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Sample records for transmutation fuel development

  1. Development of Metallic Fuels for Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

    2015-09-01

    Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

  2. Definition of Technology Readiness Levels for Transmutation Fuel Development

    International Nuclear Information System (INIS)

    Jon Carmack; Kemal O. Pasamehmetoglu

    2008-01-01

    To quantitatively assess the maturity of a given technology, the Technology Readiness Level (TRL) process is used. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Transmutation fuel development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the transmutation fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Transuranic Fuel Development Campaign

  3. Transmutation Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Song, T. Y.; Park, W. S.; Kim, Y. H. (and others)

    2007-06-15

    The spent fuel coming from the PWR is one of the most difficult problems to be solved for the continuous use of nuclear power. It takes a few million years to be safe under the ground. Therefore, it is not easy to take care of the spent fuel for such a long time. Transmutation technology is the key technology which can solve the spent fuel problem basically. Transmutation is to transmute long-lived radioactive nuclides in the spent fuel into short-lived or stable nuclide through nuclear reactions. The long-lived radioactive nuclides can be TRU and fission products such as Tc-99 and I-129. Although the transmutation technology does not make the underground disposal totally unnecessary, the period to take care of the spent fuel can be reduced to the order of a few hundred years. In addition to the environmental benefit, transmutation can be considered to recycle the energy in the spent fuel since the transmutation is performed through nuclear fission reaction of the TRU in the spent fuel. Therefore, transmutation technology is worth being developed in economical aspect. The results of this work can be a basis for the next stage research. The objective of the third stage research was to complete the core conceptual design and verification of the key technologies. The final results will contribute to the establishment of Korean back end fuel cycle policy by providing technical guidelines.

  4. Development of CERMET fuels for minor actinides transmutation

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Naestren, C.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

    2006-01-01

    The sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmutation fuels are linked with the core design and safety parameters, the fuel properties and the ease of reprocessing. This study concerns the properties of metals as matrices, with the particular case of Mo. To improve the neutronic characteristics, enriched molybdenum (Mo-92) is required. To overcome the high enrichment cost, it is proposed to recover the matrix by pellet dissolution, and to recycle it for further use. Irradiation programmes are also planned to examine the in-reactor properties of the material. Based on the current status of the research, the results are promising, but irradiation results are still missing. (authors)

  5. EFTTRA, a European collaboration for the development of fuels and targets for the transmutation

    International Nuclear Information System (INIS)

    Babelot, J.F.; Muehling, G.; Prunier, C.; Rome, M.

    1994-12-01

    In the frame of the research programmes on the transmutation of long lived nuclides, many experimental or theoretical investigations have to be carried out within European collaborations, owing mainly to the costs of such studies. Therefore, a group named 'Experimental Feasibility of Targets for Transmutation' (EFTTRA), has been formed, with participants from CEA (France). ECN (The Netherlands), EDF (France), KFK (Germany) and ITU (European Commission), to organise joint experiments for the study of materials for the transmutation. So far, it was decided to focus the work on the transmutation of 99 Tc (metal), of 129 I (compound), and of Am (in an inert matrix). Irradiations will take place in parallel in the Phenix fast reactor in France, and in the high flux thermal reactor HFR in the Netherlands. These experiments, together with the related post-irradiation examinations, constitute the first phase of the EFTTRA collaboration. In subsequent phases, EFTTRA will contribute to the development of fuels and targets. (orig.)

  6. Impact of Transmutation Scenarios on Fuel Transportation

    International Nuclear Information System (INIS)

    Saturnin, A.; Duret, B.; Allou, A.; Jasserand, F.; Fillastre, E.; Giffard, F.X.; Chabert, C.; Caron-Charles, M.; Garzenne, C.; Laugier, F.

    2015-01-01

    Minor actinides transmutation scenarios have been studied in the frame of the French Sustainable Radioactive Waste Management Act of 28 June 2006. Transmutation scenarios supposed the introduction of a sodium-cooled fast reactor fleet using homogeneous or heterogeneous recycling modes for the minor actinides. Americium, neptunium and curium (MA) or americium alone (Am) can be transmuted together in a homogeneous way embedded in FR-MOX fuel or incorporated in MA or Am-Bearing radial Blankets (MABB or AmBB). MA transmutation in Accelerator Driven System has also been studied while plutonium is being recycled in SFR. Assessments and comparisons of these advanced cycles have been performed considering technical and economic criteria. Transportation needs for fresh and used transmutation fuels is one of these criteria. Transmutation fuels have specific characteristics in terms of thermal load and neutron emissions. Thermal, radiation and criticality constraints have been taken into account in this study to suggest cask concepts for routine conditions of transport, to estimate the number of assemblies to be transported in a cask and the number of annual transports. Comparison with the no transmutation option, i.e. management of uranium and plutonium in SFRs, is also presented. Regarding these matters, no high difficulties appear for assemblies with limited content of Am (homogeneous or heterogeneous recycling modes). When fuels contain curium, technical transport uncertainties increase because of the important heat release requiring dividing fresh fuels and technological innovations development (MABB and ADS). (authors)

  7. FCRD Transmutation Fuels Handbook 2015

    Energy Technology Data Exchange (ETDEWEB)

    Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloys containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling

  8. Recycling and transmutation of spent fuel as a sustainable option for the nuclear energy development

    International Nuclear Information System (INIS)

    Maiorino, Jose R.; Moreira, Joao M.L.

    2013-01-01

    The objective of this paper is to discuss the option of recycling and transmutation of radioactive waste against Once-through Fuel Cycle (OTC) based on uranium feed under the perspective of sustainability. We use a qualitative analysis to compare OTC with closed fuel cycles based on studies already performed such as the Red Impact Project and the comparative study on accelerator driven systems and fast reactors for advanced fuel cycles performed by the Nuclear Energy Agency. The results show that recycling and transmutation fuel cycles are more attractive than the OTC from the point of view of sustainability. The main conclusion is that the decision about the construction of a deep geological repository for spent fuel disposal must be reevaluated. (author)

  9. Application of the pyrochemical DOS, developed by the CEA, within reprocessing of CERCER transmutation fuel targets

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, E.; Ducasse, T.; Bertrand, M. [CEA, Centre de Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, SMCS, LDPS, F-30207 Bagnols-sur-Ceze (France); Miguirditchian, M. [CEA, Centre de Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, SMCS, LCPE, F-30207 Bagnols-sur-Ceze (France)

    2016-07-01

    Pyrochemical technology using high-temperature molten salts and molten metal media presents a potential interest for an overall separation and transmutation strategy for long-lived radionuclides. Within the frame of the two French acts on radioactive waste management, a pyrochemical research/development program was launched at the CEA Marcoule in the late 90's. The second step is the actinides back-extraction, which consists in a liquid/liquid oxidative stripping of the An from aluminium matrix into molten chloride media. The DOS process has been successfully demonstrated for treatment of oxide type fuels within the last years: the core of the process has been already assessed and the studies have shown high selectivity and a quantitative recovery of actinides. Within the framework of the SACSESS European research program, the pyrochemical activities focused on applications of the DOS process to reprocess CERCER transmutation targets. These particular types of fuels consist of a mixture of minor actinides (MA) oxides diluted in an inert (oxide MgO) matrices. The behaviour of matrices material was first investigated regarding the solubility in the fluoride salt, starting from both oxide powders or sintered pellets. The saturation of Mg in the salt could be estimated at ∼ 3 wt%. Regarding the reductive extraction, as expected no Mg was reduced by the metallic phase. The present work also highlights that Mg has low impact on the extraction efficiency of U as long as the salt is not saturated. Once the saturation occurs, the efficiency starts to decrease. So we recommend recycling the salt when Mg saturation is reached.

  10. Transmutation Fuel Campaign Description and Status

    International Nuclear Information System (INIS)

    Jon Carmack; Kemal O. Pasamehmetoglu

    2008-01-01

    This report contains a technical summary package in response to a Level 2 milestone in the transmutation fuel campaign (TFC) management work-package calling for input to the Secretarial decision. At present, the form of the Secretarial decision package is not fully defined, and it is not clear exactly what will be required from the TFC as a final input. However, it is anticipated that a series of technical and programmatic documents will need to be provided in support of a wider encompassing document on GNEP technology development activities. The TFC technical leadership team provides this report as initial input to the secretarial decision package which is being developed by the Technical Integration Office (TIO) in support of Secretarial decision. This report contains a summary of the TFC execution plan with a work breakdown structure, high level schedule, major milestones, and summary description of critical activities in support of campaign objectives. Supporting documents referenced in this report but provided under separate cover include: (1) An updated review of the state-of-the art for transmutation fuel development activities considering national as well as international fuel research and development testing activities. (2) A definition of the Technology Readiness Level (TRL) used to systematically define and execute the transmutation fuel development activities

  11. Development of nitride fuel and pyrochemical process for transmutation of minor actinides

    International Nuclear Information System (INIS)

    Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo; Uno, Masayoshi

    2010-01-01

    Nitride fuel cycle for transmutation of minor actinides has been investigated under the double-strata fuel cycle concept. Mononitride solid solutions containing minor actinides have been prepared and characterised. Thermo-physical properties, such as thermal expansion, heat capacity and thermal diffusivity, have been measured by use of minor actinide nitride and burn-up simulated nitride samples. Irradiation behaviour of nitride fuel has been examined by irradiation tests. Pyrochemical process for treatment of spent nitride fuel has been investigated mainly by electrochemical measurements and nitride formation behaviour in pyrochemical process has been studied for recycled fuel fabrication. Recent results of experimental study on nitride fuel and pyrochemical process are summarised in the paper. (authors)

  12. Transmutation Fuels Campaign FY-09 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2009-09-01

    This report summarizes the fiscal year 2009 (FY-08) accomplishments for the Transmutation Fuels Campaign (TFC). The emphasis is on the accomplishments and relevance of the work. Detailed description of the methods used to achieve the highlighted results and the associated support tasks are not included in this report.

  13. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

  14. Transmutation of radioactive waste: Effect on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Rasmussen, N.C.; Pigford, T.H.

    1997-01-01

    A committee of the National Research Council reviewed three concepts for transmuting radionuclides recovered from the chemical reprocessing of commercial light-water-reactor (LWR) fuel: LWR transmutation reactors fueled with recycled actinides, advanced liquid-metal reactors (ALMRs), and accelerator-driven subcritical reactors for transmutation of waste (ATW). The concepts were evaluated in terms of: (1) the extent to which waste disposal would benefit from transmutation, (2) time required to reduce the total inventory of radionuclides in the waste and fuel cycle, (3) the complexity of the overall transmutation system, (4) the extent of new development required, and (5) institutional and economic problems of operating such systems. Transmutation could affect geologic disposal of waste by reducing the inventory of transuranics (TRUs), fission products, and other radionuclides in the waste. Reducing the inventory of transuranics does not necessarily affect radiation doses to people who use contaminated ground water if the dissolution rate of transuranics in waste is controlled by elemental solubilities. However, reducing inventories of Am and Pu would decrease potential hazards from human intrusion. The likelihood for underground nuclear criticality would also be reduced. The long-lived fission products Tc-99, I-129, Cs-135 and others typically contribute most to the long-term radiation doses to future populations who use contaminated water from the repository. Their transmutation requires thermal or epithermal neutrons, readily available in LWR and ATW transmutors. ALMR and LWR transmutors would require several hundred years to reduce the total transuranic inventory by even a factor of 10 at constant electric power, and thousands of years for a hundred-fold reduction. For the same electrical power, the ATW could reduce total transuranic inventory about tenfold more rapidly, because of its very high thermal-neutron flux. However, extremely low process losses would be

  15. Oxide fuels and targets for transmutation

    International Nuclear Information System (INIS)

    Sudreau, F.; Bonnerot, J.M.; Warin, D.; Gaillard-Groleas, G.; Ferroud-Plattet, M.P.

    2007-01-01

    Full text of publication follows. Direction 1 of the French Act dated 30 December 1991 on the management of high-level, long-lived radioactive waste involves exploring solutions designed to separate long-lived radionuclides from the spent fuel and to transmute them under neutron flux into shorter half-lives or stable elements. In the French research programme conducted by CEA, these radionuclides are mainly minor actinides (americium, neptunium and curium) and fission products (particularly caesium, iodine and technetium). Within this context, this paper aims at illustrating the vast programme that CEA has performed in order to demonstrate the scientific and technical feasibility of minor actinide transmutation. An important part of the research was carried out in collaboration with French research (CNRS) and industrial (EDF, AREVA) organisations, and also in the framework of international co-operation programmes with the European Institute for Transuranium Elements in Karlsruhe (ITU), the US Department of Energy (DOE), the Japanese Atomic Energy Research Institute (now JAEA) and Central Research Institute of Electric Power Industry (CRIEPI) and the Russian Ministry for Atomic Energy (ROSATOM). Such research made it possible to evaluate the capacity of MOX fuels to be used as a support for minor actinide transmutation (homogeneous method). Simulations of pressurised water reactor (PWR) fuels have revealed the limits of this transmutation method, which are mainly related to the pressurization of the fuel rods and the formation of high active californium. On the contrary, for sodium-cooled fast reactor fuels possibly designed with large expansion plenums a first experimental demonstration of the transmutation of americium and neptunium has been successful in the Phenix reactor. Various studies designed to demonstrate the theoretical and experimental feasibility of transmutation using an inert support (heterogeneous method) have been carried out in HFR (EFTTRA

  16. Technical meeting on 'Review of solid and mobile fuels for partitioning and transmutation systems'. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    The topics covered during the Meeting were divided into two Sessions. Session 1 - Qualification of Solid and Mobile Fuels delt with: Neutronic, fuel and material properties of a molten salt transmuter; and Preliminary analysis of transmutation fuels for KALIMER. Session 2 - Reactor Physics and Safety Characteristics of Transmutation Systems based on Solid and Mobile Fuel Types included the following: Activity in NEA for P and T area; IAEA activities in the area of partitioning and transmutation; The R and D activity in Brazil: A conceptual fast energy amplifier ADS cooled by helium double stata Th/U fuel cycle; Closed fuel cycle and contemporary tendencies of the nuclear facilities development; Current Russian activities in P and T area; Pyrochemical reprocessing and nuclear spent fuel disposal project; Fuel selection criteria specific for double stratum minor actinide burners.

  17. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  18. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  19. Program on fuels for transmutation: present status and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Combustibles

    2001-07-01

    The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

  20. Program on fuels for transmutation: present status and prospects

    International Nuclear Information System (INIS)

    Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S.

    2001-01-01

    The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

  1. 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)

  2. Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

    2010-06-15

    Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

  3. Separations technology development to support accelerator-driven transmutation concepts

    International Nuclear Information System (INIS)

    Venneri, F.; Arthur, E.; Bowman, C.

    1996-01-01

    This is the final report of a one-year Laboratory-Directed Research and Development (LDRD) Project at the Los Alamos National Laboratory (LANL). This project investigated separations technology development needed for accelerator-driven transmutation technology (ADTT) concepts, particularly those associated with plutonium disposition (accelerator-based conversion, ABC) and high-level radioactive waste transmutation (accelerator transmutation of waste, ATW). Specific focus areas included separations needed for preparation of feeds to ABC and ATW systems, for example from spent reactor fuel sources, those required within an ABC/ATW system for material recycle and recovery of key long-lived radionuclides for further transmutation, and those required for reuse and cleanup of molten fluoride salts. The project also featured beginning experimental development in areas associated with a small molten-salt test loop and exploratory centrifugal separations systems

  4. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  5. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  6. Multi-faceted evaluation for nuclear fuel cycles with transmutation

    International Nuclear Information System (INIS)

    Nishihara, Kenji

    2015-03-01

    Environment impact, economy and proliferation resistance were estimated for nuclear fuel cycles involving transmutation by fast reactor and accelerator-driven system in equilibrium state. As a result, the transmutation scenario using only fast reactor was superior to the scenarios combined with accelerator-driven system in all estimation, but the differences were insignificant. (author)

  7. Progress on the Application of Metallic Fuels for Actinide Transmutation

    International Nuclear Information System (INIS)

    Kennedy, J. Rory; Fielding, Randall; Janney, Dawn; Mariani, Robert; Teague, Melissa; Egeland, Gerald

    2015-01-01

    Full text of publication follows: Idaho National Laboratory (INL) is developing actinide bearing alloy metallic fuels intended for effecting the transmutation of long-lived isotopes in fast reactor application as part of a partitioning and transmutation strategy. This presentation will report on progress in three areas of this effort: demonstration of the fabrication of fuels under remote (hot cell) conditions directly coupled to the product from the Pyro-processing of spent fuel as part of the Joint Fuel Cycle Studies (JFCS) collaboration with the Korean Atomic Energy Research Institute (KAERI); the chemical sequestration of lanthanide fission products to mitigate fuel-cladding-chemical-interaction (FCCI); and transmission electron microscopy (TEM) and atom probe tomography (APT) studies on the as-cast microstructure of the metallic fuel alloy. For the JFCS efforts, we report on the implementation of the Glove-box Advanced Casting System (GACS) as a prototype casting furnace for eventual installation into the INL Hot Fuel Examination Facility (HFEF) where the recycled fuel will be cast. Results from optimising process parameters with respect to fuel characteristics, americium volatility, materials interaction, and lanthanide fission product carry over distribution will be discussed. With respect to the lanthanide carry over from the Pyro-processing product, encouraging studies on concepts to chemically sequester the FCCI promoting lanthanides within the fuel matrix thus inhibiting migration and interaction with the cladding will be presented. Finally, in relation to advanced modelling and simulation efforts, detailed investigations and interpretation on the nano-scale as cast microstructure of possible recycle fuel composition containing U, Pu, Am, Np as well as carry-over lanthanide species will be discussed. These studies are important for establishing the initial conditions from which advanced physics based fuel performance codes will run. (authors)

  8. Development of transmutation technologies of radioactive waste by actinoid hydride

    International Nuclear Information System (INIS)

    Konashi, Kenji; Matsui, Hideki; Yamawaki, Michio

    2001-01-01

    Two waste treatment methods, geological disposal and transmutation, have been studied. The transmutation method changes long-lived radioactive nuclides to short-lived one or stabilizes them by nuclear transformation. The transmutation by actinoid hydride is exactly alike that transformation method from actinoid disposal waste to Pu fuel. For this object, OMEGA project is processing now. The transmutation is difficult by two causes such as large amount of long-lived radioactive nuclides and not enough development of control technologies of nuclear reaction except atomic reactor. The transmutation using actinoid hydride has merits that the amount of actinoid charged in the target increases and the effect of thermal neutrons on fuel decreases depending on homogeneous transmutation velocity in the target. Development of stable actinoid hydride under the conditions of reactor temperature and irradiation environment is important. The experimental results of U-ZrH 1.6 are shown in this paper. The irradiation experiment using Th hydride has been proceeding. (S.Y.)

  9. Fuel design for the U.S. accelerator driven transmutation system

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hayes, S. L.; Crawford, D. C.; Pahl, R. G.; Tsai, H.

    2002-01-01

    The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed

  10. Fuel Design for the U.S. Accelerator Driven Transmutation System

    International Nuclear Information System (INIS)

    Meyer, M.K.; Hayes, S.L.; Crawford, D.C.; Pahl, R.G.; Tsai, H.

    2002-01-01

    The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed. (authors)

  11. Transmutation studies in France, R and D programme on fuels and targets

    International Nuclear Information System (INIS)

    Boidron, M.; Chauvin, N.; Garnier, J.C.; PIllon, S.; Vambenepe, G.

    2001-01-01

    For the management of high level and long-lived radioactive waste, a large and continuous research and development effort is carried out in France, to provide a wide range of scientific and technical alternatives along three lines, partitioning and transmutation, disposal in deep geological formations and long term interim surface or subsurface storage. For the line one, and in close link with the partitioning studies, research is carried out to evaluate the transmutation potential of long-lived waste in appropriate reactors configurations (scenarios) relying on current technologies as well as innovative reactors. Performed to evaluate the theoretical feasibility of the Pu consumption and waste transmutation from the point of view of the reactor cores physics to reach the equilibrium of the material fluxes (i.e. consumption = production) and of the isotopic compositions of the fuels, these studies insure the 'scientific' part of the transmutation feasibility. For the technological part of the feasibility of waste transmutation in reactors, a large programme on fuel development is underway. This includes solutions based on the advanced concepts for plutonium fuels in PWR and the development of specific fuels and targets for transmutation in fast reactors in the critical or sub-critical state. For the waste transmutation in fast reactors, an important programme has been launched to develop specific fuels and targets with experiments at various stages of preparation in different experimental reactors including Phenix. Composite fuels as well as particle fuels are considered. This programme is presented and recent results concerning the preparation of the experiments, the characterisation of the compounds properties, the thermal and mechanical modelling and the behaviour of U free fuels are given. (author)

  12. Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

    International Nuclear Information System (INIS)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-01-01

    In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

  13. Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

    International Nuclear Information System (INIS)

    Washington, J.; King, J.; Shayer, Z.

    2017-01-01

    Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO_2, Pu_3Si_2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO_2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO_2), Pu_0_._3_1ZrH_1_._6Th_1_._0_8, and PuZrO_2MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B_4C, CdO, Dy_2O_3, Er_2O_3, Eu_2O_3, Gd_2O_3, HfO_2, In_2O_3, Lu_2O_3, Sm_2O_3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO_2MgO (8 wt% Pu) target fuel with a coating of Lu_2O_3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium

  14. Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Shayer, Z., E-mail: zshayer@mines.edu [Department of Physics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

    2017-03-15

    Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO{sub 2}, Pu{sub 3}Si{sub 2}, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO{sub 2}) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO{sub 2}), Pu{sub 0.31}ZrH{sub 1.6}Th{sub 1.08}, and PuZrO{sub 2}MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B{sub 4}C, CdO, Dy{sub 2}O{sub 3}, Er{sub 2}O{sub 3}, Eu{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, HfO{sub 2}, In{sub 2}O{sub 3}, Lu{sub 2}O{sub 3}, Sm{sub 2}O{sub 3}, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO{sub 2}MgO (8 wt% Pu) target

  15. Accelerator-driven transmutation of spent fuel elements

    Science.gov (United States)

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  16. French fuel cycle strategy and partitioning and transmutation programme

    International Nuclear Information System (INIS)

    Pradel, Ph.

    2007-01-01

    The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel sources. How we deal with radioactive waste is crucial in this context. The production of nuclear energy in France has been associated, since its inception, with the optimisation of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. The public's concern regarding long-term waste management led the French government to prepare and pass the Law of December 1991, requesting in particular the study for fifteen years of solutions to minimising even further the quantity and the hazardousness of final waste, via partitioning and transmutation. At the end of these fifteen years of research, it is considered that partitioning techniques which have been validated on real solutions are at disposal. Indeed, aqueous process for separation of minor actinides from the PUREX raffinate has been brought to a point where there is reasonable assurance that industrial deployment can be successful. A key experiment has been the kilogram-scale successful trials in the CEA-Marcoule Atalante facility in 2005 and this result, together with the results obtained in the frame of the successive European projects, constitutes a considerable step forward. For transmutation, CEA has conducted programmes proving the feasibility of the elimination of minor actinides and fission products: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. Scenario studies have also allowed assessing the feasibility, at the level of cycle and fuel facilities, and the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending of

  17. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

  18. Transmutation of long-lived nuclides in the fuel cycle of Brest-type reactors

    International Nuclear Information System (INIS)

    Lopatkin, A.V.; Orlov, V.V.; Filin, A.I.

    2001-01-01

    Transmutation of long-lived nuclides produced as a result of nuclear generation, should be set up proceeding from the principle of reasonable sufficiency, expressed as radiation equivalence between the radwaste sent to disposal and source natural uranium. In this case, introduction of fast reactors of new generation (such as BREST or other reactors based on similar philosophy) will resolve transmutation problems even with the thermal-to-fast reactor capacity ratio of 2:1. The authors of the 'Strategy of nuclear power development in Russia' foresee, and substantiate their prediction, that fast reactors of the new generation will account for no less than 2/3 of nuclear capacity in future large-scale nuclear power sector. Fast reactors will be the basis of a transmutation fuel cycle, which will remove the need of creating additional transmutation facilities. (author)

  19. Development of long-lived radionuclide transmutation technology -Development of nuclear transmutation technology-

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myung Chan; Jung, Woo Tae; Koh, Duk Joon; Kim, Jung Doh; Kil, Choong Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Based on the performance assessment of current reactor nuclear design codes, CASMO-3, LEOPARD, CITATION could be used for the simulation of transmutation, but further improvements are required on the reliability of cross sections of MA or FP and the accuracy of burnup model. Our simulation results based on the calculation by using CASMO-3 and NEM-3D (developed at Seoul National University) showed that transmutation efficiency for Am was high but Np and Cm elements were found to be hard to transmute. In our calculation, micro depletion calculations with burnup variation were done separately. Possibility of MA and FP transmutation with hard and fast neutrons was reported to be greater but detail calculation will be done in next year. 44 figs, 31 tabs, 17 refs. (Author).

  20. Fuels and materials for transmutation. A status report

    International Nuclear Information System (INIS)

    2005-01-01

    The safe and efficient management of spent fuel from the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from reactors currently operating will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors. Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium. When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred to a thousand years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus be done in controlled environments having timescales of centuries rather than millennia stretching beyond 10 000 years. Transmutation is one of the means being explored to address the disposal of transuranic elements. To achieve this, advanced reactor systems, appropriate fuels, separation techniques and associated fuel cycle strategies are required. This report describes the current status of fuel and material technologies for transmutation and suggests technical R and D issues that need to be resolved. It will be of particular interest to nuclear fuel and material scientists involved in the field of partitioning and transmutation (P and T), and in advanced fuel cycles in general. (author)

  1. Transmutation Strategy Using Thorium-Reprocessed Fuel ADS for Future Reactors in Vietnam

    Directory of Open Access Journals (Sweden)

    Thanh Mai Vu

    2013-01-01

    Full Text Available Nuclear power is believed to be a key to the energy security for a developing country like Vietnam where the power demanding increases rapidly every year. Nevertheless, spent nuclear fuel from nuclear power plants is the source of radiotoxic and proliferation risk. A conceptual design of ADS utilizing thorium fuel as a based fuel and reprocessed fuel as a seed for nuclear waste transmutation and energy production is proposed as one of the clean, safe, and economical solutions for the problem. In the design, 96 seed assemblies and 84 blanket assemblies were inserted into the core to make a heterogeneous subcritical core configuration. Introducing thorium fuel into the core offers an effective way to transmute plutonium and minor actinide (MA and gain energy from this process. Transmutation rate as a function of burnup is estimated using MCNPX 2.7.0 code. Results show that by using the seed-blanket designed ADS, at 40 GWd/t burnup, 192 kg of plutonium and 156 kg of MA can be eliminated. Equivalently, 1  ADS can be able to transmute the transuranic (TRU waste from 2  LWRs. 14 units of ADS would be required to eliminate TRUs from the future reactors to be constructed in Vietnam.

  2. Development of TRU transmuters for optimization of the global fuel cycle. Final Report for the NERI Project

    International Nuclear Information System (INIS)

    Lee, John C.

    2009-01-01

    This final report summarizes the research activities during the entire performance period of the NERI grant, including the extra 9 months granted under a no-cost time extension. Building up on the 14 quarterly reports submitted through October 2008, we present here an overview of the research accomplishments under the five tasks originally proposed in July 2004, together with citations for publications resulting from the project. The AFCI-NERI project provided excellent support for two undergraduate and 10 graduates students at the University of Michigan during a period of three years and nine months. Significant developments were achieved in three areas: (1) Efficient deterministic fuel cycle optimization algorithms both for PWR and SFR configurations, (2) Efficient search algorithm for PWR equilibrium cycles, and (3) Simplified Excel-based script for dynamic fuel cycle analysis of diverse cycles. The project resulted in a total of 8 conference papers and three journal papers, including two that will be submitted shortly. Three pending publications are attached to the report

  3. Multiple Tier Fuel Cycle Studies for Waste Transmutation

    International Nuclear Information System (INIS)

    Hill, R.N.; Taiwo, T.A.; Stillman, J.A.; Graziano, D.J.; Bennett, D.R.; Trellue, H.; Todosow, M.; Halsey, W.G.; Baxter, A.

    2002-01-01

    As part of the U.S. Department of Energy Advanced Accelerator Applications Program, a systems study was conducted to evaluate the transmutation performance of advanced fuel cycle strategies. Three primary fuel cycle strategies were evaluated: dual-tier systems with plutonium separation, dual-tier systems without plutonium separation, and single-tier systems without plutonium separation. For each case, the system mass flow and TRU consumption were evaluated in detail. Furthermore, the loss of materials in fuel processing was tracked including the generation of new waste streams. Based on these results, the system performance was evaluated with respect to several key transmutation parameters including TRU inventory reduction, radiotoxicity, and support ratio. The importance of clean fuel processing (∼0.1% losses) and inclusion of a final tier fast spectrum system are demonstrated. With these two features, all scenarios capably reduce the TRU and plutonium waste content, significantly reducing the radiotoxicity; however, a significant infrastructure (at least 1/10 the total nuclear capacity) is required for the dedicated transmutation system. (authors)

  4. Multiple tier fuel cycle studies for waste transmutation

    International Nuclear Information System (INIS)

    Hill, R.N.; Taiwo, T.A.; Stillman, J.A.; Graziano, D.J.; Bennett, D.R.; Trellue, H.; Todosow, M.; Halsey, W.G.; Baxter, A.

    2002-01-01

    As part of the U.S. Department of Energy Advanced Accelerator Applications Program, a systems study was conducted to evaluate the transmutation performance of advanced fuel cycle strategies. Three primary fuel cycle strategies were evaluated: dual-tier systems with plutonium separation, dual-tier systems without plutonium separation, and single-tier systems without plutonium separation. For each case, the system mass flow and TRU consumption were evaluated in detail. Furthermore, the loss of materials in fuel processing was tracked including the generation of new waste streams. Based on these results, the system performance was evaluated with respect to several key transmutation parameters including TRU inventory reduction, radiotoxicity, and support ratio. The importance of clean fuel processing (∼0.1% losses) and inclusion of a final tier fast spectrum system are demonstrated. With these two features, all scenarios capably reduce the TRU and plutonium waste content, significantly reducing the radiotoxicity; however, a significant infrastructure (at least 1/10 the total nuclear capacity) is required for the dedicated transmutation system

  5. Fuels and targets for the transmutation of high activity long lived radioactive wastes

    International Nuclear Information System (INIS)

    Pillon, S.; Warin, D.

    2010-01-01

    The authors present and comment the different strategies which can be adopted to transmute minor actinides (concerned reactors, in fast breeder reactors, in accelerator driven systems or ADS), and the chemical composition of transmutation fuels (actinide compounds, inert matrices, fuels and targets). They describe the behaviour of refractory ceramic fuels during their service life under irradiation with their different damage origins (neutrons, fission by-products, alpha particles), the fabrication of transmutation fuels and targets through different processes (metallurgical, co-precipitate, sol-gel, wax, infiltration of radioactive materials, VIPAC/SPHEREPAC) and the reprocessing or recycling of these transmutation fuels and targets

  6. Transmutation Fuel Fabrication-Fiscal Year 2016

    Energy Technology Data Exchange (ETDEWEB)

    Fielding, Randall Sidney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grover, Blair Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    ABSTRACT Nearly all of the metallic fuel that has been irradiated and characterized by the Advanced Fuel Campaign, and its earlier predecessors, has been arc cast. Arc casting is a very flexible method of casting lab scale quantities of materials. Although the method offers flexibility, it is an operator dependent process. Small changes in parameter space or alloy composition may affect how the material is cast. This report provides a historical insight in how the casting process has been modified over the history of the advanced fuels campaign as well as the physical parameters of the fuels cast in fiscal year 2016.

  7. Fuels and targets for incineration and transmutation of actinides: the ITU programme

    International Nuclear Information System (INIS)

    Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D.

    2000-01-01

    The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

  8. Fuels and targets for incineration and transmutation of actinides: the ITU programme

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D. [Eurpean Commission, Joint Research Centre, Institute for Transuranium Elements, Kurlsruhe (Germany)

    2000-07-01

    The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

  9. Transmutation technologies to solve the problem of long-term spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Hosnedl, P.; Valenta, V.; Blahut, O.

    2000-01-01

    The paper gives a brief description of the transmutation process for actinides and long-lived fission products which are present in spent nuclear fuel. Transmutation technologies can solve the problem of long-term spent nuclear fuel storage and reduce the requirements for storage time and conditions. The basic data and requirements for the detailed design of the transmutor are summarized, and the views upon how to address the fuel purification and dry reprocessing issues are discussed. The results of activities of SKODA JS are highlighted; these include, for instance, the fluoride salt-resistant material MONICR, test loops, and electrowinners. The preliminary design of the transmutor is also outlined. Brief information regarding activities in the field of transmutation technologies in the Czech Republic and worldwide is also presented. The research and design activities to be developed for the whole design of the demonstration and basic units are summarized. It is emphasized that SKODA JS can join in international cooperation without constraints. The Attachment presents a simple assessment of how the radioactivity balance can be reduced, based on the actinide and long-lived fission product transmutation half-lives, is presented in the Attachment. (author)

  10. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

  11. Development of nuclear transmutation technology

    International Nuclear Information System (INIS)

    Park, Won Seok; Song, Tae Young; Yoo, Jae Kwon; Choi, Byung Ho; Shin, Hee Sung; Gil, Chung Sup; Kim, Jung Do.

    1997-08-01

    A basic characteristics and neutronic code development for accelerator driven subcritical reactor have been performed. In the field of basic characteristic study, the world-wide technical trends for a subcritical reactor has been investigated and some new directions for the subcritical system development were investigated. For the analysis of subcritical reactor core, a Montecarlo depletion code was developed by combining LAHET code with ORIGEN2 code. In addition, one-point kinetics equation for subcritical reactor programmed in order to analyze the dynamic behavior of subcritical core. (author). 26 tabs., 49 figs

  12. Development of nuclear transmutation technology for transuranic elements

    International Nuclear Information System (INIS)

    Mukaiyama, Takehiko

    1996-01-01

    Partitioning and Transmutation (P-T) of long-lived radioactive nuclides is conceived as the technology to improve the high-level radioactive waste management. This report discusses the incentives of P-T, generation of long-lived nuclides in fission reactors, nuclear transmutation technologies, R and D activities of the partitioning and transmutation technology development programs at JAERI and in the world. (author)

  13. Design and safety studies on the European Facility for Industrial Transmutation (EFIT) with CERMET fuel

    International Nuclear Information System (INIS)

    Chen, X.N.; Rineiski, A.; Liu, P.; Matzerath Boccaccini, C.; Flad, M.; Gabrielli, F.; Maschek, W.; Morita, K.

    2008-01-01

    European R and D for ADS design and fuel development is driven in the 6 th FP of the EU by the EUROTRANS Programme [1]. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation in an Accelerator Driven System. The EFIT development, the European Facility for Industrial Transmutation, aims at a generic conceptual design of a full transmuter. A key issue of the R and D work is the choice of an adequate fuel to be used in an Accelerator Driven Transmuter (ADT) like EFIT. Various fuel forms have been assessed. CERCER and CERMET fuels, specifically with the matrices MgO and Mo, have finally been selected and are now under closer investigation. Within EUROTRANS, a special domain named 'AFTRA', is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel data base for the core design of the EFIT. The EFIT concept has to be optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. The final recommendation on fuels by AFTRA gave a ranking of these fuels based on the mentioned criteria. The composite CERMET fuel (Pu 0.5 ,Am 0.5 )O 2-x - Mo (with the isotope 92 Mo comprising 93% of the molybdenum) has been recommended as the primary candidate for the EFIT. This CERMET fuel fulfils adopted criteria for fabrication and reprocessing, and provides excellent safety margins. Disadvantages include the cost for enrichment of 92 Mo and a lower specific transmutation rate of minor actinides, because of the higher neutron absorption cross-section of the matrix. The composite CERCER fuel (Pu 0.4 ,Am 0.6 )O 2-x - MgO has therefore been recommended as a backup solution as it might offer a higher consumption rate of minor actinides, and can be manufactured for a lower unit cost. This paper is in fact a sequel to our last paper [2

  14. On the use of spinel-based nuclear fuels for the transmutation of actinides

    International Nuclear Information System (INIS)

    Konings, R.J.M.; Bakker, K.; Boshoven, J.G.; Hein, H.; Huntelaar, M.E.; Zhang, H.; Meeldijk, J.D.; Woensdregt, C.F.

    1997-01-01

    The properties of spinel-based nuclear fuels for the transmutation of actinides are investigated. The results of laboratory experiments, thermodynamic calculations and irradiations in the High Flux Reactor (HFR) at Petten are presented, and allow us to evaluate the potential of spinel as an inert matrix for fuels and targets for transmutation. (author)

  15. Status of nuclear transmutation study

    International Nuclear Information System (INIS)

    Takizuka, Takakazu

    1999-01-01

    JAERI is carrying out R and Ds on partitioning and transmutation under the OMEGA Program. The R and Ds include the design study of accelerator-driven transmutation systems and the development of transmutation experimental facilities. Accelerator-driven systems have received much interests due to their potential role as dedicated transmuters in the nuclear fuel cycle for minimizing long-lived waste. Principles of accelerator-driven system, its history, JAERI proposed system concepts, and the experimental program are overviewed. (author)

  16. Effects of actinide compositional variability in the US spent fuel inventory on partitioning-transmutation systems

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

    1992-01-01

    Partitioning and transmutation (P-T) is an advanced waste management concept by which certain undesirable nuclides in spent fuel are first isolated (partitioned) and later destroyed (transmuted) in a nuclear reactor or other transmutation device. There are wide variabilities in the nuclide composition of spent fuel. This implies that there will also be wide variabilities in the transmutation device feed. As a waste management system, P-T must be able to accept (all) spent fuel. Variability of nuclide composition (i.e., the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure transuranic (TRU) nuclides recovered from discharged lightwater reactor (LWR) spent fuel in critical or near-critical cores. To date, all transmutation system core analyses assume invariant nuclide concentrations for startup and recycle cores. Using the US Department of Energy's (DOE's) Characteristics Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactors). The variability of the infinite multiplication factor (k ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems

  17. Development and application of new parameters for TRU transmutation effectiveness

    International Nuclear Information System (INIS)

    Han, Chi Young

    2005-02-01

    Four new parameters (incineration branching ratio, incineration rate, incineration time, and incineration buckling) have been developed to evaluate quantitatively the TRU transmutation effectiveness and applied to transmutation of uranium and TRU. From the incineration branching ratio, it is possible to analyze the main contributors to fission reaction for transmutation of a target nuclide. From the incineration rate, it is available to evaluate the transmutation effectiveness in the viewpoint of a relative incineration rate to incineration potential of a target nuclide and its family. This parameter is also used to calculate the incineration time and incineration buckling together with the incineration branching ratio. The incineration time makes it possible to discuss more practically the transmutation speed instead of the existing other parameters. The incineration buckling can be used to evaluate the time behavior of the incineration rate and also employed to support the results from the incineration time. Taking into account the transmutation effectiveness and potential of uranium and TRU derived by using the parameters and an existing neutron economy parameter, it was noted that the thermal neutron energy is very preferable from the transmutation effectiveness point of view, on the other hand the fast neutron energy is effective from the transmutation potential. Applying them to the typical critical and subcritical TRU burners, it is indicated that the critical reactor containing fertile uranium undergoes effectively the selective TRU transmutation on the present fast spectrum. It was also noted that the uranium-free subcritical reactor could be operated effectively on a little softer spectrum due to the larger neutron excess in the present spectrum. It is expected that the new parameters developed in this study and the results are directly applicable to practical transmutation reactor design, in particular accelerator-driven transmutation reactor

  18. Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

    International Nuclear Information System (INIS)

    Schitthelm, Oliver

    2012-01-01

    Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

  19. Radioactive Wastes Generated From JAERI Partitioning-Transmutation Fuel Cycle

    International Nuclear Information System (INIS)

    Nakayama, Shinichi; Morita, Yasuji; Nishihara, Kenji

    2003-01-01

    Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI's processes. Long-lived radionuclides such as 14 C and 59 Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (authors)

  20. Waste transmutation with minimal fuel cycle long-term risk

    Energy Technology Data Exchange (ETDEWEB)

    Slessarev, I.; Salvatores, M.; Uematsu, M. [Direction des Reacteurs Nucleaires, Cadarache (France)

    1995-10-01

    Hybrid systems (source-driven subcritical reactors), are investigated at CEA, mainly from a conceptual point of view, in order to assess their potential to transmute radioactive wastes (mainly long-lived fission products, LLFP) and their potential to insure a minimal long-term radiological risk related both to the fuel inventory inside the system and to the full fuel cycle (mass flows, reprocessing transport, waste disposal). The physics of these systems has been explored and work is in progress both in the field of basic data and INC code validation, in the frame of international collaborations and in the field of conceptual design studies. The most interesting feature of subcritical source-driven system is related to the possibility to obtain an {open_quotes}excess{close_quotes} of neutrons per fission, which can be used to reduce the long-term radiological risk. A specific example will be discussed here.

  1. Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

    International Nuclear Information System (INIS)

    Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

    2007-01-01

    Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

  2. Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

    International Nuclear Information System (INIS)

    Hoggett-Jones, C.; Robbins, C.; Gettinby, G.; Blythe, S.

    2002-01-01

    An inventory modelling and impact assessment system to investigate the potential effects of partitioning and transmutation is proposed. It is founded on a mass based inventory analysis using the principles of basic nuclear physics and the international standards for assessing radiological health effects. It is specific to the back-end of the nuclear fuel cycle and is applied to four alternative spent fuel management strategies. The system accounts for the dynamic nature of post-irradiation scenarios and is being used to develop software for use within the nuclear power industry. Four example waste-disposal options are considered using the method. Impact assessments and parameter sensitivity analyses are presented

  3. Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Hoggett-Jones, C. E-mail: craig@stams.strath.ac.uk; Robbins, C.; Gettinby, G.; Blythe, S

    2002-03-01

    An inventory modelling and impact assessment system to investigate the potential effects of partitioning and transmutation is proposed. It is founded on a mass based inventory analysis using the principles of basic nuclear physics and the international standards for assessing radiological health effects. It is specific to the back-end of the nuclear fuel cycle and is applied to four alternative spent fuel management strategies. The system accounts for the dynamic nature of post-irradiation scenarios and is being used to develop software for use within the nuclear power industry. Four example waste-disposal options are considered using the method. Impact assessments and parameter sensitivity analyses are presented.

