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Sample records for actinides transmutation fuel

  1. Transmutation of minor actinide using thorium fueled BWR core

    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 6th 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

  2. Reducing the impact of used fuel by transmuting actinides in a CANDU reactor

    With world stockpiles of used nuclear fuel increasing, the need to address the long term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes in CANDU reactors to reduce the decay heat period. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle facilitates the fabrication and handling of active fuels. Online refueling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation in CANDU reactors, including both recent and past activities. The transmutation schemes that are presented reflect several different partitioning schemes and include both homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. (author)

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

    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.

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

    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)

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

    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)

  6. Vaporisation of candidate nuclear fuels and targets for transmutation of minor actinides

    Gotcu-Freis, P., E-mail: p.gotcu@tudelft.nl [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Delft University of Technology, Faculty of Applied Sciences, Mekelweg 15, 2629 JB Delft (Netherlands); Hiernaut, J.-P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Colle, J.-Y., E-mail: jean-yves.colle@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Naestren, C.; Carretero, A. Fernandez; Konings, R.J.M. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany)

    2011-04-15

    The thermal stability and high temperature behaviour of candidate fuels and targets for transmutation of minor actinides has been investigated. Zirconia-based solid solution, MgO-based CERCER and molybdenum-based CERMET fuels containing Am and/or Pu in various concentrations were heated up to 2700 K in a Knudsen cell coupled with a quadrupole mass spectrometer, to measure their vapour pressure and vapour composition. The results reveal that the vaporisation of the actinides from the samples is not only determined by the thermodynamics of the system but is also related to the dynamic evolution of multi-component mixtures with complex composition or microstructure.

  7. Vaporisation of candidate nuclear fuels and targets for transmutation of minor actinides

    The thermal stability and high temperature behaviour of candidate fuels and targets for transmutation of minor actinides has been investigated. Zirconia-based solid solution, MgO-based CERCER and molybdenum-based CERMET fuels containing Am and/or Pu in various concentrations were heated up to 2700 K in a Knudsen cell coupled with a quadrupole mass spectrometer, to measure their vapour pressure and vapour composition. The results reveal that the vaporisation of the actinides from the samples is not only determined by the thermodynamics of the system but is also related to the dynamic evolution of multi-component mixtures with complex composition or microstructure.

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

    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)

  9. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    Washington, Jeremy A.

    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 near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

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

    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

  11. Minor Actinide Transmutation Performance in Fast Reactor Metal Fuel. Isotope Ratio Change in Actinide Elements upon Low-Burnup Irradiation

    Metal fuel alloys containing 5 wt% or less minor actinide (MA) and rare earth (RE) were irradiated in the fast reactor Phénix. After nondestructive postirradiation tests, a chemical analysis of the alloys irradiated for 120 effective full power days was carried out by the inductively coupled plasma - mass spectrometry (ICP-MS) technique. From the analysis results, it was determined that the discharged burnups of U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, and U-19Pu-10Zr-5MA were 2.17, 2.48, and 2.36 at.%, respectively. Actinide isotope ratio analyses before and after the irradiation experiment revealed that Pu, Am, and Cm nuclides added to U-Pu-Zr alloy and irradiated up to 2.0 - 2.5 at.% burnups in a fast reactor are transmuted properly as predicted by ORIGEN2 calculations. (author)

  12. Actinide and fission product partitioning and transmutation

    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.)

  13. Actinide and fission product partitioning and transmutation

    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.)

  14. Actinide and fission product separation and transmutation

    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.)

  15. Actinide and fission product separation and transmutation

    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.)

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

    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 ZrO2 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 ZrO2. 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 would require an

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

    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

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

    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

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

    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

  20. Actinide and fission product partitioning and transmutation

    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.)

  1. Actinide and fission product partitioning and transmutation

    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. Actinide transmutation in nuclear reactors

    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)

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

    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 keff 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: 241Pu, 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

  4. Use of fast reactors for actinide transmutation

    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

  5. Minor actinide transmutation on PWR burnable poison rods

    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 keff 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 keff markedly. The PWR keff 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

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

    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)

  7. High flux transmutation of fission products and actinides

    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

  8. A thermodynamic study of actinide oxide targets/fuels for americium transmutation

    A thermodynamic study was performed on the systems Am-O, AmOx-MgO, AmOx-MgAl2O4, Pu-Mg-O and U-Mg-O. Both experimental work (X-ray analyses, oxygen potential measurements etc.) and calculations on the phase diagrams involved were made. The reaction between americium oxide and spinel is expected to form the compound AmAlO3. Isothermal sections have been calculated for AmOx-(MgO, Al2O3), Pu-Mg-O and U-Mg-O at 2000 K using the software package ''Thermo-Calc''. Thermodynamic equilibrium data were used to predict the behaviour of actinide oxides in a reactor. The implication of the results for the technological application is discussed, with emphasis on the effects of the high oxygen potential of AmO2 as compared to the conventional fuel, i.e. UO2. (author)

  9. PC code STAR. Show transmutation of actinides in reactors

    A program is made named STAR (acronym for Show Transmutation of Actinides in Reactors), which solves analytically the differential equations describing buildup and removal (by decay and transmutation) of nuclides irradiated in a constant neutron flux. The model and algorithm according to which STAR solves the differential equations are explained. Also a short description of the data library is given. STAR is validated with the ORIGEN-S fuel depletion code and runs on IBM compatible PCs and DEC alpha workstations. (orig.)

  10. Actinide transmutation in nuclear reactors

    This report has also been published as a PhD thesis. It discusses the reduction of the transuranics part of nuclear waste. Requirements and criteria for efficient burning of transuranics are developed. It is found that a large reduction of transuranics produced per unit of energy is possible when the losses in reprocessing are small and when special transuranics burner reactors are used at the end of the nuclear era to reduce the transuranics inventory. Two special burner reactors have been studied in this thesis. In chapter 3, the Advanced Liquid Metal Reactor is discussed. A method has been developed to optimize the burning capability while complying to constraints imposed on the design for safety, reliability, and economics. An oxide fueled and metallic fueled ALMR have been compared for safety and transuranics burning. Concluded is that the burning capability is the same, but that the higher thermal conductivity of the metallic fuel has a positive effect on safety. In search for a more effective waste transmuter, a modified Molten Salt Reactor was designed for this study. The continuous refueling capability and the molten salt fuel make a safe design possible without uranium as fuel. A four times faster reduction of the transuranics is possible with this reactor type. The amount of transuranics can be halved every 10 years. The most important conclusion of this work is that it is of utmost importance in the study of waste transmutation that a high burning is obtained with a safe design. In future work, safety should be the highest priority in the design process of burner reactors. (orig.)

  11. Potential Benefits and Impacts of Advanced Nuclear Fuel Cycles with Actinide Partitioning and Transmutation

    This report provides a comparative analysis of different studies performed to assess the potential impact of partitioning and transmutation (P and T) on different types of geological repositories for radioactive waste in various licensing and regulatory environments. Criteria, metrics and impact measures have been analysed and compared with the goal of providing an objective comparison of the state of the art to help shape decisions on options for future advanced fuel cycles. P and T allows a reduction of the inventory of the emplaced materials which can have a significant impact on the repository. Such a reduction can also make the uncertainty about repository performance less important both during normal evolution and in the case of disruptive scenarios. While P and T will never replace the need for waste repositories, it has the potential to significantly improve public perception regarding the ability to effectively manage radioactive waste by largely reducing the transuranic (TRU) waste masses to be stored and, consequently, to improve public acceptance of the geological repositories. Both issues are important for the future sustainability of nuclear power

  12. Actinide transmutation in the advanced liquid metal reactor (ALMR)

    The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) sponsored fast reactor design based on the Power Reactor, Innovative Small Module (PRISM) concept originated by General Electric. The current reference design is a 471 MWt modular reactor loaded with ternary metal fuel. This paper discusses actinide transmutation core designs that fit the design envelope of the ALMR and utilize spent LWR fuel as startup material and makeup. Actinide transmutation may be accomplished in the ALMR by using either a breeding or burning configuration. Lifetime actinide mass consumption is calculated as well as changes in consumption behaviour throughout the lifetime of the reactor. Impacts on system operational and safety performance are evaluated in a preliminary fashion. (author). 3 refs, 6 figs, 3 tabs

  13. Actinide and fission product separation and transmutation

    NONE

    1991-07-01

    The first international information exchange meeting on actinide and fission product separation and transmutation, took place in Mito in Japan, on 6-8 November 1990. It starts with a number of general overview papers to give us some broad perspectives. Following that it takes a look at some basic facts about physics and about the quantities of materials it is talking about. Then it proceeds to some specific aspects of partitioning, starting with evolution from today commercially applied processes and going on to other possibilities. At the end of the third session it takes a look at the significance of partitioning and transmutation of actinides before it embarks on two sessions on transmutation, first in reactors and second in accelerators. The last session is designed to throw back into the discussion the main points which need to be looked at when considering future work in this area. (A.L.B.)

  14. Actinide and fission product separation and transmutation

    The first international information exchange meeting on actinide and fission product separation and transmutation, took place in Mito in Japan, on 6-8 November 1990. It starts with a number of general overview papers to give us some broad perspectives. Following that it takes a look at some basic facts about physics and about the quantities of materials it is talking about. Then it proceeds to some specific aspects of partitioning, starting with evolution from today commercially applied processes and going on to other possibilities. At the end of the third session it takes a look at the significance of partitioning and transmutation of actinides before it embarks on two sessions on transmutation, first in reactors and second in accelerators. The last session is designed to throw back into the discussion the main points which need to be looked at when considering future work in this area. (A.L.B.)

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

    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 99Tc and 129I. 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

  16. Evaluation of actinide partitioning and transmutation

    After a few centuries of radioactive decay the long-lived actinides, the elements of atomic numbers 89-103, may constitute the main potential radiological health hazard in nuclear wastes. This is because all but a very few fission products (principally technetium-99 and iodine-129) have by then undergone radioactive decay to insignificant levels, leaving the actinides as the principal radionuclides remaining. It was therefore at first sight an attractive concept to recycle the actinides to nuclear reactors, so as to eliminate them by nuclear fission. Thus, investigations of the feasibility and potential benefits and hazards of the concept of 'actinide partitioning and transmutation' were started in numerous countries in the mid-1970s. This final report summarizes the results and conclusions of technical studies performed in connection with a four-year IAEA Co-ordinated Research Programme, started in 1976, on the ''Environmental Evaluation and Hazard Assessment of the Separation of Actinides from Nuclear Wastes followed by either Transmutation or Separate Disposal''. Although many related studies are still continuing, e.g. on waste disposal, long-term safety assessments, and waste actinide management (particularly for low and intermediate-level wastes), some firm conclusions on the overall concept were drawn by the programme participants, which are reflected in this report

  17. Minor actinides transmutation strategies in sodium fast reactors

    In minor actinides transmutation strategies for fast spectrum reactors, different possibilities regarding the core loading are considered. We study both homogeneous patterns (HOM) with various minor actinides (MA) content values and heterogeneous schemes (HET) with higher percentages of MA (Np, Am and Cm) at the periphery of reactor. We analyze the capability of transmutation of each design and the reactivity coefficients such as the Doppler constant, void worth and the fraction of delayed neutrons. The EVOLCODE2 code is the computational tool used in this study. It is based on MCNPX and ORIGEN/ACAB codes and allows carrying out burn-up calculations to get the isotopic evolution of fuel composition. Among the three strategies studied (HOM 2.5 %, HOM 4% and HET 20 %) for a possible design of a Sodium Cooled Fast Breeder Reactor, the one with better transmutation results is the HOM 4%, which shows higher absolute and relative values (12 Kg-MA/TWe, 29% respectively). Concerning transmutation in blankets with 20% MA content, results show a very little or no transmutation values when considering Np, Am and Cm together, though a positive small value for Np and Am is obtained

  18. Why Faster is Better : On Minor Actinide Transmutation in Hard Neutron

    Westlén, Daniel

    2007-01-01

    In this thesis, options for efficient transmutation of transuranium elements are discussed. The focus is on plutonium, americium and curium mainly because of their long-term contribution to the radiotoxicity of spent nuclear fuel. Two innovative helium-cooled core designs are proposed, dedicated to the transmutation of actinides. The performance of the more promising of the two is studied in realistic transient fuel cycle scenarios. During the 1150 day irradiation cycle, a minor actinide cons...

  19. Chemical separation and nuclear transmutation of by-product actinides

    The paper presents the most important results and conclusions of the assessment studies carried out by the Joint Research Centre-Ispra and by other organizations on the advanced waste disposal strategy based on chemical separation of By-product Actinides (BPA's) from high level liquid waste (HLLW) and their transmutation in nuclear reactors. The technological developments required for the implementation of this strategy have been identified: they concern mainly fuel reprocessing, BPA recovery from all important waste streams and fuel refabrication. After consideration of different strategies for BPA transmutation, the homogeneous recycling in FBR's appears to be most suitable due to its transmutation rate and the compatibility of BPA's with its fuel cycle. The fuel cycle with transmutation has been compared with an advanced reference fuel cycle on the basis of costs and risks. The large effort required for the development and implementation of this new fuel cycle, the increased costs operating the fuel cycle compared with the marginal benefits in the long-term risk of geological disposal, make this strategy not very attractive

  20. FCRD Transmutation Fuels Handbook 2015

    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

  1. FCRD Transmutation Fuels Handbook 2015

    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

  2. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    The irradiation of Th232 breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U238. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

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

    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

  4. Sensitivity analysis of minor actinides transmutation to physical and technological parameters

    Kooyman Timothée

    2015-01-01

    Full Text Available Minor actinides transmutation is one of the three main axis defined by the 2006 French law for management of nuclear waste, along with long-term storage and use of a deep geological repository. Transmutation options for critical systems can be divided in two different approaches: (a homogeneous transmutation, in which minor actinides are mixed with the fuel. This exhibits the drawback of “polluting” the entire fuel cycle with minor actinides and also has an important impact on core reactivity coefficients such as Doppler Effect or sodium void worth for fast reactors when the minor actinides fraction increases above 3 to 5% depending on the core; (b heterogeneous transmutation, in which minor actinides are inserted into transmutation targets which can be located in the center or in the periphery of the core. This presents the advantage of decoupling the management of the minor actinides from the conventional fuel and not impacting the core reactivity coefficients. In both cases, the design and analyses of potential transmutation systems have been carried out in the frame of Gen IV fast reactor using a “perturbation” approach in which nominal power reactor parameters are modified to accommodate the loading of minor actinides. However, when designing such a transmutation strategy, parameters from all steps of the fuel cycle must be taken into account, such as spent fuel heat load, gamma or neutron sources or fabrication feasibility. Considering a multi-recycling strategy of minor actinides, an analysis of relevant estimators necessary to fully analyze a transmutation strategy has been performed in this work and a sensitivity analysis of these estimators to a broad choice of reactors and fuel cycle parameters has been carried out. No threshold or percolation effects were observed. Saturation of transmutation rate with regards to several parameters has been observed, namely the minor actinides volume fraction and the irradiation time

  5. Status report on actinide and fission product transmutation studies

    The management of radioactive waste is one of the key issues in today's political and public discussions on nuclear energy. One of the fields that looks into the future possibilities of nuclear technology is the neutronic transmutation of actinides and of some most important fission products. Studies on transmutation of actinides are carried out in various countries and at an international level. This status report which gives an up-to-date general overview of current and planned research on transmutation of actinides and fission products in non-OECD countries, has been prepared by a Technical Committee meeting organized by the IAEA in September 1995. 168 refs, 16 figs, 34 tabs

  6. Calculations of the actinide transmutation with HELIOS for fuels of light water reactors; Calculos de la transmutacion de actinidos con HELIOS para combustibles de reactores de agua ligera

    Francois L, J.L.; Guzman A, J.R. [UNAM-FI, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jlfl@fi-b.unam.mx

    2006-07-01

    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{sub 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% {delta}k/k, which was also among other participants. For the isotopic concentrations: {sup 241}Pu, {sup 242} Pu and {sup 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 {sup 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

  7. Assessment of Partitioning Processes for Transmutation of Actinides

    To obtain public acceptance of future nuclear fuel cycle technology, new and innovative concepts must overcome the present concerns with respect to both environmental compliance and proliferation of fissile materials. Both these concerns can be addressed through the multiple recycling of all transuranic elements (TRUs) in fast neutron reactor. This is only possible through a process known as partitioning and transmutation scheme (P and T) as this scheme is expected to reduce the long term radio-toxicity as well as the radiogenic heat production of the nuclear waste. Proliferation resistance of separated plutonium could further be enhanced by mixing with self-generated minor actinides. In addition, P and T scheme is expected to extend the nuclear fuel resources on earth about 100 times because of the recycle and reuse of fissile actinides. Several Member States are actively pursuing the research in the field of P and T and consequently several IAEA publications have addressed this topic. The present coordinated research project (CRP) focuses on the potentials in minimizing the residual TRU inventories of the discharged nuclear waste and in enhancing the proliferation resistance of the future civil nuclear fuel cycle. Partitioning approaches can be grouped into aqueous- (hydrometallurgical) and pyroprocesses. Several aqueous processes based on sequential separation of actinides from spent nuclear fuel have been developed and tested at pilot plant scale. In view of the proliferation resistance of the intermediate and final products of a P and T scheme, a group separation of all actinides together is preferable. The present CRP has gathered experts from different organisations and institutes actively involved in developing P and T scheme as mentioned in the list of contributors and also taken into consideration the studies underway in France and the UK. The scientific objectives of the CRP are: To minimize the environmental impact of actinides in the waste stream; To

  8. Minor actinides partitioning and transmutation technology in France

    The global energy context pleads in favour of a sustainable development of nuclear energy. It is a technology with a future 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. From the start, the CEA has devoted considerable effort to management of the back end of the cycle. It furnished the process and techniques used in the La Hague facility to extract the re-usable materials, uranium and plutonium, and condition the resulting waste. Towards the end of the 1960's, it developed the process of vitrification for highly active waste that has become the world reference. French industry was responsible for the introduction of standard international practices with respect to waste conditioned during the processing of spent fuels. The specifications for the packages are approved by more than ten countries across the world.The law of 1991 specifically gave new momentum to the research into waste by requesting exploration not only into deep geological storage repositories, but also into reducing the quantity and toxicity of the long-life radioactive elements present in the waste by separation and transmutation and studying their conditioning for long-term disposal. All of the above-mentioned research has been conducted in close collaboration with partners from industry (ANDRA, EDF, COGEMA, FRAMATOME), with the CNRS (France) and various universities.A review of the situation ten years on indicates a number of significant results that have changed the prospects for nuclear waste management. The paper will focus on separation and transmutation R and D programme and main results at CEA. Over the past 10 years the CEA has been conducting a massive research programme on enhanced separation, supported by broad international co-operation. This year, 2001, saw some vital progress. Based on

  9. Scenarios for Minor Actinides Transmutation in the Frame of the French Act for Waste Management

    In the frame of the French Act for waste management, options of minor actinides (MA) transmutation are studied, based on a scenario of a 60 GWe SFR fleet deployment from 2040 to 2100. The advantages and drawbacks of different transmutation options are evaluated. The transmutation of all MA or only of americium is considered, in homogeneous mode (MA bearing fuel in all the core) or in heterogeneous mode (MA bearing radial blankets). Scenarios have been optimized to limit the impacts of MA transmutation on fuel cycle, with a reduction of the initial MA content in core in homogeneous mode to mitigate the effect on reactivity coefficients and a reduction of the fuel decay heat for transportation in heterogeneous mode. The sensitivity of results to the SFR core design is evaluated by considering a homogeneous core (SFR V2B) or a new heterogeneous core with a significant gain on sodium void effect (CFV). (author)

  10. The nuclear waste issue: towards an assessment of the partitioning and transmutation of actinides

    First of all, this paper describes recent regulatory, scientific and technical developments in France concerning the management of high level, long-lived radioactive waste. In this context, which culminated in parliament's adoption of the radioactive waste management bill, it analyses, from an industrial viewpoint, the motivation for separating and transmuting minor actinides, as well as the technical and economic consequences for waste management. The new law underlines the need to reduce the harmfulness of radioactive waste and research into separation and transmutation is continuing with a view to 'an assessment in 2012 of the industrial perspectives for these technologies' that makes allowance for the developments made with new reactors. It is admitted that the waste will not be eradicated by transmutation and therefore that the geological disposal is inevitable. But destroying the waste, however partially, is worthy of consideration if it helps to simplify disposal and reduce costs. It may also merely reduce safety assessment uncertainty or increase calculation margins. The results of previous studies have shown that research has to focus on the minor actinides. On the one hand, they account for practically all the radiotoxicity of waste and on the other hand, with the exception of two fission products with relatively short half-lives for which suitable interim storage is to be envisaged, they are the main contributors to the thermal load. We therefore examine the possible consequences of management through actinide separation and transmutation on the fuel cycle as a whole, and the advantages to be gained for disposal or release. For example, americium makes a considerable contribution to the thermal load and transmutation would appear possible. Conversely, curium is known to be difficult to transmute and would complicate target or fuel fabrication operations. The potential savings to be made in disposal costs should also make allowance for future optimization

  11. Evaluation of actinide partitioning and transmutation in light-water reactors

    Advanced Fuel Cycle Initiative (AFCI) studies were made to evaluate the feasibility of multicycle transmutation of plutonium and the minor actinides (MAs) in light-water reactors (LWRs). Results showed that significant repository benefits, cost reductions, proliferation resistance, and effective use of facilities can be obtained. Key advantages are shown to be made possible by processing 30-year-decayed spent fuel rather than the more traditional 5-year-decayed fuel. (authors)

  12. Overall assessment of actinide partitioning and transmutation for waste management purposes

    A program to establish the technical feasibility and incentives for partitioning (i.e., recovering) actinides from fuel cycle wastes and then transmuting them in power reactors to shorter-lived or stable nuclides has recently been concluded at the Oak Ridge National Laboratory. The feasibility was established by experimentally investigating the reduction that can be practicably achieved in the actinide content of the wastes sent to a geologic repository, and the incentives for implementing this concept were defined by determining the incremental costs, risks, and benefits. Eight US Department of Energy laboratories and three private companies participated in the program over its 3-year duration. A reference fuel cycle was chosen based on a self-generated plutonium recycle PWR, and chemical flowsheets based on solvent extraction and ion-exchange techniques were generated that have the potential to reduce actinides in fuel fabrication and reprocessing plant wastes to less than 0.25% of those in the spent fuel. Waste treatment facilities utilizing these flowsheets were designed conceptually, and their costs were estimated. Finally, the short-term (contemporary) risks from fuel cycle operations and long-term (future) risks from deep geologic disposal of the wastes were estimated for cases with and without partitioning and transmutation. It was concluded that, while both actinide partitioning from wastes and transmutation in power reactors appear to be feasible using currently identified and studied technology, implementation of this concept cannot be justified because of the small long-term benefits and substantially increased costs of the concept

  13. Transmutation rates of technetium 99 and iodine 129 in the CANDU actinide burner

    Transmutation rates for the two long-lived fission products technetium 99 and iodine 129 have been calculated for the CANDU Actinide Burner that operates with weapons grade plutonium in an inert matrix as fuel. These transmutation rates are compared with those obtained for the current natural uranium CANDU and for LWRs and FBRs. The higher thermal flux and the softer neutron spectrum of the CANDU Actinide Burner, which is a result of its lower fissile requirements can provide net transmutation half lives as short as 14 y for technetium 99 and 2 y for iodine 129. It is assumed that the iodine 129 can be irradiated as a solution in heavy water. The shorter half life for iodine 129 is due to the large volume of moderator and reflector available that leads to negligible self shielding of the iodine 129 cross section. (author) 1 fig., 2 tabs., 2 refs

  14. Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste

    An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities

  15. Gas core reactors for actinide transmutation and breeder applications. Annual report

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

  16. Minor actinide transmutation in ADS: the EFIT core design

    Accelerator-Driven-Systems represent one of the possible future strategies for transmuting minor actinides. EFIT, the conceptual industrial burner designed in EUROTRANS IP, is an ADS of about 400 MWth, fuelled by MA and Pu in inert matrix, cooled by lead (673-753 K) and sustained by a 800 MeV proton of some 15 mA. It features the MA fission (42 kg/TWhth) while maintaining a zero net balance of Pu and a negligible keff swing during the cycle. Three radial zones, differing in pin diameter or in inert matrix percentage have been defined in order to maximize the average power density together with the flattening of the assembly coolant outlet temperatures. Thermal-hydraulic analyses have been performed and show acceptable maximum temperatures: 1672 K peak fuel temperature (disintegration at 2150 K) and 812 K peak cladding temperature in nominal conditions (max 823 K). The behaviour of the core power, the temperature and the reactivity during the Unprotected Loss Of Flow transient (ULOF) has been studied as well by obtaining: a peak fuel temperature of 1860 K, a peak cladding temperature of 1030 K, a power increase of 2% removed by natural circulation. (authors)

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

    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)

  18. Preparation of TRU fuel and target materials for the transmutation of actinides by gel co-conversion

    In the fabrication of fuel containing transuranium (TRU) elements, flow sheets and techniques that allow a shielded and/or remote fabrication will probably need to be applied. One approach, which has been demonstrated on the laboratory and semi prototype scale, is the wet fabrication route of co-precipitation of the matrix element uranium and plutonium to form either dense spherical particles or to produce hybrid pellets made from pressed gel microspheres. The ceramic material produced holds the TRU-elements (Pu, Np, Am) homogeneously distributed in the matrix. In conjunction with the Department d'Etudes des Combustibles of the French Commissariat a l'Energie Atomique (CEA-DEC) in Cadarache, PSI is further developing a mixed nitride ceramic and mixed oxide with high concentrations (up to 50%) of plutonium with the aim of a joint irradiation test of transuranium elements in the French PHENIX reactor. (author) 2 figs., 3 tabs., 16 refs

  19. Gas core reactors for actinide transmutation. [uranium hexafluoride

    Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

    1979-01-01

    The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

  20. Georgia Institute of Technology research on the gas core actinide transmutation reactor (GCATR)

    The Gas Core Actinide Transmutation Reactor (GCATR) offers several advantages including (1) the gaseous state of the fuel may reduce problems of processing and recycling fuel and wastes, (2) high neutron fluxes are achievable, (3) the possibility of using a molten salt in the blanket may also simplify the reprocessing problem and permit breeding, (4) the spectrum can be varied from fast to thermal by increasing the moderation in the blanket so that the trade-off of critical mass versus actinide and fission product burnup can be studied for optimization, and (5) the U233-Th cycle, which can be used, appears superior to the U235-Pu cycle in regard to actinide burnup. The program at Georgia Tech is a study of the feasibility, design, and optimization of the GCATR

  1. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  2. Transmutation of nuclear waste. State-of-the-art national and international research and strategy studies on partitioning and transmutation of actinides and fission products

