducah Laser Enrichment Facility (PLEF) adjacent to the Energy Department site. The facility will be a commercial uranium enrichment production facility under a Nuclear Regulatory Commission (NRC) license. DOE’s inventory of depleted uranium is safely stored in approximately 65,000 specialized storage cylinders at the Department’s Paducah and Portsmouth (Ohio) sites. The Paducah plant was constructed in the 1950s to enrich uranium for national security applications, and later enriched uranium for commercial nuclear power generation. The Energy Department resumed control of the plant enrichment facilities in 2014 after the operator ceased gaseous-diffusion enrichment operations in 2013. GLE is a joint business venture of GE (51%), Hitachi (25%) and Cameco (24%). Earlier this year GE Hitachi announced its desire to reduce its equity interest in GLE and in April signed a term sheet with Silex giving the Australian company an exclusive option to acquire GE Hitachi’s entire 76% interest in GLE. In 2012, the US NRC granted GLE a combined construction and operating licence for a laser enrichment plant of up to 6 million separative work units at Wilmington, North Carolina. GLE has successfully demonstrated the concept in a test loop at Global Nuclear Fuel’s Wilmington fuel fabrication facility but has not yet decided whether to proceed with a full-scale commercial plant there. (NucNet): Russia is considering asking foreign partners to join its development of the Generation IV SVBR 100 reactor design, but has denied reports that the cost of the project has more than doubled. The original cost of the project was put at 15bn rubles (€209m, $226m) and this has not changed, Rosatom said. The SVBR 100 is one of six designs chosen by the Generation IV International Forum (GIF) for its program of research and development into the next generation nuclear energy systems. GIF said the SVBR 100 is a lead-cooled fast reactor which features a fast neutron spectrum, high temperature operation, and cooling by molten lead or lead-bismuth. It would have multiple applications including production of electricity, hydrogen and process heat. Molten Salt Reactors: IAEA to Establish New Platform for Collaboration Experts from 17 countries laid the foundations last week for enhanced international cooperation on a technology that promises to deliver nuclear power with a lower risk of severe accidents, helping to decrease the world’s dependence on fossil fuels and mitigate climate change. “It is the first time a comprehensive IAEA international meeting on molten salt reactors has ever taken place,” said Stefano Monti, Head of the Nuclear Power Development Section at the IAEA. “Given the interest of Member States, the IAEA could provide a platform for international cooperation and information exchange on the development of these advanced nuclear systems.” Molten salt reactor technology has attracted private funding over the last few years, and several reactor concepts are under development. One area under research is the compatibility between the salt coolant and the structural materials and, for some designs, the chemical processes related to the associated fuel cycle, Monti said. The challenges are not only technical. Nuclear regulators will need to review existing safety regulations to see how these can be modified, if necessary, to fit molten salt reactors, since they differ significantly from reactors in use today, said Stewart Magruder, senior nuclear safety officer at the IAEA. Participants, including researchers, designers and industry representatives, emphasized the need for an international platform for information exchange. “While the United States is actively developing both technology and safety regulations for molten salt reactors, the meeting is an important platform to exchange knowledge and information with Member States not engaged in the existing forums,” said David Holcomb from the Oak Ridge National Laboratory. Molten salt reactors, nuclear power reactors that use liquid salt as primary coolant or a molten salt mixture as fuel, have many favorable characteristics for nuclear safety and sustainability. The concept was developed in the 1960s, but put aside in favor of what has become mainstream nuclear technology since. In recent years, however, technological advances have led to growing interest in molten salt technology and to the launch of new initiatives. The technology needs at least a decade of further intensive research, validation and qualification before commercialization. Molten salt reactors operate at higher temperatures, making them more efficient in generating electricity. In addition, their low operating pressure can reduce the risk of coolant loss, which could otherwise result in an accident. Molten salt reactors can run on various types of nuclear fuel and use different fuel cycles. This conserves fuel resources and reduces the volume, radiotoxicity and lifetime of high-level radioactive waste. To help speed up research, it is essential to move from bilateral to multilateral cooperation, said Chen Kun from the Shanghai Institute of Applied Physics of the Chinese Academy of Sciences. “It is the first time China has the opportunity to share knowledge with India, Indonesia and Turkey on this technology.” Indonesia is considering building its first nuclear power plant with molten salt reactor design, said Bob Soelaiman Effendi from Indonesia Thorium Energy Community. (WNN) China and the UK have signed a joint R&D agreement which created their Joint Research and Innovation Centre (JRIC) to be opened soon in Manchester, England. Initial work is expected to include developing advanced manufacturing methods. JRIC will support innovation in nuclear research and development through UK-China collaboration. This will develop, it said, “leading-edge research and innovative technologies which will support safe and reliable nuclear energy around the globe.” With NNL and CNNC each owning a 50% share, they will jointly pay for the centre’s research and development expenses and plan to invest 422 million yuan ($65.1 million) over a five-year period, CNNC said. (WNN) The UK’s Nuclear Advanced Manufacturing Research Centre (AMRC) said it has signed a new agreement with the US Nuclear Infrastructure Council (USNIC) to work together on research and development to support the UK’s civil nuclear program. The memorandum of understanding was signed by Jay Shaw, senior business development manager for the Nuclear AMRC, and David Blee, executive director of USNIC, during a visit to the Nuclear AMRC on 10/26.