  4. Description of Transmutation Library for Fuel Cycle System Analyses

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Hoffman, Edward A.

    2010-01-01

    This report documents the Transmutation Library that is used in Fuel Cycle System Analyses. This version replaces the 2008 version.(Piet2008) The Transmutation Library has the following objectives: (1) Assemble past and future transmutation cases for system analyses. (2) For each case, assemble descriptive information such as where the case was documented, the purpose of the calculation, the codes used, source of feed material, transmutation parameters, and the name of files that contain raw or source data. (3) Group chemical elements so that masses in separation and waste processes as calculated in dynamic simulations or spreadsheets reflect current thinking of those processes. For example, the CsSr waste form option actually includes all Group 1A and 2A elements. (4) Provide mass fractions at input (charge) and output (discharge) for each case. (5) Eliminate the need for either ''fission product other'' or ''actinide other'' while conserving mass. Assessments of waste and separation cannot use ''fission product other'' or ''actinide other'' as their chemical behavior is undefined. (6) Catalog other isotope-specific information in one place, e.g., heat and dose conversion factors for individual isotopes. (7) Describe the correlations for how input and output compositions change as a function of UOX burnup (for LWR UOX fuel) or fast reactor (FR) transuranic (TRU) conversion ratio (CR) for either FR-metal or FR-oxide. This document therefore includes the following sections: (1) Explanation of the data set information, i.e., the data that describes each case. In no case are all of the data presented in the Library included in previous documents. In assembling the Library, we return to raw data files to extract the case and isotopic data, into the specified format. (2) Explanation of which isotopes and elements are tracked. For example, the transition metals are tracked via the following: two Zr isotopes, Zr-other, Tc99, Tc-other, two Mo-Ru-Rh-Pd isotopes, Mo

  5. Microgamma Scan System for analyzing radial isotopic profiles of irradiated transmutation fuels

    International Nuclear Information System (INIS)

    Hilton, Bruce A.; McGrath, Christopher A.

    2008-01-01

    The U. S. Global Nuclear Energy Partnership / Advanced Fuel Cycle Initiative (GNEP/AFCI) is developing metallic transmutation alloys as a fuel form to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. A micro-gamma scan system is being developed to analyze the radial distribution of fission products, such as Cs-137, Cs-134, Ru-106, and Zr-95, in irradiated fuel cross-sections. The micro-gamma scan system consists of a precision linear stage with integrated sample holder and a tungsten alloy collimator, which interfaces with the Idaho National Laboratory (INL) Analytical Laboratory Hot Cell (ALHC) Gamma Scan System high purity germanium detector, multichannel analyzer, and removable collimators. A simplified model of the micro-gamma scan system was developed in MCNP (Monte-Carlo N-Particle Transport Code) and used to investigate the system performance and to interpret data from the scoping studies. Preliminary measurements of the micro-gamma scan system are discussed. (authors)

  6. Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

    2017-01-15

    Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

  7. Actinide nitride ceramic transmutation fuels for the Futurix-FTA irradiation experiment

    International Nuclear Information System (INIS)

    Voit, St.; McClellan, K.; Stanek, Ch.; Maloy, St.

    2007-01-01

    Full text of publication follows. The transmutation of plutonium and other minor actinides is an important component of an advanced nuclear fuel cycle. The Advanced Fuel Cycle Initiative (AFCI) is currently considering mono-nitrides as potential transmutation fuel material on account of the mutual solubility of actinide mono-nitrides as well as their desirable thermal characteristics. The feedstock is most commonly produced by a carbothermic reduction/nitridisation process, as it is for this programme. Fuel pellet fabrication is accomplished via a cold press/sinter approach. In order to allow for easier investigation of the synthesis and fabrication processes, surrogate material studies are used to compliment the actinide activities. Fuel compositions of particular interest denoted as low fertile (i.e. containing uranium) and non-fertile (i.e. not containing uranium) are (PuAmNp) 0.5 U 0.5 N and (PuAm) 0.42 Zr 0.58 N, respectively. The AFCI programme is investigating the validity of these fuel forms via Advanced Test Reactor (ATR) and Phenix irradiations. Here, we report on the recent progress of actinide-nitride transmutation fuel development and production for the Futurix-FTA irradiation experiment. Furthermore, we highlight specific cases where the complimentary approach of surrogate studies and actinide development aid in the understanding complex material issues. In order to allow for easier investigation of the fundamental materials properties, surrogate materials have been used. The amount of surrogate in each compound was determined by comparing both molar concentration and lattice parameter mismatch via Vegard Law. Cerium was chosen to simultaneously substitute for Pu, Am and Np, while depleted U was chosen to substitute for enriched U. Another goal of this work was the optimisation of added graphite during carbothermic reduction in order to minimise the duration of the carbon removal step (i.e. heat treatment under H 2 containing gas). One proposed

  8. Technology readiness of partitioning and transmutation toward closed fuel cycle in Japan

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Kurata, Masaki; Morita, Yasuji; Tsujimoto, Kazufumi; Minato, Kazuo; Koyama, Shin-ichi

    2011-01-01

    This paper treats technology readiness level (TRL) assessment of Partitioning and Transmutation (P-T) toward closed fuel cycle in JAPAN. The purpose is providing clarified information related to the current maturity of the partitioning and transmutation technologies by applying the methodology of TRL, parallel to attempting to establish common indications among relating technology area. The methodology should be one of useful communication tools between specialists and management level, and also among countries interested in the P-T technologies. The generic TRL in this study is based on the GNEP (Global Nuclear Energy Partnership)'s definition: TRL 3 shows the status that critical function is proved and elemental technologies are identified, TRL 4 represents that relating technologies are validated at bench scale in laboratory environment, and TRL 5 achieves the completion of development related to the subsystem and elemental technologies. Detailed indications are established through discussion of the relating specialists. Reviewed technological area includes P-T and minor actinide (MA) cycle: Fast Breeder Reactor (FBR) and Accelerator driven system (ADS) for MA transmutation, partitioning processes, and MA-bearing fuels. The assessments reveal that TRL spreads around TRL 3 to TRL 4 because each system requires more the development of elemental technologies. Transmutation core of FBR is assessed to be TRL 4 in that MA bearing integral test is required additionally, and ADS becomes TRL 3 because the elemental technologies were identified and the requirements were specified. Consequently, the common key issue is how the nuclear calculation methodology will be validated for MA-bearing-fuelled core, since several percentages of MA changes the void reactivity and the Doppler Effect significantly, which are inherently important in reactor safety. It should be that critical experiments with several kg of americium or more are difficult in the existing experimental

  9. Post-irradiation examinations of THERMHET composite fuels for transmutation

    Science.gov (United States)

    Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

    2003-07-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

  10. Post-irradiation examinations of THERMHET composite fuels for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Noirot, J. E-mail: jnoirot@cea.fr; Desgranges, L.; Chauvin, N.; Georgenthum, V

    2003-07-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl{sub 2}O{sub 4} spinel inert matrix and around 40% weight of UO{sub 2} to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

  11. Post-irradiation examinations of THERMHET composite fuels for transmutation

    International Nuclear Information System (INIS)

    Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

    2003-01-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2 O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour

  12. Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

    Science.gov (United States)

    Uyttenhove, W.; Sobolev, V.; Maschek, W.

    2011-09-01

    A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

  13. Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

    International Nuclear Information System (INIS)

    Uyttenhove, W.; Sobolev, V.; Maschek, W.

    2011-01-01

    A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

  14. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    2009-09-01

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

  15. Basic plan of partitioning and transmutation technology development

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo; Ozawa, Masaki

    2003-04-01

    Basic plan of partitioning and transmutation technology development has been made in more detail and concrete manner in terms of development goal, nuclides to be portioned and to be transmuted, and development schedule, based on the pre-evaluation results of the Research Evaluation Committee on Research and development of partitioning and transmutation technology for long life nuclides' held in August 2000. A step by step approach, consists of three steps, to reach the goal of partitioning and transmutation technology has been adopted under the recognition that the partitioning and transmutation technology development should be progressed steadily as a long term them. The first step is supposed to be able to attain within about 5 years by the present technology and on the extension of it. Such researches as collective separation of TRU, MA/Ln effective separation, and irradiation experiment of iodine and technetium. The second step is such a goal that is expected to be able to realize the engineering feasibility, within about 15 years, through the progress of science technology in future, although the engineering feasibility is not sufficiently foreseen at present. It will need revolutionary technology or breakthrough. Nuclides to be partitioned and to be transmuted have been selected in view points of 'radioactivity and radio-toxicity', 'geological repository', and 'effective utilization', corresponding to the each step of the development goal. Collaboration with other research organizations and with universities in the world should be pursued. Especially, such collaborations with France, with which information exchange on JOYO/PHENIX irradiation experiments is progressing, and with USA, which has recently developed positive activities in this field, are strongly expected. (author)

  16. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

    International Nuclear Information System (INIS)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-01-01

    Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

  17. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2013-07-01

    Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

  18. A review of reprocessing, partitioning, and transmutation of spent nuclear fuel and the implications for Canada

    International Nuclear Information System (INIS)

    Jackson, D.P.

    2006-01-01

    The current status of the reprocessing, partitioning, and transmutation of used nuclear fuel are reviewed in the context of assessing the possible application of these technologies to used CANDU fuel. The status of commercial reprocessing is briefly surveyed and recent progress in world R and D programs on the transmutation of FP's and actinides using Accelerator Driven Systems is summarized. The implications of reprocessing for Canada are explored from the point of view of a long strategy for managing used CANDU fuel in terms of the costs of initiating reprocessing domestically at some time in the future including public and occupational radiation doses, and the wastes generated. (author)

  19. Concept and experimental studies on fuel and target for minor actinides and fission products transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C; Guerin, Y [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Combustibles; Salvatores, M [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires; Zaetta, A [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Reacteurs

    1994-12-31

    High activity long-lived radionuclides in nuclear wastes, namely minor actinides (americium and neptunium) are in large amount generated by current nuclear reactive. The destruction of these radionuclides is a part of the French SPIN (Partitioning and Burning) program consistent with the determination to send a minimum amount of harmful products for final storage. Transmutation concepts are defined for neptunium and americium taking into account fuel cycle strategies. Neptunium destruction does not pose any major problems. It`s a by-product of uranium consumption, as plutonium and in despite of a slight gamma activity due to the protactinium 233 it`s quite easy to handle. Diluting neptunium in the mixed oxide fuels (MOX) should not be an obstacle for fabrication, in-pile behaviour and reprocessing either. Consequently we make the proposal of homogeneous mode of neptunium in MOX which should be soon explored in the experimental OSIRIS reactor and in the Phenix and Superphenix reactors. The analysis is more complex for the multi isotope americium. Its destruction is difficult because of gamma radioactivity which complicates fabrication. Experiments in Phenix and calculation showed that Phenix reactor offers a good potential for americium incineration, but similar data do not exist for PWR. It will remain a well known difficulty for fabrication and reprocessing. In this case we have to put a real new face to the fabrication flow-sheet of americium compounds and we propose to develop the heterogeneous mode. Targets choice are defined in term of: -safety, considering fuel reaction with cladding and water sodium, -transmutation rate, limited by target behaviour, in FR`s (Phenix), PWR`s (OSIRIS) and HFR (Petten), -reprocessing, checking the solubility of such targets by Purex process. So, at the beginning of our program the account has been on improving fuel and targets properties related to safety and fuel cycle. (authors). 4 figs.

  20. Development of a fast reactor for minor actinides transmutation - (1) Overview and method development - 5092

    International Nuclear Information System (INIS)

    Takeda, T.; Usami, S.; Fujimura, K.; Takakuwa, M.

    2015-01-01

    The Ministry of Education, Culture, Sports, Science and Technology in Japan has launched a national project entitled 'technology development for the environmental burden reduction' in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmutation and volume reduction of minor actinides (MA) with long-lived radioactivity and high decay heat contained in high level radioactive wastes by using sodium cooled fast reactors. We are developing MA transmutation core concepts which harmonize efficient MA transmutation with core safety. To accurately design the core concepts we have improved calculation methods for estimating the transmutation rate of individual MA nuclides, and estimating and reducing uncertainty of MA transmutation. The overview of the present project is first described. Then the method improvement is presented with numerical results for a minor-actinide transmutation fast reactor. The analysis is based on Monju reactor data. (authors)

  1. Transmutation research and fuel cycle (report on discussion at Research Reactor Institute, Kyoto University)

    International Nuclear Information System (INIS)

    Yamana, Hajimu

    1999-01-01

    A symposium was held on a topic of 'Transmutation Research' on Dec. 21 and 22, 1999 at Research Reactor Institute, Kyoto University. This meeting was held as a joint-meeting of KUR's specialist meeting and Tokyo University's activity supported by the Grant-in-Aid for Scientific Research of Ministry of Education, Sport and Culture of Japan. This paper describes the overview of the discussions of this joint-meeting, and interprets their significance. Major themes discussed are, needed discussions on the transmutation research, policy and concepts of the organizations doing transmutation researches, a view from university side, transmutation researches in the oversea countries, opinions from various standpoints of the nuclear fuel cycle, conclusive discussions. 'the meanings of the transmutation research should be discussed together with the geological disposal and fast reactor system', 'transmutation may be a cooperative option for the disposal, thus, they should not be in a independent relation', and Balance evaluation will be needed' are the examples of the conclusive remarks of this meeting. (author)

  2. Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

    International Nuclear Information System (INIS)

    Hong, Seong Hee; Kim, Myung Hyun

    2016-01-01

    DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

  3. Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

  4. Results from the characterisation of the Futurix-FTA metal alloy transmutation fuels

    International Nuclear Information System (INIS)

    Rory Kennedy, J.; O'Holleran, Th.; Keiser, D.

    2007-01-01

    Full text of publication follows. Idaho National Laboratory has been developing and irradiation testing a number of fuels and fuel types for actinide transmutation as part of the Advanced Fuel Cycle Initiative (AFCI). Fuel types under consideration include both fertile (fast reactor systems) and fertile-free (accelerator-driven systems) metallic alloys. Most recently, fuel fabrication was completed and the fuel pins shipped to the fast flux Phenix reactor in Marcoule, France for irradiation testing as part of the FUTURIX-FTA experiment: an international experiment involving the USA, France, the European Commission and Japan. The metal alloy fuels for this experiment are the low-fertile U-29Pu-4Am-2Np-30Zr and the non-fertile Pu-12Am-40Zr. The fresh fuels have been fully characterised for chemical composition, phase, microstructure, thermal behaviour and fuel-cladding-chemical-interaction (FCCI). Preliminary FCCI results raised some safety concerns with respect to the formation of low melting phases and cladding degradation, which could preclude a fuel from consideration. Results from diffusion couple experiments between the non-fertile fuel Pu-12Am-40Zr and the ferritic HT9 and 422 stainless steels (SS) used in the AFC experiments in the ATR reactor (USA) compared to the austenitic AIM1 SS used in the FUTURIX-FTA experiments in the Phenix reactor (France) indicate significant inter-diffusion with the AIM1 SS. Up to about a 30-fold increase in the diffusion of iron (and accompanying Ni and Cr) into the fuel at 650 C was observed compared to the 422 SS studies. Comparable studies between the low-fertile U-29Pu-4Am-2Np-30Zr fuel alloy and the AIM1 SS show virtually no inter-diffusion. The Fe (along with small amounts of Ni and Cr) appears as small precipitates in the fuel alloy with only minor concentrations identified in the fuel alloy matrix. These results will be discussed in terms of mechanisms of the inter-diffusion and the difference in behaviour between the

  5. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

    International Nuclear Information System (INIS)

    Rodríguez, Iván Merino; Álvarez-Velarde, Francisco; Martín-Fuertes, Francisco

    2014-01-01

    Highlights: • Four fuel cycle scenarios have been analyzed in resources and economic terms. • Scenarios involve Once-Through, Pu burning, and MA transmutation strategies. • No restrictions were found in terms of uranium and plutonium availability. • The best case cost and the impact of their uncertainties to the LCOE were analyzed. - Abstract: Four European fuel cycle scenarios involving transmutation options (in coherence with PATEROS and CP-ESFR EU projects) have been addressed from a point of view of resources utilization and economic estimates. Scenarios include: (i) the current fleet using Light Water Reactor (LWR) technology and open fuel cycle, (ii) full replacement of the initial fleet with Fast Reactors (FR) burning U–Pu MOX fuel, (iii) closed fuel cycle with Minor Actinide (MA) transmutation in a fraction of the FR fleet, and (iv) closed fuel cycle with MA transmutation in dedicated Accelerator Driven Systems (ADS). All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for 200 years, looking for long term equilibrium mass flow achievement. The simulations were made using the TR E VOL code, capable to assess the management of the nuclear mass streams in the scenario as well as economics for the estimation of the levelized cost of electricity (LCOE) and other costs. Results reveal that all scenarios are feasible according to nuclear resources demand (natural and depleted U, and Pu). Additionally, we have found as expected that the FR scenario reduces considerably the Pu inventory in repositories compared to the reference scenario. The elimination of the LWR MA legacy requires a maximum of 55% fraction (i.e., a peak value of 44 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation) or an average of 28 units of ADS plants (i.e., a peak value of 51 ADS units). Regarding the economic analysis, the main usefulness of the provided economic results is for relative comparison of

  6. Development of high level radwaste treatment and conversion technology. Transmutation technology development

    Energy Technology Data Exchange (ETDEWEB)

    Park, Won Suk; Song, T Y; Kim, Y H

    2001-03-01

    The final disposition of spent fuel has been, and continues to be, an important issue of nuclear industry. The conceptual design for the accelerator driven transmutation system HYPER is scheduled to be completed by the year of 2006. As the first step for the conceptual design, a study to determine 1. sub-critical core characteristics, 2. fuel concept, 3. coolant system concept, 4. spallation target concept for the HYPER was performed from 1997 to 2000. Most of concept studies were done on the literature basis. The major objectives of the study is to give a guid-line for the second stage research which will be performed during 2001-2003. In addition, the technologies related with TRU-Zr fuel and Pb-Bi coolant can be utilized for the future nuclear reactor development such as generation 4.

  7. Development of high level radwaste treatment and conversion technology. Transmutation technology development

    International Nuclear Information System (INIS)

    Park, Won Suk; Song, T. Y.; Kim, Y. H.

    2001-03-01

    The final disposition of spent fuel has been, and continues to be, an important issue of nuclear industry. The conceptual design for the accelerator driven transmutation system HYPER is scheduled to be completed by the year of 2006. As the first step for the conceptual design, a study to determine 1. sub-critical core characteristics, 2. fuel concept, 3. coolant system concept, 4. spallation target concept for the HYPER was performed from 1997 to 2000. Most of concept studies were done on the literature basis. The major objectives of the study is to give a guid-line for the second stage research which will be performed during 2001-2003. In addition, the technologies related with TRU-Zr fuel and Pb-Bi coolant can be utilized for the future nuclear reactor development such as generation 4

  8. The advanced fuel cycle initiative: the future path for advanced spent fuel treatment and transmutation research in the United States

    International Nuclear Information System (INIS)

    Herczeg, J.W.

    2003-01-01

    The U. S. Department of Energy (DOE) has invested over USD 100 million in transmutation research and development over the past three years. The programme has evolved from an accelerator based transmutation programme to a multi-tier reactor and accelerator based programme. These changes have resulted in a significant re-focus of the research and development programme as well as a name change to reflect the new direction. The Advanced Accelerator Application (AAA) programme is now renamed the Advanced Fuel Cycle Initiative (AFCI). Research completed by the AAA programme in Fiscal Year 2002 points to a multi-phased AFCI Programme consisting of two elements that would be conducted in parallel as part of an integrated research effort: an intermediate-term technology element (AFCI Series One), which emphasises advanced technical enhancements to the current commercial nuclear power infrastructure; and a long term technology element (AFCI Series Two), which will require the introduction of next-generation nuclear energy systems to reduce the toxicity of nuclear waste. (author)

  9. Development of fluoride reprocessing technology for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    Uhlir, J.; Hosnedl, P.; Matal, O.

    2000-01-01

    At present, the transmutation of spent nuclear fuel is considered a prospective alternative conception with respect to the current conception based on the non-reprocessed spent fuel disposal into a deep geological repository. The Czech research and development programme in the area of partitioning is directed primarily on the development of the fuel cycle technology for the accelerator - driven subcritical reactor with a liquid fuel based on fluoride melts. The final objective of the research programme is the development of pyrochemical technologies suitable for a continuous or semi-continuous separation process which would allow practically perfect utilization of the transmutation potentialities of the reactor system. The present research is directed particularly on the development of suitable fluoride separation methods the target of which is the removal of the uranium component from spent nuclear fuel and on the research of the electro-separation procedures and further on the development of appropriate construction materials and equipment for the technology of fluoride salt melts. (authors)

  10. The sphinx project: experimental verification of design inputs for a transmuter with liquid fuel based on molten fluorides

    International Nuclear Information System (INIS)

    Hron, M.; Uhlir, J.; Vanicek, J.

    2002-01-01

    The current proposals for high-active long-lived (more then 10 4 years) waste from spent nuclear fuel disposal calls forth an increasing societal mistrust towards nuclear power. These problems are highly topical in the Czech Republic, a country which is operating nuclear power and accumulating spent fuel from PWRs and is further located on an inland and heavily populous Central European region. The proposed project, known under the acronym SPHINX (SPent Hot fuel Incineration by Neutron flux) deals with a solution to some of the principle problems through a very promising means of radioactive waste treatment. In particular, high-level wastes from spent nuclear fuel could be treated using this method, which is based on the transmutation of radionuclides through the use of a nuclear reactor with liquid fuel based on molten fluorides (Molten Salt Transmutation Reactor - MSTR) which might be a subcritical system driven by a suitable neutron source. Its superiority also lies in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect. After the first three-year stage of Research and Development which has been focused mostly on computer analyses of neutronics and corresponding physical characteristics, the next three-year stage of this programme will be devoted to experimental verification of inputs for the design of a demonstration transmuter using molten fluoride fuel. The Research and Development part of the SPHINX project in the area of fuel cycle of the MSTR is focused in the first place on the development of suitable technology for the preparation of an introductory liquid fluoride fuel for MSTR and subsequently on the development of suitable fluoride pyrometallurgical technology for the separation of the transmuted elements from the non-transmuted ones. The idea of the introductory fuel preparation is based on the reprocessing of PWR spent fuel using the Fluoride Volatility Method

  11. Development of neutron-transmutation-doped germanium bolometer material

    International Nuclear Information System (INIS)

    Palaio, N.P.

    1983-08-01

    The behavior of lattice defects generated as a result of the neutron-transmutation-doping of germanium was studied as a function of annealing conditions using deep level transient spectroscopy (DLTS) and mobility measurements. DLTS and variable temperature Hall effect were also used to measure the activation of dopant impurities formed during the transmutation process. In additioon, a semi-automated method of attaching wires on to small chips of germanium ( 3 ) for the fabrication of infrared detecting bolometers was developed. Finally, several different types of junction field effect transistors were tested for noise at room and low temperature (approx. 80 K) in order to find the optimum device available for first stage electronics in the bolometer signal amplification circuit

  12. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  13. Recent research and development activities on partitioning and transmutation of radioactive nuclides in Japan

    International Nuclear Information System (INIS)

    Minato, K.; Ikegami, T.; Inoue, T.

    2005-01-01

    In Japan, research and development activities for partitioning and transmutation (P and T) have been promoted under the OMEGA programme for more than 15 years. These activities were reviewed by the Atomic Energy Commission in Japan in 2000. In accordance with the results of the review, three institutes, the Japan Atomic Energy Research Institute (JAERI), the Japan Nuclear Cycle Development Institute (JNC) and the Central Research Institute of Electric Power Industry (CRIEPI), are continuing the research and development on the P and T technology. This report summarises the recent activities in Japan by these institutes. JAERI is engaging in the research and development on the Double-strata Fuel Cycle concept consisting of the partitioning process of the high-level waste and the dedicated transmutation cycle using the accelerator driven system (ADS) fuelled with the minor actinide (MA) nitride fuel. JNC and CRIEPI are engaging in the research and development on the P and T technology using commercialized fast reactors (FR), where JNC is mainly in charge of the MOX fuel and the aqueous reprocessing, while CRIEPI is mainly in charge of the metallic fuel and the dry reprocessing. The research and development activities on FR are organised under the Feasibility Study on Commercialized Fast Reactor Cycle Systems. (authors)

  14. International development within the spent nuclear fuel cycle

    International Nuclear Information System (INIS)

    Aggeryd, I.; Broden, K.; Gelin, R.

    1990-06-01

    The report gives a survey of the newest international development of the fuel processing and the spent nuclear fuel cycle. The transmutation technology of long lived nuclides is discussed in more details. (K.A.E)

  15. Calculation characterization of spent fuel hazard related to partitioning and transmutation of minor actinides and fission products

    International Nuclear Information System (INIS)

    Gerasimov, A. S.; Bergelson, B. R.; Tikhomirov, G.V.; Volovik, A.I. . E-mail of corresponding author: geras@itep.ru; Gerasimov, A.S.)

    2005-01-01

    Radiotoxicity is one of important characteristics of radwaste hazard. Radiotoxicity of actinides and fission products from spent fuel of VVER-1000 reactor for processes of burnup, long-term storage, and transmutation is discussed. (author)

  16. The role of Z-pinch fusion transmutation of waste in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Smith, James Dean; Drennen, Thomas E.; Rochau, Gary Eugene; Martin, William Joseph; Kamery, William; Phruksarojanakun, Phiphat; Grady, Ryan; Cipiti, Benjamin B.; Wilson, Paul Philip Hood; Mehlhorn, Thomas Alan; Guild-Bingham, Avery; Tsvetkov, Pavel Valeryevich

    2007-01-01

    The resurgence of interest in reprocessing in the United States with the Global Nuclear Energy Partnership has led to a renewed look at technologies for transmuting nuclear waste. Sandia National Laboratories has been investigating the use of a Z-Pinch fusion driver to burn actinide waste in a sub-critical reactor. The baseline design has been modified to solve some of the engineering issues that were identified in the first year of work, including neutron damage and fuel heating. An on-line control feature was added to the reactor to maintain a constant neutron multiplication with time. The transmutation modeling effort has been optimized to produce more accurate results. In addition, more attention was focused on the integration of this burner option within the fuel cycle including an investigation of overall costs. This report presents the updated reactor design, which is able to burn 1320 kg of actinides per year while producing 3,000 MWth

  17. Two-step B/T (burning and/or transmutation) method for self-completed nuclear fuel cycle with thermal and fast B/T reactors

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto, M.R.; Marsodi, M.R.

    1995-01-01

    The total cost minimization for P and T (partitioning and transmutation) treatment with appropriate recycle period through out-core optimization was examined in order to find the possibility of P and T treatment of minor actinides (MA) and/or long lived fission products (LLFP) and the technology to be improved and/or developed in self-completed nuclear fuel cycle. The P and T should be done for B/T (burning and/or transmutation) treatment based on three criteria, and the grouping was closely related to the effectiveness of Two-Step B/T Method in B/T treatment. (authors)

  18. HILBILAC development for accelerator-driven transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Pirozhenko, V.; Plink, O. [Moscow Radiotechnical Institute (Russian Federation)

    1995-10-01

    High-Intensity Low-Beta Ion Linac (HILBILAC) is intended for acceleration of ion beams with current of about 1 A and higher. The CW HILBILAC with beam current of 2l50 mA is under development at MRTI. Concept of parameters choice is presented along with results of beam dynamics and resonator parameters calculations. A pulse prototype HILBILAC-TEST will have to be constructed and tested for the CW accelerator development, its scheme and parameters are presented.

  19. Synthesis of the studies on fuels and transmutation targets (fabrication, design, irradiation damage and dissolution) realized in the framework of the Bataille law

    International Nuclear Information System (INIS)

    Pillon, S.

    2004-12-01

    This document presents the different studied fuels and targets for the transmutation of the minor actinides and of the long life fission products for PWR/EPR and Fast neutron Reactor/EFR of today technology; the results of studies on the behavior under ions irradiation and in experimental nuclear reactor; the knowledge in terms of design, simulation and sizing; the development in terms of fabrication; the knowledge on the dissolution aptitude of these fuels and targets. (A.L.B.)

  20. U.S. advanced accelerator applications program: plans to develop and test waste transmutation technologies

    International Nuclear Information System (INIS)

    Van Tuyle, G.; Bennett, D.; Arthur, E.; Cappiello, M.; Finck, P.; Hill, D.; Herczeg, J.; Goldner, F.

    2001-01-01

    The primary mission of the U.S. Advanced Accelerator Applications (AAA) Program is to establish a national nuclear technology research capability that can demonstrate accelerator-based transmutation of waste and conduct transmutation research while at the same time providing a capability for the production of tritium if required. The AAA Program was created during fiscal year 2001 from the Accelerator Transmutation of Waste (ATW) Program and the Accelerator Production of Tritium (APT) Project. This paper describes the new AAA Program, as well as its two major components: development and testing of waste transmutation technologies and construction of an integrated accelerator-driven test facility (ADTF). (author)

  1. Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

    International Nuclear Information System (INIS)

    Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

    2018-01-01

    Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

  2. Effects of actinide compositional variability in the U.S. spent fuel inventory on partitioning-transmutation systems

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

    1993-01-01

    The partitioning and transmutation concept (P-T) has as a mission the reduction by many orders of magnitude of certain undesirable nuclides in the waste streams. Given that only a very small fiction of spent fuel can be rejected by a P-T enterprise, a P-T system must therefore be capable of accommodating a wide range of spent fuel characteristics. Variability of nuclide composition (i.e. the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure TRU nuclides recovered from discharged LWR fuel in critical or near-critical cores. To date, all transmutation system core analyses assume nonvariable nuclide concentrations for startup and recycle cores. Using the Department of Energy (DOES) Characteristic Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactor). In addition to quantifying the variability of nuclide composition in current and projected LWR fuel discharge, the variability of the infinite multiplication factor (K ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems. It is shown that actinide compositional variations are potentially significant and warrant further investigation. (authors)

  3. Design considerations and evaluations of an accelerator-driven fluid fuel transmuter

    Energy Technology Data Exchange (ETDEWEB)

    Lizana, P.; Lypsch, F.; Phlippen, P.W. [Institute for Safety Research and Reactor Technology, Juelich (Germany)

    1995-10-01

    A fluid fuel transmuter is proposed on the basis of circulating lead forming the fluid carrier material for long-lived actinides. Thermalization of neutrons is achieved by the use of graphite in the blanket leading to low actinide concentrations, typically around 100 g/l. An eigenvalue of 0.95 is aimed at and the extraneous source neutrons are provided by the interaction of 1.6 GeV protons with a central lead target (spallation process). Fuel depletion and neutron transport calculations are discussed with a view to the technical feasibility and possible advantageous design modifications.

  4. Fuel and target programs for the transmutation at Phenix and other reactors

    International Nuclear Information System (INIS)

    Gaillard-Groleas, G.

    2002-01-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  5. Minor actinide transmutation in accelerator driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike [IANUS, TU Darmstadt (Germany)

    2015-07-01

    Transmutation of radioactive waste, the legacy of nuclear energy use, gains rising interest. This includes the development of facilities able to transmute minor actinides (MA) into stable or short-lived isotopes before final disposal. The most common proposal is to use a double-strata approach with accelerator-driven-systems (ADS) for the efficient transmutation of MA and power reactors to dispose plutonium. An ADS consists of a sub-critical core that reaches criticality with neutrons supplied by a spallation target. An MCNP model of the ADS system Multi Purpose Research Reactor for Hightech Applications will be presented. Depletion calculations have been performed for both standard MOX fuel and transmutation fuel with an increased content of minor actinides. The resulting transmutation rates for MAs are compared to published values. Special attention is given to selected fission products such as Tc-99 and I-129, which impact the radiation from the spent fuel significantly.

  6. Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

  7. Concept of a subcritical transmutation system with fast neutron spectrum and liquid fuel

    International Nuclear Information System (INIS)

    Tittelbach, S.

    2002-11-01

    The annual amount of nearly 9500 t of spent fuel from worldwide industrial nuclear energy utilization has to be disposed as high level waste. The retention of nuclear waste from the biosphere has to be assured until the radiological risk decreases to tolerable levels. The long-term radiological risk of spent fuel is dominated by actinide elements, i.e. plutonium, americium and curium. It is intended to reduce this amount of high level waste by Partitioning and Transmutation, so that the radiotoxicity of the disposed waste falls short of the reference value of fresh fuel decaying naturally after about thousand years. For this time period the retention of high level waste can be assured by technical means. The scope of this work is the design of a subcritical fast transmutation system with liquid metal cooling and liquid metal fuel. The lead bismuth eutectic has been choosen as the liquid metal coolant and fuel carrier. To dissolve at least 3 at% of transuran elements, a minimum fuel temperature of 600 C is required. The calculations were carried out with a fuel composition, which results from two plutonium recycling steps in a thorium fuel cycle. Two homogeneous and two heterogeneous blankets have been designed and evaluated leading to one preferred heterogeneous blanket design, which has been investigated in more detail. This blanket design merges the positive properties of a solid fuel system (better control of fuel and reactivity because of smaller and closed fuel volumina) and a liquid fuel system (continous charge and discharge or extraction of fission products). The blanket design is based on the core design of fast breeder liquid metal reactors. It consists of hexagonal fuel elements housing up to six annular shaped fuel cylinders. The hexagonal shape of the fuel elements leads to three fuel zones positioned concentrically around the central spallation target. There is a strong heterogeneous distribution of power and heat flux in this blanket design. Besides

  8. Reprocessing of gas-cooled reactor particulate graphite fuel in a multi-strata transmutation system

    International Nuclear Information System (INIS)

    Laidler, J.J.

    2001-01-01

    Spent nuclear fuel discharged for light water reactors (LWRs) contains significant quantities of plutonium and other transuranic elements. Recent practice in Europe and Japan has been to recover the plutonium from spent fuel and recycle it to LWRs in the form of mixed uranium-plutonium oxide (MOX) fuel. Irradiation of the recycle fuel results in the generation of further plutonium and an increase in the isotopic concentration of the higher isotopes of plutonium, those having much lover fission cross sections than 239 Pu. This restricts plutonium recycle to one or two cycles, after which use of the plutonium becomes economically unfavorable. Recycle of the highly-transmuted plutonium in fast spectrum reactors can be an efficient method of fissioning this plutonium as well as other minor transuranics such as neptunium, americium and perhaps even curium. Those minor transuranics that are not conveniently burned in a fast reactor can be sent to an accelerator driven subcritical transmutation device for ultimate destruction. The preceding describes what has become known as a 'dual strata' or 'multi-strata' system. It is driven by the incentives to realize the maximum amount of energy from nuclear fuel and to eliminate the discharge of radio-toxic transuranic elements to the environment. Its implementation will be dependent in the long run upon the economic viability of the system and on the value placed by society on the elimination of radio-toxic materials that can conceivably be used in the manufacture of weapons of mass destruction. (author)

  9. Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A., E-mail: abanades@etsii.upm.es [ETSII/Universidad Politecnica de Madrid, J.Gutierrez Abascal, 2-28006 Madrid (Spain); Garcia, C.; Garcia, L. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba); Escriva, A.; Perez-Navarro, A. [Instituto de Ingenieria Energetica, Universidad Politecnica de Valencia, C.P. 46022 Valencia (Spain); Rosales, J. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba)

    2011-06-15

    Highlights: > Utilization of Accelerator Driven System (ADS) for Hydrogen production. > Evaluation of the potential use of gas-cooled ADS for a sustainable use of Uranium resources by transmutation of nuclear wastes, electricity and Hydrogen production. > Application of the Sulfur-Iodine thermochemical process to subcritical systems. > Application of CINDER90 to calculate burn-up in subcritical systems. - Abstract: The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45-46%, either in the form of Hydrogen, electricity, or both.

  10. Study on partitioning and transmutation as a possible option for spent fuel management within a nuclear phase-out scenario

    Energy Technology Data Exchange (ETDEWEB)

    Fazion, C.; Rineiski, A.; Salvatores, M.; Schwenk-Ferrero, A.; Romanello, V.; Vezzoni, B.; Gabrielli, F. [Karlsruhe Institute of Technology - KIT, Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2013-07-01

    Most Partitioning and Transmutation (PT) studies implicitly presuppose the continuous use of nuclear energy. In this case the development of new facilities or the modification of the fuel cycle can be justified in the long-term as an important feature in order to improve sustainability by minimizing radioactive waste and reducing the burden at waste disposal. In the case of a country with nuclear energy phase-out policy, the PT option might have also an important role for what concerns the final disposal strategies of the spent fuel. In this work three selected scenarios are analyzed in order to assess the impact of PT implementation in a nuclear energy phase out option. The scenarios are: -) Scenario 1: Identification of Research/Development activities needs for a technological development of PT while postponing the decision of PT implementation; -) Scenario 2: Isolated application of PT in a phase-out context; and -) Scenario 3: Implementation of PT in a European context. In order to facilitate the discrimination among the 3 scenarios, a number of figures of indicators have been evaluated for each scenario. The selected indicators are: the mass of High Level Waste (HLW), Uranium inventory, thermal output of HLW, Radiotoxicity, Fuel cycle secondary waste associated to the PT operation, and Facility capacity/number requirements. The reduction, due to PT implementation, of high level wastes masses and their associated volumes can be significant. For what concerns the thermal output and radiotoxicity a significant impact can be also expected. However, when assessing these two indicators the contribution coming from already vitrified waste should also not be neglected. Moreover, an increase of secondary waste inventory is also expected. On the contrary, the increase of fission product inventories due to the operation of the transmutation system has a relatively limited impact on the fuel cycle.

  11. Partitioning and transmutation. Current developments - 2007. A report from the Swedish reference group on P-T-research

    International Nuclear Information System (INIS)

    Ahlstroem, Per-Eric; Blomgren, Jan; Eriksson, Marcus; Seltborg, Per; Wallenius, Jan; Westlen, Daniel

    2007-06-01

    This report is written on behalf of the Swedish reference group for research on partitioning and transmutation. The reference group has been assembled by SKB and its members represent the teams that are active in this field at Swedish universities. The present report summarises the progress in the field through the years 2004-2006. A prerequisite for transmutation by irradiation with neutrons is that the nuclides to be transmuted are separated (partitioned) from the other nuclides in the spent fuel. In particular the remaining uranium must be taken away unless you want to produce more plutonium and other transuranium elements. Separation of the various elements can at least in principle be achieved by mechanical and chemical processes. Currently there exist some large scale facilities for separation of uranium and plutonium from the spent fuel-reprocessing plants. These can, however, not separate the minor actinides - neptunium, americium and curium - from the high level waste that goes to a repository. Plutonium constitutes about 90% of the transuranium elements in fuel from light water reactors. The objective of current research on partitioning is to find and develop processes suitable for separation of the heavier actinides (and possibly some long-lived fission products) on an industrial scale. The objective of current research on transmutation is to define, investigate and develop facilities that may be suitable for transmutation of the aforementioned long-lived radionuclides. The research on partitioning has made important progress in recent years. In some cases one has succeeded to separate americium and curium. Many challenges remain however. Within hydrochemistry one has achieved sufficiently good distribution and separation factors. The focus turns now towards development of an operating process. The search for ligands that give sufficiently good extraction and separation will continue but with less intensity. The emphasis will rather be on improving

  12. Partitioning and transmutation. Current developments - 2007. A report from the Swedish reference group on P-T-research

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstroem, Per-Eric (ed.) [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Blomgren, Jan [Uppsala Univ. (Sweden). Dept. of Neutron Research; Ekberg, Christian; Englund, Sofie; Fermvik, Anna; Liljenzin, Jan-Olov; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden); Eriksson, Marcus; Seltborg, Per; Wallenius, Jan; Westlen, Daniel [Royal Inst. of Technology, Stockholm (Sweden)

    2007-06-15

    This report is written on behalf of the Swedish reference group for research on partitioning and transmutation. The reference group has been assembled by SKB and its members represent the teams that are active in this field at Swedish universities. The present report summarises the progress in the field through the years 2004-2006. A prerequisite for transmutation by irradiation with neutrons is that the nuclides to be transmuted are separated (partitioned) from the other nuclides in the spent fuel. In particular the remaining uranium must be taken away unless you want to produce more plutonium and other transuranium elements. Separation of the various elements can at least in principle be achieved by mechanical and chemical processes. Currently there exist some large scale facilities for separation of uranium and plutonium from the spent fuel-reprocessing plants. These can, however, not separate the minor actinides - neptunium, americium and curium - from the high level waste that goes to a repository. Plutonium constitutes about 90% of the transuranium elements in fuel from light water reactors. The objective of current research on partitioning is to find and develop processes suitable for separation of the heavier actinides (and possibly some long-lived fission products) on an industrial scale. The objective of current research on transmutation is to define, investigate and develop facilities that may be suitable for transmutation of the aforementioned long-lived radionuclides. The research on partitioning has made important progress in recent years. In some cases one has succeeded to separate americium and curium. Many challenges remain however. Within hydrochemistry one has achieved sufficiently good distribution and separation factors. The focus turns now towards development of an operating process. The search for ligands that give sufficiently good extraction and separation will continue but with less intensity. The emphasis will rather be on improving

  13. Fuel and target programs for the transmutation at Phenix and other reactors; Programmes combustibles et cibles pour la transmutation dans Phenix et autres reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard-Groleas, G

    2002-07-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  14. Fuel cycle of fast reactor Brest with non-proliferation, transmutation of long-lived nuclides and equivalent disposal of radioactive waste

    International Nuclear Information System (INIS)

    Lopatkin, A.V.; Orlov, V.V.