    Since 1991 the Netherlands Energy Research Foundation (ECN) in Petten, Netherlands, runs a programme on recycling and transmutation of actinides and long-lived fission products that are present in the spent fuel from nuclear power generation. This programme, which is known under the Dutch acronym RAS, is concentrated on the following topics: reactor physics and scenario studies for transmutation, non-proliferation, thorium cycle, irradiations in the High Flux Reactor at Petten, chemical and material studies of fuels and targets, radiological effects and risks. In the present paper a short description of the achievements of the RAS programme is given. Next, the status of the international research on recycling of actinides and fission products is described. Strategies and (innovative) fuel cycle technology required for the recycling of plutonium, minor actinides and fission products are discussed and their possibilities and limits are identified. Also the potential of future options with low actinide production (thorium cycles, accelerators) is considered. Recommendations for future research in this field are given, taking into account the results of a review by a national committee of experts from government, science and industry. The future work should concentrate on: advanced partitioning methods for trivalent actinides, for which a break-through is required, transmutation of actinides using inert matrices as support (non-fissionable materials), studies using 100% MOX-PWRs, HWRs, HTRs and fast burners, innovative systems for future 'clean' energy production using thorium cycle and/or accelerators. It is emphasized that the radiological effects of all new concepts to be developed for recycling and transmutation should be analysed adequately. 6 figs., 14 tabs., 97 refs

  3. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit

  4. Fabrication and characterisation of composite targets for the transmutation of actinides

    Transmutation of transuranic elements separated from spent fuel is a way to reduce the toxicity of long-lived nuclides in the waste before disposal. Plutonium and the minor actinides (MA) are reintroduced into the fuel cycle for further irradiation and incineration. Currently CERMET fuel forms, in which a ceramic actinide is dispersed in a matrix, are considered for MA transmutation. In a first step, PuO2 beads are produced by a sol gel method in which a Pu nitrate solution is converted to solid, dust-free, particles. These porous beads are then infiltrated with an americium nitrate solution to the incipient wetness point and calcined to give the (PuAm)O2 beads, which are blended with a metal matrix and compacted and sintered to form the final fuel pellet. The matrix used is molybdenum due to its high thermal conductivity and low neutron capture cross section, if it is enriched in 92Mo. In this work, optimization of the bead porosity is investigated to achieve a higher Am content by infiltration. Addition of carbon to the mother solution in the sol gel step increases the bead porosity but it also changes both bead and final fuel pellet microstructure. A surrogate fuel, with cerium simulating the actinides has been fabricated and its mechanical stability and bead characteristics investigated as a function of carbon content and thermal treatment. The characterization of the surrogate fuel by ceramography, density, porosity, bead-quality, etc., is a necessary step in the process optimization, to be transferred to the production of the actinide samples. This process is now at an advanced stage and is being used for the production of fuels for irradiation tests in the Phenix (Futurix) and HFR-Petten (HELIOS) reactors. In parallel, studies on the dissolution of the fuel pellets, with the aim of dissolving the Mo-matrix while keeping the CeO2 beads intact, have been initiated. Thus, Mo can be recycled for further fuel fabrication either from production scraps or from the

  5. Neutronic study regarding transmutation fuel research at Jules Horowitz Reactor

    In order to estimate the possibilities for transmutation experiments at the Jules Horowitz Reactor several ideas for neutronic and fuel behaviour studies are investigated at CEA Cadarache. Naturally an exact replication of the burning of minor actinides in fast reactors, as expected in most transmutation scenarios, is impossible, but some key transmutation parameters can be investigated in a MTR neutron spectrum. In this paper a parametric study regarding fuel damage by He and fission products in AmUO2 is presented. By varying flux level, uranium enrichment and americium content of the sample in the JHR reflector a He production to fission ratio comparable to reference samples in the core of a SFR can be achieved. The calculations were done with the depletion code DARWIN2.2 using JEF2.2 data and spectra from a TRIPOLI model of JHR and an ERANOS model for the SFR respectively. (author)

  6. Actinide partitioning-transmutation program. Final report. VII. Long-term risk analysis of the geologic repository (appendix)

    The Chemical Technology Division of ORNL has prepared a set of documents that evaluate a partitioning-transmutation (PT) fuel cycle relative to a reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted to shorter-lived materials, thereby reducing waste toxicity. Data pertaining to the long-term risk analysis of waste generated from the PT fuel cycle are presented

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

    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 feed 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 ORIGEN 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

  8. The economics of transmutation fuel cycles

    The fuel cycle cost of any transmutation system is one of the major components of the total cost of electricity generated by that system. The fuel cycle cost was estimated for an 1800 MWth actinide burning reactor (ABR) design developed by MIT and INEEL. The fuel is of metallic material composed of 25-30% of TRU and 70-75% Zr. The cost calculations were based on the cost estimates of fuel reprocessing and manufacturing facilities similar to those discussed in the ATW road-mapping effort. An assumption was made that 10 ABRs will be serviced by the fuel separations and manufacturing facilities, and that the fuel will be discharged at a burnup of 70 MWD/kg of total metal (TRU + Zr). A nominal capacity factor of 80% was assumed for operations of the reactor and electric plant system. An analysis was performed to examine the sensitivity of the fuel cycle cost to key factors, specifically to the unit costs of the front-end components of the fuel cycle and the reactor capacity factor (in effect fuel burnup). The results show that the fuel cycle cost of the reference ABR will be about 11 Mills/kWhe, much higher than that of existing LWR nuclear power plants at around 6 Mills/kWhe. The fuel cycle cost has small (< 14%) sensitivity to a ±15% variation in each of the following unit costs: LWR fuel reprocessing, ABR fuel reprocessing and ABR fuel fabrication. The variation of fuel cycle cost is found to be 3 Mills/kWhe for capacity factor variation from 70 to 95%. Therefore, means to reduce the fuel cycle cost would be needed to improve the economic competitiveness of the ABR compared to other electricity generation systems. This work suggests two possible ways to reduce the fuel cycle cost. One is scaling up the production capacity of the fuel separation and manufacturing facilities, perhaps to service 15 ABRs. The second is increasing the discharge burnup, perhaps to 100 ∼ 125 MWD/kg of total metal, which will cut the cost down proportionally. Additionally, the cost of

  9. Denaturing of plutonium by transmutation of minor-actinides for enhancement of proliferation resistance

    Feasibility study for the plutonium denaturing by utilizing minor-actinide transmutation in light water reactors has been performed. And the intrinsic feature of proliferation resistance of plutonium has been discussed based on IAEA's publication and Kessler's proposal. The analytical results show that not only 238Pu but also other plutonium isotopes with even-mass-number have very important role for denaturing of plutonium due to their relatively large critical mass and noticeably high spontaneous fission neutron generation. With the change of the minor-actinide doping ratio in U-Pu mix oxide fuel and moderator to fuel ratio, it is found that the reactor-grade plutonium from conventional light water reactors can be denatured to satisfy the proliferation resistance criterion based on the Kessler's proposal but not to be sufficient for the criterion based on IAEA's publication. It has been also confirmed that all the safety coefficients take negative value throughout the irradiation. (author)

  10. IAEA Activities on Assessment of Partitioning Processes for Transmutation of Actinides

    In these days of nuclear renaissance, appropriate management of radioactive materials arising from the nuclear fuel cycle back end is one of the most important issues related to the long term sustainability of nuclear energy. The present practice in the back end of the closed fuel cycle involves the recovery of uranium and plutonium from spent fuel by the aqueous based PUREX process for reuse in reactors and the conditioning of reprocessing waste into a form suitable for long term storage. The waste contains mainly fission products and transuranium elements immobilized in glass matrix. However, advanced fuel cycles incorporating partitioning of actinides along with minor actinides and their subsequent transmutation (P and T) in a fast neutron energy spectrum could be proliferation resistant and at the same time reduce the waste radiotoxicity by many orders of magnitude. Considering the importance of P and T on long term sustainability, the International Atomic Energy Agency has initiated many collaborative research programs in this area as part of our advanced fuel cycle activities. This paper presents the current and future activities on advanced partitioning methods, highlighting the challenges associated with these processes, fuel manufacturing techniques suitable for integration with reprocessing facility and the IAEA's minor actinide data base (MADB), as a part of integrated nuclear fuel cycle information system (iNFCIS). (authors)

  11. Performances on actinide transmutation based accelerator-driven systems (ADS) at CIEMAT

    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 and 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 the participation and the data analysis of the most advanced experiments in the field and international benchmarks. (authors)

  12. Conceptual design study of an accelerator-based actinide transmutation plant with sodium-cooled solid target/core

    Research and development works on accelerator-based nuclear waste transmutation are carried out at JAERI under the national program OMEGA. The preliminary design of the proposed minor actinide transmutation plant with a solid target/core is described. The plant consists of a high intensity proton accelerator, spallation target of solid tungsten, and subcritical core loaded with actinide alloy fuel. Minor actinides are transmuted by fast fission reactions. The target and core are cooled by the forced flow of liquid sodium coolant. Thermal energy is recovered to supply electricity to power its own accelerator. The core with an effective multiplication factor of about 0.9 generates. The thermal power of 820 MW by using a 1.5 GeV proton beam with a current of 39 mA. The average burnup is about 8%, about 250 kg of actinides, after one year operation at an 80% of load factor. With the conventional steam turbine cycle, electric output of about 246 MW is produced. The design of the transmutation plant with sodium-cooled solid target/core is mostly based on the well-established technology of current LMFRs. Advantages and disadvantages of solid target/core are discussed. Recent progress in the development of intense proton accelerator, the development of simulation code system, and the spallation integral experiment is also presented. (author)

  13. Transmutations of nuclear waste. Progress report RAS programme 1995: Recycling and transmutation of actinides and fission products

    This report describes the progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1995, which is the second year of the 4-year programme 1994-1997. An extensive listing of reports and publications from 1991 to 1995 is given. Highlights in 1995 were: -The completion of the European Strategy Study on Nuclear Waste Transmutation as a result of which the understanding of transmutation of plutonium, minor actinides and long-lived fission products in thermal and fast reactors has been increased significantly. Important ECN contributions were given on Am, 99Tc and 129I transmutation options. Follow-up contracts have been obtained for the study of 100% MOX cores and accelerator-based transmutation. - Important progress in the evaluation of CANDU reactors for burning very large amounts of transuranium mixtures in inert matrices. - The first RAS irradiation experiment in the HFR, in which the transmutation of technetium and iodine was examined, has been completed and post-irradiation examination has been started. - A joint proposal of the EFTTRA cooperation for the 4th Framework Programme of the EU, to demonstrate the feasibility of the transmutation of americium in an inert matrix by an irradiation in the HFR, has been granted. - A bilateral contract with CEA has been signed to participate in the CAPRA programme, and the work in this field has been started. - The thesis work on Actinide Transmutation in Nuclear Reactor Systems was succesfully defended. New PhD studies on Pu burning in HTGR, on nuclear data for accelerator-based systems, and on the SLM-technique for separation of actinides were started. - A review study of the use of the thorium cycle as a means for nuclear waste reduction, has been completed. A follow-up of this work is embedded in an international project for the 4th Framework Programme of the EU. (orig./DG)

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

    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

  15. Neutronics design of transmutation of minor actinides in a fusion reactor

    A concept of transmutation of Minor Actinide (MA) nuclear waste based on the spherical torus (ST) tokamak reactor, FDTR, is put forward. A set of plasma parameter was decided suitable for the ST transmuting nuclear waste blanket. The 2-D neutron transport code TWODANT, 3-D Monte Carlo code MCNP-4B and 1-D burn-up calculation code BISON3.0 and their associated data libraries are used to calculate the transmutation rate, the energy multiplication factor and the tritium breeding rate of the transmutation blanket. The calculation results of the system parameters and the actinide series isotopes for different operation times are also given. The engineering feasibility of the center-post of FDTR is investigated. Relevant results are also given. A preliminary neutronics calculation based on ST transmutation blanket shows that proposed system has high transmuting ability for MA wastes

  16. Design concepts and process analysis for transmuter fuel manufacturing

    The large-scale deployment of remote fabrication and re-fabrication processes (approx. 100 tons of Minor Actinides (MA) annually) will be required for all transmutation scenarios. Process automation has the potential to decrease the cost of remote fuel fabrication and to make transmutation a more economically viable process. The paper describes the design of hot cell fuel manufacturing processes using robotic equipment in hot cells. The dynamics of the robots and the objects handled by them are analyzed in detail using state of the art software tools. In addition to the evaluation and testing of normal assembly operations, the 3D simulation provides for a comprehensive analysis of normal work flows and atypical events such as collisions. The results permit a detailed analysis of the robotic assembly process in terms of forces, torques, and accidents. Detailed simulation results for several operations are presented. (author)

  17. Transmutation of nuclear waste. Status report RAS programme 1993: Recycling and transmutation of actinides and fission products

    The term ''nuclear transmutation'' means a conversion of long-lived radioactive nuclides into short-lived or stable nuclides and ''recycling'' means re-use of fissile material to generate energy in power reactors. With these two processes a reduction of the radiotoxicity and of its duration may be achieved, thus reducing the potential hazard to future generations. Firstly, the report gives a survey of the present situation regarding nuclear waste: its components, how the waste is produced in current LWR and possible options for interim and final storage. Then the objective of the RAS programme, the working methods and the state of the art of the research are considered. Two chapters deal with preliminary results of national and international research. A rather tentative prediction for the future is formulated. Some conclusions are drawn: It seems to be in the best interests of the Netherlands to continue the established line of reprocessing nuclear waste, should new reactors be introduced. It may be advisable to make international agreements so that in the future fission products will contain as few traces of transuranic actinides and long-lived components as possible. Consequently, nuclear waste would become cleaner in terms of long-lived components. For the transmutation of products separated in foreign countries, the Netherlands could pursue an active policy, perform research and also consider the use of MOX fuel in future Dutch reactors. Further contributions towards the solution of these problems can only be made by the Netherlands on an international level. As such, the research and study performed within the framework of the RAS-programme represents a useful international contribution. The possibilities offered by the HFR are particularly of great value. Finally, the choice of a new generation of nuclear reactors should be made not based only on the safety aspects, but also on the extent of waste production and on the transmutation possibilities (application

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

    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

  19. Present status of research activities on transmutation of actinides in Japan

    In Japan, the idea to make use of transmutation for the final disposal method of HLW was first examined by Ichimiya, Amano, Hamada et al., when the Japan Atomic Industry forum had organized a study committee for HLW treatment in 1973. This article has the scope to outline the present research activities on transmutation of actinides in Japan

  20. Status of fuel transmutation programmes in Japan and France. Lessons drawn from results

    Arai, Y.; Pillon, S

    2004-07-01

    France and Japan are currently developing a comprehensive and complementary programme focusing on the transmutation of minor actinides (MA: Np, Am, Cm) and fission products (FP: Tc, I, Cs) in fast breeder reactors (FBR). A summary of current MA-fuel transmutation programmes in France and Japan is provided in this paper, covering objectives, results and perspectives, with emphasis placed on the complementary effort of the two countries. (authors)

  1. Status of fuel transmutation programmes in Japan and France. Lessons drawn from results

    France and Japan are currently developing a comprehensive and complementary programme focusing on the transmutation of minor actinides (MA: Np, Am, Cm) and fission products (FP: Tc, I, Cs) in fast breeder reactors (FBR). A summary of current MA-fuel transmutation programmes in France and Japan is provided in this paper, covering objectives, results and perspectives, with emphasis placed on the complementary effort of the two countries. (authors)

  2. IAEA activity on partitioning and transmutation of actinides and fission products

    In 1990, the IAEA received a request from Member States to review the status of research and development on partitioning and transmutation of actinides and fission products. In response to this request the Advisory Group Meeting (AG) was held in the fall of 1991. AG advised the Agency to play an active role in coordinating international activities in this area. A series of meetings that followed identified considerable interest among many Member States and international organizations in the P and T options as a potential complement to the reference concepts of the back-end of nuclear fuel cycle. Inherent difficulties for the Agency to actively explore this programme were identified including non-proliferation concerns from some Member States about partitioning technology and possible duplication of effort in other international organizations, especially OECD/NEA. But, there remain fundamental questions to be addressed on the objectives of and motivations for P and T and it is clear that some common international understanding would be necessary. In order to contribute to the solution of this problem, and considering the existence of programmes being implemented by OECD/NEA, the Agency has initiated a new CRP entitled 'Safety, environmental and non-proliferation aspects of partitioning and transmutation of actinides and fission products' (1994-1998). This presentation will explain about this Agency's new CRP and how the Agency's work is co-ordinated with other international activities. (author)

  3. Transmutation of minor actinides in a Candu thorium borner

    denaturized the new 233U fuel with 238U. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k∞= ∼ 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the 2nd year and remains above k∞ > 1.06 for ∼ 20 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Nuclear waste actinides can also be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. In the third phase, two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO2) + 5% minor actinides MAO2 and 2) 95% ThO2 + 5% MAO2 + 5% UO2. The latter allows a higher degree of nuclear safeguarding thorough denaturing the new 233U fuel with 238U. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k∞ > 1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k∞ =∼1.06 for ∼ 10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Finally, in the fourth phase, a CANDU reactor fueled with a mixed fuel made of thoria (ThO2) and the totality of nuclear waste actinides has been investigated. The mixed fuel composition has been varied in radial direction to achieve a uniform power distribution and fuel burn up in the fuel bundle. The best fuel compositions with respect to power

  4. Fabrication of fuel and recycling of minor actinides in fast reactors

    Somers, Joseph

    2010-01-01

    Fuels for future fast reactors will not only produce energy, but they must also actively contribute to the minimisation of long lived wastes produced by these, and other reactor systems. The fuels must incorporate minor actinides (MA = Np, Am, Cm) for neutron transmutation into short lived isotopes. Within Europe oxide fuels are favoured. Transmutation can be considered in homogeneous or heterogeneous reactor recycle modes (i.e. in fuels or targets, respectively). Fabrication of such fuels...

  5. Measurements of minor actinides cross sections for transmutation

    The existing reactors produce two kinds of nuclear waste: the fission products and heavy nuclei beyond uranium called minor actinides (Americium and Curium isotopes). Two options are considered: storage in deep geological site and/or transmutation by fast neutron induced fission. These studies involve many neutron data. Unfortunately, these data bases have still many shortcomings to achieve reliable results. The aim of these measurements is to update nuclear data and complement them. We have measured the fission cross section of 243Am (7370 y) in reference to the (n,p) elastic scattering to provide new data in a range of fast neutrons (1-8 MeV). A statistical model has been developed to describe the reaction 243Am (n,f). Moreover, the cross sections from the following reactions have been be extracted from these calculations: inelastic scattering 243Am (n,n') and radiative capture 243Am (n,γ) cross sections. The direct measurements of neutron cross sections are often a challenge considering the short half-lives of minor actinides. To overcome this problem, a surrogate method using transfer reactions has been used to study few isotopes of curium. The reactions 243Am (3He, d)244Cm, 243Am (3He, t)243Cm and 243Am (3He, α)242Am allowed to measure the fission probabilities of 243,244Cm and 242Am. The fission cross sections of 242,243Cm (162,9 d, 28,5 y) and 241Am (431 y) have been obtained by multiplying these fission probabilities by the calculated compound nuclear neutron cross section relative to each channel. For each measurement, an accurate assessment of the errors was realized through variance-covariance studies. For measurements of the reaction 243Am(n,f), the analysis of error correlations allowed to interpret the scope of these measures within the existing measurements. (author)

  6. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    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

  7. Transmutation of nuclear waste. Status report RAS programme 1994: Recycling and transmutation of actinides and fission products

    This report describes the status and progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1994, which is the first year of the second 4-year programme. This programme is outlined and a short progress report is given over 1994, including a listing of 23 reports and publications over the year 1994. Highlights of 1994 were: The completion of long-lived fission-product transmutation studies, the initiation of small-scale demonstration experiments in the HFR on Tc and I, the issue of reports on the potential of the ALMR (Advanced Liquid Metal Reactor) for transmutation adn the participation and international cooperation on irradiation experiments with actinides in inert matrices. The remaining chapters contain more extended contributions on recent developments and selected topics, under the headings: Benefits and risks of partitioning and transmutation, Perspective of chemical partitioning, Inert matrices, Evolutionary options (MOX), Perspective of heavy water reactors, Perspective of fast burners, Perspective of accelerator-based systems, Thorium cycle, Fission-product transmutation, End scenarios, and Executive summary and recommendations. (orig.)

  8. Special scientific programme on use of high energy accelerators for transmutation of actinides and power production

    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

  9. Recovery of actinides from spent nuclear fuel by pyrochemical reprocessing

    The Partitioning and Transmutation (P and T) strategy is based on reduction of the long-term radiotoxicity of spent nuclear fuel by recovery and recycling of plutonium and minor actinides, i.e. Np, Am and Cm. Regardless if transmutation of actinides is conceived by a heterogeneous accelerator driven system, fast reactor concept or as integrated waste burning with a homogenous recycling of all actinides, the reprocessed fuels used are likely to be significantly different from the commercial fuels of today. Because of the fuel type and the high burn-up reached, traditional hydrometallurgical reprocessing such as used today might not be the most adequate method. The main reasons are the low solubility of some fuel materials in acidic aqueous solutions and the limited radiation stability of the organic solvents used in extraction processes. Therefore, pyrochemical separation techniques are under development worldwide, usually based on electrochemical methods, reductive extraction in a high temperature molten salt solvent or fluoride volatility techniques. The pyrochemical reprocessing developed in ITU is based on electrorefining of metallic fuel in molten LiCl-KCl using solid aluminium cathodes. This is followed by a chlorination process for the recovery of actinides from formed actinide-aluminium alloys, and exhaustive electrolysis is proposed for the clean-up of salt from the remaining actinides. In this paper, the main achievements in the electrorefining process are summarised together with results of the most recent experimental studies on characterisation of actinides-aluminium intermetallic compounds. U, Np and Pu alloys were investigated by electrochemical techniques using solid aluminium electrodes and the alloys formed by electrodeposition of the individual actinides were analysed by XRD and SEM-EDX. Some thermodynamic properties were determined from the measurements (standard electrode potentials, Gibbs energy, enthalpy and entropy of formation) as well as

  10. Transmutation Fuels Campaign FY-09 Accomplishments Report

    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.

  11. Protected Plutonium Production (P3) by transmutation of minor actinides for peace and sustainable prosperity

    'Protected Plutonium Production (P3)' 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 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, 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 238Pu or 242Pu 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. The new evaluation function of attractiveness is applied for assessing the existing plutonium criteria as summarized in the following, (a) weapon grade plutonium (b) plutonium with 30% fraction of 240Pu (c) plutonium with 6% fraction of 238Pu (d) plutonium exempt from safeguards. Since both proliferation resistant plutonium compositions (b) and (c) give almost the same value of attractiveness, plutonium is categorized by following well accepted terminology, weapon grade, usable, practically unusable and exempt as shown. It is concluded based on the new evaluation function 'Attractiveness' that P3 mechanism by the transmutation of MA is very effective to improve the proliferation resistance of plutonium. In the conference, the fundamentals of P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of

  12. Fuel cycle of actinide burner-reactor. Review of investigations by > program

    The problem of long-lived minor-actinides (Np, Am, Cm) transmutation is one of major part of problem of nuclear power ecological safety. The problem of Pu surpluses burning-out adjoins to this problem. Existing and perspective reactor systems could be used for it, but task of optimum organization of the external closed cycle for actinide burner reactor becomes the important aspect of transmutation problem. Since 1992, SSC RIAR has proposed the demonstration program-concept DOVITA (Dry reprocessing, Oxide fuel, Vibropac, Integral, Transmutation of Actinides), which should demonstrate opportunities of new technologies for realization of the optimized fuel cycle for actinide burner reactor. The brief review of study on DOVITA program for 5 years is given in this paper. (J.P.N.)

  13. Status of the French research program for actinides and fission products partitioning and transmutation

    currently presented to French Ministries of Research and Industry and to the National Parliament which plans to pass a new waste management law in 2006 asking for new prospects for P and T further implementation. 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 in 2001 the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process. Then, the 2002-2005 program has encompassed technological demonstration of the selected liquid-liquid process, with representative equipment which have been set up for this purpose in new shielded cells inside the Atalante facility. CEA also conducted programmes proving the feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation test in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. The scenario studies aimed at examining the possibilities of reducing significantly the final waste inventory and at quantifying the inventories of plutonium, minor actinides and certain long-lived fission products in various nuclear-power-plant geometries; they also allowed to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. (author)

  14. Transmutation Fuel Campaign Description and Status

    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

  15. Fabrication of actinide mononitride fuel

    Fabrication of actinide mononitride fuel in JAERI is summarized. Actinide mononitride and their solid solutions were fabricated by carbothermic reduction of the oxides in N2 or N2-H2 mixed gas stream. Sintering study was also performed for the preparation of pellets for the property measurements and irradiation tests. The products were characterized to be high-purity mononitride with a single phase of NaCl-type structure. Moreover, fuel pins containing uranium-plutonium mixed nitride pellets were fabricated for the irradiation tests in JMTR and JOYO. (author)

  16. NSC-WPFC task force on potential benefits and impacts of advanced fuel cycles with actinide partitioning and transmutation (WPFC/TFPT)

    A study has been performed by a Task Force within the Working Party on Scientific Issues of the Fuel Cycle (WPFC) of the OECD NEA Nuclear Science Committee with the objective to gather and analyze results of different studies performed to assess the potential impact of P and T on different types of repositories in different licensing and regulatory environments. The present paper summarizes the approach and the main finding of that study and reviews the extent to which P and T will impact geological disposal depending on the disposal environment and the details of the P and T approach. (author)

  17. A measurement of actinide neutron transmutations with accelerator mass spectrometry in order to infer neutron capture cross sections

    Bauder, William K.

    Improved neutron capture cross section data for transuranic and minor actinides are essential for assessing possibilities for next generation reactors and advanced fuel cycles. The Measurement of Actinide Neutron TRAnsmutation (MANTRA) project aims to make a comprehensive set of energy integrated neutron capture cross section measurements for all relevant isotopes from Th to Cf. The ability to extract these cross sections relies on the use of Accelerator Mass Spectrometry (AMS) to analyze isotopic concentrations in samples irradiated in the Advanced Test Reactor (ATR). The AMS measurements were performed at the Argonne Tandem Linear Accelerator System (ATLAS) and required a number of key technical developments to the ion source, accelerator, and detector setup. In particular, a laser ablation material injection system was developed at the electron cyclotron resonance ion source. This system provides a more effective method to produce ion beams from samples containing only 1% actinide material and offers some benefits for reducing cross talk in the source. A series of four actinide measurements are described in this dissertation. These measurements represent the most substantial AMS work attempted at ATLAS and the first results of the MANTRA project. Isotopic ratios for one and two neutron captures were measured in each sample with total uncertainties around 10%. These results can be combined with a MCNP model for the neutron fluence to infer actinide neutron capture cross sections.

  18. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Bourg Stéphane; Geist Andreas; Narbutt Jerzy

    2015-01-01

    Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities strivin...

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

    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.