Chakin V.,Karlsruhe Institute of Technology | Reimann J.,Karlsruhe Institute of Technology | Moeslang A.,Karlsruhe Institute of Technology | Latypov R.,Research Institute of Atomic Reactors | Obukhov A.,Research Institute of Atomic Reactors
Progress in Nuclear Energy | Year: 2012

Beryllium will be used as a neutron multiplier in Helium Cooled Pebble Bed (HCPB) DEMO blankets. The beryllium thermal conductivity is determining the maximum pebble bed temperature and, therefore, is very important for blanket design. Different grades of beryllium discs were neutron-irradiated at temperatures between 343 and 673 K and at fluences up to 1.6 × 10 23 cm -2. At lower irradiation temperatures a significant drop of the beryllium thermal conductivity occurs even after small neutron fluences. With increasing neutron fluence, further moderate decreases of the conductivity are observed. With increasing irradiation temperature, the thermal conductivity further decreases. If the thermal conductivity of the irradiated beryllium is known, the conductivity of irradiated beryllium pebble beds can be assessed using the model suggested in this study. © 2011 Elsevier Ltd. All rights reserved.

Zilberman B.Y.,Khlopin Radium Institute | Chistyakov V.M.,Research Institute of Atomic Reactors
Radiochemistry | Year: 2016

α-Radiolysis of tributyl phosphate in Sintin n-paraffin diluent in equilibrium with HNO3 solutions at single “internal” irradiation from the extracted 238Pu was studied. The radiation-chemical yields (molecules/100 eV) of butyl hydrogen phosphates (BHP), carboxylic acids, carbonyl compounds, and nitro compounds upon irradiation of 20% TBP in the treated Sintin in equilibrium with 3 M HNO3 were 0.4 (at dibutyl hydrogen phosphate to monobutyl dihydrogen phosphate ratio HDBP: H2MBP = 4.3), 1.4, 0.2–0.3, and 0.2–0.3, respectively. The degradation and oxidation processes occur more deeply than under γ-irradiation. A simple volumetric method for determining carboxylic acids in the extract was developed. In the course of irradiation, the Pu(IV) oxidation state in the extract does not change, and its retention is due to the interaction with BHP at the ratio BHP: Pu = 2 in stripping with 0.02 M HNO3 and BHP: Pu = 4 in stripping with Fe(II). The retention can be eliminated by the displacing action of Np(IV). © 2016, Pleiades Publishing, Inc.

Zhang J.,University of Michigan | Livshits T.S.,RAS Institute of Geology and Mineralogy | Lizin A.A.,Research Institute of Atomic Reactors | Hu Q.,University of Michigan | Ewing R.C.,University of Michigan
Journal of Nuclear Materials | Year: 2010

Garnet, A3B2X3O12, has a structure that can incorporate actinides. Hence, the susceptibility of the garnet structure to radiation damage has been investigated by comparing the results of self-radiation damage from α-decay of 244Cm and a 1 MeV Kr2+ ion irradiation. Gradual amorphization with increasing fluence was observed by X-ray diffraction analysis and in situ transmission electron microscopy. The critical dose, Dc, for an yttrium-aluminum garnet (Y3Al5O12) doped with 3 wt.% 244Cm is calculated to be 0.4 displacements per atom (dpa). While the doses obtained by ion irradiation experiments of garnets with different compositions (Y2.43Nd0.57)(Al4.43Si 0.44)O12, (Ca1.64Ce0.41Nd 0.42La0.18Pr0.18Sm0.14Gd 0.04)Zr1.27Fe3.71O12, and (Ca 1.09Gd1.23Ce0.43)Sn1.16Fe 3.84O12, varied from 0.29 to 0.55 dpa at room temperature. The similarity in the amorphization dose at room temperature and critical temperature of the different garnet compositions suggest that the radiation response for the garnet structure is structurally constrained, rather than sensitive to composition, which is the case for the pyrochlore structure-type. © 2010 Elsevier B.V. All rights reserved.

Momotov V.N.,Research Institute of Atomic Reactors | Erin E.A.,Research Institute of Atomic Reactors | Chistyakov V.M.,Research Institute of Atomic Reactors
Radiochemistry | Year: 2015

The kinetics of reduction of U(VI) with zinc amalgam at H2SO4 concentrations of 0.5-4.0 M, initial UO 2 2+ concentrations of (2.08-40.00) × 10-3 M, and temperatures of 295.0-325.0 K was studied by spectrophotometry. The kinetic scheme of the UO 2 2+ transformation, the rate law, and the reduction rate constants were determined. The activation energy E a and the thermodynamic functions of formation of the activated complex (Gibbs energy ΔG ≠, enthalpy ΔH ≠, and entropy ΔS ≠ of activation) were calculated. © 2015 Pleiades Publishing, Inc.