    2001-01-01

    The declared objectives in the fuel cycle of fast reactor BREST achieved by the following measures. Proliferation resistance of the fuel cycle being developed for BREST reactors is provided along two lines: reactors physics and design features; spent fuel reprocessing technology excluding plutonium separation at all process stages. Surplus neutrons produced in a chain reaction in a fast reactor without uranium blanket and the high flux of fast neutrons, allow efficient transmutation of not only all actinides in the core but also long-lived fission products (I, Te) in lead blanket by leakage neutrons without detriment to the inherent safety of this reactor. (author)

  15. Studies on the safety and transmutation behaviour of innovative fuels for light water reactors; Untersuchungen zum Sicherheits- und Transmutationsverhalten innovativer Brennstoffe fuer Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schitthelm, Oliver

    2012-07-01

    Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its {sup 238}U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

  16. Advanced orient cycle, for strategic separation, transmutation and utilization of nuclides in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Ozawa, M.; Fujita, R.; Koyama, S.; Suzuki, T.; Fujii, Y.

    2007-01-01

    which enables to directly recover pure Cm as well as pure Am with minimum number of reprocessing separation steps is reported in another paper. The recent experiments indicated that strong adsorption of 1 06Ru and 1 25Sb was observed under the diluted HCl medium, thereby completely 1 06Ru-free feed dissolver solution was obtained. The CEE separation step will follow this IX step for further purification and fabrication of RMFP material for their utilization. Based on those technologies, the Trinitarian Research and Development project (Advanced ORIENT Cycle) on partitioning, transmutation and utilization of actinides and fission products will be developed to realize ultimate reducing long-term radio toxicity in the radioactive wastes. Actinides, LLFP ( 1 35Cs, etc), MLFP ( 9 0Sr, 1 37Cs) and RMFP shall be separated to the level of isotope as well as element. The CEE process will be added for utilization of RMFP. The RMFP, one of the products of Ad. ORIENT Cycle, would be expected to be a 'FP-catalyst' to circulate between nuclear and hydrogen / fuel cell energy systems, and thereby contributing to save the natural precious metal resources

  17. Actinide and fission product separation and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

  18. Safety assesment on radioactive waste from the partitioning and transmutation fuel cycle

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Kim, Sung Gi; Park, Won Suk

    2000-12-01

    A preliminary study on the quantitative effect of the partition and transmutation on the permanent disposal of HLW, which means the spent fuel in view of current Korean situation, was carried out. Two approaches in quantitative way are considered to be available for evaluating the deterministic influence of P and T strategy on the long-term disposal of this HLW are assessments of waste toxicity indices (TIs) and the repository performance assessments (PAs). TI is measures of the intrinsic radiotoxicity of the wastes and does not incorporate any detailed consideration of the feature, event and processes (FEPs) which might be lead to the release of the nuclides from the waste disposed of in the repository and the transport to and through the biosphere. Whereas, PA, which treated as main topic of present study, does include consideration of such FEPs even though it could not fully comprehensive at the current stage of R andD on geological disposal. Through the study, after reviewing the PA approaches which considered by some countries, relative advantages in case P and T will be performed before disposal over direct permanent disposal. Even though P and T could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and even minimize the repository area through the proper handling of nuclides whose decay heat is significant and further produce the electricity, it should overcome the such major disadvantages as problems technically exposed during developing and improving the P and T system, economic point of view, and public acceptance in view of environment-friendly issues. In this regard some relevant issues are also discussed to show the direction for further studies

  19. Monte Carlo method in ADS transmutation reactor coolant and the research of optimal placement of the fuel

    International Nuclear Information System (INIS)

    Niu Yunlong; Wei Qianglin; Liu Yibao; Wang Aixing; Zhang Peng

    2014-01-01

    This paper calculated the effects of different coolants to neutron energy spectrum in different position of the transmutation reactor by Monte Carlo N-Particle Transport Code (MCNP5). After having chosen the coolant and particular parameters, different nuclides in fuel rods of the transmutation reactor were calculated and compared. According to the actual situation, nuclides of 99 Tc and 241 Am were chosen and compared. Then the nonuniform-arrangement scheme of different spent fuels were proposed. By comparison of the diagram, it is found that it is more effective to promote the neutron utilization in the reactor by the non-uniform arrangement scheme, which is more reasonable than traditional uniform one. Thus, it would be helpful for transmutation technology by the application of the scheme. (authors)

  20. Emerging applications of advanced fuels for energy generation and transmutation. Overview of IAEA activities

    International Nuclear Information System (INIS)

    Pong Eil Juhn; Arkhipov, V.

    1996-01-01

    Nuclear power generation is an established part of the world's electricity mix. However, the highly radioactive waste generated during power production is of great concern of public perception of nuclear energy. In order for nuclear power to realize its full potential as a major energy source for the entire world, there must be a safe and effective way to deal with this waste. Therefore, science must come to the rescue in the form of new, more effective technology aimed at reducing the amount of long-lived radioactive waste and eliminating nuclear weapons' grade material through transmutation of these isotopes in fission reactors or accelerators. In the framework of IAEA activities on the use of this new technologies the Agency has periodically review and assess the current status of the new fuel cycles, its applications worldwide, its economic benefits, and its perceived advantages vis-a-vis other nuclear fuel cycles. (author)

  1. Inert matrix fuels for incineration of plutonium and transmutation of americium

    International Nuclear Information System (INIS)

    Matzke, Hj.

    2000-01-01

    In conventional U-based nuclear fuels, both Pu and higher actinides (mainly Am, but also Np and Cm) are formed by neutron capture reactions and α- or β-decay. If a strategy of reprocessing is adopted as in some European nations and in Japan, the separated Pu can be recycled as (U, Pu)O 2 (or mixed-oxide-MOX) fuel. The high-level liquid waste of reprocessing is presently vitrified. However, the alternative of separating the minor actinides from the fission products (partitioning) and subsequent transmutation in existing reactors or in new dedicated actinide burners is widely studied as a possible means to reduce the radiotoxicity of the waste

  2. Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

    International Nuclear Information System (INIS)

    Han Song; Yang Yongwei

    2007-01-01

    Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

  3. Summary of twelfth session of the AER Working Group F - 'Spent Fuel Transmutations' and third meeting of INPRO Project RMI - 'Meeting energy needs in the period of raw materials insufficiency during the twenty first century'

    International Nuclear Information System (INIS)

    Lelek, V.

    2010-01-01

    Information about the development in the problems spent fuel transmutation and future nuclear reactors development during the last years 2009-2010. Some critical views on the coming works tendencies are given by the coordinator of works within AER cooperation. (Author)

  4. Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2010-01-01

    The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

  5. Nuclear transmutation of actinides other than fuel as a radioactive waste management scheme

    International Nuclear Information System (INIS)

    Cecille, L.; Hage, W.; Hettinger, H.; Mannone, F.; Mousty, F.; Schmidt, E.; Sola, A.; Huber, B.; Koch, L.

    1977-01-01

    The bulk of fission products in the high-level waste (HLW) decays to innocuous hazard levels within about 600 years. Actinide waste and a few fission products however represent a potential risk up to some hundreds of thousand of years. An alternative to the disposal of the whole HLW in geological formations is its fractionation, a nuclear transmutation of long-lived isotopes in fission reactors and a geological disposal of the other components. This solution would decrease the potential long-term risks of the geological waste disposal and would also accomodate to the demand of public opinion. The results of studies related to this management scheme are outlined with special reference to areas, where additional effort is required for realistic cost/benefit evaluations. Reactor physics calculations demonstrated the feasibility of actinide incineration in thermal and fast reactors. Obtained transmutation rates are sufficiently high to garantee acceptably small actinide inventories in the reactor in the case of self-generated actinide recycling. It appears that fast breeders could be used as transmutation devices without major additional reactor devlopment work. The thermal power rating of actinide fuel elements and the contribution of actinides and of minor amounts of lanthanide impurities to the neutron economy of the reactor has been evaluated. Sensitivity studies indicated that the results are dependent on the reactor operation mode and on the accuracy of the nuclear data. These calculations permitted the identification of isotopes for which cross section masurements and improved theoretical methods are required. The chemical separation of actinides from the HLW with the envisaged decontamination factors is being studied by solvent extraction and precipitation techniques using waste simulates and samples of high activity waste from European reprocessing plants. Up to now, the obtained results do not yet allow a definitive judgement on the feasibility of actinides

  6. Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

    Science.gov (United States)

    Claux, B.; Souček, P.; Malmbeck, R.; Rodrigues, A.; Glatz, J.-P.

    2017-08-01

    One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

  7. Nuclear waste transmutation

    International Nuclear Information System (INIS)

    Salvatores, M.; Girard, C.; Delpech, M.; Slessarev, I.; Tommasi, J.

    1994-01-01

    Waste management strategies foresee the use of a deep geological repository either for final disposal of irradiated fuel or, after reprocessing and reuse of U and Pu for final disposal of long-lived radio-active materials. In the second case, partitioning and transmutation of these materials can be considered to reduce the impact of radiation on man due to the storage. On the basis of the SPIN programme developed by CEA in this field, the main features of transmutation is presented. The goal to achieve and the criteria to use are quite difficult to establish. The rights para-meters to characterize the risk are the potential radiotoxicity in the the repository and the residual radiotoxicity at the outlet. Transmutation studies in CEA used the potential radiotoxicity which is based on well-known parameters and less precise hazardous factors. The second point to appreciate the trans- mutation interest is to dispose of a criteria for the radio-radiotoxicity reduction. As there is no general agreement, we try to have a toxicity as low as possible within reasonable technical limits. To reduce the long term radio- toxicity, Pu, minor actinides and some long-lived fission products have to be transmuted. To assess the feasibility of such trans-mutation in reactors or advanced systems, one has to consider constraints on neutronic balance, safety, fuel cycle, technology , economy. Taking in account the main conclusions of this analysis, parametric studies of homogeneous and heterogenous transmutation permit a choice of promising solutions. Goals are to use every long-lived element with a minimized production of other long- lived elements in order to obtain an appreciable radiotoxicity reduction. It implies multi recycling of Pu which favours fast neutron reactors and different strategies of multi recycling for Np, Am, Cm. Multi recycling makes the results strongly dependant of losses. Researches to obtain the high partitioning efficiency needed are in progress. Calculations

  8. Radioactive waste partitioning and transmutation within advanced fuel cycles: Achievements and Challenges

    International Nuclear Information System (INIS)

    Salvatores, M.; Palmiotti, G.

    2011-01-01

    In the last decades, numerous studies have been performed in order to identify appropriate 'Partitioning and Transmutation' (P and T) strategies, aiming to the reduction of the burden on a geological storage (see, among many others, Salvatores, 2005). P and T strategies are very powerful and unique tools to reduce drastically the radiotoxicity level of the wastes and to reduce the time needed to reach the reference level (from ∼100,000 years to few hundred years, i.e. comparable to the period in which technological and engineering means allow reasonably to control the radioactivity confinement). Moreover, P and T allows, in principle, also the reduction of the residual heat in a geological repository, with a potential significant impact on the repository size and characteristics. The first requirement of P and T strategies is the deployment of spent fuel reprocessing techniques (aqueous or dry), which are both in the continuity of today technologies (e.g. as implemented at La Hague in France, where Pu is separated up to 99.9%) or which represent innovative, adapted approaches (e.g. pyrochemistry). The requirement is to extend the performance of Pu separation to 99.9% also to Np, Am and Cm kept together or separated and in any case decontaminated from the lanthanides as much as possible. The separated TRU should then be 'transmuted' (or 'burned') in a neutron field. The essential mechanism is to fission them, transforming them into much shorter lived or stable fission products. However, the fission process is always in competition with other processes, and, in particular, with neutron capture, which does eliminate isotope A, but transforms it into isotope A+1, which can still be radioactive. Isotope A+1 can in turn be fissioned or transmuted into isotope A+2, and so on. The neutron field has to be provided by a fission reactor. The requirement for this (dedicated) reactor is to be able to privilege the fission process with respect to the capture process and to be

  9. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  10. The development of the market for neutron transmutation doped silicon

    International Nuclear Information System (INIS)

    Herzer, H.; Vieweg-Gutberlet, G.

    1984-01-01

    Neutron transmutation doped silicon was introduced to the electronic device market in the 1975-1976 time period. Today, neutron transmutation doping is definitely a mature technology applied mainly to semiconductor power devices. There is no doubt that the power device sector will remain the major consumer of NTD silicon in the near future. This paper examines the possible application of NTD silicon to other areas of the semiconductor market, and concludes that the need for NTD silicon will continue to grow and will expand into other applications. Consequently, unless new reactor capacities become available by the end of the decade, NTD silicon applications will probably be limited mainly to power and sensor devices

  11. High intensity proton linear accelerator development for nuclear waste transmutation

    International Nuclear Information System (INIS)

    Mizumoto, M.; Hasegawa, K.; Oguri, H.; Ito, N.; Kusano, J.; Okumura, Y.; Murata, H.; Sakogawa, K.

    1997-01-01

    A high-intensity proton linear accelerator with an energy of 1.5 GeV and an average current of 10 mA has been proposed for various engineering tests for the transmutation system of nuclear waste by JAERI. The conceptual and optimization studies for this accelerator performed for a proper choice of operating frequency, high b structure, mechanical engineering considerations and RF source aspects are briefly described

  12. Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

    Science.gov (United States)

    Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

    2011-07-01

    The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

  13. Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

    International Nuclear Information System (INIS)

    Turchi, P.E.; Kaufman, L.; Fluss, M.J.

    2008-01-01

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER

  14. Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P E; Kaufman, L; Fluss, M J

    2008-11-10

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  15. Partitioning and transmutation. Current developments - 2010. A report from the Swedish reference group for PT-research

    Energy Technology Data Exchange (ETDEWEB)

    Blomgren, Jan (ed.) (Swedish Centre for Nuclear Technology, SKC, Stockholm (Sweden)); Karlsson, Fred (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Pomp, Stephan (Uppsala Univ., Uppsala, Dept. of Physics and Astronomy, Div. of Applied Nuclear Physics (Sweden)); Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Skarnemark, Gunnar (Nuclear Chemistry, Dept. of Chemical and Biological Engineering, Chalmers Univ. of Technology, Goeteborg (Sweden)); Wallenius, Janne; Zakova, Jitka (Reactor Physics Div., Physics Dept., Royal Inst. of Technology, Stockholm (Sweden)); Grenthe, Ingemar; Szabo, Zoltan (School of Chemical Science and Engineering, Royal Inst. of Technology, Stockholm (Sweden))

    2010-01-15

    The research and development on methods for partitioning and transmutation (P and T) of long-lived radionuclides in spent nuclear fuel has attracted considerable interest during the last decade. The main objective of P and T is to eliminate or at least substantially reduce the amount of such long-lived radionuclides that has to go to a deep geological repository for final disposal. The objective of current research on partitioning is to find and develop processes suitable for separation of the heavier actinides (and possibly some long-lived fission products) on an industrial scale. The objective of current research on transmutation is to define, investigate and develop facilities that may be suitable for transmutation of the long-lived radionuclides. The research on partitioning has made important progress in recent years. In some cases one has succeeded to separate americium and curium. Many challenges remain however. Within hydrochemistry one has achieved sufficiently good distribution and separation factors. The focus turns now towards development of an operating process. The search for ligands that give sufficiently good extraction and separation will continue but with less intensity. The emphasis will rather be on improving stability against hydrolysis and radiolysis. This may be achieved either by additives to the solvent or by selection of a proper solvent. The development of processes and equipment must be intensified. Pyrochemical research is looking into methods for recovery of uranium and for separating fission products with large neutron cross sections. The objective is to avoid separation of plutonium from other transuranium elements and thus simplify the proliferation issue. The future work is focused on improved selectivity and on technical development. Design of processes and equipment is difficult due to the aggressive properties of the melts and the relatively high temperatures required. The fabrication of fuel for transmutation and the

  16. Partitioning and transmutation. Current developments - 2010. A report from the Swedish reference group for PT-research

    International Nuclear Information System (INIS)

    Blomgren, Jan; Karlsson, Fred; Pomp, Stephan; Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Skarnemark, Gunnar; Wallenius, Janne; Zakova, Jitka; Grenthe, Ingemar; Szabo, Zoltan

    2010-01-01

    The research and development on methods for partitioning and transmutation (P and T) of long-lived radionuclides in spent nuclear fuel has attracted considerable interest during the last decade. The main objective of P and T is to eliminate or at least substantially reduce the amount of such long-lived radionuclides that has to go to a deep geological repository for final disposal. The objective of current research on partitioning is to find and develop processes suitable for separation of the heavier actinides (and possibly some long-lived fission products) on an industrial scale. The objective of current research on transmutation is to define, investigate and develop facilities that may be suitable for transmutation of the long-lived radionuclides. The research on partitioning has made important progress in recent years. In some cases one has succeeded to separate americium and curium. Many challenges remain however. Within hydrochemistry one has achieved sufficiently good distribution and separation factors. The focus turns now towards development of an operating process. The search for ligands that give sufficiently good extraction and separation will continue but with less intensity. The emphasis will rather be on improving stability against hydrolysis and radiolysis. This may be achieved either by additives to the solvent or by selection of a proper solvent. The development of processes and equipment must be intensified. Pyrochemical research is looking into methods for recovery of uranium and for separating fission products with large neutron cross sections. The objective is to avoid separation of plutonium from other transuranium elements and thus simplify the proliferation issue. The future work is focused on improved selectivity and on technical development. Design of processes and equipment is difficult due to the aggressive properties of the melts and the relatively high temperatures required. The fabrication of fuel for transmutation and the

  17. Actinide partitioning and transmutation program progress report, October 1, 1976--March 31, 1977

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Tedder, D.W.

    1977-01-01

    Experimental work on the 16 tasks comprising the Actinide Partitioning and Transmutation Program was initiated at the various sites. This work included the development of conceptual material balance flowsheets which define integrated waste systems supporting an LWR fuel reprocessing plant and a mixed (U-Pu) oxide fuel refabrication plant. In addition, waste subsystems were defined for experimental evaluation. Computer analysis of partitioning-transmutation, utilizing an LMFBR for transmutation, was completed for both constant and variable waste actinide generation rates

  18. Calculations of the actinide transmutation with HELIOS for fuels of light water reactors

    International Nuclear Information System (INIS)

    Francois L, J.L.; Guzman A, J.R.

    2006-01-01

    In this work a comparison of the obtained results with the HELIOS code is made and those obtained by other similar codes, used in the international community, respect to the transmutation of smaller actinides. For this the one it is analyzed the international benchmark: 'Calculations of Different Transmutation Concepts', of the Nuclear Energy Agency. In this benchmark two cell types are analyzed: one small corresponding to a PWR standard, and another big one corresponding to a PWR highly moderated. Its are considered two types of burnt of discharge: 33 GWd/tHM and 50 GWd/tHM. The following types of results are approached: the k eff like a function of the burnt one, the atomic densities of the main isotopes of the actinides, the radioactivities in the moment in that the reactor it is off and in the times of cooling from 7 up to 50000 years, the reactivity by holes and the Doppler reactivity. The results are compared with those obtained by the following institutions: FZK (Germany), JAERI (Japan), ITEP (Russia) and IPPE (Russian Federation). In the case of the eigenvalue, the obtained results with HELIOS showed a discrepancy around 3% Δk/k, which was also among other participants. For the isotopic concentrations: 241 Pu, 242 Pu and 242m Am the results of all the institutions present a discrepancy bigger every time, as the burnt one increases. Regarding the activities, the discrepancy of results is acceptable, except in the case of the 241 Pu. In the case of the Doppler coefficients the discrepancy of results is acceptable, except for the cells with high moderation; in the case of the holes coefficients, the discrepancy of results increases in agreement with the holes fraction increases, being quite high to 95% of holes. In general, the results are consistent and in good agreement with those obtained by all the participants in the benchmark. The results are inside of the established limits by the work group on Plutonium Fuels and Innovative Fuel Cycles of the Nuclear

  19. The Molten Salt Fast Reactor as Highly Efficient Transmutation System

    International Nuclear Information System (INIS)

    Merk, B.; Rohde, U.; Scholl, S.

    2013-01-01

    Conclusion and future steps: • MSFR offers very attractive features for efficient transmutation; • significant advantages due to liquid fuel and online refuelling and reprocessing; • significant developments are required on the way to application; • system is very promising for transmutation; • development of a safety approach for liquid fuel reactors (RSWG); • investigation of possibilities to solve the “last transmuter” problem (ICAPP2013) – as future for countries envisaging nuclear phase out or no transition to fast reactor fleet for energy production; • establishing of a strong group “MSFR for transmutation”; • development of a transmutation optimized design

  20. Actinide transmutation using inert matrix fuels versus recycle in a low conversion fast burner reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, M.R.; Schneider, E.A.; Recktenwald, G.; Cady, K.B. [The Department of Mechanical Engineering, The University of Texas at Austin, 1 University Station, C2200, Austin, 78712 (United States)

    2009-06-15

    Reducing the disposal burden of the long lived radioisotopes that are contained within spent uranium oxide fuel is essential for ensuring the sustainability of nuclear power. Because of their non-fertile matrices, inert matrix fuels (IMFs) could allow light-water reactors to achieve a significant burn down of plutonium and minor actinides that are that are currently produced as a byproduct of operating light-water reactors. However, the extent to which this is possible is not yet fully understood. We consider a ZrO{sub 2} based IMF with a high transuranic loading and show that the neutron fluence (and the subsequent fuel residence time required to achieve it) present a practical limit for the achievable actinide burnup. The accumulation of transuranics in spent uranium oxide fuel is a major obstacle for the sustainability of nuclear power. While commercial light-water reactors (LWR's) produce these isotopes, they can be used to transmute them. At present, the only viable option for doing this is to partly fuel reactors with mixed oxide fuel (MOX) made using recycled plutonium. However, because of parasitic neutron capture in the uranium matrix of MOX, considerable plutonium and minor actinides are also bred as the fuel is burned. A better option is to entrain the recycled isotopes in a non-fertile matrix such as ZrO{sub 2}. Inert matrices such as these were originally envisioned for burning plutonium from dismantled nuclear weapons [1]. However, because they achieve a conversion ratio of zero, they have also been considered as a better alternative to MOX [2-6]. Plutonium and minor actinides dominate the long term heat and radiological outputs from spent nuclear fuel. Recent work has shown that that IMFs can be used to reduce these outputs by at least a factor of four, on a per unit of energy generated basis [6]. The degree of reduction is strongly dependent on IMF burnup. In principle, complete transmutation of the transuranics could be achieved though this

  1. Preliminary Analysis of High-Flux RSG-GAS to Transmute Am-241 of PWR’s Spent Fuel in Asian Region

    Science.gov (United States)

    Budi Setiawan, M.; Kuntjoro, S.

    2018-02-01

    A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

  2. Actinide partitioning-transmutation program final report. III. Transmutation studies

    International Nuclear Information System (INIS)

    Wachter, J.W.; Croff, A.G.

    1980-07-01

    Transmutation of the long-lived nuclides contained in fuel cycle wastes has been suggested as a means of reducing the long-term toxicity of the wastes. A comprehensive program to evaluate the feasibility and incentives for recovering the actinides from wastes (partitioning) and transmuting them to short-lived or stable nuclides has been in progress for 3 years under the direction of Oak Ridge National Laboratory (ORNL). This report constitutes the final assessment of transmutation in support of this program. Included are (1) a summary of recent transmutation literature, (2) a generic evaluation of actinide transmutation in thermal, fast, and other transmutation devices, (3) a preliminary evaluation of 99 Tc and 129 I transmutation, and (4) a characterization of a pressurized-water-reactor fuel cycle with and without provisions for actinide recovery and transmutation for use in other parts of the ORNL program. The principal conclusion of the report is that actinide transmutation is feasible in both thermal and fast reactors, subject to demonstrating satisfactory fuel performance, with relatively little impact on the reactor. It would also appear that additional transmutation studies are unwarranted until a firm decision to proceed with actinide transmutation has been made by the responsible authorities

  3. The cross sections of reactions resulting in transmutation of long-lived radionuclides of exhausted nuclear fuel exposed to fast neutrons

    International Nuclear Information System (INIS)

    Konodeev, A.Yu.; Korovin, Yu.A.; Erview, K.

    1993-01-01

    Research is at present concerned with the possible transmutation of long-lived radionuclides of spent nuclear fuel in the flux of fast neutrons from neutron generators which are distinguished by their energy spectrum and density of the flux generated. For this purpose one must know the cross sections of the nuclear reactions resulting in the transmutation and formation of new long-lived radionuclides due to the irradiation. The transmutation rate of radioisotope irradiated with neutrons have a known energy spectrum is determined by calculating the transmutation cross section which is equal to the sum of the cross sections of neutron reactions causing conversion of a particular isotope into another after the decay of short-lived residual nuclei. The presently available neutron cross section data of long-lived radionuclides, i.e., the products of the fission of nuclear fuel, are insufficient for research on these effects as transmutations occur in a flux of high-energy neutrons. This paper presents the cross sections of reactions accounting for the transmutation of the most important long-lived radionuclides of exhausted nuclear fuel during its irradiation with neutron having energies of up to 100 MeV. The neutron cross sections were calculated for 79 Se, 90 Sr, 93 Zr, 99 Tc, 107 Pd, 126 Sn, 129 I, 135 Cs and 137 Cs with a half-life ≥30 years

  4. Transmutation of radioactive wastes: how and why?

    International Nuclear Information System (INIS)

    Patarin, L.

    2004-01-01

    After having evoked the natural or spontaneous transmutation of natural or artificial radioactive atoms, the author describes how this transmutation is technically obtained, indicates the two main families of atoms present in a used nuclear fuel and for which transmutation is to be investigated (long-lived fission residues or products, and transuranium elements) and of which the behaviour in neutron fluxes must be explored. He discusses the industrial means required for artificial transmutation. He discusses the interest of performing such a transmutation

  5. Nuclear Fuel Cycle Evaluation and Screening Findings on Partitioning and Transmutation

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Taiwo, T.A.; Gehin, J.C.; Jubin, R.; Todosow, M.

    2015-01-01

    A Nuclear Fuel Cycle Evaluation and Screening (E and S) study has recently been completed in the United States. The study considered the entire fuel cycle, included considerations for both once-through and recycle fuel cycle options, evaluated a set of 40 fuel cycles that allowed a comprehensive assessment of fuel cycle performance, identified a relatively small number of promising fuel cycle options that have the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States, and allowed the identification of research and development (R and D) activities needed to support the development of the promising fuel cycle options. Nine high-level criteria (Nuclear Waste Management, Proliferation Risk, Nuclear Material Security Risk, Safety, Environmental Impact, Resource Utilisation, Development and Deployment Risk, Institutional Issues, and Financial Risk and Economics) and associated metrics were used in the study to compare the performance of nuclear fuel cycle options to that of the current fuel cycle practiced in the United States. The study also evaluated a number of fuel cycle characteristics that may have the potential to impact future R and D directions. These included for example: 1) The fuel resources used, i. e., uranium and/or thorium. 2) Impact of extremely high burnup fuels. 3) Minor actinide recycle. 4) The impact of losses during separations (partitioning). 5) Critical versus subcritical (externally-driven) systems for material irradiation. 6) Impact of spectrum of irradiation system, i.e., fast, thermal or intermediate. 7) Waste generation reduction, all of which were quantified in the study. The E and S study has implemented a framework that can be used now and in the future to objectively inform on the potential of alternative nuclear fuel cycles, providing decision-makers and others with perspective on fuel cycle capabilities. (authors)

  6. Effects of an LMR-based partitioning-transmutation system on US nuclear fuel cycle health risk

    International Nuclear Information System (INIS)

    Michaels, G.E.; Reich, W.J.

    1992-01-01

    Health risks for the current US nuclear fuel cycle and for an illustrative partitioning and transmutation (P-T) fuel cycle based on Liquid Metal Reactor (LMR) technology are calculated and compared. Health risks are calculated for all non-reactor fuel cycle steps, including reprocessing, transportation, and high-level waste (HLW) disposal. Uranium mining and milling health risks have been updated to include recent occupational injury and death statistics, and the radiological health risk to the general public posed by the uranium mining overburden. In addition, the radiological health risks for transportation have been updated to include latent cancer fatalities associated with both normal transport and accidents. Given the assumptions of the study, it is shown that the deployment of an LMR-based P-T system is expected to reduce overall nuclear fuel cycle health risk

  7. Effect of spectral characterization of gaseous fuel reactors on transmutation and burning of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Fung, C.; Anghaie, S. [Florida Univ., Wilmington, NC (United States)

    2007-07-01

    Gaseous Core Reactors (GCR) are fueled with stable uranium compounds in a reflected cavity. The spectral characteristics of neutrons in GCR systems could shift from one end of the spectrum to the other end by changing design parameters such as reflector material and thickness, uranium enrichment, and the average operational temperature and pressure. The rate of actinide generation, transmutation, and burnup is highly influenced by the average neutron energy in reactor core. In particular, the production rate and isotopic mix of plutonium are highly dependent on the neutron spectrum in the reactor. Other actinides of primary interest to this work are neptunium-237 and americium-241 due to their pivotal impact on high-level nuclear waste disposal. In all cavity reactors including GCR's, the reflector material and thickness are the most important design parameters in determining the core spectrum. The increase in the gaseous fuel pressure and enrichment results in relative shift of neutron population toward energies greater than 2 eV. Reflector materials considered in this study are beryllium oxide, lithium hydride, lithium deuteride, zirconium carbide, graphite, lead, and tungsten. Results of the study suggest that the beryllium oxide and tungsten reflected GCR systems set the lower (softest) and upper (hardest) limits of neutron spectra, respectively. The inventory of actinides with half-lives greater than 1000 years can be minimized by increasing neutron flux level in the reactor core. The higher the neutron flux, the lower the inventory of these actinides. The majority of the GCR designs maintained a flux level on the order of 10{sup 15} cm{sup -2}*s{sup -1} while the PWR flux is one order of magnitude lower. The inventory of the feeder isotopes to Np{sup 237} including U{sup 237}, Pu{sup 241}, and Am{sup 241} decreases with relative shift of neutron spectrum toward higher energies. This is due to increased resonance absorption in these isotopes due to higher

  8. Advances in conceptual design of a gas-cooled accelerator driven system (ADS) transmutation devices to sustainable nuclear energy development

    International Nuclear Information System (INIS)

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia; Castells, Escriva; Abanades

    2011-01-01

    The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In this paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made. On the other hand, the heterogeneity of the TRISO particles inside the pebbles can not be negligible. In this paper, a study of the power density distribution inside the pebbles by means of a detailed simulation of the TRISO fuel particles and using an homogeneous composition of the fuel is addressed. (author)

  9. Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    2015-06-01

    The benefits of partitioning and transmutation (P and T) have now been established worldwide and, as a result, many countries are pursuing R and D programmes to advance the technologies associated with P and T. In this context, the OECD Nuclear Energy Agency (NEA) has organised a series of biennial information exchange meetings to provide experts with a forum to present and discuss state-of-the-art developments in the field of partitioning and transmutation since 1990. The OECD Nuclear Energy Agency Information Exchange Meeting on Actinides and Fission Products Partitioning and Transmutation is a forum for experts to present and discuss the state-of-the-art development in the field of P and T. Thirteen meetings have been organised so far and held in Japan, the United States, France, Belgium, Spain, the Republic of Korea and the Czech Republic. This 13. meeting was hosted by Seoul National University (Seoul, Republic of Korea) and was organised in co-operation with the International Atomic Energy Agency (IAEA) and the European Community (EC). The meeting covered strategic and scientific developments in the field of P and T such as: fuel cycle strategies and transition scenarios, the role of P and T in the potential evolution of nuclear energy as part of the future energy mix; radioactive waste management strategies; transmutation fuels and targets; advances in pyro and aqueous separation processes; P and T specific technology requirements (materials, spallation targets, coolants, etc.); transmutation systems: design, performance and safety; impact of P and T on the fuel cycle; fabrication, handling and transportation of transmutation fuels. A total of 103 presentations (39 oral and 64 posters) were discussed among the 110 participants from 19 countries and 2 international organisations. The meeting consisted of one plenary session where national and international programmes were presented followed by 5 technical sessions: - Fuel Cycle Strategies and Transition

  10. Development of nuclear transmutation technology - A study on accelerator-driven transmutation of long-lived radionuclide

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; Chung, Kie Hyung; Hong, Sang Hee; Hwang, Il Soon; Park, Byung Gi; Yang, Hyung Lyeol; Kim, Duk Kyu; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The objective of this study is to help establish the long-range nuclear waste disposal strategy through the investigations and comparisons of various= concepts of the accelerator-driven nuclear waste transmutation reactors, which have been suggested to replace the geological waste disposal due to the technical uncertainties in the long-time scale. Nuclear data, categorized in high -and low-energy neutron cross-sections, were investigated and the structures, principles, and recent progresses of proton linac were reviews, Also the accelerator power for transmutation and the economics were referred, The comparison of the transmutation concepts concentrated on two: Japanese OMEGA program of alloy fuelled system, Minor actinide molten salt system, and Eutectic alloy system and American ATW program of aqueous system and molten salt system. From the comparative study, a state-of-art of the technology has been identified as a concept employing proton-accelerate of 800 {approx} 1600 MeV with 100 mA capacity combined with liquid lead target, molten salt blanket and on-line chemical separation using centrifuge and electrowinning technology. 34 refs., 25 tabs., 64 figs. (author)

  11. TRANSMUTED EXPONENTIATED EXPONENTIAL DISTRIBUTION

    OpenAIRE

    MEROVCI, FATON

    2013-01-01

    In this article, we generalize the exponentiated exponential distribution using the quadratic rank transmutation map studied by Shaw etal. [6] to develop a transmuted exponentiated exponential distribution. Theproperties of this distribution are derived and the estimation of the model parameters is discussed. An application to real data set are finally presented forillustration

  12. The development and application of silicon neutron transmutation doping (NTD) technology in china

    International Nuclear Information System (INIS)

    Qiao Chenyang; Sun Zhiyong; Ke Guotu, Lu Cungang; Shen Feng; Chen Huiqiang

    2009-01-01

    The research and development history of silicon Neutron Transmutation Doping (NTD) technology and its applications at home and abroad are introduced in this paper. The advantages of NTD, compared with conventional technology of doping, are narrated. The principle of NTD as well as the implementation of the main procedures related to Si NTD is explained. The market demand tendency is prospected, and the advanced measures on NTD quality control are described. (authors)

  13. The Impact of Partitioning and Transmutation on the Risk Assesment of a Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Amrani, Naima

    2006-01-01

    Partitioning and transmutation of radioactive and long lived component from the highly radioactive waste stream in order to reduce or probably eliminate their radiotoxic inventory was the important option for the nuclear waste management. The principal radionuclides contribution to the long term radiotoxic inventory is mostly due to Pu, minor actinides and some long-lived fission products. The conditioning operation can present artificial barriers which are potentially capable of confining the radionuclides within their package for thousands of years. After this time nothing can be predicted. The solubility of the actinides (except Np) is generally low whereas the long lived fission product. particularly 137 Cs, 129 I and in some case 99 Tc, display high mobility In the geosphere. Conditioning of separated long-lived nuclides in appropriate matrices which could serve as irradiation matrix in a delayed transmutation option is a possible outcome for the next decades. The general strategy of introducing Partitioning and Transmutation as an additional waste management option is based on the radiological benefit which is expected from such an option. The short term impact of partitioning would be to reduce long-term radiotoxic inventory of the resulting HLW at the expense of an increase of the operational requirements for the nuclear facilities concerned. Fast neutron spectrum devices (FR or ADS facilities) are more efficient than current LWRs for recycling and transmuting long-lived radionuclides

  14. A Fast Numerical Method for the Calculation of the Equilibrium Isotopic Composition of a Transmutation System in an Advanced Fuel Cycle

    Directory of Open Access Journals (Sweden)

    F. Álvarez-Velarde

    2012-01-01

    Full Text Available A fast numerical method for the calculation in a zero-dimensional approach of the equilibrium isotopic composition of an iteratively used transmutation system in an advanced fuel cycle, based on the Banach fixed point theorem, is described in this paper. The method divides the fuel cycle in successive stages: fuel fabrication, storage, irradiation inside the transmutation system, cooling, reprocessing, and incorporation of the external material into the new fresh fuel. The change of the fuel isotopic composition, represented by an isotope vector, is described in a matrix formulation. The resulting matrix equations are solved using direct methods with arbitrary precision arithmetic. The method has been successfully applied to a double-strata fuel cycle with light water reactors and accelerator-driven subcritical systems. After comparison to the results of the EVOLCODE 2.0 burn-up code, the observed differences are about a few percents in the mass estimations of the main actinides.

  15. Current status on research and development of accelerator-driven system and nuclear transmutation technology in Asian countries

    International Nuclear Information System (INIS)

    Pyeon, Cheol Ho

    2013-01-01

    This status report describes the current status on research and development (R and D) of accelerator-driven system (ADS) and nuclear transmutation techniques (NTT), including nuclear data, accelerator techniques, Pb-Bi target, fuel technologies and reactor physics, in East Asian countries: Japan, Korea and China. The report also includes all presentation materials presented in 'the 10th International Workshop on Asian Network for ADS and NTT (ADS+NTT 2012)' held at the Kyoto University Research Reactor Institute, Osaka, Japan on 6th and 7th December, 2012. The objective of this workshop is to make actual progress of ADS R and D especially in East Asian countries, as well as in European countries, through sharing mutual interests and conducting the information exchange each other. The 5 of 27 papers presented at the entitled report and meeting are indexed individually. (J.P.N.)

  16. Nuclear transmutations

    International Nuclear Information System (INIS)

    Mikulaj, V.

    1992-01-01

    Two types of nuclear transmutations are outlined, namely the radioactive transmutations and nuclear reactions. The basic characteristics are given of radioactive transmutations (gamma transmutations and isomeric transitions, beta, alpha transmutations, spontaneous fission and spontaneous emission of nucleons), their kinetics and the influence of the physical and chemical state of the radionuclide on the transmutation rate. The basic characteristics are described of nuclear reactions (reactions of neutrons including fission, reactions induced by charged particles and photons), their kinetics, effective cross sections and their mechanism. Chemical reactions caused by nuclear transmutations are discussed (recoil energy, properties of hot atoms, Szilard-Chalmers effect). A brief information is given on the behavior of radionuclides in trace concentrations. (Z.S.) 2 tabs., 19 figs., 12 refs

  17. Reactor physical program in the frame of the MSR-SPHINX transmuter concept development

    International Nuclear Information System (INIS)

    Hron, M.; Mikisek, M.

    2008-01-01

    In the frame of the R and D program for the Molten Salt Reactor (MSR) - SPHINX (Spent Hot fuel Incinerator by Neutron flux) concept, which has been under development in the Czech Republic as an actinide burner in resonance neutron spectrum and a radionuclide transmuter in a well-thermalized neutron spectrum, and namely its reactor physical part, the relatively broad experimental activities have been involved in the program, recently, which will serve for a validation of computer codes and verification of design inputs for designing of a demonstration unit of the MSR-type. The experimental program, which has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors, will be in the proposed next stage of the program focused on a large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0, which may allow to modify it to experimental zero power salt reactor SR-0. There has been a preparatory stage of the project called EROS started in the year 2006 and new experiments with MSR-type zones irradiated by cyclotron based neutron source are planned at the end of 2007 and should go on in the year 2008. There will be a brief description of the so far prepared and performed experimental programs introduced in the paper. (authors)

  18. Nuclear waste transmutation and related innovative technologies

    International Nuclear Information System (INIS)

    2002-01-01

    The main topics of the summer school meeting were 1. Motivation and programs for waste transmutation: The scientific perspective roadmaps; 2. The physics and scenarios of transmutation: The physics of transmutation and adapted reactor types. Impact on the fuel cycle and possible scenarios; 3. Accelerator driven systems and components: High intensity accelerators. Spallation targets and experiments. The sub critical core safety and simulation physics experiments; 4. Technologies and materials: Specific issues related to transmutation: Dedicated fuels for transmutation. Fuel processing - the role of pyrochemistry. Materials of irradiation. Lead/lead alloys. 5. Nuclear data: The N-TOF facility. Intermediate energy data and experiments. (orig./GL)

  19. Role of accelerators in the Czech national transmuter project

    International Nuclear Information System (INIS)

    Bem, P.; Kugler, A.

    1999-01-01

    The problem of spent nuclear fuel from the so far operated PWRs has become a crucial issue in the Czech Republic. The first attempts to solve this problem by a final deposit of spent fuel into a suitable geological formation have been shown not to be fully acceptable. Therefore, the revival of nuclear transmutation technology application for nuclear incineration of nuclear waste and spent fuel in particular was welcomed. A realistic national project started to be developed in 1996. The four major nuclear research institutions of the country formed a consortium focused on an adoption of the world-wide experience and a development of a national project of a transmutation technology (experimental transmuter LA-0) or an efficient participation in the international effort in that field. Because the LA-0 transmuter concept of subcritical reactor with liquid fuel based on molten fluorides driven by an external neutron source has been adopted, the R and D effort has been focused on three regions. The first is devoted to the problem of a suitable neutron source, the second to a pre-conceptual design of a blanket for burning of actinides contained in spent fuel from PWRs. The third region is devoted to the utilisation of the experience from a specific field of dry (fluorine) reprocessing of spent fuel and a preparation of liquid fuel in the form of molten fluorides for the transmuter LA-0. (R.P.)

  20. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group

    International Nuclear Information System (INIS)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.J.; Laidler, J.J.; McDeavitt, S.M.; Thompson, M.; Toth, L.M.; Williamson, M.; Willit, J.L.