  20. Safety and environmental aspects of partitioning and transmutation of actinides and fission products. Proceedings of a technical committee meeting held in Vienna, 29 November - 2 December 1993

    There is considerable interest in many countries in the partitioning and transmutation of long lived radionuclides as a potential complement to the closed fuel cycle. Recognizing this, the IAEA organized a Technical Committee Meeting on Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products, to review the current status of progress of national and international programmes and identify the most important directions of co-operation. The results of the Technical Committee meeting are presented in this document. Refs, figs and tabs

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

    Smith, James Dean; Drennen, Thomas E. (Hobart & William Smith College, Geneva, NY); Rochau, Gary Eugene; Martin, William Joseph; Kamery, William (Hobart & William Smith College, Geneva, NY); Phruksarojanakun, Phiphat (University of Wisconsin, Madison, WI); Grady, Ryan (University of Wisconsin, Madison, WI); Cipiti, Benjamin B.; Wilson, Paul Philip Hood (University of Wisconsin, Madison, WI); Mehlhorn, Thomas Alan; Guild-Bingham, Avery (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX)

    2007-10-01

    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.

  2. Impact of minor actinide transmutation options on geological disposal: The French case

    Within the framework of June 28, 2006 waste management French Act, an assessment of industrial perspectives of partitioning and transmutation of actinides is provided in 2012. These studies must be carried out in tight connection with GENIV systems development. In this perspective, CEA asked the French waste management Agency (Andra) to assess the impact of high and intermediate level waste as produced by various transmutation options, on the sizing of a geological repository. Andra used repository architectures similar to those employed in the Cigéo project which is under development for current NPPs. Results allow to compare the underground footprint and the excavated volume for each option ; the impact of the interim storage duration is also assessed. Solutions are proposed to optimize the footprint of the repository. An analysis of the advantages and drawbacks of transmutation options is proposed. (author)

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

    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.)

  4. Seed and blanket ADS using thorium–reprocessed fuel: Parametric survey on TRU transmutation performance and safety characteristics

    Highlights: • The seed and blanket ADS reference configuration without MA enrichment shows a large reactivity swing due to burnup. • The larger amount of MA introducing into the core, the more TRUs are transmuted. • With larger MA content, the smaller reactivity swing in the system can be achieved. • Increasing core size by introducing more thorium and reprocessed fuel assemblies into the system reduces reactivity swing. - Abstract: Conceptual designs of accelerator driven systems (ADS) that utilize thorium fuel as blanket and reprocessed fuel as seed for fission reaction in order to transmute the transuranic elements in spent nuclear fuel and produce energy from thorium utilization was proposed. In order to analyze the TRU transmutation performance of the system while the safety margin during the operation is satisfied, a parametric survey is done in this study. The impact of minor actinide (MA) amount and the reactor size on transmutation efficiency and safety characteristics of the ADS is investigated using MCNPX code using the ENDF/B-VII. As the results, with increasing MA content, TRU transmutation rate is increased while void reactivity of the system becomes less negative. Besides, increasing the core size by introducing more thorium and reprocessed fuel assemblies into the system reduces the TRU transmutation rate and reactivity swing due to burnup of the system. In order to match the efficient transmutation purpose and safety features during operation, higher MA content is suggested to introduce into the system

  5. Engineering assessment studies on the JRC's actinides partitioning processes for transmutation

    Three conceptual processes have been studied and investigated for the feasibility of removing actinides from high active waste. Two of the flowsheets rely completely on counter current techniques for the actinides separation namely the TBP and HDEHP processes, whereas the third process OXAL, uses a precipitation technique in the first instance followed by dissolution of the actinides and rare-earths (RE) for further treatment using a modified HDEHP process. Many important factors such as 'direct' or 'delayed', concentrated or unconcentrated HAW, storage time, activity and heat release levels, solvent irradiation DF's, safety and steady-state recycling conditions for U-LWR, Pu-LWR and FBRs for possible transmutation scenarios have been taken into consideration

  6. Description of Transmutation Library for Fuel Cycle System Analyses

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

    2010-08-01

    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: • Assemble past and future transmutation cases for system analyses. • 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. • 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. • Provide mass fractions at input (charge) and output (discharge) for each case. • 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. • Catalog other isotope-specific information in one place, e.g., heat and dose conversion factors for individual isotopes. • 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: • 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. • 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

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

    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.

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

    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.)

  9. Work within the coordinated programme on environmental evaluation and hazard assessment of the separation of the actinides from the high-level waste from fuel reprocessing followed by either transmutation of separate disposal

    Reduction of plutonium losses in non-high-level wastes streams from fuel reprocessing is important in the conventional post-fission operations and would be essential in any actinide partitioning alternatives. The balance of input process streams and output waste streams in present reprocessing technology is compared to the balance after process modifications based on recent experimental developments. The results are showing that by the introduction of new electrochemical redox processes and non-salt-forming process chemicals the routinely generated intermediate-level waste streams from PUREX reprocessing can be avoided. Plutonium-bearing waste streams can be extensively recycled within the chemical processing

  10. The nuclear fuel cycle for transmutation: a critical review

    This review presents a critical common FZK and CEA discussion of the transmutation possibilities of actinide nuclei and of fission products as Tc and I in reactors (PWRs and FBRs) and in accelerator-driven subcritical configurations. The activities in the Research Center Karlsruhe in the chemical area are briefly discussed. Activities in the chemical area at CEA are presented elsewhere at this conference. The alternate waste disposal with transmutation is compared to the direct disposal option, as seen from the FZK point of view. Work in France on this point is still underway according to a law, voted in the French Parliament in 1991. The aim of this study is to evaluate, how the short-term and long-term risks of nuclear waste, including both direct disposal and transmutation scenarios, realistically could be minimized. (authors)

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

    European R and D for ADS design and fuel development is driven in the 6th 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 (Pu0.5,Am0.5)O2-x - Mo (with the isotope 92Mo 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 92Mo and a lower specific transmutation rate of minor actinides, because of the higher neutron absorption cross-section of the matrix. The composite CERCER fuel (Pu0.4,Am0.6)O2-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] in this

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

    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 TREVOL 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

  13. Actinide partitioning-transmutation program final report. VI. Short-term risk analysis of reprocessing, refabrication, and transportation: appendix

    The Chemical Technology Division of the Oak Ridge National Laboratory has prepared a set of documents that evaluate a Partitioning-Transmutation (PT) fuel cycle relative to a Reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted to shorter-lived materials, thereby reducing the waste toxicity. This report compares the two fuel cycles on the basis of the short-term radiological and nonradiological risks they present to the public and to workers. The accidental radiological risk to the public is analyzed by estimating the probabilities of sets of accidents; the consequences are calculated using the CRAC code appropriately modified for the material composition. Routine radiological risks to the public are estimated from the calculated release amounts; the effects are calculated using the CRAC code. Radiological occupational risks are determined from prior experience, projected standards, and estimates of accident risk. Nonradiological risks are calculated from the number of personnel involved, historical experience, and epidemiological studies. The result of this analysis is that the short-term risk of PT is 2.9 times greater than that of the Reference cycle, primarily due to the larger amount of industry. This conclusion is strongly dominated by the nonradiological risk, which is about 150 times greater than the radiological risk. The absolute risk as estimated for the fuel cycle portions considered in this report is 0.91 fatalities/GWe-year for the PT cycle and 0.34 fatalities/GWe-year for the Reference cycle. This should be compared with Inhaber's estimate of 1.5 for nuclear and 150 for coal. All of the risks assumed here are associated with the production of one billion watts of electricity (GWe) per year

  14. Actinide partitioning-transmutation program final report. VI. Short-term risk analysis of reprocessing, refabrication, and transportation: appendix

    Fullwood, R.R.; Jackson, R.

    1980-01-01

    The Chemical Technology Division of the Oak Ridge National Laboratory has prepared a set of documents that evaluate a Partitioning-Transmutation (PT) fuel cycle relative to a Reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted to shorter-lived materials, thereby reducing the waste toxicity. This report compares the two fuel cycles on the basis of the short-term radiological and nonradiological risks they present to the public and to workers. The accidental radiological risk to the public is analyzed by estimating the probabilities of sets of accidents; the consequences are calculated using the CRAC code appropriately modified for the material composition. Routine radiological risks to the public are estimated from the calculated release amounts; the effects are calculated using the CRAC code. Radiological occupational risks are determined from prior experience, projected standards, and estimates of accident risk. Nonradiological risks are calculated from the number of personnel involved, historical experience, and epidemiological studies. The result of this analysis is that the short-term risk of PT is 2.9 times greater than that of the Reference cycle, primarily due to the larger amount of industry. This conclusion is strongly dominated by the nonradiological risk, which is about 150 times greater than the radiological risk. The absolute risk as estimated for the fuel cycle portions considered in this report is 0.91 fatalities/GWe-year for the PT cycle and 0.34 fatalities/GWe-year for the Reference cycle. This should be compared with Inhaber's estimate of 1.5 for nuclear and 150 for coal. All of the risks assumed here are associated with the production of one billion watts of electricity (GWe) per year.

  15. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Bourg Stéphane

    2015-12-01

    Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

  16. Advanced Recycling Reactor with Minor Actinide Fuel

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  17. Proceedings of the Twelfth Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation

    Partitioning and transmutation (P and T) is one of the key technologies for reducing the radiotoxicity and volume of radioactive waste produced by the nuclear power industry. Recent developments indicate the advantages to be realised by embedding P and T strategies into advanced fuel cycles considering both waste management and economic issues. In this context, the OECD Nuclear Energy Agency (NEA) has been organising 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. Previous meetings were held in Mito (Japan) in 1990, at ANL (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, in Mito (Japan) in 2008, in San Francisco (United States) in 2010 and have been co-sponsored by the European Commission (EC) and the International Atomic Energy Agency (IAEA). The 12. Information Exchange Meeting was held in Prague, Czech Republic on 24-27 September 2012, hosted by the Radioactive Waste Repository Authority (RAWRA). The workshop comprised a plenary session on national and international programmes followed by technical sessions and a poster session covering various aspects of P and T. The information exchange meetings on P and T form a part of NEA programme of work in the field of advanced nuclear fuel cycles. The titles of the eight technical sessions are: International and National Programmes; Fuel Cycle Strategies and Transition Scenarios; Impact of P and T on Geological Disposal; Transmutation Systems: Design, Performance and Safety; Pyro and Aqueous Separation Processes; Transmutation Fuels and Targets; Transmutation Physics, Experiments and Nuclear Data; Economics of P and T. Poster session contributions to this meeting are also available at http

  18. Removal of actinides from selected nuclear fuel reprocessing wastes

    The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables

  19. Analyses in Support of Z-Pinch IFE and Actinide Transmutation - LLNL Progress Report for FY-06

    Meier, W R; Moir, R W; Abbott, R

    2006-09-19

    This report documents results of LLNL's work in support of two studies being conducted by Sandia National Laboratories (SNL): the development of the Z-pinch driven inertial fusion energy (Z-IFE), and the use of Z-pinch driven inertial fusion as a neutron source to destroy actinides from fission reactor spent fuel. LLNL's efforts in FY06 included: (1) Development of a systems code for Z-IFE and use of the code to examine the operating parameter space in terms of design variables such as the Z-pinch driver energy, the chamber pulse repetition rate, the number of chambers making up the power plant, and the total net electric power of the plant. This is covered in Section 3 with full documentation of the model in Appendix A. (2) Continued development of innovative concepts for the design and operation of the recyclable transmission line (RTL) and chamber for Z-IFE. The work, which builds on our FY04 and FY05 contributions, emphasizes design features that are likely to lead to a more attractive power plant including: liquid jets to protect all structures from direct exposure to neutrons, rapid insertion of the RTL to maximize the potential chamber rep-rate, and use of cast flibe for the RTL to reduce recycling and remanufacturing costs and power needs. See Section 4 and Appendix B. (3) Description of potential figures of merit (FOMs) for actinide transmutation technologies and a discussion of how these FOMs apply and can be used in the ongoing evaluation of the Z-pinch actinide burner, referred to as the In-Zinerator. See Section 5. (4) A critique of, and suggested improvements to, the In-Zinerator chamber design in response to the SNL design team's request for feedback on its preliminary design. This is covered in Section 6.

  20. The Use of Molybdenum-Based Ceramic-Metal (CerMet) Fuel for the Actinide Management in LWRs

    The technical and economic aspects of the use of molybdenum depleted in the isotope 95Mo (DepMo) for the transmutation of actinides in a light water reactor are discussed. DepMo has a low neutron absorption cross section and good physical and chemical properties. Therefore, DepMo is expected to be a good inert matrix in ceramic-metal fuel. The costs of the use of DepMo have been assessed, and it was concluded that these costs can be justified for the transmutation of the actinides neptunium, americium, and plutonium

  1. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    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.

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

    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 238U 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.

  3. Nuclear data for new fuel cycles and waste transmutation

    At the present rate of consumption, the known uranium resources on earth will be sufficient to continue its use as fuel for nuclear energy production for no longer then about 200 more years. Other fuels such as oil and gas probably face an even shorter life before exhaustion of known resources. At the same time the actual rate of radioactive waste production associated with this energy generation requires technical solutions that will permit the continuous support of economic growth and improved quality of life, together with a minimal impact to environment. Worldwide new research programmes for radioactive waste management have recently been undertaken. In France, at the French National Research Council (CNRS), various groups are presently working on the development of new technologies to reduce the present inventory of nuclear wastes and simultaneously on more efficient ways to use the thorium natural resources to generate nuclear energy while minimizing waste production during many centuries. At the Centre d'Etudes Nucleaires in Bordeaux we have undertaken a programme of cross-section measurements for neutron-induced reactions with energies ranging from a few meV to 6 MeV. We have in particular measured the (n, γ) and/or the (n,fission) cross-sections for three nuclei (232 Th, 233 Pa and 233 U) playing an important role for the Th-U cycle. This fuel offers a number of advantages (in particular a factor 100 less minor actinides produced) compared to the commonly used uranium fuel in light water pressurized reactors. Furthermore we have also remeasured the transmutation rate for the long-lived fission product 129 I which have a half live of 1.6x107 years. The various experiments have been performed at the 4 MV Van de Graaff accelerator in Bordeaux and at the Tandem accelerator of the Institut de Physique Nucleaire in Orsay. This work was partly supported by the CNRS programme PACE (Programme Aval du Cycle Electronucleaire) as well as by the Conseil Regional d

  4. Definition of Technology Readiness Levels for Transmutation Fuel Development

    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

  5. Enhancing VVER annular proliferation resistance fuel with minor actinides

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the

  6. Comparative assessment of the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ads

    A preliminary comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides has been performed in the case of ANSALDO's Energy Amplifier Demonstration Facility based on molten lead-bismuth eutectic cooling, classical MOX-fuel technology and operating at 80 MWth. The neutronic calculations presented in this paper are a result of a state-of-the-art computer code package, EA-MC, developed by C. Rubbia and his group at CERN. Both high-energy particle interactions and low-energy neutron transport are treated with a sophisticated method based on a full Monte Carlo simulation, together with modern nuclear data libraries. Detailed Monte Carlo transport calculations were performed for different types of external neutron sources: D-D and D-T fusion sources and proton induced spallation neutron sources. The fuel core was described on a pin-by- pin basis allowing for detailed scans of the main neutronic properties, e.g. neutron flux spectra and power density distributions. (author)

  7. Actinide partitioning-transmutation program final report. VII. Long-term risk analysis of the geologic repository

    This report supports the overall assessment by Oak Ridge National Laboratory of actinide partitioning and transmutation by providing an analysis of the long-term risks associated with the terminal storage of wastes from a fuel cycle which incorporates partitioning and transmutation (P-T) and wastes from a cycle which does not. The system model and associated computer code, called AMRAW (Assessment Method for Radioactive Waste), are used for the analysis and are applied to the Los Medanos area in southeastern New Mexico. Because a conservative approach is used throughout, calculated results are believed to be consistently higher than reasonable expectations from actual disruptive incidents at the site and therefore are not directly suited for comparison with other analyses of the particular geologic location. The assessment is made with (1) the probabilistic, or risk, mode that uses combinations of reasonable possible release incidents with their probability of occurrence distributed and applied throughout the assessment period, and (2) the consequence mode that forces discrete release events to occur at specific times. An assessment period of 1 million years is used. The principal results are: (1) In all but the expulsive modes, 99Tc and 129I completely dominate cumulative effects based on their transport to man through leaching and movement with groundwater, effecting about 33,000 health effects (deaths) over the 1 million years; (2) P-T has only limited effectiveness in reducing long-term risk from a radionuclide waste repository under the conditions studied, and such effectiveness is essentially confined to the extremely unlikely (probability of occurrence 10-12/year) expulsive events; (3) Removal or immobilization of 99Tc and 129I might provide benefits sufficiently tangible to warrant special consideration

  8. AECL/U.S. INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors Fuel Requirements and Down-Select Report

    William Carmack; Randy Fielding; Pavel Medvedev; Mitch Meyer

    2005-08-01

    This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Euratom Joint Proposal 1.8 entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Light-Water Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light-water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.

  9. U.S./EURATOM INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in LWRs -- Fuel Requirements and Down-Select Report

    William Carmack; Randy Fielding; Pavel Medvedev; Mitch Meyer

    2005-08-01

    This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Canada Joint Proposal entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.

  10. Thermal and hydraulic test plan of TRU fuel element for transmutation process

    JAERI is developing processes to partition long-lived transuranic elements (TRU) from high-level radioactive waste and transmutation processes to transform TRU into shorter-lived or stable nuclides under the OMEGA program. To promote developments of transmutation processes, thermal and hydraulic tests were planed to optimize a fuel element of an actinide burner fast reactor (ABR) cooled by helium gas. Along the test plan, a simulated fuel element in which simulated fuel particles were filled up in the porous annular space of 11.7mm in gap width and of 600mm in length was manufactured experimentally, and also a test apparatus which could circulate helium gas or nitrogen gas at a maximum flow rate of 400 m3/h under 1 MPa was designed and fabricated. Hydraulic performance of the test apparatus was confirmed through preliminary operations. This paper presents mainly a thermal and hydraulic test plan of the fuel element for developing ABR core design, outlines of the simulated fuel element and the test apparatus, and preliminary operation results. (author)

  11. Research and development for the fabrication of minor actinide-bearing fuel materials and technologies

    The transmutation of minor actinides (MA) in 4th-generation reactors can be envisioned in homogeneous or heterogeneous mode. Minor actinide-bearing blankets (MABB) -- fuel for heterogeneous transmutation comprising 10 to 20% MA dispersed in a UO2 matrix -- are largely unknown and warrant further research on the fabrication and properties of the materials, on their evolution under self-irradiation, and on their behavior in the reactor. This article summarizes progress in co conversion of uranium-americium compounds by oxalate precipitation or ion exchange resins. It also describes current R&D on MABB fabrication by powder metallurgy or spherical particle metallurgy. The fabrication processes in teleoperated shielded cells are discussed together with the technologies applicable to MABB fabrication equipment. (author)

  12. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E. [Idaho National Engineering and Environment Lab., Advanced Nuclear Energy, Idaho (United States)

    2001-07-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  13. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  14. Comparison of different options for minor actinide transmutation in the frame of the French law for waste management

    In the frame of the French Act for waste management which has been passed by French Parliament on June 28th, 2006, it is requested to obtain in 2012 an assessment of industrial perspectives of partitioning and transmutation of long-lived elements. These studies must be carried out in tight connection with GENIV systems development. The expected results must include the evaluation of technical and economic scenarios taking into account the optimization options between the minor actinide transmutation processes, their interim storage and geological disposal, including an analysis of several criteria. In this perspective, the CEA has established a working group named 'GT TES' (Working Group on Technical and Economic Scenarios) involving EDF and AREVA to define scenarios, the various criteria to evaluate them, to conduct these evaluations and then to highlight the key results. The group also relied on ANDRA for the geological storage studies. The scenarios evaluations take place in the French context. The nuclear energy production is supposed to remain constant during the scenarios and equal to 430 TWhe/year in accordance with the current French nuclear power installed capacity of 60 GW(e). The deployment of the first Sodium-cooled Fast Reactor (SFR) starts in 2040, considering that at this date the SFR technology should be mature. Several management schemes of minor actinides have been studied: Plutonium recycling in SFR (minor actinides are sent to the waste). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in homogeneous mode ('Hom.'). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in heterogeneous mode ('Het.'). Plutonium recycling in SFR and minor actinide transmutation in Accelerator-Driven-System (ADS). The criteria used to analyze these different scenarios, should take into account the viewpoint of scientists, industrials, administrations, and the general public. They are listed below: Inventories and

  15. Sol gel chemistry applied to the synthesis of actinide-based compounds for the fabrication of advanced fuels

    The chemistry of the sol-gel process is based on hydroxylation and condensation of molecular precursors and can be used for the elaboration of advanced nuclear fuel or transmutation targets. On the one hand, some fundamental studies are conducted, based on complexation reactions to modulate and control the reactivity of the different cations (Zr(IV) and minor actinides) prior to hydrolysis and condensation step. The purpose of this work is to obtain hetero poly-condensation in order to form homogenous compounds with a controlled microstructure. On the other hand, internal gelation process, one of the important sol-gel routes for the preparation of actinides microspheres (the dedicated design for advanced nuclear fuel or transmutation targets) is developed. Investigations are currently carried out to study the gelation behaviour of solutions containing actinides (III) or (IV) in comparison with the more well known behaviour of U(VI) studied during the development of process for beads production (1960 - 1990). (authors)

  16. Build-Up Of Actinides In Irradiated Fuel Rods Of The ETRR-1 Reactor

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on PC computer to provide the required calculations. The fuel element of 10% 235U enrichment of ETRR-1 reactor was taken as an example for calculations using BAC code. The results are compared with other calculations for the ETRR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% '235U enrichment for ETRR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated

  17. Build up of actinides in burnt fuel rods of the ET-RR-1 reactor

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% 235U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% 235U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated. (author)

  18. Thermal-hydraulics of actinide burner reactors

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  19. Picolinamides, a possible family of actinide extractants, for the one step actinide (VI), (IV) and (III) separation from the spent nuclear fuels dissolution liquors

    The separation of all the actinides U, Np, Pu, Am and Cm from the spent fuels and their further transmutation become one alternative to the deep ground repository storage of radioactive wastes. To achieve this aim, new chemical separation processes, more convenient than the PUREX process, are necessary. Especially, extractants able to separate in one single step the actinide (VI), (IV) and (III) from acidic nitrate solutions, leaving the lanthanides (III) in the effluents are highly suitable. Results in the search of such extractants are presented. 1 tab., 6 refs

  20. Plutonium Management, Minor Actinides Partitioning and Transmutation R and D in France

    Jean-Marc Cavedon (CEA, France) then presented the developments concerning Plutonium management and minor actinides P and T research and development in France. By the 1991 law on high-level long-lived radioactive waste a research programme was launched in the areas: (i) geological disposal, (ii) conditioning and long-term storage, and (iii) radiotoxicity reduction by P and T. The results of the work in these areas will be presented to the French Government and Parliament in 2006. The control of Plutonium stocks generated by the French PWRs is proposed to increase Plutonium consumption in reactors and minimise radioactive waste production, and requires the recycling of actinides, especially Plutonium. In the long term, CEA intends to develop a new technology based on gas cooled reactors and their associated fuel cycle, including multiple recycling of Plutonium. The advantages of this development consist in the optimisation of the use of natural resources and the concentration of Plutonium in limited quantities of fuel rods. If needed, the minor actinides could also be recycled. The planned CEA developments depend on new fuel types and will lead to novel waste types (light glasses) with a reduction of long-term radiotoxicity. Radiotoxicity reductions by a factor of 3 to 5 are expected for Plutonium recycling scenarios, and by up to a factor of a few hundreds for Plutonium and minor actinides recycling scenarios. This gain is nearly independent on the reactor type used, but needs about 100 years of application to become effective in terms of making a difference in the total waste inventory to be disposed of

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

    The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O2-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.

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

    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)

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

    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.)

  4. TRU transmutation type BWR fuel assembly

    The BWR fuel assembly is formed by bundling a plurality of fuel rods and a water channel disposed at the center of the assembly by a plurality of spacers. An upper tie plate and a lower tie plate are disposed to upper and lower portions of the fuel rods and the water channel respectively. An upper end plug of the water channel is attached detachably to a cylindrical main body of the water channel. A zircaloy tube incorporating TRU nuclides is contained and secured in the water channel. The zircaloy tube has such a structure as capable of incorporating and sealing oxides or metal materials containing TRU nuclides. Since the zircaloy tube containing TRU nuclides is contained not in fuel region but in the water rod, the loaded uranium amount of fuels is not reduced but the reactivity can be ensured. (I.N.)

  5. Depletion analysis on long-term operation of the conceptual Molten Salt Actinide Recycler and Transmuter (MOSART) by using a special sequence based on SCALE6/TRITON

    Highlights: ► An automatic computation and control sequence has been developed for MSR neutronics and depletion analyses. ► The method was developed based on a series of stepwise SCALE6/TRITON calculations. ► A detailed reexamination of the MOSART operation in 30 years was performed. ► Clean-up scenarios of fission products have a significant impact on the MOSART operation. - Abstract: A special sequence based on SCALE6/TRITON was developed to perform fuel cycle analysis of the Molten Salt Actinide Recycler and Transmuter (MOSART), with emphasis on the simulation of its dynamic refueling and salt reprocessing scheme during long-term operation. MOSART is one of conceptual designs in the molten salt reactor (MSR) category of the Generation-IV systems. This type of reactors is distinguished by the use of liquid fuel circulating in and out of the core, which offers many unique advantages but complicates the modeling and simulation of core behavior using conventional reactor physics codes. The TRITON control module in SCALE6 can perform reliable depletion and decay analysis for many reactor physics applications due to its problem-dependent cross-section processing and rigorous treatment of neutron transport. In order to accommodate a simulation of on-line refueling and reprocessing scenarios, several in-house programs together with a run script were developed to integrate a series of stepwise TRITON calculations; the result greatly facilitates the neutronics analyses of long-term MSR operation. Using this method, a detailed reexamination of the MOSART operation in 30 years was performed to investigate the neutronic characteristics of the core design, the change of fuel salt composition from start-up to equilibrium, the effects of various salt reprocessing scenarios, the performance of actinide transmutation, and the radiotoxicity reduction

  6. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: (1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs; (2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs; (3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs; and (4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs

  7. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  8. A deep burn fuel management strategy for the transmutation of light water reactors waste in the gas turbine-modular helium reactor

    We have investigated the waste actinide burnup capabilities of the gas turbine modular helium reactor (GT-MHR), similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo continuous energy burnup code (MCB), an extension of Monte Carlo N-particle transport code (MCNP) developed at the Royal Institute of Technology in Stockholm and the University of Science and Technology in Cracow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present studies, the GT-MHR is fueled with the transuranic actinides contained in light water reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible. The driver fuel (DF) of the GT-MHR uses fissile isotopes (e.g. 239Pu and 241Pu), previously generated in the LWRs, and maintains criticality conditions in the GT-MHR. After an irradiation of three years, the spent driver fuel is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). Transmutation Fuel mainly contains non-fissile actinides that undergo neutron capture and transmutation during the subsequent three-year irradiation in the GT-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. The destruction of more than 94% of 239Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of proliferation. (author)

  9. Study of the fuel behavior, safety characteristics and transmutation performance of a gas cooled accelerator driven system (ADS)

    The neutronic behavior of an ADS system based on gas cooling is examined in this work by using the simulation tools MCNPX and ORIGEN. The main character of the MCNPX code is the use of the Monte-Carlo method allowing a high dimensional simulation of the physical processes. The whole model of the core is represented in 3 dimensional zones including the target structure, which provides the initial spallation neutrons for the chain reaction in the fuel zone. At the beginning, MOX fuel with 19.5 wt. Pu/(Pu+U) is loaded in order to investigate the technical feasibility of a test facility. The fuel assemblies are replaced step by step with Plutonium and minor actinides (PuMa) uranium free fuel according to a loading and shuffling pattern. The designed test facility consists of 120 fuel assemblies each 91 fuel rods which are arranged around the spallation target. For a thermal power of 100 MW the burn-up and transmutation rate is studied. The first results for the MOX and partially PuMa fuel loaded core are presented in this paper. For the PuMa fuel two compositions are investigated. Both fuel types chosen for the analysis demonstrate the capability of the incineration of americium. The simulations show that the initial composition has significant influence on the transmutation rate. The deployment of MOX type fuel in the ADS core causes a considerable consumption of Pu but also a significant generation of americium

  10. Fast Reactor Systems and Innovative Fuels for Minor Actinides Homogeneous Recycling

    This work is focused on the performance of critical fast reactor systems aimed at the transmutation of minor actinides (Np, Am, Cm) homogeneously dispersed in the MOX driver fuel. In particular, the paper deals with two scenarios in the 2050 time horizon, at first evaluating an extension of once-through fuel cycle strategy, hence introducing fast reactors in a closed fuel cycle strategy beyond 2030. The synergistic use of the DESAE and NFCSS scenario codes permitted to evaluate key indicators for natural resources usage, waste management, proliferation issues, and fuel cycle infrastructures needs. The paper aims at discussing the sustainability of a high development of nuclear energy to promote a transition to a low-carbon energy future. Finally, the results of scenarios analysis are discussed in the light of the ongoing studies moving ahead in the development of innovative fuels for minor actinides transmutation (e.g., PELGRIMM EU projects), where ENEA is actively involved on the track of related past activities. (author)

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

    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.