Momotov V.N.,Research Institute of Atomic Reactors | Erin E.A.,Research Institute of Atomic Reactors | Chistyakov V.M.,Research Institute of Atomic Reactors
Radiochemistry | Year: 2014

The influence exerted on the accuracy of coulometric titration by the composition of the supporting electrolyte solution in the titration step, by the way of stirring of the supporting electrolyte solution, by the weight of the aliquot of the solution being titrated, by the weight of the substance being titrated in the aliquot, by the kind of reductant, and by the solution compositions in the steps of oxidation and reduction of U-Pu mixtures was studied by spectrophotometry and coulometry. The optimum conditions for performing the analysis were found. © 2014 Pleiades Publishing, Inc.

Sorokin A.A.,Central Research Institute of Structural Materials prometey | Margolin B.Z.,Central Research Institute of Structural Materials prometey | Kursevich I.P.,Central Research Institute of Structural Materials prometey | Minkin A.J.,Central Research Institute of Structural Materials prometey | Neustroev V.S.,Research Institute of Atomic Reactors
Journal of Nuclear Materials | Year: 2014

Tensile properties of austenitic stainless steels used for pressure vessel internals of WWER type reactors (18Cr-10Ni-Ti steel and its weld metal) in the initial and irradiated conditions were investigated. Based on the presented original investigations and generalization of the available experimental data the dependences of yield strength and ultimate strength on a neutron damage dose up to 108 dpa, irradiation temperature range 320-450 C and test temperature range 20-450 C were obtained. The method of determination of the stress-strain curve parameters was proposed which does not require uniform elongation of a specimen as an input parameter. The dependences was proposed allowing one to calculate the stress-strain curve parameters for 18Cr-10Ni-Ti steel and its weld metal for different test temperatures, different irradiation temperatures and doses. The dependences were obtained to describe the fracture strain decrease under irradiation at a temperature range 320-340 C when irradiation swelling is absent. © 2013 Published by Elsevier B.V.

Pavlov S.V.,Research Institute of Atomic Reactors
Russian Journal of Nondestructive Testing | Year: 2011

An ultrasonic method for the detection of leaking fuel elements in the structure of the fuel assembly of BB Cyrillic capital letter EP-1000 water-water power reactors is considered. The method is intended for the detection of water passing through a cladding flaw into a fuel element. The calculation and experimental study data for the acoustic channel of the method and the results of testing on radiated fuel assemblies from the Kalinin and Balakovo Nuclear Power Plants are given. The high sensitivity of the developed method is shown experimentally. The boundaries of its applicability are determined. © 2011 Pleiades Publishing, Ltd.

Livshits T.S.,RAS Institute of Geology and Mineralogy | Lizin A.A.,Research Institute of Atomic Reactors | Zhang J.M.,University of Michigan | Ewing R.C.,University of Michigan
Geology of Ore Deposits | Year: 2010

The stability of synthetic REE-aluminate garnets irradiated by accelerated Kr2+ ions and affected by alpha decay of 244Cm (T1/2 = 18. 1 yr) has been studied. The dose of irradiation sufficient for the complete disordering of the aluminate garnet structure is 0. 40-0. 55 displacements per atom. This value increases with rising temperature due to the increasing intensity of recovery from radiation damage to the lattice by heating. The critical temperature above which the structure of REE-aluminate is not damaged by radiation is 550°C. The amorphization dose for aluminates with garnet structure is two to three times higher than of that previously studied ferrites; the critical temperature of both is similar. In resistance to radiation, aluminate garnets do not yield to zirconolite and exceed titanate pyrochlore. Heating to 250°C does not lead to substantial recovery from radiation defects in the garnet structure. The radiation impact on matrices of real actinide (An) wastes is lower than that related to ion irradiation and 244Cm doping, and this facilitates a higher radiation resistance of garnets containing HLW. © 2010 Pleiades Publishing, Ltd.

Osipenko A.,Research Institute of Atomic Reactors | Maershin A.,Research Institute of Atomic Reactors | Smolenski V.,RAS Institute of High Temperature Electrochemistry | Novoselova A.,RAS Institute of High Temperature Electrochemistry | And 2 more authors.
Journal of Electroanalytical Chemistry | Year: 2011

This work presents the electrochemical study of Cm(III) in fused 3LiCl-2KCl eutectic in the temperature range 723-923 K. Transient electrochemical techniques such as cyclic, square wave voltammetries, and chronopotentiometry have been used in order to investigate the reduction mechanism of curium ions up to metal. The results obtained show that the reduction reaction takes place in a single step Cm(III)+3ē⇒Cm(0). The diffusion coefficient of Cm 3+ ions was determined by cyclic voltammetry at different temperatures by applying Berzins-Delahay equation. The validity of the Arrhenius law was also verified and the activation energy for diffusion was found to be 28.21 kJ mol-1. The apparent standard electrode potential of the redox couple Cm(III)/Cm(0) was found by chronopotentiometry at several temperatures. The thermodynamic properties of curium trichloride have also been calculated. © 2010 Elsevier B.V. All rights reserved.

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