    1999-01-01

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD and D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years

  1. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

    1999-08-12

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

  2. Efficiency Of Transuranium Nuclides Transmutation

    International Nuclear Information System (INIS)

    Kazansky, Yu.A.; Klinov, D.A.; Semenov, E.V.

    2002-01-01

    One of the ways to create a wasteless nuclear power is based on transmutation of spent fuel nuclides. In particular, it is considered that the radioactivity of the nuclear power wastes should be the same (or smaller), than radioactivity of the uranium and the thorium extracted from entrails of the Earth. The problem of fission fragments transmutation efficiency was considered in article, where, in particular, the concepts of transmutation factor and the ''generalised'' index of biological hazard of the radioactive nuclides were entered. The transmutation efficiency has appeared to be a function of time and, naturally, dependent on nuclear power activity scenario, from neutron flux, absorption cross-sections of the nuclides under transmutation and on the rate of their formation in reactors. In the present paper the efficiency of the transmutation of transuranium nuclides is considered

  3. Partitioning and transmutation: Radioactive waste management option

    International Nuclear Information System (INIS)

    Stanculescu, A.

    2005-01-01

    Growing world population with increasing energy needs, especially in the developing countries, Threat of global warming due to CO 2 emissions demands non-fossil electricity production. Nuclear will have to be part of a sustainable mix of energy production options Figures show that 350 GWe worldwide capacity is 'nuclear'. Present worldwide spent fuel (containing high Pu inventory) and HLW would need large repositories. In view of the previous facts this lecture deals Partitioning and transmutation as radioactive waste management option. Partitioning and transmutation (P and T) is a complex technology i.e. advanced reprocessing, and demand transuranics fuel fabrication plants, as well as innovative and/or dedicated transmutation reactors. In addition to U, Pu, and 129 I, 'partitioning' extracts from the liquid high level waste the minor actinides (MA) and the long-lived fission products (LLFP) 99-Tc, 93-Zr, 135-Cs, 107-Pd, and 79-Se). 'Transmutation' requires fully new fuel fabrication plants and reactor technologies to be developed and implemented on industrial scale. Present LWRs are not suited for MA and LLFP transmutation (safety consideration, plant operation, poor incineration capability). Only specially licensed LWRs can cope with MOX fuel; for increased Pu loadings (up to 100%), special reactor designs (e.g., ABB80+) are required; a combination of these reactor types could allow Pu inventory stabilization. Long-term waste radiotoxicity can be effectively reduced only if transuranics are 'incinerated' through fission with very hard neutron spectra. New reactor concepts (dedicated fast reactors, Accelerator Driven Systems (ADS), fusion/fission hybrid reactors) have been proposed as transmuters/incinerators. Significant Pu+MAs incineration rates can be achieved in symbiotic scenarios: LWR-MOX and dedicated fast reactors; fast neutron spectrum ADS mainly for MA incineration; very high thermal flux ADS concepts could also provide a significant transuranics

  4. Transmutation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.c [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Lahodova, Z. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Klupak, V. [Nuclear Research Institute Rez plc (Czech Republic); Sus, F. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Kucera, J. [Research Centre Rez Ltd. (Czech Republic); Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic); Kus, P.; Marek, M. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic)

    2011-03-11

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

  5. Transmutation detectors

    International Nuclear Information System (INIS)

    Viererbl, L.; Lahodova, Z.; Klupak, V.; Sus, F.; Kucera, J.; Kus, P.; Marek, M.

    2011-01-01

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

  6. R and D activities for partitioning and transmutation in Korea

    International Nuclear Information System (INIS)

    Jae-Hyung, Yoo; Won-Seok, Park

    2003-01-01

    According to the long-term plan of nuclear technology development, KAERI is conducting a research and development project of transmutation with the objective of key technology development in the areas of partitioning and transmutation system. The research and development activities for partitioning and transmutation of long-lived radionuclides are introduced in this work. The studies of partitioning are focused on the electrorefining and electrowinning, which are aimed at investigating the thermodynamic properties of electrodeposition behaviours as well as the separation efficiency. As for the transmutation system, the HYPER (HYbrid Power Extraction Reactor) combined by a proton accelerator and a sub-critical reactor is being studied in KAERI as a prominent candidate facility in the future. Some conceptual studies are being conducted to develop key elemental systems of the sub-critical reactor such as the core, TRU fuel, proton target, and the cooling system. The conceptual design of the HYPER system will be completed by 2006. (author)

  7. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  8. Status of SFR Metal Fuel Development

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

    2013-01-01

    Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

  9. Partitioning and transmutation (P and T) 1997. Status report

    International Nuclear Information System (INIS)

    Enarsson, Aasa; Landgren, A.; Liljenzin, J.O.; Skaalberg, M.; Spjuth, L.; Gudowski, W.; Wallenius, J.

    1998-05-01

    Research on and the evaluation of partitioning and transmutation are currently in progress in many industrial countries due to its potential as a long-term, sustainable energy source with low environmental impact and due to its ability to destroy many long-lived nuclides. The cost of the research and development work on partitioning and transmutation is considered to be so great that international co-operation is required. With respect to Sweden, we recommend a balanced research work on both partitioning and transmutation technology. Within the area of partitioning, it is above all a question of locating new reagents which can be used to simplify the necessary partitioning processes and minimize the losses. The requirements with respect to high selectivity and minor losses will be significantly higher in a recirculating system based on transmutation than in the reprocessing facilities of today where only uranium and plutonium are recovered. If the utilized reagents can be easily destroyed, by dry or wet incineration and conversion into non-complex gaseous chemical compounds, this will open up good opportunities for the recovery of the radionuclides. From a purely technical standpoint, it would seem that a combination of different types of reactor systems would give the best possible transmutation efficiency. While existing light water reactors can be utilized for increased plutonium incineration, there is currently consensus about the view that reactors with high-energy neutrons are necessary to achieve a sufficiently high transmutation efficiency for neptunium, americium, curium and certain fission products. By allowing an accelerator-based neutron source to drive a subcritical heavy metal-cooled reactor, the potential for transmutation of fission products is increased, at the same time that satisfactory safety margins are achieved for certain fuel types with a low share of delayed neutrons and a high heat conductivity. Regardless of what types of systems are

  10. Partitioning and transmutation (P and T) 1997. Status report

    Energy Technology Data Exchange (ETDEWEB)

    Enarsson, Aasa; Landgren, A.; Liljenzin, J.O.; Skaalberg, M.; Spjuth, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Gudowski, W.; Wallenius, J. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    1998-05-01

    Research on and the evaluation of partitioning and transmutation are currently in progress in many industrial countries due to its potential as a long-term, sustainable energy source with low environmental impact and due to its ability to destroy many long-lived nuclides. The cost of the research and development work on partitioning and transmutation is considered to be so great that international co-operation is required. With respect to Sweden, we recommend a balanced research work on both partitioning and transmutation technology. Within the area of partitioning, it is above all a question of locating new reagents which can be used to simplify the necessary partitioning processes and minimize the losses. The requirements with respect to high selectivity and minor losses will be significantly higher in a recirculating system based on transmutation than in the reprocessing facilities of today where only uranium and plutonium are recovered. If the utilized reagents can be easily destroyed, by dry or wet incineration and conversion into non-complex gaseous chemical compounds, this will open up good opportunities for the recovery of the radionuclides. From a purely technical standpoint, it would seem that a combination of different types of reactor systems would give the best possible transmutation efficiency. While existing light water reactors can be utilized for increased plutonium incineration, there is currently consensus about the view that reactors with high-energy neutrons are necessary to achieve a sufficiently high transmutation efficiency for neptunium, americium, curium and certain fission products. By allowing an accelerator-based neutron source to drive a subcritical heavy metal-cooled reactor, the potential for transmutation of fission products is increased, at the same time that satisfactory safety margins are achieved for certain fuel types with a low share of delayed neutrons and a high heat conductivity. Regardless of what types of systems are

  11. Enhancing TRU burning and Am transmutation in Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Kochendarfer, Richard A.; Moriwaki, Hiroyuki; Kunishima, Shigeru

    2011-01-01

    Research highlights: → This ARR is an oxide fueled sodium cooled reactor based on innovative technologies to destruct TRU. → TRU burning core is designed to burn TRU at 28 kg/TW th h, adding moderator pins of B 4 C (Enriched B-11). → Am transmutation core can transmute Am at 34 kg/TW th h, adding uranium free AmN blanket to TRU burning core. → The TRU burning core improves TRU burning by 40-50% than the previous core. → The Am transmutation core can transmute Am effectively, keeping the void reactivity acceptable. - Abstract: This paper presents about conceptual designs of Advanced Recycling Reactor (ARR) focusing on enhancement in transuranics (TRU) burning and americium (Am) transmutation. The design has been conducted in the context of the Global Nuclear Energy Partnership (GNEP) seeking to close nuclear fuel cycle in ways that reduce proliferation risks, reduce the nuclear waste in the US and further improve global energy security. This study strives to enhance the TRU burning and the Am transmutation, assuming the development of related technologies in this study, while the ARR based on mature technologies was designed in the previous study. It has followed that the provided TRU burning core is designed to burn TRU at 28 kg/TW th h, by adding moderator pins of B 4 C (Enriched B-11) and the Am transmutation core will be able to transmute Am at 34 kg/TW th h, by locating Am blanket of AmN around the TRU burning core. It indicates that these concepts improve TRU burning by 40-50% than the previous core and can transmute Am effectively, keeping the void reactivity acceptable.

  12. Development of advanced technological systems for accelerator transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Batskikh, G.I.; Bondarev, B.I.; Durkin, A.P. [Russian Academy of Sciences, Moscow (Russian Federation)] [and others

    1995-10-01

    A development concept of the accelerator nuclear energy reactors is considered for energy generation and nuclear power plant waste conversion into short-lived nuclides along with the requirements imposed on the technological systems necessary for implementation of such projects. The state of art in the field is discussed.

  13. Transmutation Capability of a Once-Through Molten-Salt and Other Transmuting Reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Lowenthal, M.; Barnes, D.; Kawasaki, D.; Kimball, D.; Matsumoto, H.; Sagara, H.; Vietez, E.R.

    2002-01-01

    A preliminary assessment is done of the transmutation characteristics of three reactor technologies: a multi-batch liquid metal (LM) cooled transmuter, a once-through molten-salt (MS) transmuter and a pebble bed (PB) transmuter. It was found that for the same fractional transmutation and same k eff drop with burnup (Δk effBU ), lead-bismuth offers smaller peak-to-average core power density, and it requires a smaller pumping power but a larger and heavier core than a sodium cooled transmuter. 99 Tc cannot effectively serve as a burnable absorber to reduce Δk effBU of LM transmuters. However, addition of thorium can greatly flatten k eff and almost double the fractional transmutation of the LWR spent fuel from ∼20% to ∼40%. If the 'once-through' MS transmuter is operated with continuous complete removal of fission products, it can achieve ∼85% fractional transmutation provided that the equilibrium concentration of actinides in the MS can reach 4 mole %. If the fission products are not actively removed, the fractional transmutation is reduced to ∼75%. The fractional transmutation of a PB transmuter can exceed 40%. More thorough analysis is required to better quantify the transmutation capability of the different transmuter technologies. (authors)

  14. Transmutation of nuclear waste in nuclear reactors

    International Nuclear Information System (INIS)

    Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

    1996-03-01

    The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

  15. Nuclear Wastes: Technologies for Separations and Transmutation

    National Research Council Canada - National Science Library

    .... The committee examines the currently used "once-through" fuel cycle versus different alternatives of separations and transmutation technology systems, by which hazardous radionuclides are converted...

  16. Use of pyroprocessing in the preparation of spent fuel for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Laidler, J.J. [Argonne National Laboratory, IL (United States)

    1995-10-01

    Pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} of spent fuel offers a number of unique advantages, particularly in comparison with coventional aqueous solvent extraction techniques. Pyroprocessing, in its present version, has been under development for nearly ten years. Emphasis has been placed on the development of a simple, efficient process that features compact, inexpensive process equipment. Particular attention has been paid to the minimization of waste volumes and to the development of a process that is both economic and environmentally sound as well as proliferation-resistant.

  17. Development of a computational algorithm for the linearization of decay and transmutation chains

    International Nuclear Information System (INIS)

    Cruz L, C. A.; Francois L, J. L.

    2017-09-01

    One of the most used methodologies to solve Bate man equations, in the problem of burning, is the Tta (Transmutation Trajectory Analysis) method. In this method, a network of decays is broken down into linear elements known as trajectories, through a process known as linearization. In this work an alternative algorithm is shown to find and construct these trajectories, which considers three aspects of linearization: the information -a priori- about the elements that make up decay and transmutation network, the use of a new notation, and in the functions for the treatment of text strings (which are common in most programming languages). One of the main advantages of the algorithm is that can condense the information of a decay and transmutation network into only two vectors. From these is possible to determine how many linear chains can be extracted from the network and even their length (in the case they are not cyclical). Unlike the Deep First Search method, which is widely used for the linearization process, the method proposed in the present work does not have a backward routine and instead occupies a process of compilation, since completes fragments chain instead of going back to the beginning of the trajectories. The developed algorithm can be applied in a general way to the information search and to the linearization of the computational data structures known as trees. It can also be applied to engineering problems where one seeks to calculate the concentration of some substance as a function of time, starting from linear differential equations of balance. (Author)

  18. Partitioning and transmutation. Current developments - 2004. A report from the Swedish reference group on P and T-research

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstroem, Per-Eric [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Andersson, Sofie; Ekberg, Christian; Liljenzin, Jan-Olov; Nilsson, Mikael; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden); Blomgren, Jan [Uppsala Univ. (Sweden). Dept. of Neutron Research; Eriksson, Marcus; Gudowski, Waclaw; Seltborg, Per; Wallenius, Jan; Sehgal, Bal Raj [Royal Inst. of Technology, Stockholm (Sweden)

    2004-05-01

    This report summarises the work reported in the years 1998-2003 and tries to assess the prospects for future development of partitioning and transmutation (P-T) as seen from a Swedish perspective. The R and D efforts on P-T have increased somewhat during the period 1998-2003. Research on P-T has taken a prominent role internationally in the R and D on future nuclear power and nuclear fuel cycle systems. Despite the fact that partitioning and transmutation have been on the agenda for quite a few years there are still a number of issues that must be settled before the research and development can be given a clearly focused direction. Studies propose research programmes for about six years at the cost of a couple of hundred million Euros. The construction of a small ADS experimental plant is a necessary step to develop and demonstrate the concept. This experimental plant should then be followed by a demonstration plant in almost full scale. Such a plant can at the earliest be ready in the mid-2030s. A number of circumstances have, however, contributed to slower speed, less intensity and lower funding than proposed in the studies. There is no unanimous view on the objectives for partitioning and transmutation. Many see it as a way to achieve broad acceptance for nuclear power at large. Others promote it as a way to get out of the impasse for a deep repository in several countries. Others again put a strong emphasise on the proliferation aspects. There is no unanimous view on the need to develop ADS or for the role of ADS in a P-T-system. Some advocate that ADS should be used for burning of all transuranium nuclides from the present enriched uranium fuelled light water reactors. Others see the ADS as a supplement particularly suitable for burning minor actinides (americium, curium and neptunium), whereas the major part of the plutonium should be burned in light water reactors (or in fast reactors) There is no consensus among experts on which technical route to follow

  19. Partitioning and transmutation. Current developments - 2004. A report from the Swedish reference group on P and T-research

    International Nuclear Information System (INIS)

    Ahlstroem, Per-Eric; Andersson, Sofie; Ekberg, Christian; Liljenzin, Jan-Olov; Nilsson, Mikael; Skarnemark, Gunnar; Blomgren, Jan

    2004-05-01

    This report summarises the work reported in the years 1998-2003 and tries to assess the prospects for future development of partitioning and transmutation (P-T) as seen from a Swedish perspective. The R and D efforts on P-T have increased somewhat during the period 1998-2003. Research on P-T has taken a prominent role internationally in the R and D on future nuclear power and nuclear fuel cycle systems. Despite the fact that partitioning and transmutation have been on the agenda for quite a few years there are still a number of issues that must be settled before the research and development can be given a clearly focused direction. Studies propose research programmes for about six years at the cost of a couple of hundred million Euros. The construction of a small ADS experimental plant is a necessary step to develop and demonstrate the concept. This experimental plant should then be followed by a demonstration plant in almost full scale. Such a plant can at the earliest be ready in the mid-2030s. A number of circumstances have, however, contributed to slower speed, less intensity and lower funding than proposed in the studies. There is no unanimous view on the objectives for partitioning and transmutation. Many see it as a way to achieve broad acceptance for nuclear power at large. Others promote it as a way to get out of the impasse for a deep repository in several countries. Others again put a strong emphasise on the proliferation aspects. There is no unanimous view on the need to develop ADS or for the role of ADS in a P-T-system. Some advocate that ADS should be used for burning of all transuranium nuclides from the present enriched uranium fuelled light water reactors. Others see the ADS as a supplement particularly suitable for burning minor actinides (americium, curium and neptunium), whereas the major part of the plutonium should be burned in light water reactors (or in fast reactors) There is no consensus among experts on which technical route to follow

  20. LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P A; Kaufman, L; Fluss, M

    2008-12-19

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  1. An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Maeda, Shigetaka; Mitsugashira, Toshiaki

    2005-01-01

    Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the 241 Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the 240 Pu to 239 Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor

  2. Comparative study of accelerator driven system (ADS) of different transmutation scenarios for actinides in advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    Embid-Segura, M.; Gonzalez Romero, M.E.; Perez Parra, A.

    2001-01-01

    The full text follows. In recent years transmutation has raised as a complementary option to solve the problem of the long-lived radioactive waste produced in nuclear power plants. The main advantages expected from transmutation are the reduction in volume of the high level waste and a significant decrease in the long-term radiotoxicity inventory, with a probable impact in the final costs and potential risks of the geological repository. This paper will describe the evaluation of different systems proposed for actinide transmutation, their integration in the waste management process, their viability, performances and limitations. Particular attention is taking of comparing transmutation scenarios where the actinides are transmuted inside fertile (U, Th) or inert matrix. This study has been supported by ENRESA inside the CIEMAT-ENRESA collaboration for the study of long-lived isotope transmutation. (authors)

  3. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  4. Neutronics design study on a minor actinide burner for transmuting spent fuel

    International Nuclear Information System (INIS)

    Choi, Hang Bok

    1998-08-01

    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors. The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the doppler coefficient, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics. (author). 34 refs., 22 tabs., 14 figs

  5. Transmutation of long-lived nuclides

    International Nuclear Information System (INIS)

    Liang Tongxiang; Tang Chunhe

    2003-01-01

    Partitioning and transmutation of long-lived nuclides have profound benefits for economic development, global political stability and the environment. This technology would reduce nuclear waste disposal requirements, prevent proliferation and eliminate a major hurdle to the development of nuclear power. This paper reviews the advanced fuel cycle process and development of ATW in the world, and some suggestions about the R and D of nuclear power in China are proposed

  6. Minor actinides transmutation performance in a fast reactor

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2016-01-01

    Highlights: • A method for calculating MA transmutation for individual nuclides has been proposed by introducing two formulas of the MA transmutation. One corresponds to the difference of MA amounts, and the other corresponds to the sum of the fission amounts and the plutonium production amounts. • Using the method the MA transmutation was calculated for Np-237 and Am-241 in a fast reactor. The burnup period was changed from 1 year to 12 year. • For the 1 year burnup a large amount of Am-242m, Cm-242 are produced from Am-241. The total MA transmutation amount increases with burnup time, but its gradient with respect to burnup time decreases after 9 years, and the transmutation amount by overall fission increases almost linearly with burnup time. • However, after the 6 year burnup the fission contribution became large because of the large production of Pu isotopes from the original Am-241. • In addition to the homogeneous loading of the MA nuclides into the cores, a heterogeneous loading of Am-241 to the blanket region was considered. - Abstract: Results obtained in the project named “Study on Minor Actinides Transmutation using Monju Data”, which has been sponsored by the Ministry of Education, Culture, Sports, Science and Technology in Japan (MEXT) are described. In order to physically understand transmutation of individual MA nuclides in fast reactors, a new method was developed in which the MAs transmutation is interpreted by two formulas. One corresponds to the difference of individual MA nuclides amounts before and after a burnup period, and the other is the sum of amount of fission of a relevant MA nuclide and the net plutonium production from the MA nuclide during a burnup period. The method has been applied to two fast reactors with MA fuels loaded in cores homogeneously and in a blanket region heterogeneously. Numerical results of MA transmutation for the two reactors are shown.

  7. The Physics of transmutation systems : system capabilities and performances

    International Nuclear Information System (INIS)

    Finck, P. J.

    2002-01-01

    This document is complementary to a document produced by Prof. Salvatores on ''The Physics of Transmutation in Critical or Subcritical Reactors and the Impact on the Fuel Cycle''. In that document, Salvatores describes the fundamental of transmutation, through basic physics properties and general parametric studies. In the present document we try to go one step further towards practical implementation (while recognizing that the practical issues such as technology development and demonstration, and economics, can only be mentioned in a very superficial manner). Section 1 briefly overviews the possible objectives of transmutation systems, and links these different objectives to possible technological paths. It also describes the overall constraints which have to be considered when developing and implementing transmutation systems. In section 2 we briefly overview the technological constraints which need to be accounted for when designing transmutation systems. In section 3 we attempt to provide a simplified classification of transmutation systems in order to clarify later comparisons. It compares heterogeneous and homogeneous recycle strategies, and single and multi-tier systems. Section 4 presents case analyses for assessing the transmutation performance of various individual systems, starting with LWR's ((1) generic results; (2) multirecycle of plutonium; (3) an alternative: transmutation based on a Thorium fuel cycle), followed by Gas-Cooled Reactors (with an emphasis on the ''deep burn'' approach), and followed by Fast Reactors and Accelerator Driven systems ((1) generic results; (2) homogeneous recycle of transuranics; (3) practical limit between Fast Reactors and Accelerator Driven Systems) Section 5 summarizes recent results on integrated system performances. It focuses first on interface effects between the two elements of a dual tier system, and then summarizes the major lessons learned from recent global physics studies

  8. Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

    2005-08-15

    Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

  9. ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

  10. The Los Alamos accelerator driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    Venneri, F.; Williamson, M.A.; Ning, L.

    1997-01-01

    The studies carried out in the frame of the Accelerator Driven Transmutation Technology (ADTT) program developed at Los Alamos in order to solve the nuclear waste problem and to build a new generation of safer and non-proliferant nuclear power plants, are presented

  11. Partitioning and Transmutation: IAEA Activities

    International Nuclear Information System (INIS)

    Basak, U.; Monti, S.; )

    2015-01-01

    Full text of publication follows: The importance of partitioning and transmutation (P and T) processes for sustaining nuclear energy growth in the world has been realised in several countries across the world. P and T processes aim at separation and recycling of actinides including minor actinides (MAs) from the spent fuel or high-level liquid waste. The objective of these processes include reuse of separated fissile materials from spent nuclear fuels to obtain energy, enhance resource utilisation, reduce the disposal of toxic radio-nuclides and improve long-term performance of geological repositories. R and D programmes have been launched in many of the Member States to develop advanced partitioning process based on either aqueous or pyro to recover MAs along with other actinides as well as automated and remote techniques for manufacturing fuels containing MAs for the purpose of transmuting them either in fast reactors or accelerator driven hybrids. A number of Member States have been also developing such transmutation systems with the aim to construct and operate demo plants and prototypes in the next decade. The International Atomic Energy Agency has a high priority for the activities on partitioning and transmutation and regularly organises conferences, workshops, seminars and technical meetings in the areas of P and T as a part of information exchange and knowledge sharing at the international level. In the recent past, the Agency organised two technical meetings on advanced partitioning processes and actinide recycle technologies with the objective of providing a common platform for the scientists and engineers working in the areas of separation of actinides along with MAs from spent nuclear fuels and manufacturing of advanced fuels containing MAs in order to bridge the technological gap between them. In 2010, the Agency concluded a Coordinated Research Project (CRP) related to Assessment of Partitioning Processes. The Agency also conducted a first CRP on

  12. Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

    International Nuclear Information System (INIS)

    Marsodi; Mulyanto; Kitamoto, Asashi.

    1994-01-01

    Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

  13. Transmutation in ASTRID

    International Nuclear Information System (INIS)

    Grouiller, Jean-Paul; Buiron, Laurent; Mignot, Gérard; Palhier, Raphael

    2013-01-01

    Summary and future prospects for incorporating Am in ASTRID: → Potential to demonstrate the minor actinide transmutation on an industrial scale in the CFV V1 core of ASTRID: • Homogeneous concept: 2% of Am in a standard fuel; • Heterogeneous concept: 10% on UO 2 in the radial blanket. • The objective of ensuring a balance in the Am (and total minor actinides) flow in the ASTRID fuel cycle may be obtained without any impact on the design of the core and handling systems for the management of the new and spent fuel subassemblies. • Several experimental phases in ASTRID to implement different transmutation scenarios using homogeneous and heterogeneous concepts. ⇒ the availability of facilities involved in the ASTRID material cycles

  14. Research and Development of Technologies for Partitioning and Transmutation of Long-lived Nuclides in Japan - Status and Evaluation

    International Nuclear Information System (INIS)

    Sanae Aoki

    2003-01-01

    Japanese basic policy regarding disposal of high-level radioactive waste (HLW) is to solidify it into stabilized form, to store it for 30-50 years to be cooled, and to dispose of it deep to the underground (geological disposal). In Japan, reference to P and T technology for long-lived and other nuclides first appeared in the Long-term Programme for Nuclear Research, Development and Utilisation (or 'long-term nuclear programme') back in 1972. That programme noted the need for research and development in order to ensure effective processing of radioactive waste. The long-term nuclear programme issued in 1987 stated that P and T technology was very important from the viewpoint of recycling HLW and enhancing disposal efficiency. It also stated that systematic R and D would be carried out jointly by JAERI, the Power Reactor and Nuclear Fuel Development Corp. (PNC, now JNC) and others. The long-term nuclear programme issued in 1994 stated that each research institute would carry out basic studies on P and T technologies and evaluate each technology at some time in the mid- 1990's to determine how to proceed thereafter. Based on the evaluation schedule stated in this program, the AEC's Advisory Committee on Nuclear Fuel Cycle Back-end Policy investigated and considered matters concerning P and T technology for long-lived and other nuclides. In March 2000, the Committee issued a report entitled 'Research and Development of Technologies for Partitioning and Transmutation of Long-lived Nuclide Status and Evaluation Report'. A brief summary of this report is presented

  15. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  16. Minor actinide transmutation - a waste management option

    International Nuclear Information System (INIS)

    Koch, L.

    1986-01-01

    The incentive to recycle minor actinides results from the reduction of the long-term α-radiological risk rather than from a better utilization of the uranium resources. Nevertheless, the gain in generated electricity by minor actinide transmutation in a fast breeder reactor can compensate for the costs of their recovery and make-up into fuel elements. Different recycling options of minor actinides are discussed: transmutation in liquid metal fast breeder reactors (LMFBRs) is possible as long as plutonium is not recycled in light water reactors (LWRs). In this case a minor actinide burner with fuel of different composition has to be introduced. The development of appropriate minor actinide fuels and their properties are described. The irradiation experiments underway or planned are summarized. A review of minor actinide partitioning from the PUREX waste stream is given. From the present constraints of LMFBR technology a reduction of the long-term α-radiological risk by a factor of 200 is deduced relative to that from the direct storage of spent LWR fuel. Though the present accumulation of minor actinides is low, nuclear transmutation may be needed when nuclear energy production has grown. (orig.)

  17. Theoretical and Experimental Research in Neutron Spectra and Nuclear Waste Transmutation on Fast Subcritical Assembly with MOX Fuel

    Science.gov (United States)

    Arkhipkin, D. A.; Buttsev, V. S.; Chigrinov, S. E.; Kutuev, R. Kh.; Polanski, A.; Rakhno, I. L.; Sissakian, A.; Zulkarneev, R. Ya.; Zulkarneeva, Yu. R.

    2003-07-01

    The paper deals with theoretical and experimental investigation of transmutation rates for a number of long-lived fission products and minor actinides, as well as with neutron spectra formed in a subcritical assembly driven with the following monodirectional beams: 660-MeV protons and 14-MeV neutrons. In this work, the main objective is the comparison of neutron spectra in the MOX assembly for different external driving sources: a 660-MeV proton accelerator and a 14-MeV neutron generator. The SAD project (JINR, Russia) has being discussed. In the context of this project, a subcritical assembly consisting of a cylindrical lead target surrounded by a cylindrical MOX fuel layer will be constructed. Present conceptual design of the subcritical assembly is based on the core with a nominal unit capacity of 15 kW (thermal). This corresponds to a multiplication coefficient, keff= 0.945, and an accelerator beam power of 0.5 kW. The results of theoretical investigations on the possibility of incinerating long-lived fission products and minor actinides in fast neutron spectrum and formation of neutron spectra with different hardness in subcritical systems based on the MOX subcritical assembly are discussed. Calculated neutron spectra emitted from a lead target irradiated by a 660-MeV protons are also presented.

  18. A review of research and development on accelerator-driven system for transmutation of long-lived nuclear waste at JAERI

    International Nuclear Information System (INIS)

    Oigawa, H.

    2004-01-01

    The dedicated transmutation system using the accelerator driven subcritical system (ADS) has been studied in the Japan Atomic Energy Research Institute (JAERI) to reduce the burden of the final disposal of the nuclear waste. A subcritical reactor with the thermal power of 800 MW is proposed, where 250 kg of minor actinide (MA) can be transmuted annually. A superconducting linear accelerator (LINAC) with the beam power of 20-30 MW is necessary for this ADS. Lead-bismuth eutectic (LBE) is used for both the spallation target and the core coolant. Many research and development activities including the design study are under way and planned at JAERI to examine the feasibility of the ADS. In the design study, optimization of the ADS design is under way in terms of neutronics and structural feasibility. In the field of the proton accelerator, a superconducting LINAC is being developed. In the field of the LBE technology, material compatibility, thermal-hydraulics and polonium behavior are being studied. The irradiation effect of structural material to be used for the beam window is also being studied. In the field of the reactor physics of the subcritical core fueled with MA, the reliability of nuclear data is examined and the subcriticality monitoring technique is being investigated. Moreover, in the framework of J-PARC project (Japan Proton Accelerator Research Complex), JAERI plans to construct the Transmutation Experimental Facility (TEF) to demonstrate the feasibility of the ADS with using high-energy proton beam, to accumulate valuable knowledge about reactor physics and operation of ADS, and to establish a database for LBE spallation target and relevant materials. (author)

  19. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    ... to light gases then steam reform the light gases into hydrogen rich stream. This report documents the efforts in developing a fuel processor capable of providing hydrogen to a 3kW fuel cell stack...

  20. The concept of fuel cycle integrated molten salt reactor for transmuting Pu+MA from spent LWR fuels

    International Nuclear Information System (INIS)

    Hirose, Y.; Takashima, Y.

    2001-01-01

    Japan should need a new fuel cycle, not to save spent fuels indefinitely as the reusable resources but to consume plutonium and miner actinides orderly without conventional reprocessing. The key component is a molten salt reactor fueled with the Pu+MA (PMA) separated from LWR spent fuels using fluoride volatility method. A double-tiered once-through reactor system can burn PMA down to 5% remnant ratio, and can make PMA virtually free from the HAW to be disposed geometrically. A key issue to be demonstrated is the first of all solubility behavior of trifluoride species in the molten fuel salt of 7 LiF-BeF 2 mixture. (author)

  1. Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

    International Nuclear Information System (INIS)

    Aziz, F.; Kitamoto, Asashi.

    1996-01-01

    A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237 Np, 241 Am, and 243 Am burned by thermal neutrons, while in the inner region 244 Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

  2. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Sergi

    2012-01-01

    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

  3. Actinide transmutation in nuclear reactors

    International Nuclear Information System (INIS)

    Bultman, J.H.

    1995-01-01

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP)

  4. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J H

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  5. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The fourth international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Mito City in Japan, on 111-13 September 1996. The proceedings are presented in six sessions: the major programmes and international cooperation, the partitioning and transmutation programs, feasibility studies, particular separation processes, the accelerator driven transmutation, and the chemistry of the fuel cycle. (A.L.B.)

  6. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  7. Transmutation Dynamics: Impacts of Multi-Recycling on Fuel Cycle Performances

    Energy Technology Data Exchange (ETDEWEB)

    S. Bays; S. Piet; M. Pope; G. Youinou; A. Dumontier; D. Hawn

    2009-09-01

    From a physics standpoint, it is feasible to sustain continuous multi-recycle in either thermal or fast reactors. In Fiscal Year 2009, transmutaton work at INL provided important new insight, caveats, and tools on multi-recycle. Multi-recycle of MOX, even with all the transuranics, is possible provided continuous enrichment of the uranium phase to ~6.5% and also limitting the transuranic enrichment to slightly less than 8%. Multi-recycle of heterogeneous-IMF assemblies is possible with continuous enrichment of the UOX pins to ~4.95% and having =60 of the 264 fuel pins being inter-matrix. A new tool enables quick assessment of the impact of different cooling times on isotopic evolution. The effect of cooling time was found to be almost as controlling on higher mass actinide concentrations in fuel as the selection of thermal versus fast neutron spectra. A new dataset was built which provides on-the-fly estimates of gamma and neutron dose in MOX fuels as a function of the isotopic evolution. All studies this year focused on the impact of dynamic feedback due to choices made in option space. Both the equilibrium fuel cycle concentrations and the transient time to reach equilibrium for each isotope were evaluated over a range of reactor, reprocessing and cooling time combinations. New bounding cases and analysis methods for evaluating both reactor safety and radiation worker safety were established. This holistic collection of physics analyses and methods gives improved resolution of fuel cycle options, and impacts thereof, over that of previous ad-hoc and single-point analyses.

  8. Partitioning and transmutation (P and D) 1995. A review of the current state of the art

    International Nuclear Information System (INIS)

    Skaalberg, M.; Landgren, A.; Spjuth, L.; Liljenzin, J.O.; Gudowski, W.

    1995-12-01

    The recent development in the field of partitioning and transmutation (P/T) is reviewed and evaluated. Current national and international R and D efforts are summarized. Nuclear transmutation with energy production is feasible in nuclear reactors where fast and thermal breeders are the most efficient for transmutation purposes. The operation of subcritical nuclear reactors by high current proton accelerators that generate neutrons in a spallation target is also an interesting option for transmutation and energy production, that has to be more carefully evaluated. These accelerator-driven systems are probably the only solution for the transmutation of long-lived fission products with small neutron capture cross sections and actinide isotopes with small fission cross sections. The requirements on the separation chemistry in the partitioning process depends on the transmutation strategy chosen. Recent developments in aqueous based separation chemistry opens some interesting possibilities to meet some of the requirements, such as separation of different actinides and some fission products and reduction of secondary waste streams. In the advanced accelerator-driven transmutation systems proposed, liquid fuels such as molten salts are considered. The partitioning processes that can be used for these types of fuel will, however, require a long term research program. The possibility to use centrifuge separation is an interesting partitioning option that recently has been proposed. 51 refs, 7 figs, 3 tabs

  9. 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

  10. Planning the research and development necessary for accelerator transmutation of waste, leading to integrated proof of performance testing

    International Nuclear Information System (INIS)

    Bennett, D.R.; Pasamehmetoglu, K.; Finck, P.; Pitcher, E.; Khalil, H.; Todosow, M.; Hill, R.; Van Tuyle, G.; Laidler, J.; Crawford, D.; Thomas, K.

    2001-01-01

    The Research and Development (R and D) Plan for the Accelerator Transmutation of Waste (ATW) Program has been developed for the Department of Energy, Office of Nuclear Energy (DOE/NE) to serve as a focus and progressional guide in developing critical transmutation technologies. It is intended that the Plan will serve as a logical reference considering all elements of an integrated accelerator-driven transmutation system, and will maximize the use of resources by identifying and prioritizing research, design, development and trade activities. The R and D Plan provides a structured framework for identifying and prioritizing activities leading to technically-justifiable integrated Proof of Performance testing within ten years and ultimate demonstration of Accelerator Transmutation of Waste (ATW). The Plan builds from the decision objectives specified for ATW, utilizes informational input from the ATW Roadmap and programmatic System Point Design efforts, and employs the knowledge and expertise provided by professionals familiar with ATW technologies. With the firm intent of understanding what, why and when information is needed, including critical interfaces, the Plan then develops a progressional strategy for developing ATW technologies with the use of a Technology Readiness Level (TRL) scale. The TRL approach is first used to develop a comprehensive, yet generic, listing of experimental, analytical and trade study activities critical to developing ATW technologies. Technology-specific and concept-specific aspects are then laid over the generic mapping to gage readiness levels. Prioritization criteria for reducing technical uncertainty, providing information to decision points, and levering off of international collaborations are then applied to focus analytical, experimental and trade activities. (author)

  11. Metallic fuel development

    International Nuclear Information System (INIS)

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  12. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  13. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  14. Fissile fuel breeding and minor actinide transmutation in the life engine

    International Nuclear Information System (INIS)

    Sahin, Suemer; Khan, Mohammad Javed; Ahmed, Rizwan

    2011-01-01

    Progress on The National Ignition Facility (NIF) brings fusion a viable energy source in foreseeable future. Energy multiplication in a fusion-fission (hybrid) reactor could lead earlier market penetration of fusion energy for commercial utilization. Originally, scientists at the Lawrence Livermore National Laboratory (LLNL) have worked out a hybrid reactor design concept; the so-called Laser Inertial Confinement Fusion-Fission Energy (LIFE) engine, which has consisted of a spherical fusion chamber of ∼5 m diameter, surrounded by a multi-layered blanket with a beryllium multiplier zone after the first wall. However, earlier work had indicated extreme power peaks at immediate vicinity of the first wall of a hybrid assembly, if a beryllium multiplier is used. Hence, in the current work, the beryllium multiplier zone has been removed in order to mitigate fission power peaks at the vicinity of the first wall as a result of neutron moderation on beryllium. Furthermore, minor actinides (MA) will cause significant neutron multiplication under fusion neutron irradiation so that an extra beryllium multiplier will not be needed. Present work has made following modifications on the LLNL design of the original (LIFE) engine: ·Omission of beryllium multiplier. ·TRISO fuel has been suspended as micro-size particles in Flibe coolant in lieu of being dissolved in uranium salt or imbedded carbon matrix in macro-size pebbles. ·Carbide fuel is used. ·Fissionable fuel charge is kept lower than in the LLNL (LIFE) engine. The modified (LIFE) engine is kept similar to the LLNL design to a great degree in order to allow mutual feedback between two geographically separated teams towards a more advanced and improved design under consideration of totally independent views. The first wall is made of ODS (2 cm) and followed by a Li 17 Pb 83 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the Flibe molten salt

  15. Deep-Burn Modular Helium Reactor Fuel Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  16. Transmutation of high-level radioactive waste - Perspectives

    CERN Document Server

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

  17. Enhancement of actinide incineration and transmutation rates in Ads EAP-80 reactor core with MOX PuO2 and UO2 fuel

    International Nuclear Information System (INIS)

    Kaltcheva-Kouzminava, S.; Kuzminov, V.; Vecchi, M.

    2001-01-01

    Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 and PuO 2 MOX fuel elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber B4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B4 C arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second scheme H2 Zr zirconium hydride moderated zones were located in fuel elements with low enrichment (∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the reference EAP-80 core and at the same time the power density distribution remains significantly unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the inner fuel region with UO 2 and PuO 2 high enrichment plutonium fuel and the second one is the outer region with fuel elements containing zirconium hydride layer was also considered. Evolution of neutronics parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were performed using the MCNP-4B code and the SCALE 4.3 computational system. (author)

  18. Vortex transmutation.

    Science.gov (United States)

    Ferrando, Albert; Zacarés, Mario; García-March, Miguel-Angel; Monsoriu, Juan A; de Córdoba, Pedro Fernández

    2005-09-16

    Using group theory arguments and numerical simulations, we demonstrate the possibility of changing the vorticity or topological charge of an individual vortex by means of the action of a system possessing a discrete rotational symmetry of finite order. We establish on theoretical grounds a "transmutation pass" determining the conditions for this phenomenon to occur and numerically analyze it in the context of two-dimensional optical lattices. An analogous approach is applicable to the problems of Bose-Einstein condensates in periodic potentials.

  19. Target development and transmutation experiments in the frame of the EFTTRA European collaboration

    International Nuclear Information System (INIS)

    Prunier, C.; Salvatores, M.; Muehling, G.; Rome, M.

    1995-01-01

    The aim of the EFTTRA collaboration between CEA (France), ECN (The Netherlands), EDF (France), FZK (Germany), IAM and ITU (European Commission), is to organize joint experiments for the study of materials for transmutation in reactors. The work is focused on the transmutation of 99 Tc (metal), of 129 I (compound), and of Am (in an inert matrix). Irradiation experiments are taking place in parallel in the Phenix fast reactor in France, and in the high flux thermal reactor HFR in the Netherlands. Examination of iodine compounds and Tc samples, following irradiation in HFR, has started. (authors). 10 refs., 2 figs

  20. Partitioning and Transmutation of minor actinides

    International Nuclear Information System (INIS)

    Koch, L.; Wellum, R.

    1991-01-01

    The partitioning of minor actinides from spent fuels and their transmutation into short-lived fission products has been the topic of two dedicated meetings organized jointly by the European Commission and the OECD. The conclusion of the last meeting in 1980, in short, was that partitioning and transmutation of minor actinides, especially in fast reactors, seemed possible. However, the incentive, which would be a reduction of the radiological hazard to the public, was too small if long-lived fission products were not included. Furthermore this meeting showed that minor actinide targets or possible nuclear fuels containing minor actinides for transmutation had not yet been developed. The European Institute for Transuranium Elements took up this task and has carried it out as a small activity for several years. Interests expressed recently by an expert meeting of the OECD/NEA (Paris, 25 April 1989), which was initiated by the proposed Japanese project Omega, led us to the conclusion that the present state of knowledge should be looked at in a workshop environment. Since the Japanese proposal within the project Omega is based on a broader approach we needed this evaluation to assess the relevance of our present activity and wanted to identifiy additional studies which might be needed to cover possible future demands from the public. This workshop was therefore organized, and participants active in the field from EC countries, the USA and Japan were invited

  1. Actinide Partitioning and Transmutation Program. Progress report, April 1--June 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Tedder, D. W.; Blomeke, J. O. [comps.