  12. On a fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel

    A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert in inner and outer cores may lead to an effective transmutation of minor actinides and a flattered power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors. (author)

  13. Numerical analysis on reduction of radioactive actinides by recycling of nuclear fuel

    Worldwide, human growth has reached unparalleled levels historically, this implies a need for more energy, and just in 2007 was consumed in the USA 4157 x 109 kWh of electricity and there were 6 x 109 metric tons of carbon dioxide, which causes a devastating effect on our environment. To this problem, a solution to the demand for non-fossil energy is nuclear energy, which is one of the least polluting and the cheapest among non-fossil energy; however, a problem remains unresolved the waste generation of nuclear fuels. In this work the option of a possible transmutation of actinides in a nuclear reactor of BWR was analyzed, an example of this are the nuclear reactors at the Laguna Verde nuclear power plant, which have generated spent fuel stored in pools awaiting a decision for final disposal or any other existing alternative. Assuming that the spent fuel was reprocessed to separate useful materials and actinides such as plutonium and uranium remaining, could take these actinides and to recycle them inside the same reactor that produced them, so il will be reduced the radiotoxicity of spent fuel. The main idea of this paper is to evaluate by means of numeric simulation (using the Core Management System (CMS)) the reduction of minor actinides in the case of being recycled in fresh fuel of the type BWR. The actinides were introduced hypothetically in the fuel pellets to 6% by weight, and then use a burned in the range of 0-65 G Wd/Tm, in order to have a better panorama of their behavior and thus know which it is the best choice for maximum reduction of actinides. Several cases were studied, that is to say were used as fuels; the UO2 and MOX. Six different cases were also studied to see the behavior of actinides in different situations. The CMS platform calculation was used for the analysis of the cases presented. Favorable results were obtained, having decreased from a range of 35% to 65% of minor actinides initially introduced in the fuel rods, reducing the

  14. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

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

    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 (Pfus ∼ 10-100 MW, βN ∼ 2-3, Qp ∼ 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)

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

    Stacey, W. M.

    2001-02-01

    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 (Pfus approx 10-100 MW, βN approx 2-3, Qp approx 2-5, R approx 3-5 m, I approx 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-à-vis a neutron source is reviewed.

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

    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 14C and 59Ni 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)

  18. Reduction of burden for waste disposal for accelerator-driven transmutation technology. Preparing for unforeseeable future by nuclear fuel cycle for back-end

    Accelerator Driven System (ADS) is an innovative nuclear system to transmute minor actinides. By coupling transmutation technology by ADS and partitioning technology, the burden for nuclear waste disposal is expected to be largely reduced. Under the present status where the future of nuclear fuel cycle is unforeseeable, it is desirable to proceed with the research and development of ADS which can flexibly harmonize well with various options in the future. The research and development of ADS should be promoted by international and interdisciplinary collaboration. In this context, Transmutation Experimental Facility under the J-PARC phase-2 project is expected to play an important role to lead worldwide activities to cope with radioactive wastes. (author)

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

    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)

  20. Enhanced minor actinide burning core for closed fuel cycle

    This paper presents core concepts enhancing TRU burning or MA transmutation in sodium cooled reactor satisfying the void reactivity requirements. In this study, two concepts of transmutation system are considered; in the first system TRUs are burned only by ARR whose target is maximizing TRU burning. The second is a system that Pu is burned by LWR and ARR, Am is transmuted by ARR whose target is maximizing Am transmutation. Therefore some innovative and challenging technologies have been examined under the safety requirements; MA burning fuel with 50% TRU fraction, moderator pin, fuel of high Am fraction, and Am blanket. According to the detailed calculation of high TRU contained oxide core with moderator pins of 12% arranged driver fuel assemblies, the TRU conversion ratio decreases to 0.33 and the TRU burning capability is improved to 67 kg/TWeh. Deploying Am blanket which is oxide fuel with Am 50% and U 50%, the total of Am transmutation capability of oxide fueled core becomes 69 kg/TWeh. (author)

  1. Mesures de sections efficaces d'actinides mineurs d'intérêts pour la transmutation

    Kessedjian, Grégoire

    2008-01-01

    Les réacteurs actuels produisent deux types de déchets dont la gestion et le devenir soulèvent des problèmes. Il s'agit d'abord de certains produits de fission et de noyaux lourds (isotopes de l'Américium et du Curium) au-delà de l'uranium appelés actinides mineurs. Deux options sont envisagées : le stockage en site géologique profond et/ou l'incinération de ces déchets dans un flux de neutrons rapides, c'est-à-dire, la transmutation par fission. Ces études font appel à de nombreuses données ...

  2. Consultancy to review and finalize the IAEA publication 'Compendium on the use of fusion/fission hybrids for the utilization and transmutation of actinides and long-lived fission products'. Working material

    In addition to the traditional fission reactor research, fusion R and D activities are becoming of interest also to nuclear fission power development. There is renewed interest in utilizing fusion neutrons, Heavy Liquid Metals, and molten salts for innovative systems (energy production and transmutation). Indeed, for nuclear power development to become sustainable as a long-term energy option, innovative fuel cycle and reactor technologies will have to be developed to solve the problems of resource utilization and long-lived radioactive waste management. In this context Member States clearly expressed the need for comparative assessments of various transmutation reactors. Both the fusion and fission communities are currently investigating the potential of innovative reactor and fuel cycle strategies that include a fusion/fission system. The attention is mainly focused on substantiating the potential advantages of such systems: utilization and transmutation of actinides and long-lived fission products, intrinsic safety features, enhanced proliferation resistance, and fuel breeding capabilities. An important aspect of the ongoing activities is the comparison with the accelerator driven subcritical system (spallation neutron source), which is the other main option for producing excess neutrons. Apart from comparative assessments, knowledge preservation is another subject of interest to the Member States: the goal, applied to fusion/fission systems, is to review the status of, and to produce a 'compendium' of past and present achievements in this area

  3. Isotopic Transmutation and Fuel Burnup in BN-600 Hybrid Fast Reactor Core

    BN-600 fast reactor core was modeled using MCNPX computer code. The core configuration and material composition, for the hybrid design, was simulated in this model. The power generated in different zones was determined and the results were compared with published results and found acceptable. Isotopes transmutations in various zones were estimated. The uranium isotopes are major contributors to power production in this reactor, the probability of plutonium incineration will increase with the increase in the use of MOX oxide. The transmutation of minor actinides is not obvious in this configuration

  4. Recycling the actinides, the cornerstone of any sustainable nuclear fuel cycles

    The sustainability of the current nuclear fuel cycles is not completely achieved since they do not optimise the consumption of natural resource (only a very small part of uranium is burnt) and they do not ensure a complete and efficient recycling of the potential energetic material like the actinides. Promoting nuclear energy as a future energy source requires proposing new nuclear systems that could meet the criteria of sustainability in terms of durability, bearability and liveability. In particular, it requires shifting towards more efficient fuel cycles, in which natural resources are saved, nuclear waste are minimised, efficiently confined and safely disposed of, in which safety and proliferation-resistance are more than ever ensured. Such evolution will require (i) as a mandatory step, evolutionary recycling of the major actinides U and Pu up to their optimized use as energetic materials using fast neutron spectra, (ii) as an optional step, the implementation of the recycling of minor actinides which are the main contributors to the long term heat power and radiotoxicity of nuclear waste. Both options will require fast neutrons reactors to ensure an efficient consumption of actinides. In such a context, the back-end of the fuel cycle will be significantly modified: implementation of advanced treatment/recycling processes, minor-actinides recovery and transmutation, production of lighter final waste requiring lower repository space. In view of the 2012 French milestones in the framework of the 2006 Waste Management Act, this paper will depict the current state of development with regards with these perspectives and will enlighten the consequences for the subsequent nuclear waste management. (authors)

  5. Transmutation of Minor Actinides in a Spherical Torus Tokamak Fusion Reactor

    FENGKaiming; ZHANGGuoshu

    2002-01-01

    Fusion energy will be a long-term energy source. Great efforts have been devoted to fusion research in the past 50 years, and there is still a long way to go. Transmutation of high-level waste (HLW) utilizing D-T fusion neutrons is a good choice for an early application of fusion.

  6. Separation of actinides from spent nuclear fuel: A review.

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

  7. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P3 Mechanism and Methodology Development for Plutonium Categorization

    'Protected Plutonium Production (P3)' 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 237Np or 241Am with large neutron capture cross-section into the uranium fuel to enhance the production of 238Pu or 242Pu, 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 238Pu or 242Pu 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 P3 mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P3 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)

  8. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of 241Am and 237Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the 241Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  9. Calculation and Analysis of B/T (Burning and/or Transmutation Rate of Minor Actinides and Plutonium Performed by Fast B/T Reactor

    Marsodi

    2006-01-01

    Full Text Available Calculation and analysis of B/T (Burning and/or Transmutation rate of MA (minor actinides and Pu (Plutonium has been performed in fast B/T reactor. The study was based on the assumption that the spectrum shift of neutron flux to higher side of neutron energy had a potential significance for designing the fast B/T reactor and a remarkable effect for increasing the B/T rate of MA and/or Pu. The spectrum shifts of neutron have been performed by change MOX to metallic fuel. Blending fraction of MA and or Pu in B/T fuel and the volume ratio of fuel to coolant in the reactor core were also considered. Here, the performance of fast B/T reactor was evaluated theoretically based on the calculation results of the neutronics and burn-up analysis. In this study, the B/T rate of MA and/or Pu increased by increasing the blending fraction of MA and or Pu and by changing the F/C ratio. According to the results, the total B/T rate, i.e. [B/T rate]MA + [B/T rate]Pu, could be kept nearly constant under the critical condition, if the sum of the MA and Pu inventory in the core is nearly constant. The effect of loading structure was examined for inner or outer loading of concentric geometry and for homogeneous loading. Homogeneous loading of B/T fuel was the good structure for obtaining the higher B/T rate, rather than inner or outer loading

  10. Transmutation in ASTRID

    Within the scope of the French Act of 28 June 2006 on managing long-lived radioactive waste, one of the objectives of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor is to demonstrate the capacity to transmute minor actinides (MA) on an industrial scale. Demonstration scenarios currently focus on two modes of transmutation: a homogeneous mode using a standard fuel with low MA content, and a heterogeneous mode where the minor actinides are loaded in the radial blanket around the core, known as the minor-actinide-bearing blanket (MABB) concept. With different initial MA concentrations for the two modes of transmutation, we have estimated their impact on the performance and safety of the ASTRID reactor core. The consequences on the dimensions of the storage means, the handling systems and the fuel sub-assembly transport packaging are also reviewed in order to identify the limits beyond which significant design changes to the core and nuclear steam supply system (NSSS) would be required. Analysis of the results has made it possible to identify the most suitable irradiation conditions and initial contents to demonstrate transmutation in ASTRID, with the main aim of achieving a balance in the minor actinide flows without significantly changing the reactor design: • Americium (Am), a main contributor to the heat and the radiotoxicity of radioactive waste after the decay of fission products, will be treated as a top priority, • Part of the americium can be overridden by neptunium (Np) without any impact on the design and performance, • Curium (Cm) is not considered; it’s too penalising in the handling of new sub-assemblies, • Possible weight levels for the demonstration: 2% of Am in the fuel for the homogeneous mode and 10% of Am in the blanket for the heterogeneous mode. Whatever the chosen mode of transmutation, it will be necessary to conduct experimental programmes in ASTRID to validate and qualify the behaviour of

  11. Study of the burnup behavior, safety characteristics and transmutation performance of the LWRs with innovative fuel concepts

    plutonium and resulting resonance absorption rate in upper thermal energy range. With the consumption of Pu and progress of burnup the temperature coefficients becomes considerable negative causing positive safety characteristics. The simulation of the transmutation process indicates an enhanced performance of (Th/Pu)O2 and IMF fuelled cores in both, reduction of plutonium and TRU (per MW reactor power). In particular, the incineration of plutonium in these two fuel variants is about three times as much as in the MOX fuel per MWd. This is due to the replacement of U-238 (MOX) by Thorium and Molybdenum respectively. The fraction of plutonium reduction leading to higher isotopes (minor actinides) is comparable for (Th/Pu)O2 and IMF and varies from 7 to about 10%. In case of the MOX fuel, the conversion of U-238 to new plutonium causes less effective plutonium reduction. The simulation shows that approx. 17.6% of the total converted plutonium result in the production of minor actinides. (author)

  12. INERT-MATRIX FUEL: ACTINIDE ''BURNING'' AND DIRECT DISPOSAL

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers

  13. Research on Actinides in Nuclear Fuel Cycles

    The electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipment, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media

  14. Current US plans for development of fuels for accelerator transmutation of waste

    The United States is currently investigating the feasibility of proposed technologies for the Accelerator Transmutation of Waste (ATW) concept, which is funded as part of the U.S. Department of Energy's Advanced Accelerator Applications (AAA) Program. The ATW concept is proposed as a means to transmute transuranic isotopes and, perhaps, long-lived fission products removed from light water reactor spent fuel to shorter-lived fission products. To attain maximum possible transmutation rates, no fertile material (i.e., U-238 or Th-232) is to be incorporated into the fuel. Fuel forms currently proposed for ATW application include non-fertile dispersions of metal alloy or nitride fuel particles in a metal matrix, a non-fertile metal alloy, or non-fertile nitride pellets for a fast-spectrum, liquid metal-cooled transmuter, and non-fertile TRISO-coated particles dispersed in graphite compacts for a thermal-spectrum, gas-cooled transmuter. There is little or no experience with these non-fertile fuels, so an extensive fuel development program is envisioned. Current plans call for initial effort to demonstrate feasibility of the proposed fuel forms by the end of 2005, consistent with AAA program decision milestones. Feasibility research and development will consist of the following: Development of fabrication processes to demonstrate fabricability of the proposed fuel forms; Simple irradiation tests to screen samples of each fuel type for unexpected or poor performance; and Determination of intrinsic properties or characteristics (e.g., out-of pile interdiffusion behavior of fuel and constituents and thermophysical properties). If the decision is made to continue development of the ATW concept beyond 2005, then of the successful candidate forms, one or two will be selected for further development, with more extensive irradiation testing and fuel property characterization. (author)

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

    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 106Ru and 125Sb was observed under the diluted HCl medium, thereby completely 106Ru-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 (135Cs, etc), MLFP (90Sr, 137Cs) 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

  16. ALMR potential for actinide consumption

    The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) sponsored fast reactor design based on the Power Reactor, Innovative Small Module (PRISM) concept originated by General Electric. This reactor combines a high degree of passive safety characteristics with a high level of modularity and factory fabrication to achieve attractive economics. The current reference design is a 471 MWt modular reactor fueled with ternary metal fuel. This paper discusses actinide transmutation core designs that fit the design envelope of the ALMR and utilize spent LWR fuel as startup material and for makeup. Actinide transmutation may be accomplished in the ALMR core by using either a breeding or burning configuration. Lifetime actinide mass consumption is calculated as well as changes in consumption behavior throughout the lifetime of the reactor. Impacts on system operational and safety performance are evaluated in a preliminary fashion. Waste disposal impacts are discussed. (author)

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

    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.

  18. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes; Mesures des sections efficaces de capture et potentiels d'incineration des actinides mineurs dans les hauts flux de neutrons: Impact sur la transmutation des dechets

    Bringer, O

    2007-10-15

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  19. Transmutation of radioactive nuclear waste- present status and requirement for the problem-oriented nuclear data base

    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 focusing on the waste transmutation problem. (author)

  20. Transmutation of radioactive nuclear waste – present status and requirement for the problem-oriented nuclear data base

    Yu A Korovin; V V Artisyuk; A V Ignatyuk; G B Pilnov; A Yu Stankovsky; Yu E Titarenko; S G Yavshits

    2007-02-01

    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 focusing on the waste transmutation problem.

  1. Actinides reduction by recycling in a thermal reactor

    This work is directed towards the evaluation of an advanced nuclear fuel cycle in which radioactive actinides could be recycled to remove most of the radioactive material; firstly a production reference of actinides in standard nuclear fuel of uranium at the end of its burning in a BWR reactor is established, after a fuel containing plutonium is modeled to also calculate the actinides production in MOX fuel type. Also it proposes a design of fuel rod containing 6% of actinides in a matrix of uranium from the tails of enrichment, then four standard uranium fuel rods are replaced by actinides rods to evaluate the production and transmutation thereof, the same procedure was performed in the fuel type MOX and the end actinide reduction in the fuel was evaluated. (Author)

  2. Actinide management with commercial fast reactors

    Ohki, Shigeo

    2015-12-01

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  3. Actinide management with commercial fast reactors

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel

  4. Actinide management with commercial fast reactors

    Ohki, Shigeo [Japan Atomic Energy Agency, 4002, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-12-31

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  5. Impact of actinide recycle on nuclear fuel cycle health risks

    The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR)1 and Integral Fast Reactor (IF)2 technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle

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

    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)

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

    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 237Np, 241Am, and 243Am burned by thermal neutrons, while in the inner region 244Cm 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)

  8. Transmutation abilities of a 3600 MWth SFR core

    This paper presents an evaluation of the potential of transmutation of minor actinides in a 3600 MWth SFR core. Two modes of transmutation have been considered : homogeneous and heterogeneous. To be consistent with cycle scenario studies, the performances achieved in minor actinide consumption (Np+Am+Cm) for the homogeneous and heterogeneous modes are about -9.6 kg/TWhe and -5.8 kg/TWhe respectively, considering an initial loaded mass of 2.3 tons in both cases. The main conclusion of this study is that the reference design of the SFR core seems to be adapted to the transmutation of minor actinides with homogeneous mode until 3% contents. In the case of higher contents, a new design of the assembly will be necessary to take into account the degradation of the fuel properties and cladding behavior. (author)

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

    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 (YH1.8) 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)

  10. Leaching of actinide elements from simulated fuel debris into seawater

    For the prediction of the leaching behavior of actinide elements contained in the fuel debris that has arisen from the severe accident in Fukushima Daiichi Nuclear Power Station (NPS), a simulated fuel debris consisting of UO2 - ZrO2 solid solution doped with 137Cs, 237Np, 236Pu and 241Am tracers was synthesized, and agitated leaching tests were conducted for the simulated fuel debris in seawater. The synthesized simulated fuel debris was immersed and shaken in natural seawater collected at a coast 11 km away from Fukushima Daiichi NPS. The brief leaching test conditions were T = 25°C and solid-liquid ratio = 4 g/l, and the test duration was up to 31 days. The ratio of tracers leached into seawater from the simulated fuel debris by the agitated leaching test for 4 days was evaluated to be 0.09% for U, 0.01% for Np, 0.01% for Pu, 0.01% for Am and 35.39% for Cs by the α or γ spectrometry of the soluble fraction. The leaching of actinides from the real fuel debris in reactor units 1 - 3 in Fukushima Daiichi NPS is expected to be suppressed in comparison with that from normal light water reactor spent fuel. (author)

  11. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  12. Behavior of actinides in the Integral Fast Reactor fuel cycle

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides (237Np, 240Pu, 241Am, and 243Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  13. Future nuclear fuel cycles: Prospect and challenges for actinide recycling

    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. In this context, nuclear power has the worldwide potential to curtail the dependence on fossil fuels and thereby to reduce the amount of greenhouse gas emissions while promoting energy independence. How we deal with nuclear radioactive waste is crucial in this context. In France, the public's concern regarding the long-term waste management made the French Governments to prepare and pass the 1991 and 2006 Acts, requesting in particular the study of applicable solutions for still minimizing the quantity and the hazardousness of final waste. This necessitates High Active Long Life element (such as the Minor Actinides MA) recycling, since the results of fuel cycle R and D could significantly change the challenges for the storage of nuclear waste. HALL recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste and thus aiding the 'burying approach' in securing a 'broadly agreed political consensus' of waste disposal in a geological repository. This paper presents an overview of the recent R and D results obtained at the CEA Atalante facility on innovative actinide partitioning hydrometallurgical processes. For americium and curium partitioning, these results concern improvements and possible simplifications of the Diamex-Sanex process, whose technical feasibility was already demonstrated in 2005. Results on the first tests of the Ganex process (grouped actinide separation for homogeneous recycling) are also discussed. In the coming years, next steps will involve both better in-depth understanding of the basis of these actinide partitioning processes and, for the new promising

  14. Removal of actinides from nuclear fuel reprocessing wastes using an organophosphorous extractant

    By removing actinides from nuclear fuel reprocessing wastes, long term waste storage hazards are reduced. A solvent extraction process to remove actinides has been demonstrated in miniature mixer-settlers and in simulated columns using actinide feeds. Nonradioactive pilot plant results have established the feasibility of using pulse columns for the process

  15. Fuel selection criteria specific for double stratum minor actinide burners

    The scope of this presentation covers the following topics: Accelerator driven systems in the double strata fuel cycle; parametric studies of neutronic properties foe dedicated inert matrix fuels (solid solution oxides ZrO2, ThO; solid solution nitrides ZrN, HfN, YN; CERCER oxides MgO; CERMET oxides Cr, V, Mo, Mo-92, W). Fertile matrices do not improve neutronic performance of americium bearing fuels. High linear rating (high thermal conductivity + high melting temperature) improves neutronic performance as well. CERCER oxide in MgO matrix appears to be a reasonable reference fuel for Minor Actinide burning in LBE cooled ADS. Solid solution nitride in ZrN or CERMET oxide in Mo-92 matrix offer better performance, but high T stability (nitrides) and helium release (CERMET) issues need to be addressed. These fuels will be fabricated and irradiated in Phenix starting 2005

  16. Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

    In the frame of nuclear fuels studies for generation IV, carbides or oxycarbides assemblies are one of the engaged material for high temperature reactors. The design of the fuels is not yet defined but some structures are actually considered with SiC as matrix for the actinide fuel. In this work we have studied the synthesis of a multi-scale structure controlled SiC matrix using molecular silicon organometallic precursors. The aim of this work was to develop a way to obtain multi-scale SiC matrix material which could be engineered to fit in any fuel structure defined for generation IV fuels. The control of this multi-scale structure was done using several simulation methods specific of the low temperature solution synthesis of the precursor. In a first step, we have focused our effort on the synthesis of the SiC material. A first level of template was successfully done by the use of solid silica 500 nm balls. A second level of template was studied by the use of meso-porous silica, structured at a 50 nm level. At least, supra-molecular simulation in non aqueous media was considered with the difficulty to build a molecular assembly (inverse micelles). In a second step, we have functionalized the primary silane phase with actinide complexing agent in order to blend directly the actinide inside this primary phase in a controlled way. During these studies, a new one pot synthesis route to obtain the functionalized primary silane phase was developed. (authors)

  17. Establishing the design basis for a Molten Salt Demonstration Transmuter

    A Molten Salt Demonstration Transmuter is required to show the operation and design performance for closing the nuclear spent fuel cycle for PWR or WWER reactors operated in the once-through cycle mode. The remnant waste would be either permanently stored or held for secondary use. The purpose of this proposal is to establish the design basis for the Molten Salt Demonstration Transmuter. It is supposed that once-through-cycle nuclear spent fuel would be delivered to the Molten Salt Demonstration Transmuter in the standard transportable container includes 84 WWER-440 SNF assemblies each weighing 250 kg and containing 120 kg U, and about 1.2 kg of Pu and minor actinides. One assembly at a time will be withdrawn from the container and chemically processed to supply Pu and minor actinides at the rate necessary for burn-up compensation. (Authors)

  18. Scientific feasibility of long lived wastes transmutation

    This report presents the results of the works carried out by the CEA on the scientifical feasibility of long-lived wastes transmutation. A first part, based on cross-sections analysis, deals with transmutation efficiency on minor actinides and fission products with respect to the neutron spectrum and independently of the transmuting system under consideration. The intrinsic advantage of the fast neutron spectrum, with respect to the neutron status, to the transmutation levels obtained and to the low generation of higher isotopes, is shown. The limitations of minor actinide loading of about 1-2% in PWRs and of about 3-5% in FBRs, in relation with core physics, are justified. The consequences on the fuel cycle in terms of residual power and neutron and gamma sources are precised with the penalties graduation: neptunium x-type fuel, are of about 300 for PWR recycling options and of about 50 for FBRs recycling options. The conclusions drawn from this scientifical feasibility step sustain the option choices retained in transmutation technical feasibility studies. A second part treats of calculation methodology aspects. The measurements and evaluation of nuclear data, specifically performed on long-lived radioactive wastes, and which represent a key component with respect to the quality and relevance of transmutation studies, are described. The different qualification programs (analytical, physical and global) allow to conclude that neutron transmutation of long lived radioactive wastes is based on a well-mastered physics, even in the case of the neutronic field of a classic fission reactor, or in the case of the neutronic field of a subcritical booster environment supplied by an external source. The models and nuclear data used for the calculations will require a detailed validation to warrant the reliability of a project of realization. However, the same models and data allow today to perform calculations with pertinent and credible results. (J.S.)

  19. Options for treatment of legacy and advanced nuclear fuels

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  20. Partitioning and Transmutation. Annual Report 2004

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products (129I, 99Tc, 135Cs, 93Zr and 126Sn and activation products (14C and 36Cl). 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

  1. Partitioning and Transmutation. Annual Report 2004

    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.

  2. Partitioning and Transmutation. Annual Report 2005

    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, 94N 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

  3. Transmutation of actinides from light water reactors in modular high-temperature reactors for the reduction of long-lived nuclides; Verbrennung von Aktiniden aus Leichtwasserreaktoren in modularen Hochtemperaturreaktoren zur Reduzierung langlebiger Nuklide

    Meier, Astrid

    2012-05-15

    Only one of many different ways to produce electric power is the Light Water Reactor (LWR).This reactor produces high level long-lived and radiotoxic nuclides like Plutonium and Minore Actinides (Neptunium, Americium, Curium,..), which have to be safely isolated and controlled in a final storage over a long time. Thus, many projects worldwide concentrate on the transformation of these long-lived nuclides into short-lived nuclides by transmutation and fission processes. Here, mainly accelerator driven systems and Generation-IV-reactors, like the graphite moderated, Helium cooled High Temperature Reactor (HTR), are in focus of research. The main advantages of the HTR are the fuel structure, which allows high burnups and the inherent safety. In case of a Loss Of Cooling Accident (LOCA), the decay heat will be dissipated without any active cooling system. This passive heat transfer is high enough to stay below the upper temperature limit in the fuel. Therefore, the fuel structure stays intact and the fission products retain inside the fuel. In this thesis, the long-lived nuclides like Plutonium, Neptunium and Americium, extracted from the spent LWR fuel, will be reused in a fresh fuel element for the HTR. To achieve the aim of reducing these nuclides and their radiotoxicity, the HTR has to operate at the highest possible burnup. Therefore parameters, like e.g. the fuel temperature or the power density distribution and also the behaviour in case of an accident have to be comparable to the HTR loaded with uranium fuel. The European Union project ''Plutonium and Minore Actinide Waste Management'' (PuMA) is the origin for the used reference reactor geometry, the fuel structure as well as the nuclide densities in the Plutonium and Minor Actinides fuel. The reactor design of this project is almost identical to the South African reactor concept with 400 MW{sub th} thermal power and an inner graphite column (Pebble Bed Modular Reactor PBMR-400).For

  4. Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

    For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

  5. Factors affecting actinide solubility in a repository for spent fuel, 1

    The main tasks in the study were to get information on the chemical conditions in a repository for spent fuel and information on factors affecting releases of actinides from spent fuel and solubility of actinides in a repository for spent fuel. The work in this field started at the Reactor Laboratory of the Technical Research Centre of Finland (VTT) in 1982. This is a report on the effects on the main parameters, Eh, pH, carbonate, organic compounds, colloids, microbes and radiation on the actinide solubility in the nearfield of the repository. Another task has been to identify available models and reported experience from actinide solubility calculations with different codes. 167 refs

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

    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.

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

    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

  8. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  9. Outcomes on Oxide Fuel Development for Minor Actinides Recycling

    A state of the art review is given of minor actinide (MA)-bearing oxide fuel development for sodium fast reactors (SFRs) and accelerator driven systems (ADS). The homogeneous recycle option in SFRs, where small amounts of MAs are diluted in (U,Pu)O2-x driver fuels, emerges as a technically sound approach, reinforced by national and international programmes. Its technology readiness level is appropriate to implement irradiation tests from pin to bundle scale. Regarding the heterogeneous recycle option in SFRs, a comprehensive database regarding inert matrix fuels is available as the result of ~35 irradiation tests. The promising results gained with MgO, Mo and ZrO2 matrices have to be completed by post-irradiation examinations on optimized fuel microstructures. On the other hand, a first step in the long term (MA,U)O2-x fuel development process is under investigation with MARIOS and DIAMINO tests in the HFR and OSIRIS, before the implementation of prototypical irradiation tests. For ADS, very informative feedback from inert matrix fuel developments has been completed by dedicated collaborative programmes, including major irradiations for the fuel performance assessment from HELIOS and FUTURIX-FTA experiments, whose post-irradiation examinations are under way. (author)

  10. Disposal of nuclear wastes by transmutation

    A study was made of the feasibility of partition and transmutation (P-T) of actinides, 99Tc, and 129I in radioactive wastes. An incremental analysis was performed on a reference fuel cycle and a P-T fuel cycle. Short-term risks from fuel cycle operations and long-term risks from a repository were estimated for cases with and without P-T. Results show that P-T cannot be justified because of the small radiological benefits and substantially increased costs. 1 table

  11. Technical feasibility of long lived wastes transmutation

    The aim of this report is to evaluate the technical feasibility of long-lived wastes transmutation in different type of reactors and their associated cycles. This feasibility depends both on the type of waste and on the type of reactor. It is performed through scenario studies which allow to evaluate the overall steps of the fuel cycle (reactor, fabrication, storage, reprocessing) and which include the detailed studies of changes in cores design and management induced by transmutation, the impacts on fuel cycle facilities, and on reprocessing and fabrication processes. Previous scenario studies have permitted to underline the advantages and drawbacks of the different strategies. The scenarios considered in this document cover the overall options foreseeable today: a PWR-based scenario for the recycling of plutonium and americium in homogeneous mode based on the MOX UE Am assembly concept from 2020 onward; a 4. generation reactor-based scenario with fast spectrum and self recycling of actinides from 2035 onward; and a scenario where minor actinides are recycled in a specific cycle in association with subcritical systems. The document comprises also a specific chapter about the technical feasibility of the transmutation fuel which covers the overall aspects of the fuel cycle to be considered. (J.S.)

  12. Transmutation Scenarios Impacts on Advanced Nuclear Cycles (fabrication/reprocessing/transportation)

    In the frame of the French Law for waste management, minor actinides transmutation scenarios have been studied for a sodium-cooled fast reactors fleet using homogeneous or heterogeneous recycling modes. Americium, neptunium and curium can be transmuted once included together in the standard MOX fuel, or the sole Americium can be incorporated in Am-bearing radial blanket. MAs transmutation in Accelerator Driven System has also been studied while Plutonium is recycling in SFR. Assessments and comparisons of these advanced cycles have been performed in light of technical and economic aspects criteria. The purpose of this study is to present the results in terms of impacts of the transmutation scenarios on fuel cycle plants (fabrication, reprocessing) and transportations taking into account thermal, radiation and criticality parameters. Comparison with no transmutation option is also presented. (author)

  13. The Czech national R and D program of nuclear incineration of PWR spent fuel in a transmuter with liquid fuel

    The principle drawbacks of any kind of solid nuclear fuel are listed and briefly analysed in the first part of the paper. On the basis of this analysis, the liquid fuel concept and its benefits are introduced and briefly described in the following parts of the paper allowing to developed new reactor systems for nuclear incineration of spent fuel from conventional reactors and a new clean source of energy. As one of the first realistic attempts to utilize the advantages of liquid fuel, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed in [1], has been proposed for a deeper, both theoretical and experimental studies in [2]. There will be a preliminary design concept of an experimental assembly LA-0 briefly introduced in the paper which is under preparation in the Czech Republic for such a project [3]. (author)

  14. Wastes Management Through Transmutation in an ADS Reactor

    Bernard Verboomen; Giuseppe Forasassi; Nicola Cerullo; Barbara Vezzoni; Barbara Calgaro

    2008-01-01

    The main challenge in nuclear fuel cycle closure is the reduction of the potential radiotoxicity, or of the time in which that possible hazard really exists. Probably, the transmutation of minor actinides with fast fission processes is the most effective answer. This work, performed in SCK⋅CEN (Belgium) and DIMNP Pisa University, is focused on preliminary evaluation of industrial scale ADS (400 MWth, 2.5 mA) burning capability. An inert matrix fuel of minor actinides, 50% ...

  15. Transmutation of high-level radioactive waste by a charged particle accelerator

    Transmutation of minor actinides and fission products using proton accelerators has many advantages over a transmutor operated in a critical condition. The energy required for this transmutation can be reduced by multiplying the spallation neutrons in a subcritical assembly surrounding the spallation target. The authors have studied the relation between the energy requirements and the multiplication factor, k, of the subcritical assembly, while varying the range of several parameters in the spallation target. A slightly subcritical reactor is superior to a reactor with large subcriticality in the context of the energy requirement of a small proton accelerator, the extent of radiation damage, and other safety problems. To transmute the fission products, the transmutor reactor must have a good neutron economy, which can be provided by a transmutor operated by a proton accelerator. The paper discusses the use of minor actinides to improve neutronics characteristics, such as a long fuel burn-up rather than simply transmuting this valuable material

  16. EC-FP7 ARCAS: technical and economical comparison of Fast Reactors and Accelerator Driven Systems for transmutation of Minor Actinides

    The ARCAS project aims to compare, on a technological and economical basis, Accelerator Driven Systems and Fast Reactors as Minor Actinide burners. It is split in five work packages: the reference scenario definition, the fast reactor system definition, the accelerator driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarizes the status of the project and its five work packages. (author)

  17. Accelerator-driven sub-critical target concept for transmutation of nuclear wastes

    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 the components of the 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, based on the described PHOENIX Concept, 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 MWe. 19 refs., 20 figs

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

    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

  19. Perspectives of partitioning and transmutation technology

    When we explore the sustainable utilization of nuclear power, reasonable and environmentally preferable waste management is indispensable. The Partitioning and Transmutation (P and T) technology has been studied in many countries aiming at reduction of the burden for disposal of high-level radioactive waste (HLW). This technology, coupled with the geological disposal, is now regarded as a part of advanced fuel cycle, and hence various research and development (R and D) are under way. As for the partitioning process of spent fuel, various innovative extractants and methods are being studied and proposed to separate actinide and lanthanide from other fission product (FP), to separate minor actinide (MA) from lanthanide, and so on. As for the transmutation of long-lived nuclides, various types of system, such as MA loading (both homogeneous and heterogeneous concepts) to a fast reactor (FR) and dedicated transmutation of MA in an accelerator-driven system (ADS), are being studied and proposed, and respective types of MA-bearing fuel are being investigated. One of problems to proceed with R and D on this technology is in the difficulty to provide and handle a certain amount of MA. To overcome this point, international collaboration to make use of facilities and MA resources is desirable. (author)

  20. Actinide-only burnup credit for spent fuel transport

    A conservative methodology is described that would allow taking credit for burn up in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burn up verification by measurement. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k. (author)

  1. Actinides reduction by recycling in a thermal reactor; Reduccion de actinidos por reciclado en un reactor termico

    Ramirez S, J. R.; Martinez C, E.; Balboa L, H., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This work is directed towards the evaluation of an advanced nuclear fuel cycle in which radioactive actinides could be recycled to remove most of the radioactive material; firstly a production reference of actinides in standard nuclear fuel of uranium at the end of its burning in a BWR reactor is established, after a fuel containing plutonium is modeled to also calculate the actinides production in MOX fuel type. Also it proposes a design of fuel rod containing 6% of actinides in a matrix of uranium from the tails of enrichment, then four standard uranium fuel rods are replaced by actinides rods to evaluate the production and transmutation thereof, the same procedure was performed in the fuel type MOX and the end actinide reduction in the fuel was evaluated. (Author)

  2. Production and measurement of minor actinides in the commercial fuel cycle

    Stanbro, W.D. [comp.

    1997-03-01

    The minor actinide elements, particularly neptunium and americium, are produced as a normal byproduct of the operation of thermal power reactors. Because of the existence of long-lived isotopes of these elements, they constitute the major sources of the residual radiation in spent fuel or in wastes resulting from reprocessing. This has led to examinations by some countries of the possibility of separating the minor actinides from waste products. The papers found in this report address the production of minor actinides in common thermal power reactors as well as approaches to measure these materials in various media. The first paper in this volume, {open_quotes}Production of Minor Actinides in the Commercial Fuel Cycle,{close_quotes} uses calculations with the ORIGEN2 reactor and decay code to estimate the amounts of minor actinides in spent fuel and separated plutonium as a function of reactor irradiation and the time after discharge. The second paper, {open_quotes}Destructive Assay of Minor Actinides,{close_quotes} describes a number of promising approaches for the chemical analysis of minor actinides in the various forms in which they are found at reprocessing plants. The next paper, {open_quotes}Hybrid KED/XRF Measurement of Minor Actinides in Reprocessing Plants,{close_quotes} uses the results of a simulation model to examine the possible applications of the hybrid KED/XRF instrument to the determination of minor actinides in some of the solutions found in reprocessing plants. In {open_quotes}Calorimetric Assay of Minor Actinides,{close_quotes} the authors show some possible extensions of this powerful technique beyond the normal plutonium assays to include the minor actinides. Finally, the last paper in this volume, {open_quotes}Environment Measurements of Transuranic Nuclides,{close_quotes} discusses what is known about the levels of the minor actinides in the environment and ways to analyze for these materials in environmental matrices.

  3. Analysis of the minority actinides transmutation in a sodium fast reactor with uniform load pattern by the MCNPX-CINDER code; Analisis de la transmutacion de actinidos en un reactor rapido de sodio con modelo de carga homogeneo mediante el codigo MCNPX-CINDER

    Ochoa Valero, R.; Garcia-Herranz, N.; Aragones, J. M.

    2010-07-01

    The aim of this study is to evaluate the minority actinides transmutation in sodium fast reactors (SFR) assuming a uniform load pattern. It is determined the isotopic evolution of the actinides along burn, and the evolution of the reactivity and the reactivity coefficients. For that, it is used the MCNPX neutron transport code coupled with the inventory code CINDER90.

  4. Review of actinide nitride properties with focus on safety aspects

    This report provides a review of the potential advantages of using actinide nitrides as fuels and/or targets for nuclear waste transmutation. Then a summary of available properties of actinide nitrides is given. Results from irradiation experiments are reviewed and safety relevant aspects of nitride fuels are discussed, including design basis accidents (transients) and severe (core disruptive) accidents. Anyway, as rather few safety studies are currently available and as many basic physical data are still missing for some actinide nitrides, complementary studies are proposed. (author)

  5. Review of actinide nitride properties with focus on safety aspects

    Albiol, Thierry [CEA Cadarache, St Paul Lez Durance Cedex (France); Arai, Yasuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    This report provides a review of the potential advantages of using actinide nitrides as fuels and/or targets for nuclear waste transmutation. Then a summary of available properties of actinide nitrides is given. Results from irradiation experiments are reviewed and safety relevant aspects of nitride fuels are discussed, including design basis accidents (transients) and severe (core disruptive) accidents. Anyway, as rather few safety studies are currently available and as many basic physical data are still missing for some actinide nitrides, complementary studies are proposed. (author)

  6. Actinide co-conversion by internal gelation

    Suitable microstructures and homogenous microspheres of actinide compounds are of interest for future nuclear fuel or transmutation target concepts to prevent the generation and dispersal of actinide powder. Sol-gel routes are being investigated as one of the possible solutions for producing these compounds. Preliminary work is described involving internal gelation to synthesize mixed compounds including minor actinides, particularly mixed actinide or mixed actinide-inert element compounds. A parameter study is discussed to highlight the importance of the initial broth composition for obtaining gel microspheres without major defects (cracks, craters, etc.). In particular, conditions are defined to produce gel beads from Zr(IV)/Y(III)/Ce(III) or Zr(IV)/An(III) systems. After gelation, the heat treatment of these microspheres is described for the purpose of better understanding the formation of cracks after calcination and verifying the effective synthesis of an oxide solid-solution. (authors)

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

    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)

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

    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. Heavy ion induced damage in MgAl sub 2 O sub 4 , an inert matrix candidate for the transmutation of minor actinides

    Wiss, T

    1999-01-01

    Magnesium aluminum spinel (MgAl sub 2 O sub 4) is a material selected as a possible matrix for transmutation of minor actinides by neutron capture or fission in nuclear reactors. To study the radiation stability of this inert matrix, especially against fission product impact, irradiations with heavy energetic ions or clusters have been performed. The high electronic energy losses of the heavy ions in this material led to the formation of visible tracks as evidenced by transmission electron microscopy for 30 MeV C sub 6 sub 0 -Buckminster fullerenes and for ions of energy close to or higher than fission energy ( sup 2 sup 0 sup 9 Bi with 120 MeV and 2.38 GeV energy). The irradiations at high energies showed a pronounced degradation of the spinel. Additionally, MgAl sub 2 O sub 4 exhibited a large swelling for irradiation at high fluences with fission products of fission energy (here I-ions of 72 MeV) and at temperatures <= 500 deg. C. These observations are discussed from the technological point of view in ...

  10. Fusion-Fission Burner for Transuranic Actinides

    Choi, Chan

    2013-10-01

    The 14-MeV DT fusion neutron spectrum from mirror confinement fusion can provide a unique capability to transmute the transuranic isotopes from light water reactors (LWR). The transuranic (TRU) actinides, high-level radioactive wastes, from spent LWR fuel pose serious worldwide problem with long-term decay heat and radiotoxicity. However, ``transmuted'' TRU actinides can not only reduce the inventory of the TRU in the spent fuel repository but also generate additional energy. Typical commercial LWR fuel assemblies for BWR (boiling water reactor) and PWR (pressurized water reactor) measure its assembly lengths with 4.470 m and 4.059 m, respectively, while its corresponding fuel rod lengths are 4.064 m and 3.851 m. Mirror-based fusion reactor has inherently simple geometry for transmutation blanket with steady-state reactor operation. Recent development of gas-dynamic mirror configuration has additional attractive feature with reduced size in central plasma chamber, thus providing a unique capability for incorporating the spent fuel assemblies into transmutation blanket designs. The system parameters for the gas-dynamic mirror-based hybrid burner will be discussed.

  11. Production of actinide isotopes in simulated PWR fuel and their influence on inherent neutron emission

    This report describes calculations that examine the sensitivity of actinide isotopes to various reactor parameters. The impact of actinide isotope build-up, depletion, and decay on the neutron source rate in a spent-fuel assembly is determined, and correlations between neutron source rates and spent-fuel characteristics such as exposure, fissile content, and plutonium content are established. The application of calculations for evaluating experimental results is discussed

  12. Accelerator for nuclear transmutation

    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

  13. HELIOS: the new design of the irradiation of U-free fuels for americium transmutation

    D' Agata, E. [European Commission, Joint Research Centre, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Klaassen, F.; Sciolla, C. [Nuclear Research and Consultancy Group, Dept. Life Cycle and Innovations, P.O. Box 25 1755 ZG Petten (Netherlands); Fernandez-Carretero, A. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Bonnerot, J.M. [Commissariat a l' Energie Atomique, DEC/SESC/LC2I CEA-Cadarache, 13108 St. Paul lez Durance Cedex (France)

    2009-06-15

    Americium is one of the radioactive elements that mostly contribute to the radiotoxicity of the nuclear spent fuel. Transmutation of long-lived nuclides like Americium is an option for the reduction of the mass, the radiotoxicity and the decay heat of nuclear waste. The HELIOS irradiation experiment is the last evolution in a series of experiments on americium transmutation. The previous experiments, EFTTRA-T4 and T4bis, have shown that the release or trapping of helium is the key issue for the design of such kind of target. In fact, the production of helium, which is characteristic of {sup 241}Am transmutation, is quite significant. The experiment is carried out in the framework of the 4-year project EUROTRANS of the EURATOM 6. Framework Programme (FP6). Therefore, the main objective of the HELIOS experiment is to study the in-pile behaviour of U-free fuels such as CerCer (Pu, Am, Zr)O{sub 2} and Am{sub 2}Zr{sub 2}O{sub 7}+MgO or CerMet (Pu, Am)O{sub 2}+Mo in order to gain knowledge on the role of the fuel microstructure and of the temperature on the gas release and on the fuel swelling. The experiment was planned to be conducted in the HFR (High Flux Reactor) in Petten (The Netherlands) starting the first quarter of 2007. Because of the innovative aspects of the fuel, the fabrication has had some delays as well as the final safety analyses of the original design showed some unexpected deviation. Besides, the HFR reactor has been unavailable since August 2008. Due to the reasons described above, the experiment has been postponed. HELIOS should start in the first quarter of 2009 and will last 300 full power days. The paper will cover the description of the new design of the irradiation experiment HELIOS. The experiment has been split in two parts (HELIOS1 and HELIOS2) which will be irradiated together. Moreover, due to the high temperature achieved in cladding and to the high amount of helium produced during transmutation the experiment previously designed for a

  14. Calculations of different transmutation concepts. An international benchmark exercise

    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)

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

    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)

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

    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)

  17. Partitioning and transmutation of nuclear wastes. Chances and risk in research and application

    Partitioning and transmutation is focused on the transformation of long-lived radioisotopes in short-lived isotopes. The methodology could be a possibility to reduce the long.-term risk of heat developing nuclear waste in final repositories. During partitioning of spent fuel elements the uranium, plutonium and the minor actinides (neptunium, americium and curium) are separated. The remaining fission and activation products are vitrified and disposed in the final repository. During the partition process radioactive water from decontamination and washing is generated as secondary waste. The transmutation process includes the irradiation of plutonium and the minor actinides with fast neutrons resulting in stable or short-lived isotopes. The separated uranium can be used for fuel element production. The facility for transmutation is being developed and is supposed to be safer than the actual nuclear power plants. The potential risks of the technology are discussed.

  18. Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

    High level radioactive waste disposal is an issue of great importance in the discussion of the sustainability of nuclear power generation. The main contributors to the high radioactivity are the fission products and the minor actinides. The long lived fission products and minor actinides set severe demands on the arrangements for safe waste disposal. Fast reactors and accelerator driven systems (ADS) are under development in Member States to reduce the long term hazard of spent fuel and radioactive waste, taking advantage of their incineration and transmutation capability. Important R and D programmes are being undertaken in many Member States to substantiate this option and advance the basic knowledge in this innovative area of nuclear energy development. The conceptual design of the lead cooled fast reactor concept BREST-OD-300, as well as various other conceptual designs of lead/lead-bismuth cooled fast reactors have been developed to meet enhanced safety and non-proliferation requirements, aiming at both energy production and transmutation of nuclear waste. Some R and D studies indicate that the use of lead and lead-bismuth coolant has some advantages in comparison with existing sodium cooled fast reactor systems, e.g.: simplified design of fast reactor core and BOP, enhanced inherent safety, and easier radwaste management in related fuel cycles. Moreover, various ADS conceptual designs with lead and lead-bismuth as target material and coolant also have been pursued. The results to date are encouraging, indicating that the ADS has the potential to offer an option for meeting the challenges of the back end fuel cycle. During the last decade, there have been substantial advances in several countries with their own R and D programme in the fields of lead/lead-bismuth cooled critical and sub-critical concepts. coolant technology, and experimental validation. In this context, international exchange of information and experience, as well as international

  19. Impact evaluation on nuclear design characteristics and safety parameters by the presence of TRU in transmutation LMR fuels

    - Core performance evaluation about a transmutation fuel with TRU. - Evaluation for transmutation fuels depend on burnup in KALIMER core. - Suggestion of M.A. contents in a fuel for reactor safety. - Verification of REBUS code system about fuels with TRU material. - Analysis of Keff, Doppler Coeff., Coolant Density Coeff. by each M.A. isotope mass change and M.A. total amount change. - Major isotope in Changing reactivity : Am-242m, Cm-243, Cm-245. - When the content of M.A. increases, Keff is increased, Doppler Coefficient has less negative value and Coolant Density Coefficient in increased. - Core performance change with M.A. Could be predicted using effect indices from the results of each M.A. isotope mass change

  20. Partitioning and Transmutation. Annual Report 2003

    Andersson, S.; Ekberg, C.; Liljenzin, J.O.; Nilsson, M.; Skarnemark, G. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Materials and Surface Chemistry

    2004-02-01

    The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products and activation products. 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 obtain are the one between trivalent actinides and lanthanides, due to their relatively similar chemical properties, and the one between different actinides themselves. Solvent extraction is an efficient and well-known method that makes it possible to obtain 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 European Union project PARTNEW. This project was a part of the fifth framework programme and was concluded in September 2003, but the work is continued in the sixth framework programme under the acronym EUROPART (start January 2004). We mainly cooperate with the Univ. of Reading, which send us new nitrogen containing ligands for evaluation of their extraction properties. The main focus is to understand the basic chemistry of these systems but also to study some process behaviour for future full-scale plants.

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

    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)

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

    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 UO2 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. Transmutation Technology Development

    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. The effects of actinide based fuels on incremental cross sections in a Candu reactor

    The reprocessing of spent fuel such as the extraction of actinide materials for use in mixed oxide fuels is a key component of reducing the end waste from nuclear power plant operations. Using recycled spent fuels in current reactors is becoming a popular option to help close the fuel cycle. In order to ensure safe and consistent operations in existing facilities, the properties of these fuels must be compatible with current reactor designs. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU reactor. Specifically, the effect of this fuel design on the incremental cross sections related to the use of adjuster rods is investigated. The actinide concentrations studied in this work were based on extraction from thirty year cooled spent fuel and mixed with natural uranium to yield a MOX fuel of 4.75% actinide by weight. The incremental cross sections were calculated using the DRAGON neutron transport code. The results for the actinide fuel were compared to those for standard natural uranium fuel and for a slightly enriched (1% U-235) fuel designed to reduce void reactivity. Adjuster reactivity effect calculations and void reactivity simulations were also performed. The impact of the adjuster on reactivity decreased by as much as 56% with TRUMOX fuel while the CVR was reduced by 71% due to the addition of central burnable poison. The incremental cross sections were largely affected by the use of the TRUMOX fuel primarily due to its increased level of fissile material (five times that of NU). The largest effects are in the thermal neutron group where the ΣT value is increased by 46.7%, the Σny) values increased by 13.0% and 9.9%. The value associated with thermal fission, υΣf, increased by 496.6% over regular natural uranium which is expected due to the much higher reactivity of the fuel. (author)

  5. Partitioning and transmutation. Annual report 2007

    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, 94N 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

  6. Partitioning and transmutation. Annual report 2007

    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.

  7. Partitioning and Transmutation. Annual Report 2005

    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.

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

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

    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.