    1977-10-01

    Experimental work on the 16 tasks comprising the Actinide Partitioning and Transmutation Program was continued. Summaries of work are given on Purex Process modifications, actinide recovery, Am-Cm recovery, radiation effects on ion exchangers, LMFBR transmutation studies, thermal reactor transmutation studies, fuel cycle studies, and partitioning-transmutation evaluation. (JRD)

  2. Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, F.; Lahoda, E.; Wenner, M. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Lindley, B. [University of Cambridge (United Kingdom); Fiorina, C. [Polytechnic of Milan (Italy); Phillips, C. [Energy Solutions, Richland, WA (United States)

    2013-07-01

    This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principle be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)

  3. Beta transmutations in apatites with ferric iron as an electron acceptor - implication for nuclear waste form development.

    Science.gov (United States)

    Yao, Ge; Zhang, Zelong; Wang, Jianwei

    2017-09-27

    Apatite-structured materials have been considered for the immobilization of a number of fission products from reprocessing nuclear fuel because of their chemical durability as well as compositional and structural flexibility. It is hypothesized that the effect of beta decay on the stability can be mitigated by introducing an appropriate electron acceptor at the neighboring sites in the structure. The decay series 137 Cs → 137 Ba and 90 Sr → 90 Y → 90 Zr were investigated using a spin-polarized DFT approach to test the hypothesis. Apatites with compositions of Ca 10 (PO 4 ) 6 F 2 and Ca 4 Y 6 (SiO 4 ) 6 F 2 were selected as model systems for the incorporation of radionuclides Cs and Sr, respectively. Ferric iron was introduced in the structure as an electron acceptor. Electron density of states, crystal and defect structures, and energies before and after beta decay were calculated. The calculated electron density of states suggests that the extra electron is localized at the ferric iron, which changes its oxidation state and becomes ferrous iron. The crystal and defect structures were analyzed based on the volume, lattice parameters, radial distribution functions, metal cation to coordinating oxygen distances, and the metaprism twist angle of the apatite crystal structure. The results show that there are minor changes in the crystal and defect structures of CsFeCa 8 (PO 4 ) 6 F 2 with Cs + and Fe 3+ substitutions undergoing the Cs → Ba transmutation, and of Ca 3 SrY 4 Fe 2 (SiO 4 ) 6 F 2 with Sr 2+ and Fe 3+ substitutions undergoing the Sr → Y → Zr transmutations. The last decay change, from Y 3+ → Zr 4+ , causes relatively larger changes in the local defect structure around Zr involving the coordination environment but the change is not significant to the crystal structure. The results on calculated cohesive energy suggest that the transmutations Cs + → Ba 2+ and Sr 2+ → Y 3+ → Zr 4+ in both apatite compositions are energetically favorable

  4. Vortex Transmutation

    International Nuclear Information System (INIS)

    Ferrando, Albert; Garcia-March, Miguel-Angel; Zacares, Mario; Monsoriu, Juan A.; Cordoba, Pedro Fernandez de

    2005-01-01

    Using group theory arguments and numerical simulations, we demonstrate the possibility of changing the vorticity or topological charge of an individual vortex by means of the action of a system possessing a discrete rotational symmetry of finite order. We establish on theoretical grounds a 'transmutation pass rule' determining the conditions for this phenomenon to occur and numerically analyze it in the context of two-dimensional optical lattices. An analogous approach is applicable to the problems of Bose-Einstein condensates in periodic potentials

  5. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    Williams, T.

    1997-01-01

    BNFL has a long tradition of willingness to embrace technological challenge and a dedication to quality. This paper describes advances in the overall manufacturing philosophy at BNFL's Fuel Business Group and then covers how some new technologies are currently being employed in BNFL Fuel Business Group's flagship oxide complex (OFC), which is currently in its final stages of commissioning. This plant represents a total investment of some Pound 200 million. This paper also describes how these technologies are also being deployed in BNFL's MOX plant now being built at Sellafield and, finally, covers some new processes being developed for advanced fuel manufacture. (author)

  6. Canadian fuel development program

    International Nuclear Information System (INIS)

    Gacesa, M.; Young, E.G.

    1992-11-01

    CANDU power reactor fuel has demonstrated an enviable operational record. More than 99.9% of the bundles irradiated have provided defect-free service. Defect excursions are responsible for the majority of reported defects. In some cases research and development effort is necessary to resolve these problems. In addition, development initiatives are also directed at improvements of the current design or reduction of fueling cost. The majority of the funding for this effort has been provided by COG (CANDU Owners' Group) over the past 10 to 15 years. This paper contains an overview of some key fuel technology programs within COG. The CANDU reactor is unique among the world's power reactors in its flexibility and its ability to use a number of different fuel cycles. An active program of analysis and development, to demonstrate the viability of different fuel cycles in CANDU, has been funded by AECL in parallel with the work on the natural uranium cycle. Market forces and advances in technology have obliged us to reassess and refocus some parts of our effort in this area, and significant success has been achieved in integrating all the Canadian efforts in this area. This paper contains a brief summary of some key components of the advanced fuel cycle program. (Author) 4 figs., tab., 18 refs

  7. Partitioning and Transmutation - Physics, Technology and Politics

    International Nuclear Information System (INIS)

    Gudowski, W.

    2002-01-01

    Nuclear reactions can be effectively used to destroy radio toxic isotopes through transmutation processes transforming those isotopes into less radio toxic or stable ones Spent nuclear fuel, a mixture of many isotopes with some of them being highly radio toxic for many hundred thousands of years, may be effectively transmuted through nuclear reactions with neutrons. In a dedicated, well designed transmutation system one can, in principle, reduce the radiotoxicity of the spent nuclear fuel to a level, which will require isolation from the biosphere for the period of time for which engineered barriers can be constructed and licensed (not more than 1-2 thousands of years). En effective transmutation process can not be achieved without a suitable partitioning. Only partitioning of the spent nuclear fuel into predetermined groups of elements makes possible an effective use of neutrons to transmute long-lived radioactive isotopes into short-lived or stable one. However, most of the chemical separation/partitioning processes are element- not isotope-specific, therefore the transmutation of the elements with an existing isotope composition is a typical alternative for transmutation processes. Isotope-specific separation is possible but still very expensive and technologically not matured

  8. Accelerator transmutation of waste economics

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A parametric systems model of the accelerator transmutation of (nuclear) waste (ATW) is used to examine key system trade-offs and design drivers on the basis of unit costs. This model is applied primarily to a fluid-fuel blanket concept for an ATW that generates net electric power from the fissioning of spent commercial reactor fuel. An important goal of this study is the development of essential parametric trade-offs to aid in any future conceptual engineering design of an ATW that would burn spent commercial fuel and generate net electric power. As such, costing procedures and methodologies used to estimate and compare advanced nuclear power generation systems are applied. The cost of electricity required by an electrical power-generating ATW fueled with spent commercial fuels is generally found to be above that projected for other advanced fission power plants. The accelerator and the chemical plant equipment cost accounts are quantitatively identified as main cost drivers, with the capital cost of radio-frequency power dominating the former. Significant reductions of this cost differential are possible by increased blanket neutron multiplication, increased plant capacity, or increased thermal-to-electric conversion efficiency. The benefits of reduced long-lived fission products and spent commercial fuel actinides provided by the ATW approach translate into a less tangible source of revenue to be provided by a charge that must be levied on the client fission power plants being serviced. The main goal of this study, however, is not a direct cost comparison but is instead a quantitative determination of cost-based sensitivity of key cost drivers and operational modes for an ATW concept that would address the growing spent commercial fuel problem; parametric results presented focus on this goal, and a specific ATW ''straw man'' is given to achieve this main objective

  9. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    Ion, S.E.; Harrop, G.; Maricalva Gonzalez, J.

    1995-01-01

    The status of the investment and R and D programmes in the UK and Spanish fuel fabrication facilities is outlined. Due to a number of circumstances, BNFL and ENUSA have been in the forefront of capital investment, with associated commitment to engineering and scientific research and development. Carrying through this investment has allowed the embodiment of proven state of the art technologies in the design of fuel fabrication plants, with particular emphasis on meeting the future challenge of health and safety, and product quality, at an acceptable cost. ENUSA and BNFL currently supply fuel, not only to their respective 'home' markets but also to France, Belgium, Sweden, and Germany. Both organisations employ an International Business outlook and partake in focused and speculative R and D projects for the design and manufacture of nuclear fuel. (orig./HP)

  10. Accelerator driven nuclear energy and transmutation systems

    International Nuclear Information System (INIS)

    Boldeman, J.W.

    1999-01-01

    Nuclear power generation has been a mature industry for many years. However, despite the overall safety record and the great attractions of nuclear power, especially in times of concern about green house gases emissions, there continues to be some lack of public acceptance of this technology. This sensitivity to nuclear power has several elements in addition to the concern of a potential nuclear accident. These include the possible diversion of plutonium into nuclear weapon production and the concern about the long term storage of plutonium and other transuranic elements. A concept which seeks to allay these fears but still takes advantage of the nuclear fuel cycle and utilises decades of research and development in this technology, is the idea of using modern accelerators to transmute the long lived radio nuclides and simultaneously generate power. A review of the novel concepts for energy production and transmutation of isotopes will be presented. Of the various proposals, the most developed is the Energy Amplifier Concept promoted by Rubbia. The possibility of using high-energy, high-current accelerators to produce large fluxes of neutrons has been known since the earliest days of accelerator technology. E.O. Lawrence, for example, promoted the concept of producing nuclear material with such an accelerator. The Canadians in the early 50s considered using accelerators to produce fuel for their heavy water reactors and there were well advanced designs for a device called the Intense Neutron Generator. The speculative idea of using accelerator produced neutrons for the transmutation of transuranic elements (i.e. elements such as neptunium plutonium and other elements with higher Z atomic number) has also been studied extensively, notably at a number of laboratories in the US, Europe and Japan. However at this time, all facilities that have actually been constructed have been designed primarily for condensed matter studies i.e. studies of the structural properties

  11. Transmutation of transuranium elements in a gas-cooled accelerator-driven system

    International Nuclear Information System (INIS)

    Biss, Klaus Hendrik

    2014-01-01

    The peaceful usage of nuclear energy by light and boiling water reactors is connected with a buildup of long-lived high-level radioactive waste. Compared to the direct disposal, partitioning and transmutation (P and T) is considered as an effective way to reduce this waste in its quantity by converting it into short-lived radio nuclides. By that the long term radiotoxicity is reduced compared to direct disposal. Subcritical systems, which are powered by spallation processes for free neutron production to maintain the nuclear chain reaction, allow a target-oriented transmutation. As a subcritical system a gas-cooled accelerator driven system (ADS) for transmutation of transuranic elements has been modeled in this thesis to evaluate the reduction of the radio toxicity by P and T. The simulation of neutron-physical processes is based on the Monte Carlo computer program MCNPX. The development of an equilibrium core made it possible to study the transmutation and operating behavior for several fuel variations in a magnesium oxide matrix and develop a simplified burnup method. Americium as part of the fuel has a stabilizing effect on the neutron multiplication due to its conversion into plutonium during the operation. Thorium was investigated as an alternative matrix for the fuel in order to replicate the stabilizing effect of americium by the conversion of thorium in 233 U. By that a consistent operating cycle in the later P and T-process is ensured. Calculation of the nuclide composition at the end of a P and T-process leads to an expansion of the mathematical description of the mass reduction (transmutation efficiency) by the material located in the reactor. The achieved transmutation efficiency with the investigated ADS is 98.8 %. The transmutation time was examined with different operating strategies regarding the number, size and thermal power of use of transmutation facilities to determine the effort for the P and T-process depending on efficiency. It turns out

  12. Transmutations across hierarchical levels

    International Nuclear Information System (INIS)

    O'Neill, R.V.

    1977-01-01

    The development of large-scale ecological models depends implicitly on a concept known as hierarchy theory which views biological systems in a series of hierarchical levels (i.e., organism, population, trophic level, ecosystem). The theory states that an explanation of a biological phenomenon is provided when it is shown to be the consequence of the activities of the system's components, which are themselves systems in the next lower level of the hierarchy. Thus, the behavior of a population is explained by the behavior of the organisms in the population. The initial step in any modeling project is, therefore, to identify the system components and the interactions between them. A series of examples of transmutations in aquatic and terrestrial ecosystems are presented to show how and why changes occur. The types of changes are summarized and possible implications of transmutation for hierarchy theory, for the modeler, and for the ecological theoretician are discussed

  13. First results and future trends for the transmutation of long-lived radioactive wastes

    International Nuclear Information System (INIS)

    Prunier, C.; Salvatores, M.; Guerin, Y.; Zaetta, A.

    1993-01-01

    In the frame of the CEA SPIN program, a project has been set-up at the Direction of Nuclear Reactors of CEA, to study the transmutation of long-lived radioactive products (both minor actinides and fission products) resulting from the operation of current nuclear power plants. The program is focused on: transmutation in minor actinides (Np, Am) in fission reactors of known technology (both of the PWR or the fast reactor type), using the so-called ''homogeneous'' (mixed with Uranium or Uranium-Plutonium), and ''heterogeneous'' (mixed with inert matrices) recycling modes for both type of reactors. Transmutation studies in dedicated devices (both fission reactors with actinide/plutonium fuel or with high thermal flux, and particle accelerator-based systems). Fuel studies related to both homogeneous and heterogeneous recycling modes in fission reactors. For the homogeneous recycling mode, some experimental irradiations results are available from past PHENIX programs. For the heterogeneous mode, very limited experimental results are available, and new theoretical and experimental work is underway on the use of appropriate inert matrices. Basic data studies to assess the quality of existing nuclear data for fission reactor transmutation studies, future data needs of relevance, and model/data developments needed for accelerator-based systems. Strategy studies, to evaluate the consequences of the different transmutation options on the fuel cycle, according to different scenarios of nuclear power development. 7 refs., 3 figs., 5 tabs

  14. Fuel development studies

    International Nuclear Information System (INIS)

    Michel, F.

    1986-12-01

    This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series [fr

  15. Evaluation of the possibility of plutonium and minor actinides transmutation in HWR

    International Nuclear Information System (INIS)

    Ghitescu, P.; Ghizdeanu, N. B.

    2008-01-01

    Partitioning and Transmutation (P and T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. Until now, for this purpose were studied ADS and/or FBR, but not HWR. There are several developed computer codes that analyze the inventory of the radio-nuclides in spent fuel before and after transmutation. WIMSD code is a deterministic lattice spectrum code, which can analyze the reactor neutronic behaviour It also has the capacity to generate burn up and can calculate the inventory of the radio-nuclides of the spent fuel. The advantage of WIMSD code is the variety of the created geometries, together with the big amount of calculated information (K-infinite, macroscopic cross-sections, burnable material radioactive inventory etc). Starting from WIMSD code, the paper presents a model, which simulates the possibility of fuel transmutation in PHWRs. First step was to propose a model, which simulates a CANDU reactor lattice and calculate the radionuclides inventory in an irradiated CANDU fuel bundle. The results were compared with the existing experimental data from CANDU reactors and the calculated parameters were found to be in good agreement with them. After the validation, several simulations were made for PHWRs in order to establish the optimal parameters, related to the efficiency of the transmutation process. Therefore, the code was used for a new type of fuel, containing Plutonium and minor actinides that could be transmuted. The new radioactive inventories were calculated. The simulations showed that Pu content decreases up to 8% in a CANDU reactor and 25% in an ACR. Thus, ACR can reduce the Plutonium inventory from MOX fuel and could be a transmutation solution. (authors)

  16. Chemical separations schemes for partitioning and transmutation systems

    International Nuclear Information System (INIS)

    Laidler, J.

    2002-01-01

    In the initial phase of the U.S. Accelerator Transmutation of Waste (ATW) program, a single-tier system was foreseen in which the transuranics and long-lived fission products (specifically, 99 Tc and 129 I) recovered from spent LWR oxide fuel would be sent directly to an accelerator-driven transmuter reactor [1]. Because the quantity of fuel to be processed annually was so large (almost 1,500 tons per year), an aqueous solvent extraction process was chosen for LWR fuel processing. Without the need to separate transuranics from one another for feed to the transmuter, it became appropriate to develop an advanced aqueous separations method that became known as UREX. The UREX process employs an added reagent (acetohydroxamic acid) that suppresses the extraction of plutonium and promotes the extraction of technetium together with uranium. Technetium can then be efficiently removed from the uranium; the recovered uranium, being highly decontaminated, can be disposed of as a low-level waste or stored in an unshielded facility for future use. Plutonium and the other transuranic elements, plus the remaining fission products, are directed to the liquid waste stream. This stream is calcined, converting the transuranics and fission products to their oxides. The resulting oxide powder, now representing only about four percent of the original mass of the spent fuel, is reduced to metallic form by means of a pyrometallurgical process. Subsequently, the transuranics are separated from the fission products in another pyro-metallurgical step involving molten salt electrorefining

  17. Project fuel development

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1981-05-01

    The activities continued on lab-scale production of uranium-plutonium carbide fuel for the fast reactor using gelation methods, irradiation testing and performance evaluation. Whereas in earlier years a balance was maintained between research and development or with emphasis on research, 1980 was marked by a concentrated equipment development effort for an increased throughput with improved process control and product reproducability and installation of new equipment for large pin fabrication. (Auth.)

  18. Estimation of Future Demand for Neutron-Transmutation-Doped Silicon Caused by Development of Hybrid Electric Vehicle

    International Nuclear Information System (INIS)

    Kim, Myong Seop; Park, Sang Jun

    2008-01-01

    By using this doping method, silicon semiconductors with an extremely uniform dopant distribution can be produced. They are usually used for high power devices such as thyristor (SCR), IGBT, IGCT and GTO. Now, the demand for high power semiconductor devices has increased rapidly due to the rapid increase of the green energy technologies. Among them, the productions of hybrid cars or fuel cell engines are excessively increased to reduce the amount of discharged air pollution substances, such as carbon dioxide which causes global warming. It is known that the neutron-transmutation-doped floating-zone (FZ) silicon wafers are used in insulated-gate bipolar transistors (IGBTs) which control the speed of the electric traction motors equipped in hybrid or fuel cell vehicles. Therefore, inevitably, it can be supposed that the demand of the NTD silicon is considerably increased. However, it is considered likely that the irradiation capacity will not be large enough to meet the increasing demand. After all, the large irradiation capacity for NTD such as a reactor dedicated to the silicon irradiation will be constructed depending on the industrial demand for NTD silicon. In this work, we investigated the relationship between the hybrid electric vehicle (HEV) industry and the NTD silicon production. Also, we surveyed the prospect for the production of the HEV. Then, we deduced the worldwide demand for the NTD silicon associated with the HEV production. This work can be utilized as the basic material for the construction of the new irradiation facility such as NTD-dedicated neutron source

  19. Performance of a transmutation advanced device for sustainable energy application

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, C.; Rosales, J.; Garcia, L. [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Perez-Navarro, A.; Escriva, A. [Universidad Politecnica de Valencia, Valencia (Spain). Inst. de Ingenieria Energetica; Abanades, A. [Universidad Politecnica de Madrid (Spain). Grupo de Modelizacion de Sistemas Termoenergeticos

    2009-07-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  20. Performance of a transmutation advanced device for sustainable energy application

    International Nuclear Information System (INIS)

    Garcia, C.; Rosales, J.; Garcia, L.; Perez-Navarro, A.; Escriva, A.; Abanades, A.

    2009-01-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  1. Comparative study for minor actinide transmutation in various fast reactor core concepts

    International Nuclear Information System (INIS)

    Ohki, S.

    2001-01-01

    A comparative evaluation of minor actinide (MA) transmutation property was performed for various fast reactor core concepts. The differences of MA transmutation property were classified by the variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The changes of MA transmutation property by fuel and coolant types are comparatively small. The effects caused by the difference of core design are rather significant. (author)

  2. Separation of actinides and their transmutation

    International Nuclear Information System (INIS)

    Bouchard, M.; Bathelier, M.; Cousin, M.

    1978-08-01

    Neutron irradiation of long-half-life actinides for transmutation into elements with shorter half-life is investigated as a means to reduce the long-term hazards of these actinides. The effectiveness of the method is analysed by applying it to fission product solutions from the first extraction cycle of fuel reprocessing plants. Basic principles, separation techniques and transmutation efficiencies are studied and discussed in detail

  3. Metallic fuel design development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

  4. Metallic fuel design development

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

  5. Partitioning and Transmutation. Annual Report 2006

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, Isabelle; Englund, Sofie; Fermvik, Anna; Liljenzin, Jan-Olov; Neumayer, Denis; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Chemical and Biological Engineering

    2007-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ({sup 79}Se, {sup 87}Rb, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, {sup 135}Cs) and activation products ({sup 14}C, {sup 36}Cl, {sup 59}Ni, {sup 93} Zr, {sup 94} To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the EUROPART project within the European Union sixth framework program. This is a continuation of the projects we participated in within the fourth and fifth framework programmes, NEWPART and PARTNEW, respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since the basic understanding is still needed we have our main focus on the chemical processes and understanding of how they work. Work is progressing in relation to a proposal for the 7th framework programme. This proposal will be aiming at a pilot plant for separation for transmutation purposes.

  6. Capacity of burning and transmutation reactor and grouping in partitioning of HLW in self-consistent fuel recycle

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto

    1993-01-01

    The concept of capacity of B/T reactor and grouping for partitioning of HLW has been developed in order to perform self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ( 99 Te, 129 I, and 135 Cs), Group B ( 137 Cs and 90 Sr) and Group R (the partitioned remain of HLW). In this study P-T treatment were optimized for the in-core and out-core system, respectively. (author). 7 refs., 10 figs

  7. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  8. Actinide partitioning-transmutation program final report. I. Overall assessment

    International Nuclear Information System (INIS)

    Croff, A.G.; Blomeke, J.O.; Finney, B.C.

    1980-06-01

    This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of 99 Tc and 129 I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted

  9. Partitioning and Transmutation. Annual Report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Sofie; Drouet, Francois; Ekberg, Christian; Liljenzin, Jan-Olov; Magnusson, Daniel; Nilsson, Mikael; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Materials and Surface Chemistry

    2005-01-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ({sup 129}I, {sup 99}Tc, {sup 135}Cs, {sup 93}Zr and {sup 126}Sn and activation products ({sup 14}C and {sup 36}Cl). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. This separation is necessary to obtain the desired efficiency in the transmutation process in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the European Union sixth framework program project EUROPART. This is a continuation of the projects we participated in within the fourth and fifth framework programmes NEWPART and PARTNEW respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development.

  10. Partitioning and Transmutation. Annual Report 2005

    International Nuclear Information System (INIS)

    Andersson, Sofie; Ekberg, Christian; Fermvik, Anna; Hervieux, Nadege; Liljenzin, Jan-Olov; Magnusson, Daniel; Nilsson, Mikael; Retegan, Teodora; Skarnemark, Gunnar

    2006-01-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ( 79 Se, 87 Rb, 99 Tc, 107 Pd, 126 Sn, 129 I, 135 Cs) and activation products ( 14 C, 36 Cl, 59 Ni, 93 Zr, 94 N To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers Univ. of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the EUROPART project within the European Union sixth framework program. This is a continuation of the projects we participated in within the fourth and fifth framework programmes, NEWPART and PARTNEW respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since the basic understanding is still needed we have our main focus on the chemical processes and understanding of how they work

  11. Nudatra: nuclear data for transmutation in IP-Eurotrans

    International Nuclear Information System (INIS)

    Gonzalez, E.M.; Koning, A.; Leray, S.; Plompen, A.; Sanz, J.

    2007-01-01

    The objective of NUDATRA, Domain 5 of the EU Integrated Project EUROTRANS (FI6W-CT-2004- 516520), is to improve and validate the nuclear data and simulation tools required for the development and optimisation of nuclear waste transmutation, ADS dedicated transmutation systems and the associated fuel cycle. Activities are essentially aimed at supplementing the evaluated nuclear data libraries and improving the reaction models for materials in transmutation fuels, coolants, spallation targets, internal structures, and reactor and accelerator shielding, relevant for the design and optimisation of the ETD and XT-ADS. These activities are distributed over four Work Packages: Sensitivity Analysis and Validation of Nuclear Data and Simulation Tools; Low- and Intermediate-energy Nuclear Data Measurements; Nuclear Data Libraries Evaluation and Low-intermediate Energy Models; and High-energy Experiments and Modelling.The main accomplishments expected from NUDATRA are: 1) new measurements and evaluations of Pb-Bi cross-sections, i.e. inelastic, (n,xn) and isomer branching ratios (Po production); 2) new measurements and evaluations for minor actinides particularly the capture in 243 Am and fission on 244 Cm; 3) improvement of TALYS as an evaluation tool and as an a priori model for the estimation of low- and intermediate-energy reaction cross-section; 4) high-energy model improvement based on measurements, particularly for the prediction of the spallation products, and gas (H, He) production cross-sections; 5) sensitivity and uncertainty analysis of ETD fuel cycle and related covariance issues. (authors)

  12. Transmutation of Tc-99 in fission reactors

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  13. Alcohol fuels for developing countries

    International Nuclear Information System (INIS)

    Bhattacharya, Partha

    1993-01-01

    The importance of alcohol as an alternative fuel has been slowly established. In countries such as Brazil, they are already used in transport and other sectors of economy. Other developing countries are also trying out experiments with alcohol fuels. Chances of improving the economy of many developing nations depends to a large extent on the application of this fuel. The potential for alcohol fuels in developing countries should be considered as part of a general biomass-use strategy. The final strategies for the development of alcohol fuel will necessarily reflect the needs, values, and conditions of the individual nations, regions, and societies that develop them. (author). 5 refs

  14. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

  15. Evaluation of alternative partitioning/transmutation scenarios using transmutation in light-water reactors (LWRs)

    International Nuclear Information System (INIS)

    Collins, E.D.; Renier, J.P.; Del Cul, B.; Spencer, B.

    2005-01-01

    Previous Advanced Fuel Cycle Initiative (AFCI) studies were made to assess the effects of the existing accumulation of LWR spent fuel in the United States on the capability to partition/transmute actinides using existing and advanced LWR. The concept of treating the oldest fuel first indicated that significant advantages could be gained in both partitioning, transmutation, and in overall cost reduction. The processing scenarios previously evaluated assumed that (1) 2000 MT/year of spent fuel, irradiated to 45 GWd/MT and decayed for 30 years is processed; (2) recovered plutonium and 90% of the neptunium are transmuted in LWR MOX fuel; and (3) minor actinides, consisting of americium, curium, and 10% of the neptunium are transmuted in burnable poison type targets. Results of the previous study showed that significant benefits could be obtained, including (1) lower costs for partitioning and transmutation and for storage of spent fuel, (2) maintenance of proliferation resistance for the fissile plutonium in spent fuels, and (3) extended lifetime for the repository. The lower costs would be achieved primarily because no capital investment for a special transmuter reactor (fast reactor, accelerator-driven system, etc.) would be required. Instead, only existing and new LWRs would be utilized. Moreover, no new storage capacity could be needed for spent fuels and irradiated targets because the number of spent fuel assemblies would remain the same after the scenario is begun. Even though the total inventory of plutonium would rise during the early cycles, ∼98% of the plutonium would be contained in stored spent fuel and would be protected by high radiation (the Spent Fuel Standard). This is because the spent fuel would be reprocessed and re-irradiated at intervals within which the fission products, 137 Cs and 90 Sr, both with half-lives of ∼ 30 years, exist in significantly high concentrations.The lifetime of the repository would be extended significantly because all of

  16. Genetic effects of decay of fuel fission products in cells of yeast Saccharomyces cerevisial. Role of transmutations

    International Nuclear Information System (INIS)

    Shanshiashvili, T.A.; Gracheva, L.M.

    1981-01-01

    Lethal and mutagenic effects and the nature of mutations induced by 91 Y decay in cells of Saccharomyces cerevisial yeasts were studied. An inactivation probability was (9.6+-4.0)x10 -5 decaysup(-I) and mutation probability in ade 1, ade 2 genes was (0.65+-0.53)x10 -8 decaysup(-I). It is shown that 95% of the effect is caused by local transmutation aftereffects of the decay of the isotope in the cell, and only 5% - by β-radiation. It is also shown that 91 Y is one of the most effective physical matagens. Its mutagenic efficiency (M/1 nS) is the same as of 32 P, and 3.5 times higher than the mutagenic efficiency of γ-rays, and 8 times - than of neutrons with E=1.3 MeV. According to the results of the analysis of the nature of mutations, 91 Y is a nonspecific mutagen which induces mutations of different types. Among 4B analyzed ade 2 mutations there have been detected 146(35%) transitions AT→GC, 139(34%) transitions GC→AT, 64(16%) mutations of reading frame shift, 63 (15%) transversions and 1 deletion. The reasons for such significant genetic effect of 91 Y on the cell are discussed [ru

  17. Development of long-lived radionuclide transmutation technology -Development of long-lived radionuclide handling technology-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Hoh; Jung, Won Myung; Lee, Kyoo Il; Woo, Moon Sik; Cho, Kyung Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The final goals of this research are completion of design for construction of wet hot cell and auxiliary facilities, and development of main equipments and technologies for remote operation and near real time monitoring system of radioactivity of solution. This wet hot cell and technology will be used for active test of the radionuclide partitioning process and for fission Mo separation and purification process. And high level radioactive and toxic materials will be treated as the form of solution in this wet hot cell. In this R. and D., the important objectives are (1)to provide safe operation, and (2)to keep radiation exposure to staff as low as practicable, (3)to protect the environment. 34 figs, 22 tabs, 44 refs. (Author).

  18. System and safety studies of accelerator driven transmutation systems

    International Nuclear Information System (INIS)

    Gudowski, W.; Wallenius, J.; Tucek, K.; Eriksson, Marcus; Carlsson, Johan; Seltborg, P.; Cetnar, J.

    2001-05-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the department has been focused on: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features; b) analysis of ADS-dynamics c) computer code and nuclear data development relevant for simulation and optimization of ADS; d) participation in ADS experiments including 1 MW spallation target manufacturing, subcritical experiments MUSE (CEA-Cadarache). Moreover, during the reporting period the EU-project 'IABAT', co-ordinated by the department has been finished and 4 other projects have been initiated in the frame of the 5th European Framework Programme. Most of the research topics reported in this paper are referred to appendices, which have been published in the open literature. The topics, which are not yet published, are described here in more details

  19. System and safety studies of accelerator driven transmutation systems

    Energy Technology Data Exchange (ETDEWEB)

    Gudowski, W.; Wallenius, J.; Tucek, K.; Eriksson, Marcus; Carlsson, Johan; Seltborg, P.; Cetnar, J. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2001-05-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the department has been focused on: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features; b) analysis of ADS-dynamics c) computer code and nuclear data development relevant for simulation and optimization of ADS; d) participation in ADS experiments including 1 MW spallation target manufacturing, subcritical experiments MUSE (CEA-Cadarache). Moreover, during the reporting period the EU-project 'IABAT', co-ordinated by the department has been finished and 4 other projects have been initiated in the frame of the 5th European Framework Programme. Most of the research topics reported in this paper are referred to appendices, which have been published in the open literature. The topics, which are not yet published, are described here in more details.

  20. Development of CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan

    1991-12-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

  1. Neutron transmutation doped Ge bolometers

    Science.gov (United States)

    Haller, E. E.; Kreysa, E.; Palaio, N. P.; Richards, P. L.; Rodder, M.

    1983-01-01

    Some conclusions reached are as follow. Neutron Transmutation Doping (NTD) of high quality Ge single crystals provides perfect control of doping concentration and uniformity. The resistivity can be tailored to any given bolometer operating temperature down to 0.1 K and probably lower. The excellent uniformity is advantaged for detector array development.

  2. Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican; Gao Chunjing; Xu Dezheng; Li Jingjing; Zhu Xiaoxiang

    2004-01-01

    Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

  3. Transmutation technology development; thermal hydraulic power analysis and structure analysis of the HYPER target beam window

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. H.; Ju, E. S.; Song, M. K.; Jeon, Y. Z. [Gyeongsang National University, Jinju (Korea)

    2002-03-01

    A thermal hydraulic power analysis, a structure analysis and optimization computation for some design factor for the design of spallation target suitable for HYPER with 1000 MW thermal power in this study was performed. Heat generation formula was used which was evaluated recently based on the LAHET code, mainly to find the maximum beam current under given computation conditions. Thermal hydraulic power of HYPER target system was calculated using FLUENT code, structure conducted by inputting the data into ANSYS. On the temp of beam windows and the pressure distribution calculated using FLUENT. Data transformation program was composed apply the data calculated using FLUENT being commercial CFD code and ANSYS being FEM code for CFX structure analysis. A basic study was conducted on various singular target to obtain fundamental data on the shape for optimum target design. A thermal hydraulic power analysis and structure analysis were conducted on the shapes of parabolic, uniform, scanning beams to choose the optimum shape of beam current analysis was done according to some turbulent model to simulate the real flow. To evaluate the reliability of numerical analysis result, benchmarking of FLUENT code reformed at SNU and Korea Advanced Institute of Science and Technology and it was compared to CFX in the possession of Korea Atomic Energy Research Institute and evaluated. Reliable deviation was observed in the results calculated using FLUENT code, but temperature deviation of about 200 .deg. C was observed in the result from CFX analysis at optimum design condition. Several benchmarking were performed on the basis of numerical analysis concerning conventional HYPER. It was possible to allow a beam arrests of 17.3 mA in the case of the {phi} 350 mm parabolic beam suggested to the optimum in nuclear transmutation when stress equivalent to VON-MISES was calculated to be 140 MPa. 29 refs., 109 figs. (Author)

  4. JAERI R & D on accelerator-based transmutation under OMEGA program

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, T.; Nishida, T.; Mizumoto, M. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

    1995-10-01

    The overview of the Japanese long-term research and development program on nuclide partitioning and transmutation, called {open_quotes}OMEGA,{close_quotes} is presented. Under this national program, major R&D activities are being carried out at JAERI, PNC, and CRIEPI. Accelerator-based transmutation study at JAERI is focused on a dedicated transmutor with a subcritical actinide-fueled subcritical core coupled with a spallation target driven by a high intensity proton accelerator. Two types of system concept, solid system and molten-salt system, are discussed. The solid system consists of sodium-cooled tungsten target and metallic actinide fuel. The molten-salt system is fueled with molten actinide chloride that acts also as a target material. The proposed plant transmutes about 250 kg of minor actinide per year, and generates enough electricity to power its own accelerator. JAERI is proposing the development of an intense proton linear accelerator ETA with 1.5 GeV-10 mA beam for engineering tests of accelerator-based transmutation. Recent achievements in the accelerator development are described.

  5. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

  6. Scenarios for minor actinides transmutation in the framework of the French Act on Waste Management

    International Nuclear Information System (INIS)

    Coquelet-Pascal, C.; Meyer, M.; Tiphine, M.; Girieud, R.; Eschbach, R.; Chabert, C.; Garzenne, C.; Barbrault, P.; Van Den Durpel, L.; Caron-Charles, M.; Favet, D.; Arslan, M.; Caron-Charles, M.; Carlier, B.; Lefevre, J.C.

    2013-01-01

    In the framework of the French Act on Waste Management, options of minor actinides (MA) transmutation are studied, based on several scenarios of sodium fast reactor deployment. Basically, one of these scenarios considers the deployment of a 60 GWe SFR fleet in two steps (20 GWe from 2040 to 2050 and 40 GWe, as well as, from 2080 to 2100). For this scenario, the advantages and drawbacks of different transmutation options are evaluated: - transmutation of all minor actinides or only of americium; - transmutation in homogeneous mode (MA bearing fuel in all the core or just in the outer core) or in heterogeneous mode (MA bearing radial blankets). Scenarios have been optimised to limit the impacts of MA transmutation on the cycle: - reduction of the initial MA content in the core in the case of transmutation in homogeneous mode to reduce the impact on reactivity coefficients; - reduction of the number of rows of blankets and fuel decay heat in the case of transmutation in heterogeneous mode. The sensitivity of transmutation options to cycle parameters such as the fuel cooling time before transportation is also assessed. Thus, the transmutation of only americium in one row of radial blankets containing initially 10 pc % Am and irradiated during the same duration as the standard fuel assemblies appears to be a suitable solution to limit the transmutation impacts on fuel cycle and facilities. A comparison of results obtained with MA transmutation in dedicated systems is also presented with a symbiotic scenario considering ADS (accelerator-driven system) deployment to transmute MA together with a SFR fleet to produce energy. The MA inventory within the cycle is higher in the case of transmutation in ADS than in the case of transmutation in SFR. Considering the industrial feasibility of MA transmutation, it appears important to study 'independently' SFR deployment and MA transmutation. Consequently, scenarios of progressive introduction of MA options are assessed

  7. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  8. The EU research activities on partitioning and transmutation. From the 4. to the 5. framework programme

    International Nuclear Information System (INIS)

    Hugon, M.

    1999-01-01

    The European Commission is partly supporting research work on partitioning and transmutation of radioactive waste under the Fourth Framework Programme (1994-1998). This work includes nine research projects. Five strategy studies are evaluating the capabilities of various burners and fuel cycles to limit the production and even destroy the stock of actinides (plutonium and minor actinides). Two experimental projects are aiming at developing techniques for the chemical separation of actinides and two others are dealing with the investigation of transmutation of americium and long-lived fission products. The objectives of these studies are described together with the main results already obtained. The European Union should adopt the 5. Framework Programme (1998-2002) at the end of 1998. The broad lines of the research activities foreseen in partitioning and transmutation and future system under the 5. Framework Programme are briefly presented. (author)

  9. Transmutation doping of silicon solar cells

    Science.gov (United States)

    Wood, R. F.; Westbrook, R. D.; Young, R. T.; Cleland, J. W.

    1977-01-01

    Normal isotopic silicon contains 3.05% of Si-30 which transmutes to P-31 after thermal neutron absorption, with a half-life of 2.6 hours. This reaction is used to introduce extremely uniform concentrations of phosphorus into silicon, thus eliminating the areal and spatial inhomogeneities characteristic of chemical doping. Annealing of the lattice damage in the irradiated silicon does not alter the uniformity of dopant distribution. Transmutation doping also makes it possible to introduce phosphorus into polycrystalline silicon without segregation of the dopant at the grain boundaries. The use of neutron transmutation doped (NTD) silicon in solar cell research and development is discussed.

  10. Transmutation of Americium in Fast Neutron Facilities

    OpenAIRE

    Zhang, Youpeng

    2011-01-01

    In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code. The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on c...

  11. Transmutation: The Roots of the Dream.

    Science.gov (United States)

    Karpenko, Vladimir

    1995-01-01

    Examines the history of alchemical attempts at transmutation and classifies them by differing approaches and techniques. Traces the development of alchemy in Asia, Europe, and the Middle East, and compares alchemy with craftsmanship. (18 references) (DDR)

  12. Transmutation and the Global Nuclear Energy Partnership

    International Nuclear Information System (INIS)

    Bresee, James

    2007-01-01

    In the January 2006 State of the Union address, President Bush announced a new Advanced Energy Initiative, a significant part of which is the Global Nuclear Energy Initiative. Its details were described on February 6, 2006 by the U.S. Secretary of Energy. In summary, it has three parts: (1) a program to expand nuclear energy use domestically and in foreign countries to support economic growth while reducing the release of greenhouse gases such as carbon dioxide. (2) an expansion of the U.S. nuclear infrastructure that will lead to the recycling of spent fuel and a closed fuel cycle and, through transmutation, a reduction in the quantity and radiotoxicity of nuclear waste and its proliferation concerns, and (3) a partnership with other fuel cycle nations to support nuclear power in additional nations by providing small nuclear power plants and leased fuel with the provision that the resulting spent fuel would be returned by the lessee to the lessor. The final part would have the effect of stabilizing the number of fuel cycle countries with attendant non-proliferation value. Details will be given later in the paper. Commercial spent fuel recycling, pioneered in the U.S., has not been carried out since the nineteen seventies following a decision by President Carter to forego fuel reprocessing and to recommend similar practices by other countries. However, many nations have continued spent fuel reprocessing, generally using the U.S.-developed PUREX process. The latest to do so are Japan, which began operations of an 800 metric tons (tonnes) per year PUREX reprocessing plant at Rokkasho-mura in northern Honshu in 2006 and China, which recently began operations of a separations pilot plant, also using PUREX. Countries using the PUREX process, recycle the separated plutonium to light water reactors (LWRs) in a mixed plutonium/uranium oxide fuel called MOX. Plutonium recycling in LWRs, which are used for electricity production in all nuclear power nations, reduces

  13. Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

    Science.gov (United States)

    Marshalkin, V. Ye.; Povyshev, V. M.

    2017-12-01

    It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

  14. Application of variance reduction technique to nuclear transmutation system driven by accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In Japan, it is the basic policy to dispose the high level radioactive waste arising from spent nuclear fuel in stable deep strata after glass solidification. If the useful elements in the waste can be separated and utilized, resources are effectively used, and it can be expected to guarantee high economical efficiency and safety in the disposal in strata. Japan Atomic Energy Research Institute proposed the hybrid type transmutation system, in which high intensity proton accelerator and subcritical fast core are combined, or the nuclear reactor which is optimized for the exclusive use for transmutation. The tungsten target, minor actinide nitride fuel transmutation system and the melted minor actinide chloride salt target fuel transmutation system are outlined. The conceptual figures of both systems are shown. As the method of analysis, Version 2.70 of Lahet Code System which was developed by Los Alamos National Laboratory in USA was adopted. In case of carrying out the analysis of accelerator-driven subcritical core in the energy range below 20 MeV, variance reduction technique must be applied. (K.I.)