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

    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

  11. Actinides and fissions products in the spent nuclear fuel from slovak NPPs

    Five nuclear units with reactor WWER-440 are in this time in operation in Slovakia - four units in Bohunice and one in Mochovce sites. Start-up of another unit can be expected in Mochovce soon. Spent nuclear fuel was considered as a waste with final disposal in deep repository. Characterization and evaluation of burned-up is now under process of changes in connection with development of Accelerator Driven Transmutation Technology systems. Applicability of transmutation technology in particular nuclear economy depends on the amount and composition of burned fuel assemblies. This study deals with evaluation of Slovak WWER-440 spent fuel composition based on ORIGEN calculation. The paper provides calculation results of WWER-440 (1.6%, 2.4%, 3.6%, 3.82% enrichment) spent fuel inventory. Amounts and activities of important nuclides are tabled. (Authors)

  12. Irradiation effects on SiAlO(N) rare earth aluminosilicate glasses in the framework of actinides transmutation; Effets de l'irradiation sur les verres d'aluminosilicates de terres rares de type SiAlO(N) dans le contexte de la transmutation des actinides

    Dauce, R

    2003-11-15

    Actinides transmutation would permit to decrease the amount of waste to be dispose in deep geological site. However, a surrounding matrix is generally necessary after the separation of the radionuclides. Reference ceramics irradiations in the context of transmutation have been widely investigated, but no study have been performed on amorphous materials in the same conditions. The extensive study of glass evolution under heavy-ions bombardment can however permit to get insight damaging mechanisms during irradiation. The glassy compositions, which are SiAlO(N) type, were chosen for their refractoriness, their high chemical durability and excellent mechanical properties. Five compositions, in the Y-Mg-Si-Al-O(-N), Nd-Mg-Si-Al-O(-N) and La-Y-Al-O-N systems, were synthesized and characterized. A link is find between the structure of glasses and their deformation mechanism. The glasses were irradiated at GANIL (Caen), with several MeV energy heavy-ions. Their hardness decrease after bombardment, in close link with the electronic stopping power, but seems to be independent of the amount and nature of the network modifiers. This hardness decrease is more pronounced in the case of nitrogen containing glasses, and is due to a change in the glass deformation mechanism under indentation. The pristine glasses exhibit a 'normal' behavior, but the irradiated glasses are strained mainly by a densification mechanism. This change in the indentation behavior is probably due to several structural modifications. Indeed, UV-visible absorption spectroscopy shows the presence of a large amount of point defects after bombardment. Furthermore, particularly in the case of nitrogen containing glasses, the local environment of aluminum and silicon are largely disturbed, as shown by NMR and Raman spectroscopies. (author)

  13. Transmutation of high level wastes in a fusion-driven transmuter (FDT)

    This study presents the transmutations of both the minor actinides (MAs: 237Np, 241Am, 243Am and 244Cm) and the long-lived fission products (LLFPs: 99Tc, 129I and 135Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZMA and TZFP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fraction of the MA is raised from 10 to 20% stepped by 2%. The MAs are cladded with the graphite (10%) and cooled with the high-pressured helium gas for nuclear heat transfer. The volume fraction of helium is reduced from 80 to 70% depending on that of MA. Furthermore, the volume fraction of graphite is raised from 10 to 80% stepped by 5% to slow down the energy of neutrons entering into the TZFP while the volume fraction of LLFP is reduced from 80 to 10% depending on the graphite volume fraction. The calculations are performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 5 MW/m2 to estimate neutronic parameters and transmutation characteristics per D-T fusion neutron. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the 99Tc nuclide among them has the highest transmutation rate

  14. System and safety studies of accelerator driven systems and generation IV reactors for transmutation of minor actinides. Annual report 2009

    During 2009, the reactor physics division has made a design study of a source efficient ADS with nitride fuel and 15/15Ti cladding, based on the EFIT design made within the EUROTRANS project. It was shown that the source efficiency may be doubled as compared to the reference design with oxide fuel and T91 cladding. Transient analysis of a medium sized sodium cooled reactor with MOX fuel allowed to define criteria in terms of power penalty for americium introduction. It was shown that for each percent of americium added to the fuel, the linear rating must be reduced by 6% in order for the fuel to survive postulated unprotected transients. The Sjoestrand area ratio method for reactivity determination has been evaluated experimentally in the strongly heterogeneous subcritical facility YALINA-Booster. Surprisingly, it has been found that the area ratio reactivity estimates may differ by a factor of two depending on detector position. It is shown that this strong spatial dependence can be explained based on a two-region point kinetics model and rectified by means of correction factors obtained through Monte Carlo simulations. For the purpose of measuring high energy neutron cross sections at the SCANDAL facility in Uppsala, Monte Carlo simulations of neutron to proton conversion efficiencies in CsI detectors have been performed. A uranium fuel fabrication laboratory has been taken into operation at KTH in 2009. Uranium and zirconium nitride powders have been fabricated by hydridation/nitridation of metallic source materials. Sample pellets have been pressed and ZrN discs have been sintered to 93% density by means of spark plasma sintering methods

  15. System and safety studies of accelerator driven systems and generation IV reactors for transmutation of minor actinides. Annual report 2009

    Bergloef, Calle; Fokau, Andrei; Jolkkonen, Mikael; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))

    2010-03-15

    During 2009, the reactor physics division has made a design study of a source efficient ADS with nitride fuel and 15/15Ti cladding, based on the EFIT design made within the EUROTRANS project. It was shown that the source efficiency may be doubled as compared to the reference design with oxide fuel and T91 cladding. Transient analysis of a medium sized sodium cooled reactor with MOX fuel allowed to define criteria in terms of power penalty for americium introduction. It was shown that for each percent of americium added to the fuel, the linear rating must be reduced by 6% in order for the fuel to survive postulated unprotected transients. The Sjoestrand area ratio method for reactivity determination has been evaluated experimentally in the strongly heterogeneous subcritical facility YALINA-Booster. Surprisingly, it has been found that the area ratio reactivity estimates may differ by a factor of two depending on detector position. It is shown that this strong spatial dependence can be explained based on a two-region point kinetics model and rectified by means of correction factors obtained through Monte Carlo simulations. For the purpose of measuring high energy neutron cross sections at the SCANDAL facility in Uppsala, Monte Carlo simulations of neutron to proton conversion efficiencies in CsI detectors have been performed. A uranium fuel fabrication laboratory has been taken into operation at KTH in 2009. Uranium and zirconium nitride powders have been fabricated by hydridation/nitridation of metallic source materials. Sample pellets have been pressed and ZrN discs have been sintered to 93% density by means of spark plasma sintering methods

  16. Research on the actinide chemistry in Nuclear Fuel Cycle

    Fundamental technique to measure chemical behaviors and properties of lanthanide and actinide in radioactive waste is necessary for the development of pryochemical process. First stage, the electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipments, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media. In the second stage, measurement system for physical properties at pyrochemical process such as viscosity, melting point and conductivity is established, and property database at different compositions of lanthanide and actinide is collected. And, both interactions between elements and properties with different potential are measured at binary composition of actinide-lanthanide in molten salt using electrochemical/spectroscopic integrated measurement system.

  17. Status of nuclear transmutation study

    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)

  18. Releases from exotic waste packages from partitioning and transmutation

    Lee, W.W.L. [Lawrence Berkeley Lab., CA (United States); Choi, J.S. [Lawrence Livermore National Lab., CA (United States)

    1991-09-01

    Partitioning the actinides in spent nuclear fuel and transmuting them in actinide-burning liquid-metal reactors has been proposed as a potential method of reducing the public risks from geologic disposal of nuclear waste. To quantify the benefits for waste disposal of actinide burning, we calculate the release rates of key radionuclides from waste packages resulting from actinide burning, and compare them with release rates from LWR spent fuel destined for disposal at the potential repository at Yucca Mountain. The wet-drip water-contact mode has been used. Analytic methods and parameter values are very similar to those used for assessing Yucca Mountain as a potential repository. Once released, the transport characteristics of radionuclides will be largely determined by site geology. For the most important nuclides such as I-129 and {Tc}-99, which are undiminished by actinide-burning reactors, it is not surprising that actinide burning offers little reduction in releases. For important actinides such as Np-237 and Pu isotopes, which are reduced in inventory, the releases are not reduced because the release rates are proportional to solubility, rather than inventory.

  19. Release rates from partitioning and transmutation waste packages

    Partitioning the actinides in light-water reactor spent fuel and transmuting them in actinide-burning liquid-metal reactors has been proposed as a potential method for reducing the public risks from geologic disposal of nuclear waste. As a first step towards quantifying the benefits for waste disposal of actinide burning, we have calculated the release rates of key radionuclides from waste packages resulting from actinide burning, and compare them with release rates from LWR spent fuel destined for disposal at the potential repository at Yucca Mountain. The wet-drip water-contact mode has been used. Analytic methods and parameter values are very similar to those used for assessing Yucca Mountain as a potential repository. Once released, the transport characteristics of radionuclides will be largely determined by site geology. For the most important nuclides such as I-129 and Tc-99, which are undiminished by actinide-burning reactors, it is not surprising that actinide burning offers little reduction in releases. For important actinides such as Np-237 and Pu isotopes, which are reduced in inventory, the releases are not reduced because the release rates are proportional to solubility, rather than inventory

  20. Motivation for transmuting long-lived radioactive products

    In the Netherlands the efforts on waste transmutation are coordinated in a research programme called RAS. One of the aims of this RAS program is to inform the public and advise the authorities on methods for transmutation/conditioning of nuclear waste, and on techniques which are being developed. Any new way to treat waste should of course not lead to significant risks for the present population. Small risks might be accepted, but these should sufficiently be compensated for. Benefits for the present generation are related to the better exploitation of the full energy content of the actinides, which will reduce fuel costs and waste streams from mining as well as from spent fuel. Future generations might profit from the fact that the waste has been cleaned from actinides and that proliferation risks are eliminated. Another benefit could be that transmutation also could lead to a reduction of dose-risks by leakage of mobile elements such as Rn-222 and the metalloid fission products like technetium and iodine. It is shown in this paper that the balance of benefits and risks is quite different for long-lived fission products than for actinides. (author)

  1. Transmutation capabilities of generation 4 reactors

    The Generation IV reactors all have the potential to play a significant role in future scenarios dealing with transmutation of spent fuel from LWR power reactors. The nature of the flux spectrum, thermal or fast, is the major factor in the effectiveness of transmuting various transuranic isotopes. We conclude that each Generation IV reactor concept could have a role, if properly co-ordinated and supported by significant development programmes. The fast reactor concepts (liquid metal and gas-cooled) are the most effective in consumption by fission of unwanted actinides (plutonium, neptunium, americium and possibly curium). Thermal spectrum concepts (water-cooled reactors with and without inert-matrix cores, high-temperature gas-cooled reactors with and without inert-matrix cores, and liquid-salt-cooled thermal reactors) all can potentially reduce some of the minor actinides, even if only used in a single pass. When teamed up with subsequent fast-reactor irradiations to reduce higher minor actinides (specifically americium and curium), their use could result in reducing the number of fast burner reactors required, per spent-fuel-producing LWR, compared to a system of only LWRs and fast burner reactors. After listing the six main Generation IV candidates with attributes, benefits and viability concerns, this presentation will focus on one example of fast spectrum systems and two thermal spectrum systems to indicate transmuting capabilities of both types of systems. These will be used for illustrative purposes only and are not meant to give any indication of the relative importance of these systems to concepts not mentioned. Likewise, the figures and graphs in this paper are presented without alteration from the originators (see acknowledgements), and are for illustration purposes only. (authors)

  2. Actinide Partitioning-Transmutation Program Final Report. V. Preconceptual designs and costs of partitioning facilities and shipping casks (appendix 3)

    This Appendix contains cost estimate documents for the Fuels Reprocessing Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contribution to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed

  3. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided

  4. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided.

  5. Transmutation of high-level radioactive waste - Perspectives

    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.

  6. Radiation aspects of fuel handling at various variants of fuel cycle organization

    A series of radiation characteristics of fresh and spent fuel assemblies with mixed uranium-plutonium oxide (MO) fuel of high-power BN-type reactor with limited content of minor actinides in origin fuel is obtained. The assessment is given for possible radiation characteristics of experimental MO fuel assemblies simulating transmutation of minor actinides (Np, Am, Cm). The obtained results can be used for solving practical problems related with minor actinides transmutation in high-power reactor. In particular, the conclusion is made that curium transmutation is impossible (especially 224Cm) in fuel assemblies of reactor under consideration due to high level of power distribution and neutron radiation of fresh MO fuel assemblies with curium

  7. Feasibility studies of actinide recycle in LMFBRs as a waste management alternative

    A strategy of actinide burnup in LMFBRs is being investigated as a waste management alternative to long term storage of high level nuclear waste. This strategy is being evaluated because many of the actinides in the waste from spent-fuel reprocessing have half-lives of thousands of years and an alternative to geological storage may be desired. From a radiological viewpoint, the actinides and their daughters dominate the waste hazard for decay times beyond about 400 years. Actinide burnup in LMFBRs may be an attractive alternative to geological storage because the actinides can be effectively transmuted to fission products which have significantly shorter half-lives. Actinide burnup in LMFBRs rather than LWRs is preferred because the ratio of fission reaction rate to capture reaction rate for the actinides is higher in an LMFBR, and an LMFBR is not so sensitive to the addition of the actinide isotopes. An actinide target assembly recycle scheme is evaluated to determine the effects of the actinides on the LMFBR performance, including local power peaking, breeding ratio, and fissile material requirements. Several schemes are evaluated to identify any major problems associated with reprocessing and fabrication of recycle actinide-containing assemblies. The overall efficiency of actinide burnout in LMFBRs is evaluated, and equilibrium cycle conditions are determined. It is concluded that actinide recycle in LMFBRs offers an attractive alternative to long term storage of the actinides, and does not significantly affect the performance of the host LMFBR. Assuming a 0.1 percent or less actinide loss during reprocessing, a 0.1 percent loss of less during fabrication, and proper recycle schemes, virtually all of the actinides produced by a fission reactor economy could be transmuted in fast reactors

  8. Fertile-Free Fast Lead-Cooled Incinerators for Efficient Actinide Burning

    Fertile-free fast lead-cooled modular reactors are proposed as efficient incinerators of plutonium and minor actinides (MAs) for application to advanced fuel cycles devoted to transmutation. Two concepts are presented: (1) an actinide burner reactor, designed to incinerate mostly plutonium and some MAs, and (2) a minor actinide burner reactor, devoted to burning mostly minor actinides and some plutonium. These transuranics are loaded in a fertile-free Zr-based metallic fuel to maximize the incineration rate. Both designs feature streaming fuel assemblies that enhance neutron leakage to achieve favorable neutronic feedback and a double-entry control rod system that reduces reactivity perturbations during seismic events and flattens the axial power profile. A detailed neutronic analysis shows that both designs have favorable neutronic characteristics and reactivity feedback mechanisms that yield passive safety features comparable to those of the Integral Fast Reactor. A safety analysis presents the response of the burners to anticipated transients without scram on the basis of (1) the integral parameter approach and (2) simulations of thermal-hydraulic accident scenario conditions. It is shown that both designs have large thermal margins that lead to safe shutdown without structural damage to the core components for a large spectrum of unprotected transients. Furthermore, the actinide destruction rates are comparable to those of the accelerator transmutation of waste concept, and a fuel cycle cost analysis shows the potential for economical accomplishment of the transmutation mission compared to other proposed actinide-burning options

  9. Overview of activities on Pu and minor actinides fuel research in JAERI

    Arai, Yasuo; Yamashita, Toshiyuki [Japan Atomic Energy Research Inst., Tokai (Japan)

    1997-12-31

    Recent activities on Pu and minor actinides fuel research in JAERI is summarized. For oxide fuel, the solid state chemistry on U-Np-Pu-O system has been investigated. Further, Pu rock-like fuel has been developed from the viewpoint of disposing excess plutonium. For nitride fuel, research on fuel fabrication, property measurements, irradiation behavior and application to pyrochemical reprocessing has been carried out. These studies aim at contributing to the development of advanced fuel cycle and innovative fuel cycle toward the 21st century. (author). 25 refs.

  10. Impact of partitioning and transmutation on repository design

    The U.S. Department of Energy's Advanced Fuel Cycle Initiative (AFCI) program is investigating spent nuclear fuel treatment technologies that have the potential to improve the performance of the proposed geologic repository at Yucca Mountain. Separating actinides and selected fission products from spent fuel, storing some of them as low level waste and transmuting them in thermal and/or fast reactors has the potential to reduce the volume, short and long-term heat load and radiotoxicity of the high level waste destined for the repository, effectively increasing its capacity by a factor of 50 or more above the current legislative limit. (author)

  11. Nuclear Methods for Transmutation of Nuclear Waste: Problems, Perspextives, Cooperative Research - Proceedings of the International Workshop

    Khankhasayev, Zhanat B.; Kurmanov, Hans; Plendl, Mikhail Kh.

    1996-12-01

    The Table of Contents for the full book PDF is as follows: * Preface * I. Review of Current Status of Nuclear Transmutation Projects * Accelerator-Driven Systems — Survey of the Research Programs in the World * The Los Alamos Accelerator-Driven Transmutation of Nuclear Waste Concept * Nuclear Waste Transmutation Program in the Czech Republic * Tentative Results of the ISTC Supported Study of the ADTT Plutonium Disposition * Recent Neutron Physics Investigations for the Back End of the Nuclear Fuel Cycle * Optimisation of Accelerator Systems for Transmutation of Nuclear Waste * Proton Linac of the Moscow Meson Factory for the ADTT Experiments * II. Computer Modeling of Nuclear Waste Transmutation Methods and Systems * Transmutation of Minor Actinides in Different Nuclear Facilities * Monte Carlo Modeling of Electro-nuclear Processes with Nonlinear Effects * Simulation of Hybrid Systems with a GEANT Based Program * Computer Study of 90Sr and 137Cs Transmutation by Proton Beam * Methods and Computer Codes for Burn-Up and Fast Transients Calculations in Subcritical Systems with External Sources * New Model of Calculation of Fission Product Yields for the ADTT Problem * Monte Carlo Simulation of Accelerator-Reactor Systems * III. Data Basis for Transmutation of Actinides and Fission Products * Nuclear Data in the Accelerator Driven Transmutation Problem * Nuclear Data to Study Radiation Damage, Activation, and Transmutation of Materials Irradiated by Particles of Intermediate and High Energies * Radium Institute Investigations on the Intermediate Energy Nuclear Data on Hybrid Nuclear Technologies * Nuclear Data Requirements in Intermediate Energy Range for Improvement of Calculations of ADTT Target Processes * IV. Experimental Studies and Projects * ADTT Experiments at the Los Alamos Neutron Science Center * Neutron Multiplicity Distributions for GeV Proton Induced Spallation Reactions on Thin and Thick Targets of Pb and U * Solid State Nuclear Track Detector and

  12. MA-burners efficiency parameters allowing for the duration of transmutation process

    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 nrep - 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)

  13. Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

    Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

    2003-06-01

    The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

  14. Radiolytic degradation of a new diglycol-diamide ligand for actinide and lanthanide co-extraction from spent nuclear fuel

    Ossola, Annalisa; Macerata, Elena; Tinonin, Dario A.; Faroldi, Federica; Giola, Marco; Mariani, Mario; Casnati, Alessandro

    2016-07-01

    Within the Partitioning and Transmutation strategies, great efforts have been devoted in the last decades to the development of lipophilic ligands able to co-extract trivalent Lanthanides (Ln) and Actinides (An) from spent nuclear fuel. Because of the harsh working conditions these ligands undergo, it is important to prove their chemical and radiolytic stability during the counter-current multi-stage extraction process. In the present work the hydrolytic and radiolytic resistance of the freshly prepared and aged organic solutions containing the new ligand (2,6-bis[(N-methyl-N-dodecyl)carboxamide]-4-methoxy-tetrahydro-pyran) were investigated in order to evaluate the impact on the safety and efficiency of the process. Liquid-liquid extraction tests with spiked solutions showed that the ligand extracting performances are strongly impaired by storing the samples at room temperature and in the light. Moreover, the extracting efficiency of the irradiated samples resulted to be influenced by gamma irradiation, while selectivity remains unchanged. Preliminary mass spectrometric data showed that degradation is mainly due to the acid-catalysed reaction of the ligand carboxamide and ether groups with the 1-octanol present in the diluent.

  15. Neutronic and burnup studies of accelerator-driven systems dedicated to nuclear waste transmutation

    Tucek, Kamil

    2004-01-01

    Partitioning and transmutation of plutonium, americium, and curium is inevitable if the radiotoxic inventory of spent nuclear fuel is to be reduced by more than a factor of 100. But, admixing minor actinides into the fuel severely degrades system safety parameters, particularly coolant void reactivity, Doppler effect, and (effective) delayed neutron fractions. The incineration process is therefore envisioned to be carried out in dedicated, accelerator-driven sub-critical reactors (ADS). Howev...

  16. Actinide burning and waste disposal

    Here we review technical and economic features of a new proposal for a synergistic waste-management system involving reprocessing the spent fuel otherwise destined for a U.S. high-level waste repository and transmuting the recovered actinides in a fast reactor. The proposal would require a U.S. fuel reprocessing plant, capable of recovering and recycling all actinides, including neptunium americium, and curium, from LWR spent fuel, at recoveries of 99.9% to 99.999%. The recovered transuranics would fuel the annual introduction of 14 GWe of actinide-burning liquid-metal fast reactors (ALMRs), beginning in the period 2005 to 2012. The new ALMRs would be accompanied by pyrochemical reprocessing facilities to recover and recycle all actinides from discharged ALMR fuel. By the year 2045 all of the LWR spent fuel now destined f a geologic repository would be reprocessed. Costs of constructing and operating these new reprocessing and reactor facilities would be borne by U.S. industry, from the sale of electrical energy produced. The ALMR program expects that ALMRs that burn actinides from LWR spent fuel will be more economical power producers than LWRs as early as 2005 to 2012, so that they can be prudently selected by electric utility companies for new construction of nuclear power plants in that era. Some leaders of DOE and its contractors argue that recovering actinides from spent fuel waste and burning them in fast reactors would reduce the life of the remaining waste to about 200-300 years, instead of 00,000 years. The waste could then be stored above ground until it dies out. Some argue that no geologic repositories would be needed. The current view expressed within the ALMR program is that actinide recycle technology would not replace the need for a geologic repository, but that removing actinides from the waste for even the first repository would simplify design and licensing of that repository. A second geologic repository would not be needed. Waste now planned

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

    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.

  18. Comparative analysis of the efficiency of minor actinide burning in the accelerator-driven system and critical reactors within various scenarios for closing the nuclear fuel cycle

    The choice of efficient types of systems for the utilization of long-lived radioactive wastes (RW) of nuclear power is one of the highest priority concerns in nuclear sphere. The accelerator-driven systems (ADS) with heavy liquid metal coolant (HLMC) and fast neutron spectrum are considered among the most efficient nuclear devices for burning minor actinides (MA). Results of numerical studies for the optimization of characteristics of an ADS-system with lead-bismuth coolant for burning minor actinides produced in the open fuel cycle of thermal reactors have been summarized. Criteria of efficiency of MA burning have been determined (time of transmutation, etc.). It has been shown that neutronic characteristics ensuring nuclear safety in an analogous critical reactor - MA burner - are significantly inferior vs. fast critical reactor with UO2 fuel. In order to define whether or not it is justified to use ADS in different scenarios for the nuclear fuel cycle closure, a comparative study has been fulfilled on radiation and technological characteristics of spent fuel from subcritical reactor ADS and on fuel from other nuclear facilities. The VVER-1000 reactor and the variant of fast reactor with lead-bismuth coolant were chosen for the comparison. SVBR-100 reactor can be considered as a prototype of the latter facility. Two options of closing the fuel cycle have been analyzed: the variant with recycling U,Pu without MA, the variant with total recycling of U and all transuranic isotopes (Pu, Np, Am, Cm). The differences have been defined in terms of specific values of radioactivity, residual heat release, intensity of sources of neutrons and gamma-radiation of spent fuel. (author)

  19. Minior Actinide Doppler Coefficient Measurement Assessment

    Nolan E. Hertel; Dwayne Blaylock

    2008-04-10

    The "Minor Actinide Doppler Coefficient Measurement Assessment" was a Department of Energy (DOE) U-NERI funded project intended to assess the viability of using either the FLATTOP or the COMET critical assembly to measure high temperature Doppler coefficients. The goal of the project was to calculate using the MCNP5 code the gram amounts of Np-237, Pu-238, Pu-239, Pu-241, AM-241, AM-242m, Am-243, and CM-244 needed to produce a 1E-5 in reactivity for a change in operating temperature 800C to 1000C. After determining the viability of using the assemblies and calculating the amounts of each actinide an experiment will be designed to verify the calculated results. The calculations and any doncuted experiments are designed to support the Advanced Fuel Cycle Initiative in conducting safety analysis of advanced fast reactor or acceoerator-driven transmutation systems with fuel containing high minor actinide content.

  20. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  1. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  2. Transmutation-incineration potential of transuraniums discharged from PWR-UO2 spent fuel in modified PROMETHEUS fusion reactor

    This study presents the potential of the burning and/or transmutation (B/T) of transuraniums (TRUs), discharged from the pressured water reactor PWR-UO2 spent fuel, in the modified PROMETHEUS-H fusion reactor. Two different design shapes (Models A and B) were considered. The transmutation zone (TZ), which contains the mixture of TRU nuclides (10%), was located in the modified blankets. The volume fraction of Pu in the mixture is raised from 0 to 40% stepped by 10% to determine its effect on the B/T. The fuel spheres were cladded with SiC (5%) and cooled with high-pressured helium gas (85%) for nuclear heat transfer. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m2. The results bring out that: (1) the Model B transmutes the TRUs more rapidly than the Model A; (2) the effective half-lives decrease about 20 and 40% with the increase of Pu fraction in the cases of Models A and B, respectively; (3) the M values are quite high with respect to the M value of the original PROMETHEUS fusion reactor; (4) the blankets can produce substantial electricity in situ

  3. Impact of different transmutation strategies on the risk from the radioactive waste

    A comprehensive study has been carried out with the aim of quantifying and comparing on a consistent basis the actinide-related component of the long-term risk for different transmutation systems with closed fuel cycles. The analyses are based on a direct method for calculating equilibrium fuel compositions and a scheme for evaluating nuclide-specific contributions to the risk from storing the remaining radioactive waste. The comparison is performed for systems with a wide range of characteristics including conventional reactors and different accelerator-based transmuters. The results allow to draw interesting conclusions regarding the long-term risk arising from the recycling of plutonium and minor actinides, as also the inherent long-term risks associated with the uranium-plutonium and the thorium-uranium fuel cycles. (author) 2 figs., 1 tab., 8 refs

  4. Review of Integral Experiments for Minor Actinide Management

    Spent nuclear fuel contains minor actinides (MAs) such as neptunium, americium and curium, which require careful management. This becomes even more important when mixed oxide (MOX) fuel is being used on a large scale since more MAs will accumulate in the spent fuel. One way to manage these MAs is to transmute them in nuclear reactors, including in light water reactors, fast reactors or accelerator-driven subcritical systems. The transmutation of MAs, however, is not straightforward, as the loading of MAs generally affects physics parameters, such as coolant void, Doppler and burn-up reactivity. This report focuses on nuclear data requirements for minor actinide management, the review of existing integral data and the determination of required experimental work, the identification of bottlenecks and possible solutions, and the recommendation of an action programme for international co-operation. (authors)

  5. Biogeochemistry of actinides: a nuclear fuel cycle perspective

    Through an examination of the comparative behavior of the actinide elements in terrestrial and aquatic food chains, the anticipated accumulation behavior of the transuranium elements by people was described. The available data suggests that Pu, Am, and Cm will not accumulate to a greater degree than U in the skeletons of individuals exposed to environmentally dispersed activity. The nature of the contamination event, the chemical and physical associations in soils and sediments, the proximity to the contamination site - all will influence observed behavior. Because of the establishment of regulatory guidelines for limiting chronic exposure to transuranium elements, research on environmental behavior must address the question of accumulation by people. In the absence of lifetime accumulation data and the paucity of contaminated sites, approaches such as those documented in this paper may aid in understanding and evaluating the hazards of releasing actinide elements to the biosphere

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

    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

  7. Reduction of minor actinides for recycling in a light water reactor

    The aim of actinide transmutation from spent nuclear fuel is the reduction in mass of high-level waste which must be stored in geological repositories and the lifetime of high-level waste; these two achievements will reduce the number of repositories needed, as well as the duration of storage. The present work is directed towards the evaluation of an advanced nuclear fuel cycle in which the minor actinides (Np, Am and Cm) could be recycled to remove most of the radioactive material; a reference of actinides production in standard nuclear fuel of uranium at the end of its burning in a BWR is first established, after a design of fuel rod containing 6% of minor actinides in a matrix of uranium from the enrichment lines is proposed, then 4 fuel rods of standard uranium are replaced by 4 actinides bars to evaluate the production and transmutation of them and finally the minor actinides reduction in the fuel is evaluated. In the development of this work the calculation tool are the codes: Intrepin-3, Casmo-4 and Simulate-3. (Author)

  8. Research and development activities for transmutation physics experimental facility in J-PARC

    The Japan Atomic Energy Agency (JAEA) has the plan to construct Transmutation Physics Experimental Facility (TEF-P) under a framework of J-PARC (Japan Proton Accelerator Research Complex) project. TEF-P is a critical assembly which can load Minor Actinide (MA) fuels to perform reactor physics experiments for transmutation systems such as Accelerator-Driven System (ADS) or Fast Reactor (FR). The facility can also use proton beam from the J-PARC accelerator to investigate the controllability of ADS. Current status and activities for TEF-P are described. (author)

  9. Experimental investigations of actinide release from coated fuel particles for high-temperature reactors

    The migrational behaviour of actinides in the coated fuel particles proposed for high-temperature reactors is investigated experimentally. Data are described in the framework of the diffusion model. The experimental procedures are presented and the necessary computer codes are discussed. The diffusion coefficients of the actinides - plutonium, americium and curium - as well as of the fission product cesium are derived from the experimental data by a nonlinear least squares fit procedure and are presented in the form of Arrhenius lines D = Do esup(-Q/RT) for U(Th)-O2, HTI-PyC and SiC. (orig.)