  15. Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

    International Nuclear Information System (INIS)

    Silva, Clarysson Alberto da

    2009-01-01

    This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

  16. Advanced Hydrocarbon Fuel Development

    Science.gov (United States)

    Bai, S. Don; Rodgers, Stephen L. (Technical Monitor)

    2000-01-01

    As a part of a high energy density materials (HEDM) development, the hot fire tests for Quadricyclane, 1,7 Octadiyne, AFRL-1, Biclopropylidene, and CINCH (Dimethyl amino ethyl azide) have been conducted at NASA/MSFC. The first 4 materials for this task are provided from Air Force Research Laboratory at Edward Air Force Base and US Army provided CINCH. The performance of these fuels is compared with RP-1. The preliminary results of these tests are presented. The preliminary results of Quadricyclane tests indicate that the specific impulse and c-star efficiency for quadricyclane at the mixture ratio 1.94 are approximately 5 sec and 105 ft/sec better than the RP-1 at mixture ratio 1.9. The 1,7 Octadiyne test indicate that the specific impulse and c-star efficiency at the mixture ratio 2.1 are approximately -1 sec and 89 ft/sec differ than the RP-1 at mixture ratio 2.04. The Quadricyclane soot buildup at the combustor is a little more than RP-1, but detail study of soot formation is not considered at this time. There was no visual soot buildup for the 1,7 Octadiyne and AFRL-1.

  17. System and safety studies of accelerator driven transmutation systems. Annual report 1998

    International Nuclear Information System (INIS)

    Wallenius, J.; Gudowski, W.; Carlsson, Johan; Eriksson, Marcus; Tucek, K.

    1998-12-01

    This annual report describes the accelerator-driven transmutation project conducted at the Department of Nuclear and Reactor Physics at the Royal Institute of Technology. The main results are: development of the simulation tools for accelerator-driven transmutation calculations including an integrated Monte-Carlo burnup module and improvements of neutron energy fission yield simulations, processing of the evacuated nuclear data files including preparation of the temperature dependent neutron cross-sections, development of nuclear data for a medium energy range for some isotopes, development of the models and codes for radiation damage simulations, system studies for the spent fuel transmuter, based on heavy metal coolant and advanced nuclear fuel, contribution to the spallation target design being manufactured in IPPE, Obninsk, and accelerator reliability studies. Moreover a lot of efforts were put to further develop existing international collaboration with the most active research groups in the world together with educational activities in Sweden including a number of meetings and workshops and a graduate course in transmutation. This project has been conducted in close collaboration with the EU-project 'Impact of the accelerator based technologies on nuclear fission safety' - IABAT and in bilateral cooperation with different foreign research groups

  18. System and safety studies of accelerator driven transmutation systems. Annual report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Wallenius, J.; Gudowski, W.; Carlsson, Johan; Eriksson, Marcus; Tucek, K. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    1998-12-01

    This annual report describes the accelerator-driven transmutation project conducted at the Department of Nuclear and Reactor Physics at the Royal Institute of Technology. The main results are: development of the simulation tools for accelerator-driven transmutation calculations including an integrated Monte-Carlo burnup module and improvements of neutron energy fission yield simulations, processing of the evacuated nuclear data files including preparation of the temperature dependent neutron cross-sections, development of nuclear data for a medium energy range for some isotopes, development of the models and codes for radiation damage simulations, system studies for the spent fuel transmuter, based on heavy metal coolant and advanced nuclear fuel, contribution to the spallation target design being manufactured in IPPE, Obninsk, and accelerator reliability studies. Moreover a lot of efforts were put to further develop existing international collaboration with the most active research groups in the world together with educational activities in Sweden including a number of meetings and workshops and a graduate course in transmutation. This project has been conducted in close collaboration with the EU-project `Impact of the accelerator based technologies on nuclear fission safety` - IABAT and in bilateral cooperation with different foreign research groups 31 refs, 23 figs

  19. Keynote address at the international conference on nuclear waste transmutation

    International Nuclear Information System (INIS)

    Steinberg, M.

    1980-07-01

    Most of the US research effort on radioactive waste management is concentrated on its solidification and storage in geologic formations. Transmutation is an alternative. It can be applied to weapons plutonium as well to spent fuel

  20. Study of nuclear energy systems and double strata scenarios for minor actinides transmutation in ADS

    International Nuclear Information System (INIS)

    Clavel, J.B.

    2012-01-01

    The French law of 28 June 2006 regarding advanced nuclear waste management requires a scientific assessment to define future industrial strategies. The present PhD thesis was carried in this framework and concerns specifically the research axis of minor actinides transmutation. A high power Accelerator Driven System (ADS) concept is developed at SUBATECH for this purpose. A 1 GeV proton beam feeds three liquid lead-bismuth spallation targets. The Multiple Spallation Target (MUST) ADS reaches the thermal powers up to 1 GW with a high specific power. A nuclear reactor dimensioning method has been developed and applied to different double strata scenarios. In these scenarios, SFR (Sodium Fast Reactors) or PWR (Pressurized Water Reactors) power reactors produce minor actinides that will be transmuted into ADS. In each core (SFR and ADS), the plutonium multi-reprocessing strategy is performed while ADS subcritical core also multi-reprocesses minor actinides. To limit the core reactivity and improve the fuel thermal conductivity, the minor actinides fuel is mixed with MgO inert matrix. Nuclear branches with lead and sodium coolants for the ADS, have been studied for different irradiation times and two transmutation strategies have been assessed: whether whole minor actinides, whether americium only is transmuted. The thesis presents precisely the MUST ADS design methodology and the calculations to get a fuel composition at equilibrium. Then a one cycle evolution is performed and analysed for the fuel and the multiplication factor. Radiotoxicity and thermal power of the waste produced are then compared. Finally, the study of double strata scenarios is performed to analyse the plutonium and minor actinides inventories in cycle and also the waste produced according to the transmutation strategies applied and the first stratum evolution. (author)

  1. Actinide and fission product partitioning and transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The third international information exchange meeting on actinide and fission product partitioning and transmutation, took place in Cadarache France, on 12-14 December 1994. The proceedings are presented in six sessions : an introduction session, the major programmes and international cooperation, the systems studies, the reactors fuels and targets, the chemistry and a last discussions session. (A.L.B.)

  2. The status of nuclear data for transmutation calculations

    International Nuclear Information System (INIS)

    Wilson, W.B.; England, T.R.; MacFarlane, R.E.; Muir, D.W.; Young, P.G.

    1995-01-01

    At this point, the accurate description of transmutation products in a radiation environment is more a nuclear data problem than a code development effort. We have used versions of the CINDER code for over three decades to describe the transmutation of nuclear reactor fuels in radiation environments. The need for the accurate description of reactor neutron-absorption, decay-power, and decay-spectra properties have driven many AEC, ERDA, and DOE supported nuclear data development efforts in this period. The level of cross-section, decay, and fission-yield data has evolved from rudimentary to a comprehensive ENDF/B-VI library permitting great precision in reactor calculations. The precision of the data supporting reactor simulations provides a sturdy foundation for the data base required for the wide range of transmutation problems currently studied. However, such reactor problems are typically limited to neutron energies below 10 MeV or so; reaction and decay data are required for actinides of, say, 90 ≤ Z ≤ 96 neutron-rich fission products of 22 ≤ Z ≤ 72. The expansion into reactor structural materials and fusion systems extends these ranges in energy and Z somewhat. The library of nuclear data, constantly growing in breadth and quality with international cooperation, is now described in the following table

  3. Partitioning and transmutation. Annual report 2007

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, Emma; Ekberg, Christian; Englund, Sofie; Fermvik, Anna; Foreman, Mark St. J.; Liljenzin, Jan-Olov; Retegan, Teodora; Skarnemark, Gunnar; Wald, Karin (Nuclear Chemistry, Dept. of Chemical and Biological Engineering, Chalmers Univ. of Technology, Goeteborg (SE))

    2007-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products (79Se, 87Rb, 99Tc, 107Pd, 126Sn, 129I, 135Cs) and activation products (14C, 36Cl, 59Ni, 93Zr, 94Nb). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in several European frame work programmes from NEWPART in the 4th framework via PARTNEW and EUROPART to ACSEPT now in the 7th programme. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since a further investigation on basic understanding of the chemical behaviour is required, we have our main focus on the chemical processes and understanding of how they work. Due to new recruitments we will now also work on ligand design and development. This will decrease the response time between new ligands and their evaluation.

  4. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    Energy Technology Data Exchange (ETDEWEB)

    Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  5. Status of the French research programme for actinides and fission products partitioning and transmutation

    International Nuclear Information System (INIS)

    Warin, D.

    2003-01-01

    The paper focus on separation and transmutation research and development programme and main results over these ten last years. The massive research programme on enhanced separation, conducted by CEA and supported by broad international cooperation, has recently achieved some vital progress. Based on real solutions derived from the La Hague process, the CEA demonstrated the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process that can be extrapolated on the industrial scale. The CEA also conducted programmes proving the technical feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for ADS developments in order to support the MEGAPIE, TRADE and MYRRHA experiments and the future 100 MW international ADS demonstrator. Scenarios studies aimed at stabilizing the inventory with long-lived radionuclides, plutonium, minor actinides and certain long-lived fission products in different nuclear power plant parks and to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. Three French Research Groups CEA-CNRS carry out partitioning (PRACTIS) and transmutation (NOMADE and GEDEON) more basic studies. (author)

  6. Fuel cycle developments

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This article is a review of the end-of-1994 status of world uranium production and fuels processing. The major producing areas/countries of the world are discussed and the production figures for each area/country are provided. The conversion services market is also discussed, as is the enrichment services market. Each of the major enrichment services provider organizations is noted

  7. Fusion-Fission Transmutation Scheme-Efficient destruction of nuclear waste

    International Nuclear Information System (INIS)

    Kotschenreuther, M.; Valanju, P.M.; Mahajan, S.M.; Schneider, E.A.

    2009-01-01

    A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades

  8. A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

    International Nuclear Information System (INIS)

    Csom, Gyula; Feher, Sandor; Szieberthj, Mate

    2002-01-01

    Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutation capabilities are studied. The goal is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named 'multi-region' MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a 'conventional' MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on. (authors)

  9. Partitioning and Transmutation. Annual Report 2005

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Sofie; Ekberg, Christian; Fermvik, Anna; Hervieux, Nadege; Liljenzin, Jan-Olov; Magnusson, Daniel; Nilsson, Mikael; Retegan, Teodora; Skarnemark, Gunnar [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Chemical and Biological Engineering

    2006-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ({sup 79}Se, {sup 87}Rb, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, {sup 135}Cs) and activation products ({sup 14}C, {sup 36}Cl, {sup 59}Ni, {sup 93}Zr, {sup 94}N To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. These separations are necessary to obtain the desired efficiency of the transmutation process and in order not to create any unnecessary waste thus rendering the process useless. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. Chalmers Univ. of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in the EUROPART project within the European Union sixth framework program. This is a continuation of the projects we participated in within the fourth and fifth framework programmes, NEWPART and PARTNEW respectively. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. However, since the basic understanding is still needed we have our main focus on the chemical processes and understanding of how they work.

  10. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

    1991-07-01

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  11. J-PARC Transmutation Experimental Facility Programme

    International Nuclear Information System (INIS)

    Sasa, T.; Takei, H.; Saito, S.; Obayashi, H.; Nishihara, K.; Sugawara, T.; Iwamoto, H.; Yamaguchi, K.; Tsujimoto, K.; Oigawa, H.

    2015-01-01

    Since the Fukushima accident, nuclear transmutation is considered as an option for waste management. Japan Atomic Energy Agency proposes the transmutation of minor actinides (MA) in accelerator-driven system (ADS) using lead-bismuth eutectic alloy (LBE) as a spallation target and a coolant of subcritical core. To obtain the data required for ADS design, we plan the building of a transmutation experimental facility (TEF) is planned within the J-PARC project. TEF consists of an ADS target test facility (TEF-T), which will be installed 400 MeV-250 kW LBE spallation target for material irradiations, and a transmutation physics experimental facility (TEF-P), which set up a fast critical/subcritical assembly driven by low power proton beam with MA fuel to study ADS neutronics. At TEF-T, various research plans to use emitted neutrons from LBE target are discussed. The paper summarises a road-map to establish the ADS transmuter and latest design activities for TEF construction. (authors)

  12. Development of MOX fuel database

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  13. Waste management in future. Partitioning and transmutation (P and T)

    International Nuclear Information System (INIS)

    Calic, D.

    2005-01-01

    Current research and development (R and D) in radioactive waste management is mainly associated with the quantities and toxicity of high level waste and spent fuel. One of the solutions that already exists, but has not yet reached scientific and technological maturity, is the process of partitioning and transmutation (P and T). Partitioning is the selective separation of radiotoxic isotopes from reprocessing streams. After the successive partitioning has been done, the long-lived radionuclides are converted into shorter-lived or stable nuclides by process called transmutation. P and T can reduce the radiotoxic inventory of spent fuel by a factor of 100 to 1000 and can achieve the reduction of time needed to reach the radioactivity level of the uranium ore from 100,000 to 5000 years. To achieve this, the separation of plutonium, minor actinides and long-lived fission products has to be implemented as early as possible in the fuel cycle strategy. Currently, P and T is still at the research and development stage and it needs to be scaled up, before it can be introduced on an industrial scale, therefore the paper will present the current status of the development of P and T and plans for the future. (author)

  14. Fuel Cell Manufacturing Research and Development | Hydrogen and Fuel Cells

    Science.gov (United States)

    | NREL Fuel Cell Manufacturing Research and Development Fuel Cell Manufacturing Research and Development NREL's fuel cell manufacturing R&D focuses on improving quality-inspection practices for high costs. A researcher monitoring web-line equipment in the Manufacturing Laboratory Many fuel cell

  15. Fluoride partitioning R and D programme for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    Uhlir, J.; Priman, V.; Vanicek, J.

    2001-01-01

    The transmutation of spent nuclear fuel is considered a prospective alternative conception to the current conception based on the non-reprocessed spent fuel disposal into underground repository. The Czech research and development programme in the field of partitioning and transmutation is founded on the Molten Salt Transmutation Reactor system concept with fluoride salts based liquid fuel, the fuel cycle of which is grounded on pyrochemical / pyrometallurgical fluoride partitioning of spent fuel. The main research activities in the field of fluoride partitioning are oriented mainly towards technological research of Fluoride Volatility Method and laboratory research on electro-separation methods from fluoride melts media. The Czech national conception in the area of P and T research issues from the national power industry programme and from the Czech Power Company intentions of the extensive utilization of nuclear power in our country. The experimental R and D work is concentrated mainly in the Nuclear Research Institute Rez plc that plays a role of main nuclear research workplace for the Czech Power Company. (author)

  16. Waste transmutation: perspectives

    International Nuclear Information System (INIS)

    Leray, S.

    1997-01-01

    After the introduction on the source and nature of nuclear waste, this lecture analyzes the different methods proposed to transmute long-lived isotopes into stable or short-lived isotopes. It is shown that direct methods (photonuclear reactions, spallation, muon catalyzed fusion) do not lead to a sufficient transmutation rate within a reasonable cost. Only the use of hybrid systems, fusion-fission or spallation-fission, can be foreseen. (author)

  17. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

    2001-01-01

    Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

  18. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  19. Romanian nuclear fuel cycle development

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Comsa, Olivia

    1998-01-01

    Romanian decision to introduce nuclear power was based on the evaluation of electricity demand and supply as well as a domestic resources assessment. The option was the introduction of CANDU-PHWR through a license agreement with AECL Canada. The major factors in this choice have been the need of diversifying the energy resources, the improvement the national industry and the independence of foreign suppliers. Romanian Nuclear Power Program envisaged a large national participation in Cernavoda NPP completion, in the development of nuclear fuel cycle facilities and horizontal industry, in R and D and human resources. As consequence, important support was being given to development of industries involved in Nuclear Fuel Cycle and manufacturing of equipment and nuclear materials based on technology transfer, implementation of advanced design execution standards, QA procedures and current nuclear safety requirements at international level. Unit 1 of the first Romanian nuclear power plant, Cernavoda NPP with a final profile 5x700 Mw e, is now in operation and its production represents 10% of all national electricity production. There were also developed all stages of FRONT END of Nuclear Fuel Cycle as well as programs for spent fuel and waste management. Industrial facilities for uranian production, U 3 O 8 concentrate, UO 2 powder and CANDU fuel bundles, as well as heavy water plant, supply the required fuel and heavy water for Cernavoda NPP. The paper presents the Romanian activities in Nuclear Fuel Cycle and waste management fields. (authors)

  20. Transmutation of Minor Actinide in well thermalized neutron field and application of advanced neutron source (ANS)

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Hirakawa, Naohiro

    1995-01-01

    Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246 Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246 Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100 eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10 -5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation. (author)

  1. Gamma ray beam transmutation

    International Nuclear Information System (INIS)

    Imasaki, K.; Li, D.; Miyamoto, S.; Amano, S.; Motizuki, T.

    2007-01-01

    We have proposed a new approach to nuclear transmutation by a gamma ray beam of Compton scattered laser photon. We obtained 20 MeV gamma ray in this way to obtain transmutation rates with the giant resonance of 1 97Au and 1 29Iodine. The rate of the transmutation agreed with the theoretical calculation. Experiments on energy spectrum of positron, electron and neutron from targets were performed for the energy balance and design of the system scheme. The reaction rate was about 1.5∼4% for appropriate photon energies and neutron production rate was up to 4% in the measurements. We had stored laser photon more than 5000 times in a small cavity which implied for a significant improvement of system efficiency. Using these technologies, we have designed an actual transmutation system for 1 29Iodine which has a 16 million year's activity. In my presentation, I will address the properties of this scheme, experiments results and transmutation system for iodine transmutation

  2. Transmutation potential of reactor WWER-440

    International Nuclear Information System (INIS)

    Darilek, P.; Sebian, V.; Necas, V.

    2001-01-01

    Theoretical evaluation of WWER-440 transmutation potential by HELIOS - code is presented. Transmutation method proposal comprising special transmutation pins, combined FA and simple reprocessing is described. Transmutation efficiency of the method is characterized (Authors)

  3. 2nd RCM of the CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) and Technical Meeting on Low Enriched Uranium (LEU) Fuel Utilization in Accelerator Driven Sub-critical Systems. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is contributing to the generic R&D efforts in various fields common to innovative fast neutron system development, i.e., heavy liquid metal thermal hydraulics, dedicated transmutation fuels and associated core designs, theoretical nuclear reaction models, measurement and evaluation of nuclear data for transmutation, and development and validation of calculational methods and codes. Ultimately, the CRP’s overall objective is to make contributions towards the realization of a transmutation demonstration facility

  4. Transmutation of Americium in Fast Neutron Facilities

    International Nuclear Information System (INIS)

    Zhang, Youpeng

    2011-01-01

    In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code. The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on core's safety parameters. Applying the SAS4A/SASSYS transient analysis code, it is demonstrated that the power rating needs to be reduced by 6% for each percent additional americium introduction into the reference MOX fuel, maintaining 100 K margin to fuel melting, which is the most limiting failure mechanism. Safety analysis of a new Accelerator Driven System design with a smaller pin pitch-to-diameter ratio comparing to the reference EFIT-400 design, aiming at improving neutron source efficiency, was also performed by simulating performance for unprotected loss of flow, unprotected transient overpower, and protected loss-of-heat-sink transients, using neutronic parameters obtained from MCNP calculations. Thanks to the introduction of the austenitic 15/15Ti stainless steel with enhanced creep rupture resistance and acceptable irradiation swelling rate, the suggested ADS design loaded with nitride fuel and cooled by lead-bismuth eutectic could survive the full set of transients, preserving a margin of 130 K to cladding rupture during the most limiting transient. The thesis concludes that efficient transmutation of americium in a medium sized sodium cooled fast reactor loaded with MOX fuel is possible but leads to a severe power penalty. Instead, preserving transmutation rates of minor actinides up to 42 kg/TWh th , the suggested ADS design with enhanced proton source efficiency appears like a better option for americium transmutation

  5. Researches on the management of high activity and long-lived radioactive wastes. Axis 1 - separation-transmutation; Recherches sur la gestion des dechets radioactifs a haute activite et a vie longue. Axe 1 - separation-transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-11-15

    This document gathers the transparencies of seven presentations given at a technical workshop of the French nuclear energy society (SFEN) about the researches on separation-transmutation of high activity and long-lived radioactive wastes. The presentations deal with: inventory and radiotoxicity of the rad-wastes in concern; industrial experience; experience on chemical separation: molecules and processes; reactors physics and transmutation - reactors for transmutation; fuels and targets; scenarios that include transmutation; environmental impacts of these different scenarios. (J.S.)

  6. International Nuclear Energy Research Initiative Development of Computational Models for Pyrochemical Electrorefiners of Nuclear Waste Transmutation Systems

    International Nuclear Information System (INIS)

    Simpson, M.F.; Kim, K.-R.

    2010-01-01

    In support of closing the nuclear fuel cycle using non-aqueous separations technology, this project aims to develop computational models of electrorefiners based on fundamental chemical and physical processes. Spent driver fuel from Experimental Breeder Reactor-II (EBR-II) is currently being electrorefined in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory (INL). And Korea Atomic Energy Research Institute (KAERI) is developing electrorefining technology for future application to spent fuel treatment and management in the Republic of Korea (ROK). Electrorefining is a critical component of pyroprocessing, a non-aqueous chemical process which separates spent fuel into four streams: (1) uranium metal, (2) U/TRU metal, (3) metallic high-level waste containing cladding hulls and noble metal fission products, and (4) ceramic high-level waste containing sodium and active metal fission products. Having rigorous yet flexible electrorefiner models will facilitate process optimization and assist in trouble-shooting as necessary. To attain such models, INL/UI has focused on approaches to develop a computationally-light and portable two-dimensional (2D) model, while KAERI/SNU has investigated approaches to develop a computationally intensive three-dimensional (3D) model for detailed and fine-tuned simulation.

  7. Researches on the management of high activity and long-lived radioactive wastes. Axis 1 - separation-transmutation

    International Nuclear Information System (INIS)

    2005-11-01

    This document gathers the transparencies of seven presentations given at a technical workshop of the French nuclear energy society (SFEN) about the researches on separation-transmutation of high activity and long-lived radioactive wastes. The presentations deal with: inventory and radiotoxicity of the rad-wastes in concern; industrial experience; experience on chemical separation: molecules and processes; reactors physics and transmutation - reactors for transmutation; fuels and targets; scenarios that include transmutation; environmental impacts of these different scenarios. (J.S.)

  8. Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

    International Nuclear Information System (INIS)

    Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

    1990-01-01

    The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

  9. Proceedings of the Eleventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation

    International Nuclear Information System (INIS)

    2012-01-01

    Partitioning and transmutation (P and T) is one of the key technologies for reducing the radiotoxicity and volume of radioactive waste arisings. Recent developments indicate the need for embedding P and T strategies in advanced fuel cycles considering both waste management and economic issues. In order to provide experts a forum to present and discuss state-of-the-art developments in the P and T field, the OECD/NEA has been organising biennial information exchange meetings on actinide and fission product partitioning and transmutation since 1990. The previous meetings were held in Mito (Japan) in 1990, at Argonne (United States) in 1992, in Cadarache (France) in 1994, in Mito (Japan) in 1996, in Mol (Belgium) in 1998, in Madrid (Spain) in 2000, in Jeju (Korea) in 2002, in Las Vegas (United States) in 2004, in Nimes (France) in 2006 and in Mito (Japan) in 2008. They have often been co-sponsored by the European Commission (EC) and the International Atomic Energy Agency (IAEA). The 11. Information Exchange Meeting was held in San Francisco, California, United States on 1-4 November 2010, comprising a plenary session on national P and T programmes and six technical sessions covering various fields of P and T. The meeting was hosted by the Idaho National Laboratory (INL), United States. The information exchange meetings on P and T form an integral part of NEA activities on advanced nuclear fuel cycles. The meeting covered scientific as well as strategic/policy developments in the field of P and T, such as: fuel cycle strategies and transition scenarios; radioactive waste forms; the impact of P and T on geological disposal; radioactive waste management strategies (including secondary wastes); transmutation fuels and targets; pyro and aqueous separation processes; materials, spallation targets and coolants; transmutation physics, experiments and nuclear data; transmutation systems (design, performance and safety); handling and transportation of transmutation fuels; and

  10. Development of Computational Models for Pyrochemical Electrorefiners of Nuclear Waste Transmutation Systems

    International Nuclear Information System (INIS)

    Kim, K. R.; Lee, H. S.; Hwang, I. S.

    2010-12-01

    The objective of this project is to develop multi-dimensional computational models in order to improve the operation of uranium electrorefiners currently used in pyroprocessing technology. These 2-D (US) and 3-D (ROK) mathematical models are based on the fundamental physical and chemical properties of the electrorefiner processes. The validated models by compiled and evaluated experimental data could provide better information for developing advanced electrorefiners for uranium recovery. The research results in this period are as follows: - Successfully assessed a common computational platform for the modeling work and identify spatial characterization requirements. - Successfully developed a 3-D electro-fluid dynamic electrorefiner model. - Successfully validated and benchmarked the two multi-dimensional models with compiled experimental data sets

  11. Safety characteristics of potential waste transmutation systems

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1993-01-01

    For nuclear waste transmutation to alter significantly the need for geologic disposal of spent fuel from US Light-water reactors (LWRs), about 1.4% of the spent fuel (by mass) must be separated and transmuted. This includes the plutonium, the minor actinides, and four fission products: iodine. technetium, cesium and strontium. Regarding the actinides, fissioning of the plutonium, neptunium, americium, and curium generates a great deal of heat, so much so that most of the plutonium should be used to produce power. However, these actinides have some undesirable neutronic characteristics, and their utilization in reactors or subcritical (proton-accelerator) targets requires either a fast neutronic spectrum or a very high thermal-neutron flux. Transmutation of the fission products is generally by neutron capture, although this is difficult in the case of cesium and strontium. In this paper, various proposed means of transmuting the actinides and fission products are discussed, with the main focus being on the safety characteristics of each approach

  12. Proposed partitioning and transmutation of long-lived nuclear wastes

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Rawlins, J.A.

    1991-01-01

    A means of transmuting key long-lived nuclear wastes, primarily the minor actinides (Np, Am, Cm) and iodine, using a hybrid proton accelerator and sub-critical lattice, is proposed. By partitioning light water reactor (LWR) spent fuel and by transmuting key elements, such as the plutonium, the minor actinides, and a few of the long-lived fission products, some of the most significant challenges in building a waste repository can be substantially reduced. The proposed machine would transmute the minor actinides and the iodine produced by 75 LWRs, and would generate usable electricity (beyond that required to run the large accelerator) of 850 MW e . 14 refs., 10 figs

  13. Present status and history of nuclear data development for transmutation technology

    CERN Document Server

    Hasegawa, A

    2002-01-01

    A history of development of nuclear data from JENDL (Japanese Evaluated Nuclear Data Library)-1 to JENDL-3.3 and JENDL-HF (High energy File) and JENDL- Actinide File are stated. 5 nuclear data such as JENDL-1, JENDL-2, JENDL-3.1, JENDL-3.2 and JENDL-3.3 have been developed by JAERI. JENDL-1 for fast reactor (1977) has 66+6 nuclide and 15 MeV the largest energy. JENDL-2 for fast and light water reactor (1982) has 173+8 nuclide and 20 MeV. All-purpose nuclear data: JENDL-3.1 (1990) with 305+19 nuclide and 59 2nd gamma-ray data, JENDL-3.2 (1994) with 318+22 nuclide and 66 2nd gamma-ray data and JENDL-3.3 (2002) with 335+2 nuclide, 114 2nd gamma-ray data, 60 angular dependence neutron data and 20 MeV have been developed. JENDL-3.3 was opened at JAERI home page in May 2002. JENDL High Energy library consists of JENDL-HF, JENDL-Photonuclear Data File and JENDL-PKA/KERMA File. JENDL-HF includes nuclear reaction data of neutron and proton incidence, for example, total cross section, elastic scattering cross section a...

  14. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M S; Park, H S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  15. Developments of linacs for accelerator-driven transmutation technology in the USA. Revision

    International Nuclear Information System (INIS)

    Schriber, S.O.

    1997-03-01

    Interesting developments in linear accelerators have been attained over the past 45 years. The status of linear accelerators and future possibilities are described in context of demanding applications and technology maturity. Features of industrial or factory-type applications are high availability, economic operations, low investment cost and ease of running a facility. All features have been demonstrated in one manner or another at large operating facilities for the research community; within a different context that has been argued in the past to be not as demanding as for a factory installation. In addition, comments are made relative to intense beam power levels and choices that can be made for power levels below 10 MW, on the assumption that a cw beam is required

  16. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-22

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  17. FY2016 Ceramic Fuels Development Annual Highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-24

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  18. Development of a computational algorithm for the linearization of decay and transmutation chains; Desarrollo de un algoritmo computacional para la linealizacion de cadenas de decaimiento y transmutacion

    Energy Technology Data Exchange (ETDEWEB)

    Cruz L, C. A.; Francois L, J. L., E-mail: cacl@ier.unam.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico)

    2017-09-15

    One of the most used methodologies to solve Bate man equations, in the problem of burning, is the Tta (Transmutation Trajectory Analysis) method. In this method, a network of decays is broken down into linear elements known as trajectories, through a process known as linearization. In this work an alternative algorithm is shown to find and construct these trajectories, which considers three aspects of linearization: the information -a priori- about the elements that make up decay and transmutation network, the use of a new notation, and in the functions for the treatment of text strings (which are common in most programming languages). One of the main advantages of the algorithm is that can condense the information of a decay and transmutation network into only two vectors. From these is possible to determine how many linear chains can be extracted from the network and even their length (in the case they are not cyclical). Unlike the Deep First Search method, which is widely used for the linearization process, the method proposed in the present work does not have a backward routine and instead occupies a process of compilation, since completes fragments chain instead of going back to the beginning of the trajectories. The developed algorithm can be applied in a general way to the information search and to the linearization of the computational data structures known as trees. It can also be applied to engineering problems where one seeks to calculate the concentration of some substance as a function of time, starting from linear differential equations of balance. (Author)

  19. APEX accelerator cycle for transmutation of long-lived fission wastes

    International Nuclear Information System (INIS)

    Powell, J.; Steinberg, M.; Takahashi, H.; Grand, P.; Botts, T.; Kouts, H.J.C.

    1980-01-01

    Based on preliminary studies, some conclusions can be drawn concerning the Accelerator Fuel Enricher and Fission Product Exterminator (APEX). APEX-1 and APEX-2 systems can destroy TU's, 137 Cs, and 90 Sr at acceptable cost and efficiency. The principal difference between APEX-1 and APEX-2 is the in-reactor and in-circuit inventory of 137 Cs and 90 Sr. Stable and low hazard wastes can be disposed of by burial. Accelerator breeders can effectively sustain a fission reactor economy indefinitely. Military waste can be blended into commercial fuel cycle for transmutation. Accelerator and target technologies appear practical and could be developed in a few years. More detailed studies are needed to better define the technical and economic features of the LAFER and APEX cycles, so that comparative assessments can be made between these cycles, as well as with other transmutation and waste disposal concepts

  20. Waste partitioning and transmutation as a means towards long-term risk reduction

    International Nuclear Information System (INIS)

    Merz, E.R.

    1993-09-01

    It has been an idea for some time to reduce the long-term potential hazard of the waste by chemical removal of the actinides as well as some long-lived fission products and their subsequent transmutation in an intense neutron flux. Transmutation would thus shorten the required containment period of radioactive material in a repository. It is estimated, that development of such technology would take at least 40 years because facilities would be required to perform a clean actinide and fission product isolation and to fabricate the fuel elements that contained the separated nuclides. This latter requirements would involve a major expansion of new chemical process steps which are not available as yet. Development of new equipment to maintain occupational exposures as low as reasonably achievable and to minimize releases of radioactivity to the environment would also be necessary. Partitioning and transmutation should be introduced, if at all, as a long-term decision about new nuclear power technology as a future energy source. With regard to this, R and D work dealing with basic questions seems to be worthwhile, However, the introduction of partitioning and transmutation will not eliminate the need for radioactive waste disposal. (orig./HP) [de

  1. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  2. Transmutation of high level nuclear waste in an accelerator driven system: towards a demonstration device of industrial interest (EUROTRANS)

    International Nuclear Information System (INIS)

    Knebel, Joachim U.; Ait Abderrahim, Hamid; Caron-Carles, Marylise

    2010-01-01

    The Integrated Project EUROTRANS (EURopean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System) within the ongoing EURATOM 6th Framework Programme (FP6) is devoted to the study of transmutation of high-level waste from nuclear power plants. The work is focused on transmutation in an Accelerator Driven System (ADS). The objective of EUROTRANS is the assessment of the design and the feasibility of an industrial ADS prototype dedicated to transmutation. The necessary R and D results in the areas of accelerator components, fuel development, structural materials, thermal-hydraulics, heavy liquid metal technology and nuclear data will be made available, together with the experimental demonstration of the ADS component coupling. The outcome of this work will allow to provide a reasonably reliable assessment of technological feasibility and a cost estimate for ADS based transmutation, and to possibly decide on the detailed design of an experimental ADS and its construction in the future. EUROTRANS is integrating activities of 51 participants from 16 countries, within the industry (10 participants), the national research centres (20) and 17 universities. 16 universities are collectively represented by ENEN (European Nuclear Education Network). EUROTRANS is the continuation of the three FP5 Clusters FUETRA, BASTRA and TESTRA together with the PDS-XADS Project. It is a five-year project which started in April 2005

  3. Neutron transmutation doping

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Byung Jin

    2001-09-01

    HE OVERALL STATE OF THE ART RELATED WITH NEUTRON TRANSMUTATION DOPING(NCT) IS SURVEYED. ITEMS RELATED FOR THE REALIZATION OF NTD IN HANARO IS FOCUSED. IN ADDITION TO THE UNIFORM IRRADIATION AND ACHIEVING THE TARGET RESISTIVITY WHICH ARE THE MOST IMPORTANT TECHNICAL REQUIREMENTS FOR THE NTD, OTHER ITEMS SUCH AS THE FUNCTION AND ROLE OF NTD, MARKET TREND, QUALITY CONTROL ARE INCLUDED. MEANWHILE THE ONLY ADVANTAGE OF NTD IS ACHIEVING VERY HIGH UNIFORMITY OF DOPING, IT HAS SEVERAL DISADVANTAGES DUE TO THE USE OF NUCLEAR REACTOR. THEREFORE THE SEMICONDUCTOR INDUSTRY HAS CONTINUED DEVELOPMENT OF TECHNOLOGY TO REPLACE NTD, AND THE DEMAND OF NTD HAD BEEN DECREASED A LOT DURING 1990S. AS THE DEMAND FOR LARGE CRYSTAL INCREASES, HOWEVER, THE NTD DEMAND BEGAN TO INCREASE AGAIN FROM 2000. SINCE THE DEMAND FOR THE LARGER CRYSTAL WILL BE CONTINUED IN THE FUTURE, THE ROLE OF NTD WOULD BE NEEDED FOR THE LONGER TIME. IN ORDER TO MITIGATE THIS TREND OF DEMAND, THE REACTOR SHOULD BE CAPABLE OF ACCEPTING LARGE CRYSTAL AND THE EFFORT TO IMPROVE DOPING UNIFORMITY AND TO REDUCE THE COST SHOULD CONTINUED.

  4. Development of PEM fuel cell technology at international fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.J.

    1996-04-01

    The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

  5. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  6. Transmutation Scenarios Impacts on Advanced Nuclear Cycles. Fabrication, Reprocessing and Transportation

    International Nuclear Information System (INIS)

    Saturnin, A.; Sarrat, P.; Hancok, H.; Milot, J.-F.; Duret, B.; Jasserand, F.; Fillastre, E.; Giffard, F.-X.; Chabert, C.; Van Den Durpel, L.; Caron-Charles, M.; Lefevre, J.C.; Carlier, B.; Arslan, M.; Favet, D.; Garzenne, C.; Barbrault, P.

    2013-01-01

    Conclusions: First detailed assessment of plants and transportation in various transmutation scenarios. In case of curium transmutation: large difficulties and uncertainties requiring whole new technology development (more pronounced for ADS option). For Am transmutation: more feasible, still to be demonstrated on specific points for industrial extrapolation

  7. Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nuvera Fuel Cells

    2005-04-15

    The potential for fuel cell systems to improve energy efficiency and reduce emissions over conventional power systems has generated significant interest in fuel cell technologies. While fuel cells are being investigated for use in many applications such as stationary power generation and small portable devices, transportation applications present some unique challenges for fuel cell technology. Due to their lower operating temperature and non-brittle materials, most transportation work is focusing on fuel cells using proton exchange membrane (PEM) technology. Since PEM fuel cells are fueled by hydrogen, major obstacles to their widespread use are the lack of an available hydrogen fueling infrastructure and hydrogen's relatively low energy storage density, which leads to a much lower driving range than conventional vehicles. One potential solution to the hydrogen infrastructure and storage density issues is to convert a conventional fuel such as gasoline into hydrogen onboard the vehicle using a fuel processor. Figure 2 shows that gasoline stores roughly 7 times more energy per volume than pressurized hydrogen gas at 700 bar and 4 times more than liquid hydrogen. If integrated properly, the fuel processor/fuel cell system would also be more efficient than traditional engines and would give a fuel economy benefit while hydrogen storage and distribution issues are being investigated. Widespread implementation of fuel processor/fuel cell systems requires improvements in several aspects of the technology, including size, startup time, transient response time, and cost. In addition, the ability to operate on a number of hydrocarbon fuels that are available through the existing infrastructure is a key enabler for commercializing these systems. In this program, Nuvera Fuel Cells collaborated with the Department of Energy (DOE) to develop efficient, low-emission, multi-fuel processors for transportation applications. Nuvera's focus was on (1) developing fuel

  8. Transmutations for Strings

    Directory of Open Access Journals (Sweden)

    Amin Boumenir

    2008-07-01

    Full Text Available We investigate the existence and representation of transmutations, also known as transformation operators, for strings. Using measure theory and functional analytic methods we prove their existence and study their representation. We show that in general they are not close to unity since their representation does not involve a Volterra operator but rather the eigenvalue parameter. We also obtain conditions under which the transmutation is either a bounded or a compact operator. Explicit examples show that they cannot be reduced to Volterra type operators.  

  9. Accelerator-driven transmutation projects in Sweden in a European perspective

    International Nuclear Information System (INIS)

    Gudowski, Waclaw; Wallenius, J.; Tucek, K.; Ericsson, M.; Carlsson, J.; Cetnar, J.

    2001-01-01

    Accelerator driven transmutation projects in Sweden are dealt with within the country's energy policy and energy plans as well as in relation to European perspectives. ADS activities at Royal Institute of Technology (RIT) and Sweden are concerned with: Conceptual design nitride fueled ADS) of Sing-Sing core (heavily 'poisoned'; Development of nitride fuel (in a EU-frame); transients of ADS; Managing nuclear data and stimulating development dedicated data libraries; Development of Monte-Carlo burnup; Design of RVACS for ADS. Besides the mentioned, this paper describes projects in fourth Framework Programme of European Union, fifth Framework Programme of EU and Experiments related to ADS

  10. Design Concept of the Thermal Flux Island in MYRRHA for LLFP Transmutation. Present Status

    International Nuclear Information System (INIS)

    Aoust, Th.; De Raedt, Ch.; Malambu, E.; Ait Abderrahim, H.

    2002-01-01

    MYRRHA is an ADS aimed at providing protons and neutrons for various R and D applications. It consists of a proton accelerator coupled to a sub-critical fast core. While the fast neutron spectrum, obtaining in the MYRRHA core, allows the transmutation of minor actinides (MAs), the transmutation of long lived fission products (LLFPs) in MYRRHA requires the utilisation of a thermal spectrum in order to take profit of the fact that the capture cross-sections of the LLFPs are much larger in the thermal-energy domain than in the high-energy domain. A high thermal flux island inside MYRRHA is therefore being developed, consisting of an irradiation device for LLFPs (or other irradiation target material needing a thermal neutron spectrum), surrounded by assemblies filled with a lattice of ZrH 2 moderator pins, cooled by the same coolant (Pb-Bi) as the MYRRHA fuel assemblies. Thermal neutron absorbers at the periphery of the neutron island are used to minimise the influence of the thermal flux on the fast core. The study of the performances of MYRRHA, including the transmutation of the LLFPs in such a thermal island, is carried out with the aid of the code MCNPX. The transmutation performances of LLFPs are assessed by coupling the code MCNPX to an evolution calculation. The LLFPs studied are Tc-99 and I-129. (authors)

  11. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  12. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P{sup 3} Mechanism and Methodology Development for Plutonium Categorization

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Masaki [Research Laboratory for Nuclear Reactor, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo, 1528550 (Japan)

    2009-06-15

    'Protected Plutonium Production (P{sup 3})' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section into the uranium fuel to enhance the production of {sup 238}Pu or {sup 242}Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the {sup 238}Pu or {sup 242}Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P{sup 3} mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P{sup 3} mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

  13. Fuel element database: developer handbook

    International Nuclear Information System (INIS)

    Dragicevic, M.

    2004-09-01

    The fuel elements database which was developed for Atomic Institute of the Austrian Universities is described. The software uses standards like HTML, PHP and SQL. For the standard installation freely available software packages such as MySQL database or the PHP interpreter from Apache Software Foundation and Java Script were used. (nevyjel)

  14. Alternative Fuels and Sustainable Development

    DEFF Research Database (Denmark)

    Jørgensen, Kaj; Nielsen, Lars Henrik

    1996-01-01

    The main report of the project on Transportation Fuels based on Renewable Energy. The report contains a review of potential technologies for electric, hybrid and hydrogen propulsion in the Danish transport sector, including an assessment of their development status. In addition, the energy...