  10. Actinide recycle

    A multitude of studies and assessments of actinide partitioning and transmutation were carried out in the late 1970s and early 1980s. Probably the most comprehensive of these was a study coordinated by Oak Ridge National Laboratory. The conclusions of this study were that only rather weak economic and safety incentives existed for partitioning and transmuting the actinides for waste management purposes, due to the facts that (1) partitioning processes were complicated and expensive, and (2) the geologic repository was assumed to contain actinides for hundreds of thousands of years. Much has changed in the few years since then. A variety of developments now combine to warrant a renewed assessment of the actinide recycle. First of all, it has become increasingly difficult to provide to all parties the necessary assurance that the repository will contain essentially all radioactive materials until they have decayed. Assurance can almost certainly be provided to regulatory agencies by sound technical arguments, but it is difficult to convince the general public that the behavior of wastes stored in the ground can be modeled and predicted for even a few thousand years. From this point of view alone there would seem to be a clear benefit in reducing the long-term toxicity of the high-level wastes placed in the repository

  11. Minimization of actinide waste by multirecycling of thoriated fuels in an EPR

    2009-01-01

    This master’s thesis explores how to minimize the long-lived actinide waste that is produced in nuclear power plants by performing simulations of thoriated nuclear fuels in existing reactor designs. An European pressurized water reactor (EPR) assembly fueled with a mixture of thorium and highly enriched uranium (20% and 90% 235U) was simulated. The spent thoriated fuel is less active, and for a much shorter period of time, than uranium or uranium/plutonium fuels and less decay heat is gene...

  12. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-m), leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling

  13. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

  14. Development of fuels for the transmutation in the frame of the EFTTRA European collaboration

    The EFTTRA collaboration (Experimental Feasibility of Targets for Transmutation) between CEA (France), ECN (The Netherlands), EDF (France), FZK (Germany), IAM and ITU (European Commission), launched in 1992, has now reached its cruising speed: joint experiments for the study of materials for the transmutation have started in parallel in the Phenix fast reactor in France, and in the high flux thermal reactor HFR in the Netherlands. One of these experiments, concerning technetium and iodine, has been completed and the results published. The EFTTRA activities are described, in particular one experiment concerning the irradiation of a spinel matrix with 10% americium content. (author)

  15. Application of Monte Carlo techniques for propagation of cross section uncertainties to actinide inventory in ADS transmuters: comparison with sensitivity analysis

    A comprehensive study is performed in order to evaluate the impact of activation cross section uncertainties on the actinide composition of the irradiated fuel in representative ADS (Accelerator Driven System) irradiation scenarios. Some of the most recent sources/compilations of uncertainty data are used, and the results obtained from them compared. The ANL covariance matrices are taken as reference data for the calculations. The complete set of cross section uncertainties provided in the EAF2005 data library are also used for comparison purposes. In this study, the inventory code ACAB is used to analyze the following questions: impact of different correlation structures using fixed uncertainties/variances; effect of the irradiation time/burn-up on the concentration uncertainties; and applicability of Monte Carlo (MC) and sensitivity-uncertainty (SU) approaches for all the range of burn-up/irradiation times of interest in ADS designs. When comparing results of calculations using ANL versus EAR2005/UN uncertainty data, we found very significant differences in the concentration uncertainties. The applicability of both MC and SU approaches is found acceptable to deal with all the range of irradiation times

  16. Coordination chemistry for new actinide separation processes

    The amount of wastes and the number of chemical steps can be decreased by replacing the PUREX process extractant (TBP) by, N.N- dialkylamides (RCONR'2). Large amounts of deep underground storable wastes can be stored into sub-surface disposals if the long lived actinide isotopes are removed. Spent nuclear fuels reprocessing including the partitioning of the minor actinides Np, Am, Cm and their transmutation into short half lives fission products is appealing to the public who is not favorable to the deep underground storage of large amounts of long half lived actinide isotopes. In this paper coordination chemistry problems related to improved chemical separations by solvent extraction are presented. 2 tabs.; 4 refs

  17. Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

    Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

  18. Partitioning and transmutation. Annual report 2008

    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, 94N 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 on

  19. Partitioning and transmutation. Annual report 2009

    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

  20. Partitioning and transmutation. Annual report 2009

    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

  1. Partitioning and transmutation. Annual report 2008

    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

  2. Set up of an innovative methodology to measure on-line the incineration potential of minor actinides under very high neutron sources in the frame of the future prospects of the nuclear waste transmutation

    This work deals generally with the problem of nuclear waste management and especially with the transmutation of it to reduce considerably its radiotoxicity potential. The principal objective of this thesis is to show the feasibility to measure on-line the incineration potential of minor actinides irradiated under very high neutron flux. To realize this goal, we have developed fission micro-chambers able to operate, for the first time in the world, in saturation regime under a severe neutron flux. These new chambers use 235U as an active deposit. They were irradiated in the high flux reactor at Laue-Langevin Institute in Grenoble. The measurement of the saturation current delivered by these chambers during their irradiation for 26 days allowed to evaluate the burn-up of 235U. We have determined the neutron flux intensity of 1,6 1015 n.cm-2.s-1 in the bottom of the irradiation tube called 'V4'. The relative uncertainty of this value is less than 4 %. This is for the first time that such high neutron flux is measured with a fission chamber. To confirm this result, we have also performed independent measurements using gamma spectroscopy of irradiated Nb and Co samples. Both results are in agreement within error bars. Simple Deposit Fission Chambers (SDFC) as above were the reference of the new generation of fission chambers that we have developed in the framework of this thesis: Double Deposit Fission Chambers (DDFC). The reference active deposit was 235U. The other deposit was the actinide that we wanted to study (e.g. 237Np and 241Am). At the end of the thesis, we present some suggestions to ameliorate the operation of the DDFC to be exploited in other transmutation applications in the future. (author)

  3. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs

  4. Time Evolution of Selected Actinides in TRIGA MARK-II Fuel

    Study is made on the evolution of several actinides capable of undergoing fission or breeding available on the Malaysian Nuclear Agency (MNA) TRIGA MARK-II fuel. Population distribution of burned fuel in the MNA reactor is determined with a model developed using WIMS. This model simulates fuel conditions in the hottest position in the reactor, thus the location where most of the burn up occurs. Theoretical basis of these nuclide time evolution are explored and compared with the population obtained from our models. Good agreements are found for the theoretical time evolution and the population of Uranium-235, Uranium-236, Uranium-238 and Plutonium-239. (author)

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

    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.)

  6. Actinide recycle utilizing oxide and metallic fuel in prism

    PRISM is a modular, pool-type sodium-cooled fast reactor employing innovative, passive features to provide an extremely high level of public safety. The PRISM reactor design can accommodate both oxide and metallic fuel forms. A comparison of core design and performance of these forms is made for various options. These options include low fuel cycle cost options, maximum transuranic burning options, and the addition of rare earth elements to the fuel mix. (authors)

  7. Minor Actinide Burning in Thermal Reactors. A Report by the Working Party on Scientific Issues of Reactor Systems

    The actinides (or actinoids) are those elements in the periodic table from actinium upwards. Uranium (U) and plutonium (Pu) are two of the principal elements in nuclear fuel that could be classed as major actinides. The minor actinides are normally taken to be the triad of neptunium (Np), americium (Am) and curium (Cm). The combined masses of the remaining actinides (i.e. actinium, thorium, protactinium, berkelium, californium, einsteinium and fermium) are small enough to be regarded as very minor trace contaminants in nuclear fuel. Those elements above uranium in the periodic table are known collectively as the transuranics (TRUs). The operation of a nuclear reactor produces large quantities of irradiated fuel (sometimes referred to as spent fuel), which is either stored prior to eventual deep geological disposal or reprocessed to enable actinide recycling. A modern light water reactor (LWR) of 1 GWe capacity will typically discharge about 20-25 tonnes of irradiated fuel per year of operation. About 93-94% of the mass of uranium oxide irradiated fuel is comprised of uranium (mostly 238U), with about 4-5% fission products and ∼1% plutonium. About 0.1-0.2% of the mass is comprised of neptunium, americium and curium. These latter elements accumulate in nuclear fuel because of neutron captures, and they contribute significantly to decay heat loading and neutron output, as well as to the overall radio-toxic hazard of spent fuel. Although the total minor actinide mass is relatively small - approximately 20-25 kg per year from a 1 GWe LWR - it has a disproportionate impact on spent fuel disposal, and thus the longstanding interest in transmuting these actinides either by fission (to fission products) or neutron capture in order to reduce their impact on the back end of the fuel cycle. The combined masses of the trace actinides actinium, thorium, protactinium, berkelium and californium in irradiated LWR fuel are only about 2 parts per billion, which is far too low for

  8. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  9. Simulation study of an accelerator driven as a transmutation and energy generation system

    In twenty first century world is facing two issues of future concern, generation of clean energy and the protection of the environment. Oil price is soaring to a level of jeopardizing world economy and on the other hand burning fossil fuel is reaching to a point of endangering life of all creatures. The sole solution to compete the energy shortage is exploiting nuclear energy and other clean energy sources. The main concern of nuclear energy is long term radioactive waste. In principle, any nuclear reactor is able to burn and transmute minor actinides, but reactors with fast neutron spectrum must be preferred, as they allow a positive neutron gain throughout the burning process. A core dedicated to the transmutation of the minor actinides should be designed in order to minimize its self-production of actinides. A possible solution to these problems is represented by a subcritical system driven by an accelerator, which is able to safely bum and/or transmute actinides and long lived fission products, as it does not rely on delayed neutrons for control or power change and the reactivity feedbacks have only limited importance during transient response. In this study, an accelerator driven system based on Japanese design was simulated using MCNPX code to calculate neutron spectrum flux level, core sub-criticality and peaking factor

  10. Set up of an innovative methodology to measure on-line the incineration potential of minor actinides under very high neutron sources in the frame of the future prospects of the nuclear waste transmutation; Mise au point d'une methodologie innovante pour la mesure du potentiel d'incineration d'actinides mineurs sous des sources tres intenses de neutrons, dans la perspective de transmutation des dechets nucleaires

    Fadil, M

    2003-03-01

    This work deals generally with the problem of nuclear waste management and especially with the transmutation of it to reduce considerably its radiotoxicity potential. The principal objective of this thesis is to show the feasibility to measure on-line the incineration potential of minor actinides irradiated under very high neutron flux. To realize this goal, we have developed fission micro-chambers able to operate, for the first time in the world, in saturation regime under a severe neutron flux. These new chambers use {sup 235}U as an active deposit. They were irradiated in the high flux reactor at Laue-Langevin Institute in Grenoble. The measurement of the saturation current delivered by these chambers during their irradiation for 26 days allowed to evaluate the burn-up of {sup 235}U. We have determined the neutron flux intensity of 1,6 10{sup 15} n.cm{sup -2}.s{sup -1} in the bottom of the irradiation tube called 'V4'. The relative uncertainty of this value is less than 4 %. This is for the first time that such high neutron flux is measured with a fission chamber. To confirm this result, we have also performed independent measurements using gamma spectroscopy of irradiated Nb and Co samples. Both results are in agreement within error bars. Simple Deposit Fission Chambers (SDFC) as above were the reference of the new generation of fission chambers that we have developed in the framework of this thesis: Double Deposit Fission Chambers (DDFC). The reference active deposit was {sup 235}U. The other deposit was the actinide that we wanted to study (e.g. {sup 237}Np and {sup 241}Am). At the end of the thesis, we present some suggestions to ameliorate the operation of the DDFC to be exploited in other transmutation applications in the future. (author)

  11. National R and D program of nuclear incineration of PWR spent fuel in a transmuter with liquid fuel as being developed in the Czech Republic

    The principal drawbacks of any kind of solid nuclear fuel are listed and briefly analysed in the first part of the paper. On the basis of this analysis, the liquid fuel concept and its benefits are introduced and briefly described in the following parts of the paper allowing to develop new reactor systems for nuclear incineration of spent fuel from conventional reactors and a new clean source of energy. As one of the first realistic attempts to utilise the advantages of liquid fuel, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT), has been proposed for a deeper, both theoretical and experimental studies. There will be a preliminary design concept of an experimental assembly LA-0 briefly introduced in the paper which is under preparation in the Czech Republic for such a project

  12. Efttra irradiation experiments for the development of fuels and targets for transmutation

    The EFTTRA collaboration (Experimental Feasibility of Targets for Transmutation) between CEA (France), ECN (The Netherlands), EDF (France), FZK (Germany), IAM and ITU (European Commission) was launched in 1992, with the aim of performing joint experiments for the study of materials for the transmutation. Irradiations have started in parallel in the Phenix fast reactor in France, and in the high flux thermal reactor HFR the Netherlands. One of these experiments, concerning technetium and iodine, has been completed; post-irradiation examinations of the Tc metallic samples are performed by ECN, CEA and ITU, and a summary of the last results is presented. The other on-going EFTTRA experiments are described, with a report on the application of fabrication methods for matrices with up to 10% americium content. Finally, some considerations on the strategies for americium are given. (authors)

  13. Fast Burner Reactor Devoted to Minor Actinide Incineration

    This study proposes a new fast reactor core concept dedicated to plutonium and minor actinide burning by transmutation. This core has a large power level of ∼1500 MW(electric) favoring the economic aspect. To promote plutonium and minor actinide burning as much as possible, total suppression of 238U, which produces 239Pu by conversion, and large quantities of minor actinides in the core are desirable. Therefore, the 238U-free fuel is homogeneously mixed with a considerable quantity of minor actinides.From the safety point of view, both the Doppler effect and the coolant (sodium) void reactivity become less favorable in a 238U-free core. To preserve these two important safety parameters on an acceptable level, a hydrogenated moderator separated from the fuel and nuclides, such as W or 99Tc, is added to the core in the place of 238U. Tungsten and 99Tc have strong capture resonances at appropriate energies, and 99Tc itself is a long-lived fission product to be transmuted with profit.This core allows the achievement of a consumption rate of ∼100 kg/TW(electric).h of transuranic elements, ∼70 kg/TW(electric).h for plutonium (due to 238U suppression), and 30 to 35 kg/TW(electric).h for minor actinides. In addition, ∼14 kg/TW(electric).h of 99Tc is destroyed when this element is present in the core (the initial loading of 99Tc is >4000 kg in the core).The activity of newly designed subassemblies has also been investigated in comparison to standard fast reactor subassemblies (neutron sources, decay heat, and gamma dose rate). Finally, a transmutation scenario involving pressurized water reactors and minor actinide-burning fast reactors has been studied to estimate the necessary proportion of burner reactors and the achievable radiotoxicity reduction with respect to a reference open cycle

  14. Burning of actinides: A complementary waste management option?

    The TRU actinide are building up at a rate of about 90 tHM per year. Approximately 45 tHM will remain occluded in the spent fuel structures, leaving about 45 tHM available; 92% as recycled plutonium and 8% as minor actinides (neptunium, americium, curium) immobilized in vitrified waste. There is renewed interest in partitioning and transmutation (P and T), largely because of difficulties encountered throughout the world in finding suitable geologic formations in locations which are acceptable to the public. In 1988, the Japanese Atomic Energy Commission launched a very important and comprehensive R and D program. The general strategy of introducing Partitioning and Transmutation (P and T) as an alternative waste management option is based on the radiological benefit which is expected from such a venture. The selection of the actinides and long-lived fission products which are beneficial to eliminate by transmutation depends upon a number of technical factors, including hazard and decontamination factors, and the effect of geological confinement. There are two ways to approach the separation of minor actinides and long-lived fission products from reprocessing streams: by modifying the current processes in order to reroute the critical nuclides into a single solution, for example high-level liquid waste, and use this as a source for partitioning processes; and by extension of the conventional PUREX process to all minor actinides and long-lived fission products in second generation reprocessing plants. Prior to the implementation of one of these schemes, it seems obvious to improve the separation yield of plutonium from HLW within the presently running plants. Actinide P and T is not an alternative long-term waste management option. Rather, it is a complementary technique to geologic disposal capable of further decreasing the radiological impact of the fuel cycle over the very long term. 1 tab

  15. Application of partitioning/transmutation of radioactive materials in radioactive waste management

    The present waste management for already vitrified HLW has to be considered as an irreversible process for which disposal in geologic strata is the most advisable and unavoidable solution. Spent fuel discharged from nuclear power plants should be stored in engineered facilities as long as there is no definite choice of long term management of long-lived actinides. Retrievable storage in underground facilities is an alternative which could have its merits as a medium term policy. Conventional reprocessing of spent fuel is a necessary step in the reduction of the actinide content of HLW. The recycling of Pu and U into the MOX-fuel cycle is a transient solution to reduce the volume of actinide loaded waste. Spent LWR-MOX fuel with its high actinide content should be stored in engineered facilities till a safe fast reactor technology becomes available. Advanced reprocessing of (high-burn up) spent fuel with removal of U, and TRUs and production of actinide-free-vitrified-HLW is in the medium term the most defendable waste management option. Final disposal of actinide-free HLW in geologic strata is fully acceptable if the actinide content has been reduced with two (or more) orders of magnitude. Transformation of separated minor actinide concentrates into a ceramic type of waste form is beneficial from radiological point of view even if the deep disposal is the last resort. The sharply reduced solubility (compared to glass) reduces the long term environmental risk. Transformation into a matrix which could also be used as a future nuclear fuel- or target form is the most versatile option. Transmutation in a fast neutron reactor facility is the only possibility to incinerate the overall actinide (Pu+MA) inventory and to make use of the large amounts of plutonium present in spent LWR-MOX fuel. However the current FR technology with sodium cooling can, for safety reasons, not be considered for that purpose. The use of less dangerous metallic coolants is to be investigated

  16. Accelerator-driven Transmutation of Waste

    Venneri, Francesco

    1998-04-01

    Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the

  17. Simulation of fuel cycles with minor actinide management using a fast burnup calculation tool

    The paper presents a fast and flexible burnup model for fuel cycle simulations which is based on the description of the one-group cross-sections as analytic functions of the isotopic composition. This was accomplished by multi-dimensional regression based on the results of numerous core calculations. The developed model is able to determine the spent fuel composition in reasonable CPU time, and was integrated into a simplified fuel cycle model containing Gas Cooled Fast Reactors (GFR) and conventional light water reactors (LWRs). The fuel cycle simulations revealed an advantageous effect of increased minor actinide content in the GFR core on the fuel utilization parameters. In order to explore the processes that lay behind this effect the neutronics balance of the GFR was investigated in equilibrium cycle conditions. (author)

  18. The feasibility of MA transmutation in CEFR

    The feasibility of MA transmutation in CEFR (China Experimental Fast Reactor) is described. The nuclear characteristics of reference core and those of MA-loaded core are compared, the MA-transmutation amount is presented. Although the amount of MA transmutation in CEFR is limited, CEFR still has a significant role in MA fuel irradiation tests and MA transmutation technique studies. (author). 6 refs, 1 fig., 3 tabs

  19. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

    Dr. Pavel V. Tsvetkov

    2009-05-20

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  20. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  1. Fast Reactor Fuel Development in Europe

    Research and development of minor-actinide-bearing fuels in Europe has made significant progress, with a number of scoping irradiation tests made on a number of candidate fuels foreseen for fast reactors and dedicated minor actinide transmutation systems, e.g. the accelerator driven system. Currently, efforts concentrate on uranium based fuels, as the deployment of fast reactor fleets requires Pu generation in order to achieve sustainability. Both homogeneous and heterogeneous concepts for minor actinide reactor recycling are considered. In the former, the minor actinides are added in small quantities to the mixed oxide fuel, while in the latter, the minor actinides are loaded in significant quantities in UO2. Irradiation programmes to test these concepts for pellet and SPHEREPAC fuel configurations are under way. (author)

  2. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

  3. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    Bomboni, Eleonora [Department of Mechanical, Nuclear and Production Engineering (DIMNP), Via Diotisalvi 2 - 56100 Pisa (Italy)

    2008-07-01

    The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical

  4. Response of actinides to flux changes in high-flux systems

    When discussing the transmutation of actinides in accelerator-based transmutation of waste (ATW) systems, there has been some concern about the dynamics of the actinides under high transient fluxes. For a pure neptunium feed, it has been estimated that the 238Np/237Np ratio increase due to an increasing flux may lead to an unstable, positive reactivity growth. In this analysis, a perturbation method is used to calculate the response of the entire set of actinides in a general way that allows for more species than just neptunium. The time response of the system can be calculated; i.e., a plot of fuel composition and reactivity versus time after a change in flux can be made. The effects of fission products can also be included. The procedure is extremely accurate on short time scales (∼ 1000 s) for the flux levels we contemplate. Calculational results indicate that the reactivity insertions are always smaller than previously estimated

  5. Spent nuclear fuel corrosion: The application of ICP-MS to direct actinide analysis

    The ICP-MS technique has been applied to the analysis of the actinide contents of corrodant solutions from experiments performed to study the corrosion of spent nuclear fuel in simulated groundwaters. Analysis was performed directly on the solutions, without employing separation or isotope dilution techniques. The results from two analytical campaigns using natural indium and thorium internal standards are compared. Under both oxic and anoxic conditions, the U contents can be determined with good accuracy and precision. The same applies to Np and Pu under oxic conditions, where the solution concentrations range down to about 0.1 ppb. Under anoxic conditions, where solution concentrations are lower by one or two orders of magnitude, reasonable results for these two actinides can be obtained, but with much lower precision. Direct analysis of Am and Cm, however, gave unsatisfactory results, since the technique is limited by poor measurement statistics and background uncertainty

  6. Advanced fuel cycles options for LWRs and IMF benchmark definition

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  7. Fabrication and characterization of minor actinides based fuels obtained by conventional powder metallurgy

    In the frame of the 2006 second French Act related to the long-lived nuclear waste management, CEA is currently assessing minor actinides (MA) recycling in nuclear fuels for fast neutron reactors. Two main routes are investigated: homogeneous recycling, where MA are added up to several percents in MOX fuel to be used in the whole core of the reactor or heterogeneous recycling which consists in larger amount of MA in uranium oxide fuel to be specifically used on periphery of the core. Both are under progress. Regarding the homogeneous recycling, CEA is involved with JAEA and DOE in the Global Actinide Cycle International Demonstration (GACID) program. A first attempt to fabricate minor actinides (Am, Np, Cm) bearing MOX fuel pellets was successfully achieved in the ATALANTE facility by a conventional powder metallurgy process. The sintered pellets were submitted to a visual inspection where neither crack nor strain was detected. In addition, the pellets exhibit a density in the range 93-96% TD which makes them proper to the irradiation in FR. The pellets were characterized by XRD, SEM combined to image analysis and microprobe investigations. In order to test the feasibility of heterogeneous recycling, several irradiation programs are planed or under development. For instance, CEA is responsible for the pellets fabrication for the MARIOS experiment, which is conducted in the frame of the FAIRFUELS program (7. Framework Program supplied by the European Community). Both dense and tailored porosity pellets incorporating between 10 and 30 wt% of americium in uranium oxide were then fabricated. Geometrical stability versus time as well as further characterisations, including XRD and SEM analysis, are underway. (authors)

  8. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion

  9. A fast lead-cooled incinerator for economical actinide burning

    A fast lead-cooled modular reactor is proposed as an efficient incinerator of plutonium and minor actinides (MAs) for application to advanced fuel cycles devoted to transmutation. This actinide burner reactor (ABR) is loaded only with transuranics (TRU) in a fertile-free Zr-based metallic fuel to maximize the incineration rates and features (a) streaming fuel assemblies that enhance neutron leakage to achieve favorable neutronic feedbacks and (b) a double-entry control rod system that reduces reactivity perturbations during seismic events and flattens the axial power profile. A detailed neutronic analysis shows that the delayed neutron fraction is comparable to that of fast reactors and that negative reactivity feedbacks from lead voiding, Doppler, fuel thermal expansion and core radial expansion lead to safety characteristics similar to those of the Integral Fast Reactor. The ABR TRU destruction rate is the same as that of the ATW and fuel cycle cost analysis shows potential for economical accomplishment of the transmutation mission compared to other proposed actinide burning options. (author)

  10. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  11. The nuclear design optimization of a Pb-Bi alloy cooled transmuter, PEACER-300

    A core design of lead-bismuth cooled fast reactor, PEACER-300 has been investigated to maximize its transmutation capability within safety criteria. Transmutation of minor actinide under a closed recycling was analyzed with assumption on decontamination factors in pyro-reprocessing plant data at reasonably high values. To acquire high transmutation performance, feed fuel composition, P/D ratio, active core height and fuel cycle strategy were changed. For preventing the fuel meting and guaranteeing long plant life-time, the number of fuel assembly array and normal operation temperature were decided. The optimized design parameter were chosen as of a flat core shape with 50 cm of active core height and 5 m core diameter, loaded with 17 x 17 arrayed fuel assemblies. A pitch to diameter ratio is 2.2, operating coolant temperature range is 300 deg. C to 400 deg. C, and core consists of 3 different enrichment zones with one year cycle length. Performance of designed core showed a high transmutation capability with support ratio of 2.085, large negative temperature feedback coefficients, and sufficient shutdown margin with 28 B4C control assemblies. (authors)

  12. Effects of actinide burning on waste disposal at Yucca Mountain

    Release rates of 15 radionuclides from waste packages expected to result from partitioning and transmutation of Light-Water Reactor (LWR) and Actinide-Burning Liquid-Metal Reactor (ALMR) spent fuel are calculated and compared to release rates from standard LWR spent fuel packages. The release rates are input to a model for radionuclide transport from the proposed geologic repository at Yucca Mountain to the water table. Discharge rates at the water table are calculated and used in a model for transport to the accessible environment, defined to be five kilometers from the repository edge. Concentrations and dose rates at the accessible environment from spent fuel and wastes from reprocessing, with partitioning and transmutation, are calculated. Partitioning and transmutation of LWR and ALMR spent fuel reduces the inventories of uranium, neptunium, plutonium, americium and curium in the high-level waste by factors of 40 to 500. However, because release rates of all of the actinides except curium are limited by solubility and are independent of package inventory, they are not reduced correspondingly. Only for curium is the repository release rate much lower for reprocessing wastes

  13. Physical mechanism analysis of burnup actinide composition in light water reactor MOX fuel and its application to uncertainty evaluation

    Highlights: • We discuss physical mechanisms for burnup actinide compositions in LWR’s MOX fuel. • Mechanisms of 244Cm and 238Pu productions are analyzed in detail with sensitivity. • We can evaluate the indirect effect on actinide productions by nuclear reactions. • Burnup sensitivity is applied to uncertainty evaluation of nuclide production. • Actinides can be categorized into patterns according to a burnup sensitivity trend. - Abstract: In designing radioactive waste management and decommissioning facilities, understanding the physical mechanisms for burnup actinide composition is indispensable to satisfy requirements for its validity and reliability. Therefore, the uncertainty associated with physical quantities, such as nuclear data, needs to be quantitatively analyzed. The present paper illustrates an analysis methodology to investigate the physical mechanisms of burnup actinide composition with nuclear-data sensitivity based on the generalized depletion perturbation theory. The target in this paper is the MOX fuel of the light water reactor. We start with the discussion of the basic physical mechanisms for burnup actinide compositions using the reaction-rate flow chart on the burnup chain. After that, the physical mechanisms of the productions of Cm-244 and Pu-238 are analyzed in detail with burnup sensitivity calculation. Conclusively, we can identify the source of actinide productions and evaluate the indirect influence of the nuclear reactions if the physical mechanisms of burnup actinide composition are analyzed using the reaction-rate flow chart on the burnup chain and burnup sensitivity calculation. Finally, we demonstrate the usefulness of the burnup sensitivity coefficients in an application to determine the priority of accuracy improvement in nuclear data in combination with the covariance of the nuclear data. In addition, the target actinides and reactions are categorized into patterns according to a sensitivity trend

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

    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.