  15. Developing fossil fuel based technologies

    International Nuclear Information System (INIS)

    Manzoori, A.R.; Lindner, E.R.

    1991-01-01

    Some of the undesirable effects of burning fossil fuels in the conventional power generating systems have resulted in increasing demand for alternative technologies for power generation. This paper describes a number of new technologies and their potential to reduce the level of atmospheric emissions associated with coal based power generation, such as atmospheric and pressurized fluid bed combustion systems and fuel cells. The status of their development is given and their efficiency is compared with that of conventional pc fired power plants. 1 tab., 7 figs

  16. Minor actinide transmutation using minor actinide burner reactors

    International Nuclear Information System (INIS)

    Mukaiyama, T.; Yoshida, H.; Gunji, Y.

    1991-01-01

    The concept of minor actinide burner reactor is proposed as an efficient way to transmute long-lived minor actinides in order to ease the burden of high-level radioactive waste disposal problem. Conceptual design study of minor actinide burner reactors was performed to obtain a reactor model with very hard neutron spectrum and very high neutron flux in which minor actinides can be fissioned efficiently. Two models of burner reactors were obtained, one with metal fuel core and the other with particle fuel core. Minor actinide transmutation by the actinide burner reactors is compared with that by power reactors from both the reactor physics and fuel cycle facilities view point. (author)

  17. Development of alkaline fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Hibbs, Michael R.; Jenkins, Janelle E.; Alam, Todd Michael; Janarthanan, Rajeswari; Horan, James L.; Caire, Benjamin R.; Ziegler, Zachary C.; Herring, Andrew M.; Yang, Yuan; Zuo, Xiaobing; Robson, Michael H.; Artyushkova, Kateryna; Patterson, Wendy; Atanassov, Plamen Borissov

    2013-09-01

    This project focuses on the development and demonstration of anion exchange membrane (AEM) fuel cells for portable power applications. Novel polymeric anion exchange membranes and ionomers with high chemical stabilities were prepared characterized by researchers at Sandia National Laboratories. Durable, non-precious metal catalysts were prepared by Dr. Plamen Atanassovs research group at the University of New Mexico by utilizing an aerosol-based process to prepare templated nano-structures. Dr. Andy Herrings group at the Colorado School of Mines combined all of these materials to fabricate and test membrane electrode assemblies for single cell testing in a methanol-fueled alkaline system. The highest power density achieved in this study was 54 mW/cm2 which was 90% of the project target and the highest reported power density for a direct methanol alkaline fuel cell.

  18. Front-end and back-end electrochemistry of molten salt in accelerator-driven transmutation systems

    International Nuclear Information System (INIS)

    Williamson, M.A.; Venneri, F.

    1995-01-01

    The objective of this work is to develop preparation and clean-up processes for the fuel and carrier salt in the Los Alamos Accelerator-Driven Transmutation Technology molten salt nuclear system. The front-end or fuel preparation process focuses on the removal of fission products, uranium, and zirconium from spent nuclear fuel by utilizing electrochemical methods (i.e., electrowinning). The same method provides the separation of the so-called noble metal fission products at the back-end of the fuel cycle. Both implementations would have important diversion safeguards. The proposed separation processes and a thermodynamic analysis of the electrochemical separation method are presented

  19. System and safety studies of accelerator driven transmutation. Annual Report 2003

    Energy Technology Data Exchange (ETDEWEB)

    Gudowski, Waclaw; Wallenius, Jan; Tucek, Kamil [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics] [and others

    2004-12-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the Dept. of Nuclear and Reactor Physics reported here has been focused on different aspects of safety of the Accelerator-Driven Transmutation Systems and on Transmutation research in more general terms. An overview of the topics of our research is given in the Summary which is followed by detailed reports as separate chapters or subchapters. Some of the research topics reported in this report are referred to appendices, which have been published in the open literature. Topics, which are not yet published, are described with more details in the main part of this report. Main focus has been, as before, largely determined by the programme of the European projects of the 5th Framework Programme in which KTH is actively participating. In particular: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features. This activity includes even computer modeling of nuclear fuel production. Three different ADS-core concept are being investigated: Conceptual design of Pb-Bi cooled core with nitride fuel so called Sing-Sing Core developed at KTH; Pb-Bi cooled core with oxide fuel so called ANSALDO design for the European Project PDS-XADS; Gas cooled core with oxide fuel a design investigated for the European Project PDS-XADS. b) analysis of potential of advance fuels, in particular nitrides with high content of minor actinides; c) analysis of ADS-dynamics and assessment of major reactivity feedbacks; d) emergency heat removal from ADS; e) participation in ADS: MUSE (CEA-Cadarache), YALINA subcritical experiment in Minsk and designing of the subcritical experiment SAD in Dubna; f) theoretical and simulation studies of radiation damage in high neutron (and/or proton) fluxes; g) computer code and nuclear data development relevant for simulation and optimization of ADS, validation of the MCB code and sensitivity analysis; h) studies of

  20. System and safety studies of accelerator driven transmutation. Annual Report 2003

    International Nuclear Information System (INIS)

    Gudowski, Waclaw; Wallenius, Jan; Tucek, Kamil

    2004-12-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the Dept. of Nuclear and Reactor Physics reported here has been focused on different aspects of safety of the Accelerator-Driven Transmutation Systems and on Transmutation research in more general terms. An overview of the topics of our research is given in the Summary which is followed by detailed reports as separate chapters or subchapters. Some of the research topics reported in this report are referred to appendices, which have been published in the open literature. Topics, which are not yet published, are described with more details in the main part of this report. Main focus has been, as before, largely determined by the programme of the European projects of the 5th Framework Programme in which KTH is actively participating. In particular: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features. This activity includes even computer modeling of nuclear fuel production. Three different ADS-core concept are being investigated: Conceptual design of Pb-Bi cooled core with nitride fuel so called Sing-Sing Core developed at KTH; Pb-Bi cooled core with oxide fuel so called ANSALDO design for the European Project PDS-XADS; Gas cooled core with oxide fuel a design investigated for the European Project PDS-XADS. b) analysis of potential of advance fuels, in particular nitrides with high content of minor actinides; c) analysis of ADS-dynamics and assessment of major reactivity feedbacks; d) emergency heat removal from ADS; e) participation in ADS: MUSE (CEA-Cadarache), YALINA subcritical experiment in Minsk and designing of the subcritical experiment SAD in Dubna; f) theoretical and simulation studies of radiation damage in high neutron (and/or proton) fluxes; g) computer code and nuclear data development relevant for simulation and optimization of ADS, validation of the MCB code and sensitivity analysis; h) studies of

  1. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  2. Accelerator for nuclear transmutation

    International Nuclear Information System (INIS)

    Schapira, J.P.

    1984-01-01

    A review on nuclear transmutation of radioactive wastes using particle accelerators is given. Technical feasibility, nuclear data, costs of various projects are discussed. It appears that one high energy accelerator (1500 MeV, 300 mA proton) could probably handle the amount of actinides generated by the actual French nuclear program [fr

  3. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1998-01-01

    Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

  4. Fuel performance, design and development

    International Nuclear Information System (INIS)

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  5. Dynamic criteria for partitioning and transmutation

    International Nuclear Information System (INIS)

    Lu, A.H.

    1991-11-01

    This paper addresses dynamic criteria intended to optimize partitioning and transmutation (P-T) concept development supporting improved nuclear waste management. Six criteria are proposed initially and the rationale for each is briefly explained. Each criterion is used as a measure (or dimension) on which the developed concepts can be evaluated. The criteria allow the P-T concepts to be evaluated in an integral system including long-term energy needs, fuel cycle, and waste management. New criteria will be identified along with the P-T concept development, and each criterion will be realistically weighted so that it is comparable in an overall criteria evaluation. The weights are subject to change as a result of technical advancements and public perception on various issues. Incomplete criteria will result in a poor choice because important factors may not be considered when the decision is made. A successful decision on the optimal P-T system depends on the completeness of criteria (dimensions) as well as realistic weights assigned to each criterion

  6. System and safety studies of accelerator driven transmutation. Annual Report 2001

    International Nuclear Information System (INIS)

    Gudowski, W.; Wallenius, J.; Tucek, K.; Eriksson, Marcus; Carlsson, Johan; Seltborg, P.; Cetnar, J.; Chakarova, R.; Westlen, D.

    2002-03-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the Dept. of Nuclear and Reactor Physics has been focused in year 2001 on: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features; b) analysis of ADS-dynamics; c) computer code and nuclear data development relevant for simulation and optimization of ADS; d) participation in ADS experiments including 1 MW spallation target manufacturing, subcritical experiments MUSE (CEA-Cadarache) and YALINA experiment in Minsk. The Dept. is very actively participating in many European projects in the 5th Framework Programme of the European Community. Most of the research topics reported in this paper are referred to by appendices, which have been published in the open literature. The topics, which are not yet published, are described here in more details

  7. System and safety studies of accelerator driven transmutation. Annual Report 2001

    Energy Technology Data Exchange (ETDEWEB)

    Gudowski, W; Wallenius, J; Tucek, K; Eriksson, Marcus; Carlsson, Johan; Seltborg, P; Cetnar, J; Chakarova, R; Westlen, D [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2002-03-01

    The research on safety of Accelerator-Driven Transmutation Systems (ADS) at the Dept. of Nuclear and Reactor Physics has been focused in year 2001 on: a) ADS core design and development of advanced nuclear fuel optimised for high transmutation rates and good safety features; b) analysis of ADS-dynamics; c) computer code and nuclear data development relevant for simulation and optimization of ADS; d) participation in ADS experiments including 1 MW spallation target manufacturing, subcritical experiments MUSE (CEA-Cadarache) and YALINA experiment in Minsk. The Dept. is very actively participating in many European projects in the 5th Framework Programme of the European Community. Most of the research topics reported in this paper are referred to by appendices, which have been published in the open literature. The topics, which are not yet published, are described here in more details.

  8. Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Yang, Chao; Cao, Liangzhi; Wu, Hongchun; Zheng, Youqi; Zu, Tiejun

    2013-01-01

    Highlights: • A fusion fission hybrid system for MA transmutation is proposed. • The analysis of neutronics effects on the transmutation is performed. • The transmutation rate of MA reaches 86.5% by 25 times of recycling. -- Abstract: The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25

  9. Subcritical molten salt reactor with fast/intermediate spectrum for minor actinides transmutation

    International Nuclear Information System (INIS)

    Degtyarev, Alexey M.; Feinberg, Olga S.; Kolyaskin, Oleg E.; Myasnikov, Andrey A.; Karmanov, Fedor I.; Kuznetsov, Andrey Yu.; Ponomarev, Leonid I.; Seregin, Mikhail B.; Sidorkin, Stanislav F.

    2011-01-01

    The subcritical molten-salt reactor for transmutation of Am and Cm with the fast-intermediate neutron spectrum is suggested. It is shown that ∼10 such reactor-burners is enough to support the future nuclear power based on the fast reactors as well as for the transmutation of Am and Cm accumulated in the spent fuel storages. (author)

  10. Helium and fission gas behaviour in magnesium aluminate spinel and zirconia for actinide transmutation

    NARCIS (Netherlands)

    Damen, P.M.G.

    2003-01-01

    In order to reduce the long-term radiotoxicity of spent nuclear fuel, many studies are performed on partitioning and transmutation of actinides. In such a scenario, the long-lived radio-isotopes (mostly actinides) are partitioned from the nuclear waste, and subsequently transmuted or fissioned in a

  11. Impact of partitioning and transmutation on the high level waste management

    International Nuclear Information System (INIS)

    Gonzalez-Romero, Enrique-Miguel

    2010-01-01

    The contribution is structured as follows: (i) Background on partitioning and transmutation; (ii) FP6 projects: RED-IMPACT; (iii) Advanced fuel cycle scenarios; (iv) Partitioning and transmutation expected performance; (v) Impact on the HLW thermal load; (vi) Impact on the deep geological disposal; and (vii) Impact on the performance assessment of deep geological disposal. (P.A.)

  12. The U.S. Department of Energy's Advanced Fuel Cycle Initiative is evaluating potential costs and benefits of partitioning and transmutation

    International Nuclear Information System (INIS)

    Van Luik, Abraham E.

    2003-01-01

    Abe van Luik (DOE, USA) stated that the U.S. Department of Energy is interested in P and T to the extent that transmutation is technically feasible and will reduce the toxicity of the waste to a point that makes it technically and economically justified. Therefore, research on P and T strategies incorporates the evaluation of its potential costs and benefits. A progress report to Congress (in preparation) will likely state that system studies in the USA and in Europe indicate a preference for reactor based transmutation rather than accelerator-driven systems. DOE proposes isolation of Cs and Sr, the recycling of Pu and Np in LWRs, and later the recycling of minor actinides in fast reactors. The report identifies the high-level waste volume reduction, the easier management of short-term heat load, the reduction of long-term heat load and radiotoxicity, and therefore long-term dose reduction as potential benefits. The goal of ongoing work is to quantify these benefits in order to allow an assessment of which alternatives can be economically useful in increasing the repository capacity, reducing the potential hazard from the repository and reducing uncertainties associated with the performance of the repository. This may, depending on the national nuclear power scenario, delay or avoid the need for a second repository for high-level waste in the USA. Furthermore DOE has received and is evaluating a proposal for simulation-based engineering to integrate all aspects of nuclear energy including reactor technology and waste disposal

  13. Partitioning and transmutation - Technical feasibility, proliferation resistance and safeguardability

    International Nuclear Information System (INIS)

    Schenkel, R.; Glatz, J.-P.; Magill, J.; Mayer, K.

    2001-01-01

    Full text: The advantages of partitioning and transmutation (P and T) of minor actinides and selected fission products are largely discussed and described in literature. The advantages of separation of the long-lived alpha-emitters for the long-term storage of highly radioactive waste have been highlighted. After separation, these nuclides shall be transmuted by means of a dedicated reactor or accelerator driven system into shorter-lived fission products that are less hazardous. This, however, requires the development and implementation of a P and T fuel cycle, involving chemical separation of the minor actinides and the fabrication of MA containing fuels or targets. Concepts for P and T fuel cycles have been developed and technical issues are being addressed in various research programs. With the recognition of the proliferation potential associated with the minor actinides by the IAEA, also the proliferation and safeguards aspects need to be addressed. It is important to raise these points at an early stage of process development, in order to identify potential problems and to develop appropriate solutions. The oxide fuels used worldwide in thermal reactor systems for energy production are reprocessed by aqueous techniques. Therefore these systems, primarily the PUREX process, are fully developed and implemented commercially. Furthermore, the safeguards approach is fully implemented in existing facilities, covering uranium and plutonium. Pyroprocess systems have largely been associated with fast reactors and metallic fuels and their development has therefore only reached the pilot-scale stage and the feasibility of minor actinide (MA) separation still needs to be demonstrated. Hydrometallurgical and pyrochemical reprocessing should however not be considered as competing but rather as complementary technologies. For instance in a so-called double strata concept (foreseen for instance in the Japanese OMEGA project), the PUREX process (first stratum) would be

  14. Detecting neutrons by forward recoil protons at the Energy & Transmutation facility: Detector development and calibration with 14.1-MeV neutrons

    Science.gov (United States)

    Afanasev, S.; Vishnevskiy, A.; Vishnevskiy, D.; Rogachev, A.; Tyutyunnikov, S.

    2017-05-01

    As part of the Energy & Transmutation project, we are developing a detector for neutrons with energies in the 10-100 MeV range emitted from the target irradiated by a charged-particle beam. The neutron is detected by measuring the time-of-flight and total kinetic energy of the forward-going recoil proton [1] knocked out at a small angle from a thin layer of plastic scintillator, which has to be selected against an intense background created by γ quanta, scattered neutrons, and charged particles. On the other hand, neutron energy has to be measured over the full range with no extra tuning of the detector operation regime. Initial measurements with a source of 14.1-MeV neutrons are reported.

  15. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

    Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

  16. Development of portable fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Nakatou, K.; Sumi, S.; Nishizawa, N. [Sanyo Electric Co., Ltd., Osaka (Japan)

    1996-12-31

    Sanyo Electric has been concentrating on developing a marketable portable fuel cell using phosphoric acid fuel cells (PAFC). Due to the fact that this power source uses PAFC that operate at low temperature around 100{degrees} C, they are easier to handle compared to conventional fuel cells that operate at around 200{degrees} C , they can also be expected to provide extended reliable operation because corrosion of the electrode material and deterioration of the electrode catalyst are almost completely nonexistent. This power source is meant to be used independently and stored at room temperature. When it is started up, it generates electricity itself using its internal load to raise the temperature. As a result, the phosphoric acid (the electolyte) absorbs the reaction water when the temperature starts to be raised (around room temperature). At the same time the concentration and volume of the phosphoric acid changes, which may adversely affect the life time of the cell. We have studied means for starting, operating PAFC stack using methods that can simply evaluate changes in the concentration of the electrolyte in the stack with the aim of improving and extending cell life and report on them in this paper.

  17. Investigation of the feasibility of a small scale transmutation device

    Science.gov (United States)

    Sit, Roger Carson

    This dissertation presents the design and feasibility of a small-scale, fusion-based transmutation device incorporating a commercially available neutron generator. It also presents the design features necessary to optimize the device and render it practical for the transmutation of selected long-lived fission products and actinides. Four conceptual designs of a transmutation device were used to study the transformation of seven radionuclides: long-lived fission products (Tc-99 and I-129), short-lived fission products (Cs-137 and Sr-90), and selective actinides (Am-241, Pu-238, and Pu-239). These radionuclides were chosen because they are major components of spent nuclear fuel and also because they exist as legacy sources that are being stored pending a decision regarding their ultimate disposition. The four designs include the use of two different devices; a Deuterium-Deuterium (D-D) neutron generator (for one design) and a Deuterium-Tritium (D-T) neutron generator (for three designs) in configurations which provide different neutron energy spectra for targeting the radionuclide for transmutation. Key parameters analyzed include total fluence and flux requirements; transmutation effectiveness measured as irradiation effective half-life; and activation products generated along with their characteristics: activity, dose rate, decay, and ingestion and inhalation radiotoxicity. From this investigation, conclusions were drawn about the feasibility of the device, the design and technology enhancements that would be required to make transmutation practical, the most beneficial design for each radionuclide, the consequence of the transmutation, and radiation protection issues that are important for the conceptual design of the transmutation device. Key conclusions from this investigation include: (1) the transmutation of long-lived fission products and select actinides can be practical using a small-scale, fusion driven transmutation device; (2) the transmutation of long

  18. Status of the French Research on Partitioning and Transmutation

    International Nuclear Information System (INIS)

    Warin, Dominique

    2007-01-01

    The global energy context pleads in favor of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel sources. How we deal with radioactive waste is crucial in this context. The production of nuclear energy in France has been associated, since its inception, with the optimization of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. The public's concern regarding the long-term waste management made the French Government prepare and pass the December 1991 Law, requesting in particular, the study for fifteen years of solutions for still minimizing the quantity and the hazardousness of final waste, via partitioning and transmutation. At the end of these fifteen years of research, it is considered that partitioning techniques, which have been validated on real solutions, are at disposal. Indeed, aqueous process for separation of minor actinides from the PUREX raffinate has been brought to a point where there is reasonable assurance that industrial deployment can be successful. A key experiment has been the successful kilogram scale trials in the CEA-Marcoule Atalante facility in 2005 and this result, together with the results obtained in the frame of the successive European projects, constitutes a considerable step forward. For transmutation, CEA has conducted programs proving the feasibility of the elimination of minor actinides and fission products: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. Scenario studies have also allowed assessing the feasibility, at the level of cycle and fuel facilities, and the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending of the

  19. Separation of technetium from ruthenium after the accelerator transmutation of technetium

    International Nuclear Information System (INIS)

    Abney, K.D.; Schroeder, N.C.; Kinkead, S.A.; Attrep, M. Jr.

    1992-01-01

    Both civilian and defense related waste must be processed with a strategy for dealing with Tc. One solution is to remove the Tc from the waste steam and transmute the Tc to stable Ru in either a reactor or an accelerator. Before any processing of waste streams can be performed (even if transmutation is not performed) the separations chemistry from the spent fuels or the stored wastes containing Tc must be developed. This report details some of the separation schemes possible for the separation of Tc and Ru, which include the baseline ion exchange process of Roberts, Smith and Wheelwright, ozonolysis, filtration, magnetic separation, solvent extraction, electrodeposition, fluorination, and pyrolysis. 5 figs, 4 refs

  20. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  1. Dry process fuel performance technology development

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

    2006-06-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  2. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  3. New developments and prospects on COSI, the simulation software for fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.; Tiphine, M.; Krivtchik, G.; Cany, C. [Atomic Energy and Alternative Energies Commission - CEA, CEA-Cadarache, DEN, DER, SPRC, F-13108 Saint-Paul-lez-Durance (France)

    2013-07-01

    COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aim in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.

  4. 1. round table - Spent fuels composition. Back-end of the fuel cycle and reprocessing, plutonium and other nuclear materials management. 2. round table - Separation-transmutation. 3. round table - Scenarios for a long term inventory of nuclear materials and wastes

    International Nuclear Information System (INIS)

    2005-01-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the round table debates which took place at Paris on the reprocessing of spent fuels. Three aspects are discussed: the risks linked with the recovery of valorizable materials, the economical viability of the separation/transmutation option, and the future of the nuclear option in the French energy policy. Six presentations (transparencies) are attached with these proceedings which treat of: the reprocessing/recycling to the test, perspectives of future wastes, present day wastes/valorizable materials and future scenarios, critical analysis scenarios, why reprocessing spent fuels?, processing of spent fuels and recycling, separation and transmutation of long-lived radioactive wastes, thorium-uranium cycle. (J.S.)

  5. Transmutation of radioactive nuclear waste

    International Nuclear Information System (INIS)

    Toor, A; Buck, R

    2000-01-01

    Lack of a safe disposal method for radioactive nuclear waste (RNW) is a problem of staggering proportion and impact. A typical LWR fission reactor will produce the following RNW in one year: minor actinides (i.e. 237 Np, 242-243 Am, 243-245 Cm) ∼40 kg, long-lived fission products (i.e, 99 Tc, 93 Zr, 129 I, 135 Cs) ∼80 kg, short lived fission products (e.g. 137 Cs, 90 Sr) ∼50kg and plutonium ∼280 kg. The total RNW produced by France and Canada amounts to hundreds of metric tonnes per year. Obtaining a uniform policy dealing with RNW has been blocked by the desire on one hand to harvest the energy stored in plutonium to benefit society and on the other hand the need to assure that the stockpile of plutonium will not be channeled into future nuclear weapons. In the meantime, the quantity and handling of these materials represents a potential health hazard to the world's population and particularly to people in the vicinity of temporary storage facilities. In the U.S., societal awareness of the hazards associated with RNW has effectively delayed development of U.S. nuclear fission reactors during the past decade. As a result the U.S. does not benefit from the large investment of resources in this industry. Reluctance to employ nuclear energy has compelled our society to rely increasingly on non-reusable alternative energy sources; coal, oil, and natural gas. That decision has compounded other unresolved global problems such as air pollution, acid rain, and global warming. Relying on these energy sources to meet our increasing energy demands has led the U.S. to increase its reliance on foreign oil; a policy that is disadvantageous to our economy and our national security. RNW can be simplistically thought of as being composed of two principal components: (1) actinides with half lives up to 10 6 years and (2) the broad class of fission fragments with typical half lives of a few hundred years. One approach to the RNW storage problem has been to transmute the

  6. Partitioning and Transmutation. Annual Report 2002

    International Nuclear Information System (INIS)

    Andersson, S.; Ekberg, C.; Liljenzin, J.O.; Nilsson, M.; Rogues, N.; Skarnemark, G.; Oestberg, J.

    2003-01-01

    How to deal with the spent fuel from nuclear power plants is an issue that much research is attracted to in many countries around the world. Several different strategies exist for treating the waste ranging from direct disposal to reprocessing and recycling of plutonium and other long-lived nuclides. In either case the remains have to be stored for a long time to render it radio-toxically safe. One method to deal with this long-lived waste is to separate (separation) out the most long lived components and then transform them into shorter-lived ones (transmutation). Several methods exist for performing the separation for example via molten salts and through solvent extraction. The work presented here has been focused on solvent extraction. This technique is well known since many years and process scale plants have been operating for decades. The new demand is to separate chemically very similar elements from each other. Within this project this is done by new extracting agents developed for this purpose alone within the EU fifth framework programme, the PARTNEW project, particularly from the University of Reading. In this work we investigate different extraction systems for the separation of trivalent actinides from trivalent lanthanides using extraction agents following the so-called CHON (Carbon, Hydrogen, Oxygen and Nitrogen) principle. The main focus is to understand the basic chemistry involved but also some processing behaviour for use in future full scale plants

  7. Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  8. ACCELERATOR TRANSMUTATION OF WASTE TECHNOLOGY AND IMPLEMENTATION SCENARIOS

    International Nuclear Information System (INIS)

    Beller, D.; Tuyle, G. van

    2000-01-01

    During 1999, the U.S. Department of Energy, in conjunction with its nuclear laboratories, a national steering committee, and a panel of world experts, developed a roadmap for research, development, demonstration, and deployment of Accelerator-driven Transmutation of Waste (ATW). The ATW concept that was examined in this roadmap study was based on that developed at the Los Alamos National Laboratory (LANL) during the 1990s. The reference deployment scenario in the Roadmap was developed to treat 86,300 tn (metric tonnes initial heavy metal) of spent nuclear fuel that will accumulate through 2035 from existing U.S. nuclear power plants (without license extensions). The disposition of this spent nuclear reactor fuel is an issue of national importance, as is disposition of spent fuel in other nations. The U.S. program for the disposition of this once-through fuel is focused to characterize a candidate site at Yucca Mountain, Nevada for a geological repository for spent fuel and high-level waste. The ATW concept is being examined in the U.S. because removal of plutonium minor actinides, and two very long-lived isotopes from the spent fuel can achieve some important objectives. These objectives include near-elimination of plutonium, reduction of the inventory and mobility of long-lived radionuclides in the repository, and use of the remaining energy content of the spent fuel to produce power. The long-lived radionuclides iodine and technetium have roughly one million year half-lives, and they are candidates for transport into the environment via movement of ground water. The scientists and engineers who contributed to the Roadmap Study determined that the ATW is affordable, doable, and its deployment would support all the objectives. We report the status of the U.S. ATW program describe baseline and alternate technologies, and discuss deployment scenarios to support the existing U.S. nuclear capability and/or future growth with a variety of new fuel cycles

  9. Processing flowsheet for the accelerator transmutation of waste (ATW) program

    International Nuclear Information System (INIS)

    Dewey, H.; Walker, R.; Yarbro, S.

    1992-01-01

    At Los Alamos, an innovative approach to transmuting long-lived radioactive waste is under investigation. The concept is to use a linear proton accelerator coupled to a solid target to produce an intense neutron flux. The intense stream of neutrons can then be used to fission or transmute long-lived radionuclides to either stable or shorter-lived isotopes. For the program to be successful, robust chemical separations with high efficiencies (>10 5 ) are required. The actual mission, either defense or commercial, will determine what suite of unit operations will be needed. If the mission is to process commercial spent fuel, there are several options available for feed preparation and blanket processing. The baseline option would be an improved PUREX system with the main alternative being the current ATW actinide blanket processing flowsheet. 99 Tc and 129 I are more likely to reach the biosphere than the actinides. Many models have been developed for predicting how the radionuclides will behave in a repository over long time periods. The general conclusion is that the actinides will be sorbed by the soil. Therefore, over a long time period, e.g., a million years their hazard will be lessened because of radioactive decay and dispersion. However, some of the long-lived fission products are not sorbed and could potentially reach the environment over a few thousand year period. Hence, they could present a significant safety hazard. Because of limited resources, most of the priority has been focused on the actinide and technetium blanket assemblies

  10. Transmutation and accelerator driven systems

    International Nuclear Information System (INIS)

    Shapira, J.P.

    2001-01-01

    Full text: Today, countries who are presently involved in nuclear energy are facing many challenges to maintain this option open for the next few decades. Among them, management of nuclear wastes produced in nuclear reactors and in fuel cycle operations has become a very strong environmental issue among the public. In most countries with sizeable commercial nuclear programs, deep geological disposal of ultimate highly active and long-lived nuclear wastes is considered as the reference long-term management scheme. But, many questions arise on the possibility to demonstrate that such wastes can be dealt in such a way as to protect the future generations and the environment. The characteristics of nuclear wastes, the various back end policies concerning spent fuels and the nuclear wastes long-term management options will be first described. Then recent proposals, based on transmutation, especially those using accelerator driven systems (ADS) and/or thorium will be presented. Finally, the possibility for the nuclear physics community to play a part in alleviating the nuclear wastes burden will be pointed out. (author)

  11. Calculations of different transmutation concepts. An international benchmark exercise

    International Nuclear Information System (INIS)

    2000-01-01

    In April 1996, the NEA Nuclear Science Committee (NSC) Expert Group on Physics Aspects of Different Transmutation Concepts launched a benchmark exercise to compare different transmutation concepts based on pressurised water reactors (PWRs), fast reactors, and an accelerator-driven system. The aim was to investigate the physics of complex fuel cycles involving reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types. The objective was also to compare the calculated activities for individual isotopes as a function of time for different plutonium and minor actinide transmutation scenarios in different reactor systems. This report gives the analysis of results of the 15 solutions provided by the participants: six for the PWRs, six for the fast reactor and three for the accelerator case. Various computer codes and nuclear data libraries were applied. (author)

  12. Transmutation studies of minor actinides in high intensity neutron fluxes

    International Nuclear Information System (INIS)

    Fioni, G.; Bolognese, T.; Cribier, M.; Marie, F.; Roettger, S.; Faust, H.; Leconte, Ph.

    1999-01-01

    Integral measurements of nuclear data and of the transmutation potential in specific neutron fluxes, constitute the fastest and essential way to overcome to the large uncertainties present in the nuclear data libraries. In the frame of the activities of the Directorate for Science of Matter (DSM) of the French Atomic Energy Authority (CEA), a new project is proposed so as to carry out integral measurements relevant for nuclear waste transmutation systems. A new beam tube will be installed to irradiate actinides and fission fragment samples at different distances from the fuel element of the ILL reactor. Variable neutron energy spectra could then be obtained by choosing the distance between the sample and the fuel element, opening the way to the determination of the ideal physical conditions to incinerate nuclear waste in hybrid transmutation systems. (author)

  13. International collaboration, the route to fuel cycle research and development

    International Nuclear Information System (INIS)

    Tinsley, T.; Mathers, D.; Rayment, F.

    2013-01-01

    In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes

  14. International collaboration, the route to fuel cycle research and development

    Energy Technology Data Exchange (ETDEWEB)

    Tinsley, T.; Mathers, D.; Rayment, F. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2013-07-01

    In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes.

  15. Separation and transmutation. A picture of the applications in Sweden; Separation och transmutation. Belysning av tillaempning i Sverige

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil; Lindgren, Maria [Kemakta Konsult AB, Stockholm (Sweden)

    2006-11-15

    This report contains a review of the transmutation technology and an elucidation of the consequences of the use of transmutation in Sweden. Transmutation has often been mentioned in the public debate as a way of rendering harmless the high-level waste from nuclear power such that the final disposal of the waste can be substantially simplified or even completely avoided. However, it can be noted that even with an exploitation of transmutation, significant amounts of radioactive waste requiring qualified final disposal will be generated. The transmutation technology will make it possible to reduce the longevity of the high-level waste by converting primarily the transuranic elements to fission products with shorter half lives. The long-term radiotoxicity of the spent nuclear fuel is dominated by the transuranics. Hence, transmutation will lead to a substantial decrease of the long-term radiotoxicity of the spent fuel. The research on transmutation has been focussed on sub-critical so called ADS-reactors (Accelerator Driven System). In such a system protons are accelerated to very high energy levels (in the order of GeV) in an electromagnetic field. The accelerated protons are impacted on a spallation source consisting of heavy atoms, e.g. lead or a mixture of lead and bismuth. At the impact the heavy nuclei are spalled releasing a number of neutrons that can be used for fissioning the nuclei of the substances to be transmuted, primarily the transuranics. ADS-reactors are still at the research stage. It is a common view that it will take several decades before the technology has reached a maturity that allows the construction of a demonstration facility. Calculations performed at Royal Institute of Technology in Stockholm show that using the ADS-technology would allow a reduction of the inventory of transuranics in the spent fuel from Swedish reactors by 50-85% within a 50-100 years period. The goal to transmute 99% of the transuranics inventory has been achieved in

  16. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  17. Development of nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

    1995-01-01

    In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

  18. Spent fuel: prediction model development

    International Nuclear Information System (INIS)

    Almassy, M.Y.; Bosi, D.M.; Cantley, D.A.

    1979-07-01

    The need for spent fuel disposal performance modeling stems from a requirement to assess the risks involved with deep geologic disposal of spent fuel, and to support licensing and public acceptance of spent fuel repositories. Through the balanced program of analysis, diagnostic testing, and disposal demonstration tests, highlighted in this presentation, the goal of defining risks and of quantifying fuel performance during long-term disposal can be attained

  19. MARS: Story on Molten Salt Actinide Recycler and Transmuter Development by Rosatom in Co-operation with Euratom

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feynberg, O.; Gnidoi, I.; Konakov, S.; Kormilitsyn, M.; Merzliakov, A.; Surenkov, A.; Uglov, V.; Zagnitko, A.

    2015-01-01

    New design options of MOSART and MSFR systems without and with U-Th support fuelled with different compositions of transuranic elements trifluorides from spent LWR fuel both based on homogeneous cores and used fuel salts with high enough solubility for transuranic elements trifluorides are being examined within MARS (Rosatom) and EVOL (Euratom) parallel coordinated projects. The paper has the main objective of presenting the fuel cycle flexibility of the mentioned above systems while accounting technical constrains and experimental data received in this study. A brief description is given of the calculation core neutronics properties and fuel cycle scenarios as well as experimental results on key fuel salt properties, salt chemistry control and combined materials compatibility to satisfy MOSART and MSFR systems requirements. Measurements described mainly concern phase behaviour and transport properties data for selected fuel salts. As for fuel salt clean-up operations in MOSART and MSFR fuel cycles, the most uncertain its part concerning rare earth removal is discussed. Last section is focused on the compatibility of special Ni-based alloys with fuel salt selected at temperatures required for MOSART and MSFR operation. The major achievements are: (1) ability to produce and maintain a high level of purity in fuel salt, (2) effective control of the Redox potential of the salt medium in order to minimize corrosion, (3) understanding of basic corrosion mechanisms in MOSART and MSFR systems. HN80MTY alloy can be recommended for further consideration as the main container material for the fuel circuit with operating temperature up to 1 023 K required for MOSART and MSFR designs. (authors)

  20. Fuel cell development for transportation: Catalyst development

    Energy Technology Data Exchange (ETDEWEB)

    Doddapaneni, N. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

  1. Development of An Advanced JP-8 Fuel

    Science.gov (United States)

    1993-12-01

    included the Microthermal Precipitation Test (MTP), Fuel Reactor Test, Hot Liquid Process Simulator (HLPS), and Isothermal Corrosion Oxidation Test (ICOT... Microthermal Precipitation Test The impetus for this development effort was the need for a screening test that could discriminate between fuels of...varying propensity to produce thermally induced insoluble particulate material in the bulk fuel. The Microthermal Precipitation (MTP) test thermally

  2. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

    1990-02-01

    Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

  3. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

  4. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    1985-01-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  5. Nuclear waste transmutation

    International Nuclear Information System (INIS)

    Leray, S.

    1995-01-01

    Accelerators can play a role in the disposal of long-lived radioactive waste: an alternative to the storage in deep underground repositories might transmuting long-lived elements into stable or short-lived ones in subcritical systems driven by spallation neutrons. These neutrons would be produced by a high intensity, intermediate energy proton accelerator irradiating a heavy target. Similar systems have also been proposed to produce energy with a minimized waste inventory. Since a good knowledge of the spallation process is essential for designing and optimizing the target-blanket assembly, new programmes aimed at studying spallation reactions are in progress. (author). 6 figs

  6. Transmutation of Tc-99 and I-129 in fission reactors. A calculational study

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1995-03-01

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MW e CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MW e ), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MW e PWR). When all guide tubes of an UO 2 fuelled PWR are filled with Tc-99 with density of 5 g cm -3 , the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

  7. Transmutation of radioactive nuclear waste – present status and ...

    Indian Academy of Sciences (India)

    Transmutation of long-lived actinides and fission products becomes an important issue of the overall nuclear fuel cycle assessment, both for existing and future reactor systems. Reliable nuclear data are required for analysis of associated neutronics. The present paper gives a review of the status of nuclear data analysis ...

  8. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  9. The physics design of accelerator-driven transmutation systems

    International Nuclear Information System (INIS)

    Venneri, F.

    1995-01-01

    Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safer, less expensive and more environmentally sound approach to nuclear power

  10. The physics design of accelerator-driven transmutation systems

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, F. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safe, less expensive and more environmentally sound approach to nuclear power.

  11. Study of a neutronic potential of a modular fast spectrum ADS for radiotoxic waste transmutation

    International Nuclear Information System (INIS)

    Slessarev, I.; Arkhipov, V.

    1999-01-01

    The IAEA has maintained an active interest in advanced nuclear technology related to Accelerator Driven Systems (ADS). The activities carried out by the IAEA within its Programme on Emerging Nuclear Energy Systems for Energy Generation and Transmutation include preparation of status reports on advanced technologies development, conduct of technical information exchange meetings and co-operative Co-ordinated Research Programmes (CRPs). Consultancy on Accelerator Driven Systems and Related Fuel Cycles held from 1-2 February 1996 in Cadarache, France reviewed the recent progress in studies on feasibility of Accelerator Driven Systems and participants recommended the IAEA to organize a CRP to investigate the accelerator driven systems potential. In 1996 the CRP: 'Use of Th-based Fuel Cycle in Accelerator Driven Systems (ADS) to Incinerate Pu and to Reduce Long-term Waste Toxicities' was established and the first Research Coordination Meeting was convened in March 1997 at the ENEA Research Center in Bologna, Italy. Results of the first stage 'Accelerator Driven Systems (ADS)-Neutronic Benchmark' were reported to the IAEA Technical Committee Meeting on Feasibility and Motivation for Hybrid Concepts for Nuclear Energy Generation and Transmutation in Madrid, Spain, in September 1997. Results obtained during the second stage of the CRP were presented and analysed at the RCM in December 1998 in Petten, Netherlands. The meeting was attended by 21 representatives and observers from IAEA, Belarus, Czech Republic, France, Germany, Italy, the Netherlands, the Russian Federation, Spain, Sweden and USA. This paper was prepared on recommendation of the participants of the meeting in Petten, Netherlands. Accelerator Driven System (ADS) neutronic potential, i.e. capability to burnout transuranic elements (TRU) or minor actinides (MA) and Long-Lived Fission Products (LLFP) with minimum ADS park as well as the rate of LLFP transmutation (mass/energy production) for nuclear waste

  12. Development of Nuclear Fuel Remote Fabrication Technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

    2005-04-01

    The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

  13. Partitioning and transmutation. Annual report 2008

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Foreman, Mark; Naestren, Catharina; Retegan, Teodora; Skarnemark, Gunnar (Nuclear Chemistry, Dept. of Chemical and Biological Engineering, Chalmers Univ. of Technology, Goeteborg (Sweden))

    2009-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products (79Se, 87Rb, 99Tc, 107Pd, 126Sn, 129I, 135Cs) and activation products (14C, 36Cl, 59Ni, 93Zr, 94Nb). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel for different reasons. One being high cross sections for neutron capture of some elements, like the lanthanides. Other reasons may be the unintentional making of other long lived isotopes. The most difficult separations to make are those between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and those between different actinides themselves. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. In the case of a fuel with a higher burnup or possible future fuels, pyro processing may be of higher advantage due to the limited risk of criticality during the process. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in several European frame work programmes from NEWPART in the 4th framework via PARTNEW and EUROPART to ACSEPT in the present 7th programme. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. One process, the SANEX (Selective ActiNide EXtraction) is now considered to be working on a basic scale and focus has moved on to more process oriented areas. However, since further investigations on basic understanding of the chemical behaviour are required, we have our main focus on the chemical processes and understanding of how they work. Our work is now manly focussed on the so called GANEX (Group ActiNide EXtraction) process. Due to new recruitments we will now also work

  14. AECL's progress in DUPIC fuel development

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Ryz, M.A.; Lee, J.W.

    1997-01-01

    Previous papers described progress in choosing a fabrication route for the DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel cycle [1], details of the OREOX (Oxidation Reduction of Oxide fuel) process, and preliminary results of out-cell and small-scale in-cell experiments [2]. AECL's project to develop the DUPIC fuel cycle has now progressed to the stage of fabricating DUPIC fuel elements for irradiation testing in a research reactor. Because of the high radiation fields around the spent PWR fuel, all work is being done in hot cells. The equipment used for fabrication of the DUPIC fuel elements is described in this paper. The commissioning, in-cell installation and current status of the fabrication process are also described and plans for the completion of this phase of the DUPIC project are outlined. The goal of this phase of the project is demonstration of the technical feasibility of the DUPIC fuel cycle. (author)

  15. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

    1992-05-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

  16. LEU fuel development at CERCA

    International Nuclear Information System (INIS)

    Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

    1998-01-01

    The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

  17. New data libraries for transmutation studies

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Interfaculty Reactor Inst., Delft (Netherlands)

    1995-06-01

    The fuel depletion code ORIGEN-S is often used for transmutation studies. It uses three different working libraries for actinides, fission products, and light elements, which contain decay data, cross-section data and fission product yields. These data have been renewed with data based on the JEF2.2 and the EAF3 evaluated files. Furthermore, data for 201 fission products have been added to the libraries. The new data libraries are particular suitable for parameter studies and other introductory calculations. For more accurate calculations, it is advised to regularly update the cross sections of the most important actinides and fission products during the burnup sequence. (orig.).

  18. New data libraries for transmutation studies

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Interfaculty Reactor Inst., Delft (Netherlands)

    1995-12-31

    The fuel depletion code ORIGEN-S is often used for transmutations studies. It uses three different working libraries for actinides, fission products, and light elements, which contain decay data, cross-section data and fission product yields. These data have renewed with data based on the JEF2.2 and the EAF3 evaluated files. Furthermore, data for 201 fission products have been added to the libraries. The new data libraries are particular suitable for parameter studies and other introductory calculations. For more accurate calculations, it is advised to regularly update the cross sections of the most important actinides and fission products during the burnup sequence. (author) 9 refs.

  19. Pellet fueling development at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foster, C.A.; Schuresko, D.D.; Foust, C.R.; Simmons, D.W.; Beard, D.S.

    1986-09-01

    Advanced plasma fueling systems for magnetic confinement devices are being developed at the Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets at speeds in the range of 1-2 km/s and higher. Two specific concepts are under development: (1) high-speed pneumatic acceleration; and (2) mechanical (centrifugal) acceleration. Both approaches are being pursued to meet the projected pellet size and delivery rates for major near-term plasma confinement devices, such as the Tokamak Fusion Test Reactor (TFTR), Tore Supra, the Joint European Torus (JET), JT-60, and Doublet III-D (DIII-D), as well as future applications. In addition to these confinement physics related activities, ORNL is pursuing advanced technologies to achieve pellet velocities significantly in excess of the 2-km/s range already attained with pneumatic injectors and has embarked on a development program designed to explore the feasibility of fabricating and accelerating tritium pellets. This paper describes these ongoing activities

  20. Optimisation of the neutron source based on gas dynamic trap for transmutation of radioactive wastes

    Science.gov (United States)

    Anikeev, Andrey V.

    2012-06-01

    The Budker Institute of Nuclear Physics in collaboration with the Russian and foreign organizations develop the project of 14 MeV neutron source, which can be used for fusion material studies and for other application. The projected neutron source of plasma type is based on the plasma Gas Dynamic Trap (GDT), which is a special magnetic mirror system for plasma confinement. Presented work continues the subject of development the GDT-based neutron source (GDT-NS) for hybrid fusion-fission reactors. The paper presents the results of recent numerical optimization of such neutron source for transmutation of the long-lives radioactive wastes in spent nuclear fuel.

  1. Deep burn transmutation of nuclear waste

    International Nuclear Information System (INIS)

    Rodriguez, C.; Baxter, A.; McEachern, D.; Venneri, F.; Williams, D.

    2002-01-01

    Helium-cooled, graphite-moderated reactors with ceramic-coated fuel particles offer unique advantages for the destruction of transuranic materials discharged in Light Water Reactor spent fuel. This is accomplished by fission, and capture-followed-by-fission processes. Three major features make it practical: (1) ceramic-coated particles accommodate high levels of burnup in one pass, thus reducing the need for repeated reprocessing; (2) graphite moderation produces valuable opportunities for thermal and epithermal neutrons to interact with fissionable and non-fissionable materials respectively; and (3) ceramic-coated particle kernel sizes can be adjusted to control the rate of such interactions. In the transmutation scheme proposed here, virtually complete destruction of weapons-usable materials, and 95% destruction of all transuranic waste is achieved. Higher levels of destruction are possible by repeated reprocessing and recycling, but there is little incentive to do so since each reprocessing step generates new secondary waste. After transmutation, the impervious ceramic-coated fuel particles provide an ideal residual waste form. (author)

  2. The French partitioning-transmutation programme, assets and prospects

    International Nuclear Information System (INIS)

    Viala, M.; Salvatores, M.; Mouney, H.

    1997-01-01

    Partitioning-transmutation studies are covered by the 1991 French law concerning radioactive waste management. The programme is progressing with a dual approach: - What can be done in partitioning-transmutation? At what cost? In what timescale? - How can long-term gains and short-term disadvantages be qualified and quantified? The first approach concerns technical know-how. The studies based on today's technologies are continuing (reactors, fuels and targets, separation of radionuclides by solvents). The second approach involves an assessment activity, based firstly on studies of scenarios. Pertinent assessment criteria must be brought out. (authors)

  3. Transmutation of LWR waste actinides in thermal reactors

    International Nuclear Information System (INIS)

    Gorrell, T.C.

    1979-01-01

    Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

  4. Waste transmutation and public acceptance

    International Nuclear Information System (INIS)

    Pigford, T.H.

    1991-01-01

    The concept of transmuting radioactive wastes with reactors or accelerators is appealing. It has the potential of simplifying or eliminating problems of disposing of nuclear waste. The transmutation concept has been renewed vigorously at a time when national projects to dispose of high-level and transuranic waste are seriously delayed. In this period of tightening federal funds and program curtailments, skilled technical staffs are available at US Department of Energy (DOE) national laboratories and contractors to work on waste transmutation. If the claims of transmutation can be shown to be realistic, economically feasible, and capable of being implemented within the US institutional infrastructure, public acceptance of nuclear waste disposal may be enhanced. If the claims for transmutation are not substantiated, however, there will result a serious loss of credibility and an unjust exacerbation of public concerns about nuclear waste. The paper discusses the following topics: how public acceptance is achieved; the technical community and waste disposal; transmutation and technical communication; transmutation issues; technical fixes and public perception

  5. Economic assessment of partitioning, transmutation and waste reduction technologies

    International Nuclear Information System (INIS)

    Lauferts, U.; Van Heek, A.; Hart, J.

    2007-01-01

    This nuclear system study focuses on a realistic evolution of Partitioning and Transmutation technologies, which can be deployed incrementally on an industrial scale as well as on future developments such as reactors of the third and fourth generation and Accelerated Driven Systems (ADS). A set of five different fuel cycles has been selected, representing the options proposed in different European countries. Two industrial scenarios as continuation of the open nuclear fuel cycles and mono-recycling of plutonium in PWRs have been chosen as a reference. In addition, 3 more innovative cycles are considered using Fast Generation IV reactors and double strata scenarios with advanced PWR, ADS and fast reactors. This study shows, first, that closing the nuclear fuel cycle would be a useful strategy to mitigate concerns about a rapid depletion of natural uranium resources in this century. Secondly, all the 3 advanced fuel cycle strategies proposed reduce effectively the total amount of nuclear waste out of pile and consequently the need for large capacities of deep geological repositories. Thirdly, the most efficient strategy towards the mitigation of waste production is the utilization of fast reactors technology to burn plutonium and ADS to burn minor actinides

  6. A study for optimal transmutation system

    International Nuclear Information System (INIS)

    Park, W.S.; Song, T.Y.; Shin, H.S.; Park, C.K.

    1996-01-01

    Couple of transmutation systems are being under investigation to design the optimal transmutation device. Several basic studies were performed for that objectives: (1) select the radioactive nuclides to be transmuted: (2) investigate the physical characteristics of each nuclide; (3) study the most favorable neutron energy environment for the transmutation. The existing LWR and LMFBR cores were found to be not a satisfiable ones in terms of transmutation rate itself. (author). 5 refs, 2 figs, 3 tabs

  7. Physics and safety of transmutation systems. A status report

    International Nuclear Information System (INIS)

    2006-01-01

    The safe and efficient management of spent fuel from the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from currently operating reactors will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors. Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium.When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred to a thousand years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus de done in controlled environments having timescales of centuries rather than millennia stretching beyond 10 000 years. Transmutation is one of the means being explored to address the disposal of transuranic elements. To achieve this, advanced reactors systems, appropriate fuels, separation techniques and associated fuel cycle strategies are required. This status report begins by providing a clear definition of partitioning and transmutation (P and T), and then describes the state of the art concerning the challenges facing the implementation of P and T, scenario studies and specific issues related to accelerator-driven systems (ADS) dynamics and safety, long-lived fission product transmutation and the impact of nuclear data uncertainty on transmutation system design. The report will be of particular interest to nuclear scientists working on P and T issues as well as advanced fuel cycles in general. (author)

  8. Partitioning and transmutation. Annual report 2009

    International Nuclear Information System (INIS)

    Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Foreman, Mark; Loefstroem-Engdahl, Elin; Retegan, Teodora; Skarnemark, Gunnar; Spendlikova, Irena

    2010-01-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products ( 79 Se, 87 Rb, 99 Tc, 107 Pd, 126 Sn, 129 I and 135 Cs) and activation products ( 14 C, 36 Cl, 59 Ni, 93 Zr, 94 Nb). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel for different reasons. One being high neutron capture cross sections for some elements, like the lanthanides. Other reasons may be the unintentional production of other long lived isotopes. The most difficult separations to make are those between different actinides but also between trivalent actinides and lanthanides, due to their relatively similar chemical properties. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. In the case of a fuel with a higher burnup or possible future fuels, pyro processing may be of higher advantage due to the limited risk of criticality during the process. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in several European frame work programmes. These projects range from NEWPART in the 4th framework via PARTNEW and EUROPART to ACSEPT in the present 7th programme. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. One process, the SANEX (Selective ActiNide EXtraction) is now considered to be working on a basic scale and focus has moved on to more process oriented areas. However, since further investigations on basic understanding of the chemical behaviour are required, we have our main focus on the chemical processes and understanding of how they work. Our work is now manly focussed on the so called GANEX (Group ActiNide EXtraction) process. We have proposed a

  9. Partitioning and transmutation. Annual report 2009

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Foreman, Mark; Loefstroem-Engdahl, Elin; Retegan, Teodora; Skarnemark, Gunnar; Spendlikova, Irena (Nuclear Chemistry, Department of Chemical and Biological Engineering, Chalmers Univ. of Technology, Goeteborg (Sweden))

    2010-01-15

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products (79Se, 87Rb, 99Tc, 107Pd, 126Sn, 129I and 135Cs) and activation products (14C, 36Cl, 59Ni, 93Zr, 94Nb). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel for different reasons. One being high neutron capture cross sections for some elements, like the lanthanides. Other reasons may be the unintentional production of other long lived isotopes. The most difficult separations to make are those between different actinides but also between trivalent actinides and lanthanides, due to their relatively similar chemical properties. Solvent extraction is an efficient and well-known method that makes it possible to have separation factors that fulfil the highly set demands on purity of the separated phases and on small losses. In the case of a fuel with a higher burnup or possible future fuels, pyro processing may be of higher advantage due to the limited risk of criticality during the process. Chalmers University of Technology is involved in research regarding the separation of actinides and lanthanides and between the actinides themselves as a partner in several European frame work programmes. These projects range from NEWPART in the 4th framework via PARTNEW and EUROPART to ACSEPT in the present 7th programme. The aims of the projects have now shifted from basic understanding to more applied research with focus on process development. One process, the SANEX (Selective ActiNide EXtraction) is now considered to be working on a basic scale and focus has moved on to more process oriented areas. However, since further investigations on basic understanding of the chemical behaviour are required, we have our main focus on the chemical processes and understanding of how they work. Our work is now manly focussed on the so called GANEX (Group ActiNide EXtraction) process. We have proposed a novel process

  10. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

    1993-09-01

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  11. 1. round table - Spent fuels composition. Back-end of the fuel cycle and reprocessing, plutonium and other nuclear materials management. 2. round table - Separation-transmutation. 3. round table - Scenarios for a long term inventory of nuclear materials and wastes; 1. table ronde - La composition des combustibles uses. L'aval du combustible et le retraitement, la gestion du plutonium et des autres matieres nucleaires. 2. table ronde - Separation-transmutation. 3. table ronde - Scenarii pour un inventaire des matieres et des dechets nucleaires a LT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the round table debates which took place at Paris on the reprocessing of spent fuels. Three aspects are discussed: the risks linked with the recovery of valorizable materials, the economical viability of the separation/transmutation option, and the future of the nuclear option in the French energy policy. Six presentations (transparencies) are attached with these proceedings which treat of: the reprocessing/recycling to the test, perspectives of future wastes, present day wastes/valorizable materials and future scenarios, critical analysis scenarios, why reprocessing spent fuels?, processing of spent fuels and recycling, separation and transmutation of long-lived radioactive wastes, thorium-uranium cycle. (J.S.)

  12. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  13. Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

    International Nuclear Information System (INIS)

    Feng, K.M.; Huang, J.H.

    1994-01-01

    The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

  14. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  15. 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)

  16. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  17. Development of a lightweight fuel cell vehicle

    Science.gov (United States)

    Hwang, J. J.; Wang, D. Y.; Shih, N. C.

    This paper described the development of a fuel cell system and its integration into the lightweight vehicle known as the Mingdao hydrogen vehicle (MHV). The fuel cell system consists of a 5-kW proton exchange membrane fuel cell (PEMFC), a microcontroller and other supported components like a compressed hydrogen cylinder, blower, solenoid valve, pressure regulator, water pump, heat exchanger and sensors. The fuel cell not only propels the vehicle but also powers the supporting components. The MHV performs satisfactorily over a hundred-kilometer drive thus validating the concept of a fuel cell powered zero-emission vehicle. Measurements further show that the fuel cell system has an efficiency of over 30% at the power consumption for vehicle cruise, which is higher than that of a typical internal combustion engine. Tests to improve performance such as speed enhancement, acceleration and fuel efficiency will be conducted in the future work. Such tests will consist of hybridizing with a battery pack.

  18. Development of metallic fuel materials

    International Nuclear Information System (INIS)

    Kang, Young Ho; Lee, Chong Tak; Yang, Yeoung Seok; Kim, Ki Hwan; Hwang, Sung Chan; Joo, Keun Sik; Ann, Hyun Suk; Chang, Sae Jung.

    1997-09-01

    Through the control of melting and casting parameters, the sound and homogenous U-10wt.%Zr alloy could be fabricated. The yield and segregation of Zr elements were 85% and ±0.1wt.%, and the density of the alloy was about 16.6 g/cm 3 . The major phase were α-U and δ-UZr 2 . The microstructure showed the laminar structure with fiber morphology which was arranged alternatively with uranium and Zr-rich phase. This alloy will be used for KALIMER fuel material through developing the fabrication technology and the characteristics analysis. And electrorefining study was performed to separate uranium from uranium-neodymium and uranium-zirconium alloy by their different free energy for chloride formation. The liquid cadmium phase becomes the anode of the electrorefining cell. Uranium is electrolytically transported through a molten salt electrolyte to a low carbon steel cathode. The electrolyte is composed of KCl-LiCl eutectic and some UCl 3 , which are installed in the salt to facilitate the electrotransport of uranium. In pyrochemical process the reaction condition of chlorination and the maintenance its purity in preparing UCl 4 by chlorination of UO 2 is strongly dependent on the reaction temperature and time. (author).52 refs., 40 tabs., 129 figs

  19. Nuclear Fuel Design Technology Development for the Future Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

    2006-04-15

    The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

  20. An optimization methodology for heterogeneous minor actinides transmutation

    Science.gov (United States)

    Kooyman, Timothée; Buiron, Laurent; Rimpault, Gérald

    2018-04-01

    In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

  1. Nuclear fuel fabrication - developing indigenous capability

    International Nuclear Information System (INIS)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

    1997-01-01

    Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO 2 powder production are also briefly indicated. (author)

  2. Impacts of new developments in partitioning and transmutation on the disposal of high-level nuclear waste in a mined geologic repository

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Jor-Shan Choi; Halsey, W.; Pasternak, A.; Cotton, T.; Burns, J.; McCabe, A.; Colglazier, W.; Lee, W.W.L.

    1992-03-01

    During the 1970s, the United States and other countries thoroughly evaluated the options for the safe and final disposal of high-level radioactive wastes (HLW). The worldwide scientific community concluded that deep geologic disposal was clearly the most technically feasible alternative. They also ranked the partitioning and transmutation (P-T) of radionuclides among the least favored options. A 1982 report by the International Atomic Energy Agency summarized the key reasons for that ranking: ''Since the long-term hazards are already low, there is little incentive to reduce them further by P-T. Indeed the incremental costs of introducing P-T appear to be unduly high in relation to the prospective benefits.'' Recently, the delays encountered by the US geologic disposal program for HLW, along with advanced in the development of P-T concepts, have led some to propose P-T as a means of reducing the long-term risks from the radioactive wastes that require disposal and thus making it easier to site, license, and build a geologic repository. This study examines and evaluates the effects that introducing P-T would have on the US geologic disposal program

  3. Performances of actinide transmutation based accelerator-driven systems (ADS) at CIEMAT

    International Nuclear Information System (INIS)

    Embid, M.; Cano, D.; Gonzales, E.; Villamarin, D.

    2000-01-01

    The FACET group at CIEMAT is studying the properties and potentialities of several liquid metal-cooled ADS designs for actinide and fission product. transmutation. The main characteristics of these systems are the use of lead or lead-bismuth eutectic as primary coolant, moderator and fuels made by transuranics. The program has two main research lines. The first one is dedicated to the development of concepts, designs, operation models and computer simulation tools characteristics of this kind of systems. The second line includes tlte participation and tlte data analysis of the most advanced experiments in the field and international benchmarks. (authors)

  4. Special scientific programme on use of high energy accelerators for transmutation of actinides and power production

    International Nuclear Information System (INIS)

    1994-09-01

    Various techniques for the transmutation of radioactive waste through the use of high energy accelerators are reviewed and discussed. In particular, the present publication contains presentations on (i) requirements and the technical possibilities for the transmutation of long-lived radionuclides (background paper); (ii) high energy particle accelerators for bulk transformation of elements and energy generation; (iii) the resolution of nuclear energy issues using accelerator-driven technology; (iv) the use of proton accelerators for the transmutation of actinides and power production; (v) the coupling of an accelerator to a subcritical fission reactor (with a view on its potential impact on waste transmutation); (vi) research and development of accelerator-based transmutation technology at JAERI (Japan); and (vii) questions and problems with regard to accelerator-driven nuclear power and transmutation facilities. Refs, figs and tabs

  5. Nuclear transmutation by flux compression

    International Nuclear Information System (INIS)

    Seifritz, W.

    2001-01-01

    A new idea for the transmutation of minor actinides, long (and even short) lived fission products is presented. It is based an the property of neutron flux compression in nuclear (fast and/or thermal) reactors possessing spatially non-stationary critical masses. An advantage factor for the burn-up fluence of the elements to be transmuted in the order of magnitude of 100 and more is obtainable compared with the classical way of transmutation. Three typical examples of such transmuters (a subcritical ringreactor with a rotating reflector, a sub-critical ring reactor with a rotating spallation source, the socalled ''pulsed energy amplifier'', and a fast burn-wave reactor) are presented and analysed with regard to this purpose. (orig.) [de

  6. Transmuted Generalized Inverse Weibull Distribution

    OpenAIRE

    Merovci, Faton; Elbatal, Ibrahim; Ahmed, Alaa

    2013-01-01

    A generalization of the generalized inverse Weibull distribution so-called transmuted generalized inverse Weibull dis- tribution is proposed and studied. We will use the quadratic rank transmutation map (QRTM) in order to generate a flexible family of probability distributions taking generalized inverse Weibull distribution as the base value distribution by introducing a new parameter that would offer more distributional flexibility. Various structural properties including explicit expression...

  7. Romanian concern for advanced fuels development

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2001-01-01

    The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

  8. MA-burners efficiency parameters allowing for the duration of transmutation process

    International Nuclear Information System (INIS)

    Gulevich, A.; Zemskov, E.; Kalugin, A.; Ponomarev, L.; Seliverstov, V.; Seregin, M.

    2010-01-01

    Transmutation of minor actinides (MA) means their transforming into the fission products. Usually, MA-burner's transmutation efficiency is characterized by the static parameters only, such as the number of neutrons absorbed and the rate of MA feeding. However, the proper characterization of MA-burner's efficiency additionally requires the consideration of parameters allowing for the duration of the MA transmutation process. Two parameters of that kind are proposed: a) transmutation time τ - mean time period from the moment a mass of MA is loaded into the burner's fuel cycle to be transmuted to the moment this mass is completely transmuted; b) number of reprocessing cycles n rep - effective number of reprocessing cycles a mass of loaded MA has to undergo before being completely transmuted. Some of MA-burners' types have been analyzed from the point of view of these parameters. It turned out that all of them have the value of parameters too high from the practical point of view. It appears that some new approaches to MA-burner's design have to be used to significantly reduce the value of these parameters in order to make the large-scale MA transmutation process practically reasonable. Some of such approaches are proposed and their potential efficiency is discussed. (authors)

  9. MA-burners efficiency parameters allowing for the duration of transmutation process

    Energy Technology Data Exchange (ETDEWEB)

    Gulevich, A.; Zemskov, E. [Institute of Physics and Power Engineering, Bondarenko Square 1, Obninsk, Kaluga Region 249020 (Russian Federation); Kalugin, A.; Ponomarev, L. [Russian Research Center ' ' Kurchatov Institute' ' Kurchatov Square 1, Moscow 123182 (Russian Federation); Seliverstov, V. [Institute of Theoretical and Experimental Physics ul.B. Cheremushkinskaya 25, Moscow 117259 (Russian Federation); Seregin, M. [Russian Research Institute of Chemical Technology Kashirskoe Shosse 33, Moscow 115230 (Russian Federation)

    2010-07-01

    Transmutation of minor actinides (MA) means their transforming into the fission products. Usually, MA-burner's transmutation efficiency is characterized by the static parameters only, such as the number of neutrons absorbed and the rate of MA feeding. However, the proper characterization of MA-burner's efficiency additionally requires the consideration of parameters allowing for the duration of the MA transmutation process. Two parameters of that kind are proposed: a) transmutation time {tau} - mean time period from the moment a mass of MA is loaded into the burner's fuel cycle to be transmuted to the moment this mass is completely transmuted; b) number of reprocessing cycles n{sub rep} - effective number of reprocessing cycles a mass of loaded MA has to undergo before being completely transmuted. Some of MA-burners' types have been analyzed from the point of view of these parameters. It turned out that all of them have the value of parameters too high from the practical point of view. It appears that some new approaches to MA-burner's design have to be used to significantly reduce the value of these parameters in order to make the large-scale MA transmutation process practically reasonable. Some of such approaches are proposed and their potential efficiency is discussed. (authors)

  10. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  11. Nuclear Waste Separation and Transmutation Research with Special Focus on Russian Transmutation Projects Sponsored by ISTC

    International Nuclear Information System (INIS)

    Conde, Henri; Blomgren, Jan; Olsson, Nils

    2003-03-01

    High-level nuclear reactor waste is made up of relatively few long-lived radioactive species, among them plutonium, that contribute to difficulties with its storage and disposal. Separation of these species from larger waste volumes mainly constituting of uranium (about 95 %) coupled with nuclear incineration to fission products of plutonium and the so called minor actinides (Neptunium, Americium, and Curium) and transmutation of some of the long lived fission products to short lived or stable isotopes represents a viable nuclear waste management strategy to drastically reduce the time and space requirements for a bed-rock repository of the remaining waste. A remarkable increase in the international research and development on partitioning and transmutation has occurred during the recent years. The road-map report published in April 2001 by The European Technical Working Group on ADS for the development of a European demonstration facility for nuclear waste transmutation has high-lighted the ongoing European research and pointed out the need for further research. The road-map has given the different research activities a position in the ultimate goal of producing an ADS demonstrator and is guiding research planning on the national as well as on the EU level. The Advanced Accelerator Application (3A) program in the US, with the long term goals to enhance long term public safety, provide benefits for the repository, reduce proliferation risks and improve prospects for nuclear power has focused the research on nuclear waste transmutation. The reports on the 3A program indicates a change of the US former abandonment position towards reprocessing and fast reactors due to a strong incentive to eliminate the reactor plutonium and to lower the amount of high level reactor waste for the Yucca Mountain repository. The SKB's proposed research and development program for the next 3 years (FUD01) was presented by SKB in September 2001. It is proposed that the research program

  12. The Los Alamos accelerator-driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    Venneri, F.; Williamson, M.A.; Ning, L.

    1997-01-01

    In the past several years, the Los Alamos ADTT program has conducted studies of an innovative technology for solving the nuclear waste problem and building a new generation of safer and non-proliferant nuclear power plants. The ATW concept destroys higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. In this paper special attention is given to the basic design of the ATW Molten Salt concept and the safety perspective. 40 refs., 11 figs

  13. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  14. IFR fuel cycle--pyroprocess development

    International Nuclear Information System (INIS)

    Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

    1992-01-01

    The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

  15. Molybdenum-base cermet fuel development

    International Nuclear Information System (INIS)

    Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

    1987-07-01

    Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

  16. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

    The need to close the fuel cycle of fast reactors makes the development of the cycle installations (fuel fabrication, irradiated assembly conditioning before reprocessing, reprocessing and waste management) especially independent with the development of the reactor. French experience with the integrated cycle over a period of about 25 years, the tonnage of fuels fabricated (more than 100 t of mixed oxides) for the Rapsodie, Phoenix and SuperPhoenix reactors, and the tonnage of reprocessed fuel (nearly 30 t of plutonium fuel) demonstrate the control of the cycle operations. The capacities of the cycle installations in existence and under construction are largely adequate for presents needs, even including a new European EFR reactor. They include the Cadarache fuel fabrication complex, the La Hague UP2-800 reprocessing plant, and the Marcoule pilot facility. Short- and medium-term R and D programs are connected with fuel developments, with the primary objective of very high burnups. For the longer term and for a specific plant to reprocess fast reactor fuels, the programs could concern new fabrication and reprocessing systems and the study of the consequences of the reduction in fuel out-of-core time

  17. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  18. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  19. Developments in fossil fuel electricity generation

    International Nuclear Information System (INIS)

    Williams, A.; Argiri, M.

    1993-01-01

    A major part of the world's electricity is generated by the combustion of fossil fuels, and there is a significant environmental impact due to the production of fossil fuels and their combustion. Coal is responsible for 63% of the electricity generated from fossil fuels; natural gas accounts for about 20% and fuel oils for 17%. Because of developments in supply and improvements in generating efficiencies there is apparently a considerable shift towards a greater use of natural gas, and by the year 2000 it could provide 25% of the world electricity output. At the same time the amount of fuel oil burned will have decreased. The means to minimize the environmental impact of the use of fossil fuels, particularly coal, in electricity production are considered, together with the methods of emission control. Cleaner coal technologies, which include fluidized bed combustion and an integrated gasification combined cycle (IGCC), can reduce the emissions of NO x , SO 2 and CO 2 . (author)

  20. The conditions of gaseous fuels development

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    Face to the actual situation of petrol and gas oil in France, the situation of gaseous fuels appears to be rather modest. However, the aim of gaseous fuels is not to totally supersede the liquid fuels. Such a situation would imply a complete overturn which has not been seriously considered yet. This short paper describes the essential conditions to promote the wider use of gaseous fuels: the intervention of public authorities to adopt a more advantageous tax policy in agreement with the ''Clean Air''law project, a suitable distribution network for gaseous fuels, a choice of vehicles consistent with the urban demand, the development of transformation kits of quality and of dual-fuel vehicles by the car manufacturers. (J.S.)

  1. Strategic Partnerships in Fuel Cell Development

    Science.gov (United States)

    Diab, Dorey

    2006-01-01

    This article describes how forming strategic alliances with universities, emerging technology companies, the state of Ohio, the federal government, and the National Science Foundation, has enabled Stark State College to develop a $5.5 million Fuel Cell Prototyping Center and establish a Fuel Cell Technology program to promote economic development…

  2. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2013-06-28

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.

  3. The fuel cell; development and possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Van Rijnsoever, J.W.M.

    Activities on fuel cells and fuel cell development in the USA and Japan are surveyed. Possibilities for large scale application are mentioned. Attention is given to efficiency and environmental aspects. There are no problems about hazardous emissions. Besides electric power some heat is generated, which is not always a disadvantage. In many cases both are useful products. (A.V.)

  4. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    Hardy, D.G.; Wood, J.C.; Bain, A.S.

    1978-12-01

    The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

  5. Development of nuclear fuel cycle technology

    International Nuclear Information System (INIS)

    Kawahara, Akira; Sugimoto, Yoshikazu; Shibata, Satoshi; Ikeda, Takashi; Suzuki, Kazumichi; Miki, Atsushi.

    1990-01-01

    In order to establish the stable supply of nuclear fuel as an important energy source, Hitachi ltd. has advanced the technical development aiming at the heightening of reliability, the increase of capacity, upgrading and the heightening of performance of the facilities related to nuclear fuel cycle. As for fuel reprocessing, Japan Nuclear Fuel Service Ltd. is promoting the construction of a commercial fuel reprocessing plant which is the first in Japan. The verification of the process performance, the ensuring of high reliability accompanying large capacity and the technical development for recovering effective resources from spent fuel are advanced. Moreover, as for uranium enrichment, Laser Enrichment Technology Research Association was founded mainly by electric power companies, and the development of the next generation enrichment technology using laser is promoted. The development of spent fuel reprocessing technology, the development of the basic technology of atomic process laser enrichment and so on are reported. In addition to the above technologies recently developed by Hitachi Ltd., the technology of reducing harm and solidification of radioactive wastes, the molecular process laser enrichment and others are developed. (K.I.)

  6. Comparative analysis of sub-critical transmutation reactor concepts

    International Nuclear Information System (INIS)

    Chang, S. H.

    1997-01-01

    The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

  7. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs, 90 Sr, 129 I, 99 Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm 3 are required for a practical fusion-based fission product transmutation system

  8. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    International Nuclear Information System (INIS)

    Dahlfors, Marcus

    2003-05-01

    This text is a summary of two studies the author has performed within the field of 3-D Monte Carlo calculations of Accelerator Driven Systems (ADS) for transmutation of nuclear waste. The simulations were carried out with the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN. The concept studied is ANSALDOs 80 MWth Energy Amplifier Demonstration Facility based on classical MOX-fuel technology and on molten Lead-Bismuth Eutectic cooling. A review of neutron cross section sensitivity in numerical calculations of an ADS and a comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ADS are presented

  9. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dahlfors, Marcus

    2003-05-01

    This text is a summary of two studies the author has performed within the field of 3-D Monte Carlo calculations of Accelerator Driven Systems (ADS) for transmutation of nuclear waste. The simulations were carried out with the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN. The concept studied is ANSALDOs 80 MWth Energy Amplifier Demonstration Facility based on classical MOX-fuel technology and on molten Lead-Bismuth Eutectic cooling. A review of neutron cross section sensitivity in numerical calculations of an ADS and a comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ADS are presented.

  10. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  11. Development of fuel and energy storage technologies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Development of fuel cell power plants is intended of high-efficiency power generation using such fuels with less air pollution as natural gas, methanol and coal gas. The closest to commercialization is phosphoric acid fuel cells, and the high in efficiency and rich in fuel diversity is molten carbonate fuel cells. The development is intended to cover a wide scope from solid electrolyte fuel cells to solid polymer electrolyte fuel cells. For new battery power storage systems, development is focused on discrete battery energy storage technologies of fixed type and mobile type (such as electric vehicles). The ceramic gas turbine technology development is purposed for improving thermal efficiency and reducing pollutants. Small-scale gas turbines for cogeneration will also be developed. Development of superconduction power application technologies is intended to serve for efficient and stable power supply by dealing with capacity increase and increase in power distribution distance due to increase in power demand. In the operations to improve the spread and general promotion systems for electric vehicles, load leveling is expected by utilizing and storing nighttime electric power. Descriptions are given also on economical city systems which utilize wide-area energy. 30 figs., 7 tabs.

  12. Feasibility analysis of constant TRU feeding in waste transmutation system using accelerator-driven subcritical system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kun Jai; Cho, Nam Zin; Jo, Chang Keun; Park, Chang Je; Kim, Do Sam; Park, Jeong Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    It is probable that the issue of nuclear spent fuel and high-level waste can have negative impact on the future expansion of nuclear power programs. Accelerator-driven nuclear waste transmutation with constant composition TRU feeding which satisfies non-proliferation condition will help establish the long-range nuclear waste disposal strategy. In this study, current status of accelerator-driven transmutation of waste technology, and feasibility analysis of constant composition TRU feeding system were investigated. We ascertained that solid system using constant composition TRU is feasible with the the capability of transmutation. (author). 13 refs., 53 figs., 20 tabs.

  13. Transmutation of Americium in Light and Heavy Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

    2009-06-15

    There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

  14. Progress in transmutation targets from Efttra

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Warin, D.; Bonnerot, J.M.; Garzenne, C.; Scaffidi-Argentina, F.; Maschek, W.; Schram, R.; Klaassen, F.

    2007-01-01

    Since 15 years, the EFTTRA partners have organised programmes to demonstrate the feasibility of the transmutation of americium in uranium-free targets. In the related transmutation scenario, the targets are introduced in a thermal neutron zone of a fast reactor, to maximize the efficiency of transmutation. Amongst these programmes, those carried out in the HFR reactor in Petten have led to important conclusions and are still at the core of the research in that field. The analysis of the EFTTRA T4 and T4bis irradiation experiments, carried out with targets of MgAl 2 O 4 +11 wt% 241 Am, showed that the release/trapping of helium is the key issue for target design, and also demonstrated a lack of technical benefits of this material, due to a unsatisfactory in-pile behaviour in terms of irradiation damage and chemical stability. A new irradiation experiment called HELIOS is currently under fabrication and will be carried out in HFR. The in-pile behaviour of U-free fuels and targets such as (Am,Zr)O 2 , (Pu,Am,Zr)O 2 , CERCER (MgO) or CERMET (Mo) will be examined. The irradiation temperature will be high enough in some of the pins to be able to tune the release of a significant fraction of helium produced so that the material swelling can be minimized as much as reasonably possible. The HELIOS irradiation experiment is planned to be carried out in the HFR core and shall last 300 full power days starting in 2007. (authors)

  15. Safety analysis of DUPIC fuel development facility

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

    2001-01-01

    Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

  16. Advanced Fuels Campaign FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  17. Comparative studies of CERCER and CERMET fuels for EFIT from the viewpoint of core performance and safety

    International Nuclear Information System (INIS)

    Chen, X.N.; Rineiski, A.; Maschek, W.; Liu, P.; Boccaccini, C.M.; Sobolev, V.; Delage, F.; Rimpault, G.

    2011-01-01

    The European Facility for Industrial Transmutation (EFIT) has been developed within the 6. EU Framework by the EUROTRANS Program, aiming at a generic conceptual design of an accelerator driven transmuter. This paper deals with assessments of EFIT cores with CERCER and CERMET fuels from the viewpoint of core performance and safety. The conclusive remarks can be drawn as follows. Because of its much better thermal conductivity, the CERMET core can be designed by using thicker pins, so that it has the same or even better transmutation performance compared to the CERCER core. Both CERCER and CERMET fuels fulfill safety requirements. Moreover the CERMET fuel has higher fuel safety margins than the CERCER one. Preliminary analyses show that the CERMET total core power can be further increased by 50% at least without exceeding fuel and clad temperature limits. (authors)

  18. Development of PHWR fuel fabrication in Korea

    International Nuclear Information System (INIS)

    Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.

    1988-01-01

    Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel

  19. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    International Nuclear Information System (INIS)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr; Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr; Afonichkin, Valery

    2008-01-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF 3 ) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the 7 Li,Na,Be/F and 7 Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  20. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr [Russian Research Center - Kurchatov Institute, Kurchatov sq. 1, Moscow, RF, 123182 (Russian Federation); Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr [Russian Federal Nuclear Center - Institute of Technical Physics, Snezhinsk (Russian Federation); Afonichkin, Valery [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    2008-07-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF{sub 3}) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the {sup 7}Li,Na,Be/F and {sup 7}Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  1. Fuel Cell Development and Test Laboratory | Energy Systems Integration

    Science.gov (United States)

    Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

  2. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  3. VVER fuel cycle development at Slovakia

    International Nuclear Information System (INIS)

    Darilek, P.; Chrapiak, V.; Majerik, J.

    1995-01-01

    Four VVER-440 units are now under exploitation at Bohunice-site in Slovakia. Fuel cycle development of Unit No.3 and No.4 (type 213) is discussed and compared with equilibrium cycles in this paper. (author)

  4. Development of the external cooling device of increase the productivity of neutron-transmutation-doped silicon semiconductor (NTD-Si) (Joint research)

    International Nuclear Information System (INIS)

    Hirose, Akira; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi; Kameyama, Iwao; Aizawa, Ryouji; Kikuchi, Naoyuki

    2007-01-01

    Neutron-Transmutation-Doped Silicon Semiconductor (hereinafter referred as 'NTD-Si') is the best semiconductor for the power device. The needs of NTD-Si increase recently in proportion to the popularization of hybrid-cars. A fission research reactor, which is a steady state neutron source, is being expected as the best device to meet the needs. So far, we have reconsidered the existing approach which is employed for NTD-Si production works at the research reactors JRR-3, JRR-4 and JMTR of JAEA so as to meet the needs. As one of the effective measures, we found out that the productivity can be increased by incorporating a new device to cool down radioactivity of irradiated silicon ingots at the place outside the main stream from the loading of silicon ingots to the withdrawal of irradiated ingots to the existing JRR-3 Uniformity Irradiation System. Consequently, we developed and installed the device (hereinafter referred as 'external cooling device'). After an ingot was irradiated once, it is turned over manually and irradiated again in order to irradiate the ingot uniformly. With the conventional system, it was necessary to wait the radioactivity of ingot decrease less than the permissible level with holding the ingot in the irradiation equipment. It was effective to shorten the waiting period by using an external cooling device for production increase of NTD-Si. It is expected that the productivity of NTD-Si will be increased by using the external cooling device. This report mentions the design of the external cooling device and verification between its design specifications and the performance of the device completed. (author)

  5. Dual neutral particle induced transmutation in CINDER2008

    Energy Technology Data Exchange (ETDEWEB)

    Martin, W.J., E-mail: wjmarti@sandia.gov [Sandia National Laboratories, Albuquerque, NM 87185 (United States); University of New Mexico, Albuquerque, NM 87131 (United States); Oliveira, C.R.E. de; Hecht, A.A. [University of New Mexico, Albuquerque, NM 87131 (United States)

    2014-12-11

    Although nuclear transmutation methods for fission have existed for decades, the focus has been on neutron-induced reactions. Recent novel concepts have sought to use both neutrons and photons for purposes such as active interrogation of cargo to detect the smuggling of highly enriched uranium, a concept that would require modeling the transmutation caused by both incident particles. As photonuclear transmutation has yet to be modeled alongside neutron-induced transmutation in a production code, new methods need to be developed. The CINDER2008 nuclear transmutation code from Los Alamos National Laboratory is extended from neutron applications to dual neutral particle applications, allowing both neutron- and photon-induced reactions for this modeling with a focus on fission. Following standard reaction modeling, the induced fission reaction is understood as a two-part reaction, with an entrance channel to the excited compound nucleus, and an exit channel from the excited compound nucleus to the fission fragmentation. Because photofission yield data—the exit channel from the compound nucleus—are sparse, neutron fission yield data are used in this work. With a different compound nucleus and excitation, the translation to the excited compound state is modified, as appropriate. A verification and validation of these methods and data has been performed. This has shown that the translation of neutron-induced fission product yield sets, and their use in photonuclear applications, is appropriate, and that the code has been extended correctly. - Highlights: • The CINDER2008 transmutation code was modified to include photon-induced transmutation tracking. • A photonuclear interaction library was created to allow CINDER2008 to track photonuclear interactions. • Photofission product yield data sets were created using fission physics similarities with neutron-induced fission.

  6. Development of a diesel substitute fuel

    Energy Technology Data Exchange (ETDEWEB)

    Reiter, Anton; Mair-Zelenka, Philipp [Graz Univ. of Technology (Austria). Inst. of Chemical Engineering and Environmental Technology; Zeymer, Marc [OMV Refining and Marketing GmbH, Vienna (Austria). MRDI-D Product Development and Innovation

    2013-06-01

    Substitute fuels composed of few real chemical compounds are an alternative characterisation approach for conventional fuels as opposed to the traditional pseudo-component method. With the algorithm proposed in this paper the generation of such substitutes will be facilitated and well-established thermodynamic methods can be applied for physical property-data prediction. Based on some quality criteria like true boiling-point curve, liquid density, C/H ratio, or cloud point of a target fuel a surrogate which meets these properties is determined by fitting its composition. The application and capabilities of the algorithm developed are demonstrated by means of an exemplary diesel substitute fuel. The substitute mixture obtained can be generated and used for evaluation of property-prediction methods. Furthermore this approach can help to understand the effects of mixing fossil fuels with biogenic compounds. (orig.)

  7. Incentives and recent proposals for partitioning and transmutation in the United States

    International Nuclear Information System (INIS)

    Donovan, T.J.

    1995-05-01

    Partitioning and transmutation (P-T) is perhaps the most elegant means of high level waste disposal. Currently, the cost of fuel obtained from reprocessing spent fuel exceeds the cost of fuel obtained by mining. This has resulted in the once through fuel cycle dominating the US nuclear industry. Despite this fact P-T continues to be examined and debated by the US as well as abroad. The US first seriously considered P-T between approximately 1976 and 1982 but rejected the concept in favor of reprocessing. More recently, since about 1989, as a result of the once through fuel cycle and the growing problems of waste disposal, studies concerning P-T have resumed. This essay will seek to outline the incentives and goals of partitioning and transmutation as it would apply to the disposal of spent fuel in the US. Recent proposals by various US national laboratories for implementing partitioning and transmutation as a high level waste management and disposal device will also be discussed. The review will seek to examine the technical concepts utilized in each of the proposals and their feasibility. The major focus of this essay will be the transmutation methods themselves, while the partitioning methods will be discussed only briefly. This is because of the fact that partitioning methods fall under reprocessing as an already fairly well established and accepted technology while feasible methods for transmutation are still being advanced

  8. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    Stahl, D.

    1993-01-01

    In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel-development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

  9. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  10. 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