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

    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

  16. Plutonium and minor actinides recycle in equilibrium fuel cycles of pressurized water reactor

    A study on plutonium and minor actinides (MA) recycle in equilibrium fuel cycles of pressurized water reactors (PWR) has been performed. The calculation results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined plutonium and MA when uranium is discharged from the reactor. However, when uranium is totally confined, the enrichment becomes extremely high. The recycle of plutonium and MA together with discharging uranium can reduce the radio-toxicity of discharged heavy metal (HM) waste to become less than that of loaded uranium. (author)

  17. Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel

    Janney, D. E.; Keiser, D. D.

    2003-09-01

    Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to ˜ 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium.

  18. Reduction of minor actinides for recycling in a light water reactor; Reduccion de actinidos menores por reciclado en un reactor de agua ligera

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The aim of actinide transmutation from spent nuclear fuel is the reduction in mass of high-level waste which must be stored in geological repositories and the lifetime of high-level waste; these two achievements will reduce the number of repositories needed, as well as the duration of storage. The present work is directed towards the evaluation of an advanced nuclear fuel cycle in which the minor actinides (Np, Am and Cm) could be recycled to remove most of the radioactive material; a reference of actinides production in standard nuclear fuel of uranium at the end of its burning in a BWR is first established, after a design of fuel rod containing 6% of minor actinides in a matrix of uranium from the enrichment lines is proposed, then 4 fuel rods of standard uranium are replaced by 4 actinides bars to evaluate the production and transmutation of them and finally the minor actinides reduction in the fuel is evaluated. In the development of this work the calculation tool are the codes: Intrepin-3, Casmo-4 and Simulate-3. (Author)

  19. Europart: European Research programme for partitioning of minor actinides within high active wastes issuing from the reprocessing of spent nuclear fuels

    The radiotoxicity of the vitrified nuclear wastes issuing from the reprocessing of nuclear spent fuels is mainly due to the presence of minor actinides (MAs = Np, Am and Cm) in these wastes. To simplify the definition of a deep underground repository, the elimination of MAs from the wastes followed by their destruction by transmutation is a positive strategy (P and T strategy). The European Integrated Project (IP) EUROPART concerns the definition of partitioning methods for the elimination of MAs from the wastes. Two chemical domains were selected for the research: i) hydrometallurgy, ii) pyrometallurgy. The research work is organised into 9 Work Packages (WP): 5 for hydrometallurgy and 4 for pyrometallurgy. The partnership involved in the research comprises 23 partners from 10 European countries, 1 the EC-JRC-ITU, 1 partner from Japan and 1 partner from Australia. This paper will present: i) the WPs, ii) some recent results related to the WPs, iii) the partnership and the organisation of the IP. (authors)

  20. Scenario analysis for transuranic transmutation by using fast reactors

    Symbiotic fast reactor scenarios with the existing nuclear power systems have been analyzed from the viewpoint of a transuranics transmutation. In this study, a sodium-cooled fast reactor (SFR) and an accelerator driven system (ADS) are considered as representative fast reactor systems. For a comparative analysis of the fuel cycle options, the once-through fuel cycle was at first analyzed based on the current nuclear power plant construction plan and the currently operating nuclear power plants such as the pressurized water reactor (PWR) and the Canada deuterium uranium (CANDU) reactor. After setting up a once-through fuel cycle model, the SFR and ADS scenarios were modeled based on the same nuclear energy demand prediction used for the once-through fuel cycle. Then important fuel cycle parameters such as the amount of the spent fuel and corresponding plutonium, minor actinides and fission products inventories were estimated and compared with those of the once-through fuel cycle. In this fuel cycle model, the Pyro process is assumed for all the spent fuel recycling. In the process all the actinides are recovered and some fraction of the fission product is removed. The deployment fractions of the fast reactor are 25, 10 and 20% for the periods of 2030-2040, 2041-2070 and 2071-2100, respectively. In order to feed the fast reactor systems, it was also assumed that the PWR and CANDU spent fuels are reprocessed from 2025 and the fast reactor spent fuel reprocessing begins in 2035. The fuel cycle calculation was performed by the DYMOND code, which has been used for an analysis of the Generation-IV road map studies. The analysis results of the once-through fuel cycle can be summarized as follows: - The nuclear power demand is expected to grow to 25.2 GWe in the year 2100. - The total spent fuel inventory is expected to be 65000 t in 2100. - The transuranics and fission product inventories are estimated to be 660 and 2390 t, respectively, in 2100. The fast reactor cycle

  1. Design of Neptunium-bearing Fuel Assembly for Transmutation Research in CEFR

    In order to have a better understanding of irradiation performance of the fuel containing neptunium, an experimental assembly is designed for future irradiation in CEFR. There is only one fuel pin in the assembly with neptunium content of 5%. Temperature monitors and neutron fluence detectors are attached. The report presents the basic structure of the fuel pin and the assembly. (author)

  2. The concept of separation-transmutation and the management of radioactive wastes

    Alpha, beta and gamma radiation emitting radionuclide waste management problems are discussed, with particular attention to minor actinide separation. The impacts of waste separation-transmutation on the disposal of wastes and nuclear transmutation of actinides are discussed. (R.P.) 21 refs

  3. Investigation of the Feasibility of a Small Scale Transmutation Device

    Sit, Roger Carson

    2009-01-01

    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), sho...

  4. Fuel reprocessing of the fast molten salt reactor: actinides et lanthanides extraction

    The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoro-acidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered. Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF4 by F-) in reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction. (author)

  5. Advancing the scientific basis of trivalent actinide-lanthanide separations

    For advanced fuel cycles designed to support transmutation of transplutonium actinides, several options have been demonstrated for process-scale aqueous separations for U, Np, Pu management and for partitioning of trivalent actinides and fission product lanthanides away from other fission products. The more difficult mutual separation of Am/Cm from La-Tb remains the subject of considerable fundamental and applied research. The chemical separations literature teaches that the most productive alternatives to pursue are those based on ligand donor atoms less electronegative than O, specifically N- and S-containing complexants and chloride ion (Cl-). These 'soft-donor' atoms have exhibited usable selectivity in their bonding interactions with trivalent actinides relative to lanthanides. In this report, selected features of soft donor reagent design, characterization and application development will be discussed. The roles of thiocyanate, aminopoly-carboxylic acids and lactate in separation processes are detailed. (authors)

  6. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Nuttin A.

    2012-02-01

    Full Text Available The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX and uranium/plutonium mixed oxide (MOX fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  7. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Rose, S. J.; Wilson, J. N.; Capellan, N.; David, S.; Guillemin, P.; Ivanov, E.; Méplan, O.; Nuttin, A.; Siem, S.

    2012-02-01

    The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  8. Actinide isotopes compositions and neutrons emission rate calculations for irradiated research reactors fuel

    the availability of burn-up data is an essential first step in any systematic approach to the enhancement of safety, economics and performance of research reactors. A computer program has been designed to solve the system of equations describing the depletion, decay and production of uranium, plutonium and transplutonium nuclides. monte Carlo code was used to calculate the effective one group microscopic cross sections averaged over ETRR-1 fuel cell. the compositions of actinide isotopes, burn-up and neutrons emission rate have been calculated as a function of irradiation time and cooling time. results indicate that the amount of plutonium produced and neutrons emission rate are strongly dependent on the fuel burn-up

  9. The release of actinides, cesium, strontium, technetium, and iodine from spent fuel under unsaturated conditions

    Drip tests to measure radionuclide release from spent nuclear fuel are being performed at 90 degrees C at a drip rate of 0.75 mL/3.5 days; the test conditions are designed to simulate the behavior of spent fuel under the unsaturated and oxidizing conditions expected in the potential repository at Yucca Mountain. This paper presents measurements of the actinide, 137Cs, 90Sr, 99Tc, and 129I contents in the leachates after 581 days of testing at 90 degrees C. These values provide an estimate of the source term for the long-lived radionuclide release under these test conditions. Comparisons are made between our results and those of other researchers

  10. Adaptation of ICP-AES in lead cell facility in Chemistry Group, IGCAR and analysis of simulated high level waste as a part of the studies on minor actinide partitioning

    The spent fuel discharged from the nuclear reactor contains unused uranium and plutonium, and Np, Am, Cm called as minor actinides and fission products. Spent fuel is dissolved in nitric acid. U and Pu are recovered by a solvent extraction process known as PUREX process using 1.1 M TBP as extractant. The raffinate rejected is known as High Level Liquid Waste which is a complex mixture of minor actinides, corrosion products, and fission products. Partitioning of minor actinides (MA) and its transmutation is a viable strategy for the safe management of high level liquid waste (HLLW)

  11. The impact of the core configuration on safety and transmutation behavior in an accelerator driven system; Auswirkung der Brennstoffwahl auf das Transmutationsverhalten in einem beschleunigergetriebenen System

    Biss, K.; Nabbi, R.; Thomauske, B. [RWTH Aachen Univ. (Germany). Inst. fuer Nuklearen Brennstoffkreislauf (INBK)

    2012-11-01

    For the reduction of the long-term hazards of high-level wastes transmutation is one of the candidate techniques. For an effective conversion of transuranic elements, esp. minor actinides, the use of accelerator driven systems (ADS) is the favored concept. The subcritical system AGATE (advanced gas-cooled accelerator driven transmutation experiment)is a 100 MW(th) facility using a proton beam to produce the required spallation neutrons. The fuel zone includes 120 uniform fuel elements with hexagonal structure (each one with 91 fuel rods) in an annular configuration around the spallation target. Neutron flux and energy spectra are determined and averaged for each zone allowing a fast calculation of fuel element variants and geometry variations. For modeling the Monte Carlo code MCNPX 2.7 is used. The transmutation rate for pure PuMA fuel show high values for americium, but the isotope analysis shows that the largest fraction is transmuted to plutonium. The use of thorium as matrix material reduces the transmutation rate of transuranic elements but allows a long-term burnup cycle without required fuel element replacement.

  12. Studies of partitioning and transmutation

    Part 1: Current status of partitioning and transmutation: The purpose of the project covered in this report is to contribute to a watching brief exercise for the Department of the Environment, Transport and the Regions (DETR) on the subject of the Partitioning and Transmutation (P and T) of long-lived radionuclides present in high level radioactive waste (HLW). The watching brief is intended to ensure that DETR are aware of international developments and progress so that UK policy continues to be soundly based. This has been achieved by attendance at international meetings and conferences and studies of the published literature, and also by participation in the Fourth Framework R and D Programme of the European Commission (EQ) in the field of P and T (see Part 2 below). Answers have been developed to a list of questions about certain aspects of P and T, provided by the DETR; and further information has also been provided about progress in the current EC programme and elsewhere. National programmes on P and T are in progress in various countries, and the motivations for these vary. These programmes concentrate exclusively on high level waste (HLW) in spite of the environmental importance of other waste streams. P and T is not generally seen as a viable waste management strategy in the short or medium term, but as an option for the future. A considerable new impetus has been imparted to P and T research by the development of Accelerator Driven Systems (ADS) which provide high neutron fluxes suitable for transmutation. Such systems may be more effective than current fission reactors for this purpose. Good progress has also been made in the separation of actinides and long-lived fission products from HLW, using both aqueous and dry (pyrochemical) processes. P and T is more likely to be implemented in future decades as part of a radically new type of fuel cycle, probably pyrochemical, rather than as an extension of PUREX reprocessing. However, pyrochemical reprocessing

  13. Towards standardized calculation tools for the Accelerator-Driven Systems and their application to various scenarios for nuclear waste transmutation

    This thesis discusses the question of partitioning and transmutation of actinides and some long-lived fission products as a way of reducing the mass and radio-toxicity of wastes from nuclear power facilities. Numerical benchmarking and computational exercises carried out in related projects are discussed and the quantitative assessment of the advantages and drawbacks of various transmutation strategies are discussed, as is the role of Accelerator-Driven Systems (ADS) and Advanced Fast Reactors (FR) in advanced nuclear fuel cycles. According to the author, the study allows three main options in nuclear waste management - open cycle, plutonium recycling and the recycling of all actinides - to be compared. The last part of the dissertation is dedicated to two phase-out schemes employing either ASDs or critical reactors

  14. Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides

    Highlights: ► Neutronic calculations for PBMR 400 were conducted with the computer codes MCNP and MONTEBURNS 2.0. ► The criticality and burnup were investigated for reactor grade plutonium and minor actinides. ► We found that the use of these new fuels in PBMRs would reduce the nuclear waste repository significantly. -- Abstract: Time evolution of criticality and burnup grades of the PBMR were investigated for reactor grade plutonium and minor actinides in the spent fuel of light water reactors (LWRs) mixed with thoria. The calculations were performed by employing the computer codes MCNP and MONTEBURNS 2.0 and using the ENDF/B-V nuclear data library. Firstly, the plutonium–thorium and minor actinides–thorium ratio was determined by using the initial keff value of the original uranium fuel design. After the selection of the plutonium/minor actinides–thorium mixture ratio, the time-dependent neutronic behavior of the reactor grade plutonium and minor actinides and original fuels in a PBMR-400 reactor was calculated by using the MCNP code. Finally, keff, burnup and operation time values of the fuels were compared. The core effective multiplication factor (keff) for the original fuel which has 9.6 wt.% enriched uranium was computed as 1.2395. Corresponding to this keff value the reactor grade plutonium/thorium and minor actinide/thorium oxide mixtures were found to be 30%/70% and 50%/50%, respectively. The core lives for the original, the reactor grade plutonium/thorium and the minor actinide/thorium fuels were calculated as ∼3.2, ∼6.5 and ∼5.5 years, whereas, the corresponding burnups came out to be 99,000, ∼190,000 and ∼166,000 MWD/T, respectively, for an end of life keff set equal to 1.02.

  15. Irradiation test of U-free nitride fuel and progress of pyro chemistry in JAERI

    JAERI has proposed the double-strata fuel cycle for transmutation of long-lived minor actinides (MAs). The transmutation system is a Pb-Bi cooled subcritical accelerator-driven system (ADS) with MAs nitride fuel. Nitride fuel has the advantage of accommodating various MAs with a wide range of composition besides superior thermal and neutronic properties. This paper concerns the status of the irradiation test of U-free nitride fuel and recent progress of pyrochemical process for nitride fuel in JAERI. Typical characteristics of nitride fuel for the irradiation test and the present schedule are described in addition to recent experimental results relating to pyro-chemistry. (author)

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

    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

  17. Alternative Thorium fuel cycle for LWRS

    In the paper, different thorium nuclear fuel cycles are examined and compared under light water reactor conditions, especially VVER-440. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium U-233 content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. Different thorium fuel distribution in fuel assemblies is modeled - homogeneous and heterogenous. The article shows main features of VVER-440 reactor, analysed fuel assemblies and fuel cycles. Fuel cycles and fissile content in the fuel are tuned to fulfil operating conditions of VVER-440 reactor. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulation in the spent thorium fuel and its comparison to UOX open fuel cycle. (authors)

  18. Transmutation detectors

    Viererbl, L.; Lahodová, Z.; Klupák, V.; Sus, F.; Kučera, Jan; Kůs, P.; Marek, M.

    2011-01-01

    Roč. 632, č. 1 (2011), s. 109-111. ISSN 0168-9002 Institutional research plan: CEZ:AV0Z10480505 Keywords : Transmutation detector * Activation method * Neutron detector * Neutron fluence Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.207, year: 2011

  19. Inert matrix fuel concept for the rapid incineration of minor actinides harmonious with a fast reactor cycle system

    We proposed a fast reactor cycle concept that incorporates inert matrix fuels as a high-performance device for rapid incineration of minor actinides and a harmonious system with the existing fast reactor cycle technologies. R and D of minor actinides containing advanced fuels for use in fast reactors is described in relation to inert matrix fuels with MgO, Mo and Si3N4. As related technologies, burn-up characteristics of a fast reactor core loaded with the inert matrix fuel with MgO and Mo were analyzed, mainly in terms of core criticality. Fabrication tests of inert matrix fuels with MgO, Mo and Si3N4 were done by a practical process that could be adapted to the presently-used commercial manufacturing technology. Preliminary investigations for the solubility of inert matrix fuels to the HNO3 were carried out for the evaluation of applicability to existing reprocessing technology. This paper describes a part of our efforts towards the establishment of a fast reactor cycle that incorporates the minor actinides containing inert matrix fuels. (author)

  20. Status of the French Research on Partitioning and Transmutation

    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

  1. Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

    Mueller, Alex C

    2012-01-01

    While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven syste...

  2. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

  3. Parametric study of the accelerator-driven transmutation system

    A couple of parametric studies were performed for an accelerator-driven transmutation system in KAERI (Korea Atomic Energy Research Institute). For the analysis, LAHET code system developed by LANL was employed. Molten salt fuel was adopted with its chemical composition of 64NaCl3-1Pu-35MA(Minor Actinide)Cl3. The isotope compositions of Pu and Ma were determined based on the composition in the 10-year cooled spent fuel having 3.2 W% initial enrichment and 33GWD/MTU discharge burnup. Proton beam of 20 mA, 1 GeV was assumed for the neutron production by the spallation reaction with the fuel. The fuel was designed to perform multifunction such as target, coolant. From the calculation results, one proton was believed to produce about 27 neutrons and the neutron multiplication factor was found to be 0.95 for the given system condition. For the beam of 20 mA, 1 GeV, the neutron flux reached up to 1.26x1015 n/cm2 and the corresponding total thermal power was 773 MWth. It was believed that the proposed system could transmute 502 kg of MA a year. 3 refs., 3 figs., 1 tab

  4. Production, disposal, and relative toxicity of long-lived fission products and actinides in the radioactive wastes from nuclear fuel cycles

    Chapters are devoted to the following topics: predicted future development of nuclear energy in the German Federal Republic and in Western Europe, fuel cycle variations and production of fission products and actinides in the radioactive waste from reprocessed nuclear fuels, long-lived fission products and actinides in the waste streams from the reprocessing of nuclear fuels, relative toxicity index, presently preferred waste management concepts, and alternative concepts for the elimination of high-level wastes

  5. Standardisation des outils de calcul pour les ADS et leur application à différents scénarios de transmutation des déchets

    Cometto, Marco; Chawla, Rakesh

    2005-01-01

    The management of radioactive wastes from the nuclear fuel cycle has become an important issue in the development of future, more sustainable nuclear energy systems. Partitioning and transmutation (P&T) of actinides and some long-lived fission products could reduce the mass and radiotoxicity of highlevel wastes and possibly ease repository licensing requirements. Influenced by political and technological developments, an increasing number of countries employing nuclear power have become inter...

  6. Standardisation des outils de calcul pour les ADS et leur application à différents scénarios de transmutation des déchets

    Cometto, Marco

    2003-01-01

    The management of radioactive wastes from the nuclear fuel cycle has become an important issue in the development of future, more sustainable nuclear energy systems. Partitioning and transmutation (P&T) of actinides and some long-lived fission products could reduce the mass and radiotoxicity of highlevel wastes and possibly ease repository licensing requirements. Influenced by political and technological developments, an increasing number of countries employing nuclear power have become inter...

  7. Partitioning and Transmutation - Annual Report 2010 and 2011

    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 have ranged 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 recycling route, called DIAMEX (DIAmide EXtracton) / SANEX (Selective ActiNide EXtraction) is now considered to be working on a basic scale and has been proven in hot tests and focus has moved on to more process oriented areas. However, since further investigations on basic understanding of the chemical behavior are required, we have our main focus on the chemical processes and understanding of how they work. Our work is now primarily put on the so called

  8. Partitioning and Transmutation - Annual Report 2010 and 2011

    Aneheim, Emma; Ekberg, Christian; Fermvik, Anna; Foreman, Mark; Littley, Alexander; Loefstroem-Engdahl, Elin; Mabile, Nathalie; Skarnemark, Gunnar [Nuclear Chemistry, Dept. of Chemical and Biological Engineering, Chalmers Univ. of Technology, Goeteborg (Sweden)

    2013-01-15

    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 and {sup 135}Cs) and activation products ({sup 14}C, {sup 36}Cl, {sup 59}Ni, {sup 93}Zr, {sup 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 have ranged 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 recycling route, called DIAMEX (DIAmide EXtracton) / SANEX (Selective ActiNide EXtraction) is now considered to be working on a basic scale and has been proven in hot tests and focus has moved on to more process oriented areas. However, since further investigations on basic understanding of the chemical behavior are required, we have our main focus on the chemical processes and

  9. Analysis of the transmutation of actinides minority in a sodium cooled fast reactor; Analisis de la transmutacion de actinidos minoritarios en un reactor rapido refrigerado por sodio

    Ochoa Valero, R.

    2011-07-01

    Fast reactors represent a highly sustainable source of energy due to the use of a closed fuel cycle, which makes better use of natural resource and reducing the volume and heat load of high level radioactive waste.

  10. Achievements on oxide and nitride ADS fuels within the European project. EUROTRANS

    One major item in the assessment of the feasibility of Minor Actinide transmutation in ADS systems deals with the fuel, that can be described as highly innovative in comparison with fuels used in a critical core. Indeed, the fuel composition, with high concentrations of Minor Actinides and plutonium, results in significant production of helium during irradiation as well as high gamma, neutron emissions and heat in fabrication and handling stages. Studies on fuel development within the European project EUROTRANS, motivated by assessing the industrial practicability for actinides transmutation, have provided a wide range of results. CERamic-CERamic and CERamic-METallic composite fuels consisting of particles of (Pu,MA)O2 dispersed in a magnesia or molybdenum matrix were investigated as primary candidates. Fuel performance and safety of preliminary core designs for a prospective 400 MWth transmuter: the European Facility for Industrial Transmutation, were evaluated to support the project. Out-of-pile as well as in-pile experiments were carried out to gain essential knowledge on properties and behaviour under irradiation of these types of fuel. Nitride-based fuels were considered as a back-up solution, due to a limited know-how in Europe. Significant progress has nevertheless been made through a JAEA partnership and Post-Irradiation Examinations conducted on (Pu,Zr)N fuels irradiated at high linear power in the Petten High Flux Reactor. The current paper gives an overview of progress on both oxide and nitride fuels gained within the project. (author)

  11. Actinide content and associated neutron emission of WWER-440 spent fuel - calculation and validation for safeguards purposes

    A major task of international nuclear material safeguards consists in the experimental determination of the characteristics of spent nuclear fuel. In this respect, passive neutron techniques take a prominent place. The paper describes both the actinide content and the passive neutron emission of spent WWER-440 fuel for a number of operational conditions. For validation purposes the results have been compared with those independently gained by other authors for the same type of fuel. The calculations are the basis for a true to fact interpretation of the primary measurement signal, i.e. for the inference from recorded neutron radiation regarding the properties of the spent nuclear fuel. (author)

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

    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 ThO2-UO2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO2 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

  13. The analysis and handling concept of minor actinides of NPP’s waste by using Ads technology

    The contents of minor actinide elements (americium, neptunium and curium) on the spent fuel inventory from PWR operation of NPP have been calculated using Vista program. The calculation used parameters: enrichment 3.968%, power 1000 M We and burn-up is 60 M Wd/kg. The result of calculation showed that the arising of minor actinide elements on the spent fuel is 16.205 kg/year and 43.471 kg/year for PWR-UOX and PWR-MOX respectively. It is also discussed a concept of the use of ADS technology for transmuting the minor actinide elements contained in spent fuels. The result of the discussion showed that an ADS of 400 M Wth will serve 7 PWRs-UOX, and on the PWR system using UOX and MOX fuels an ADS will serve 3 PWRs. (author)

  14. The advanced liquid metal reactor actinide recycle system

    The current U.S. National Energy Strategy includes four key goals for nuclear policy: enhance safety and design standards, reduce economic risk, reduce regulatory risk, and establish an effective high-level nuclear waste program. The U.S. Department of Energy's Advanced Liquid Metal Reactor Actinide Recycle System is consistent with these objectives. The system has the ability to fulfill multiple missions with the same basic design concept. In addition to providing an option for long-term energy security, the system can be effectively utilized for recycling of actinides in light water reactor (LWR) spent fuel, provide waste management flexibility, including the reduction in the waste quantity and storage time and utilization of the available energy potential of LWR spent fuel. The actinide recycle system is comprised of (1) a compact liquid metal (sodium) cooled reactor system with optimized passive safety characteristics, and (2) pyrometallurgical metal fuel cycle presently under development of Argonne National Laboratory. The waste reduction of LWR spent fuel is accomplished by transmutation or fissioning of the longer-lived transuranic isotopes to shorter-lived fission products in the reactor. In this presentation the economical and environmental incentive of the actinide recycle system is addressed and the status of development including licensing aspects is described. 3 refs., 1 tab., 6 figs

  15. Treatment and recycling of spent nuclear fuel. Actinide partitioning - Application to waste management

    subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)

  16. Treatment and recycling of spent nuclear fuel. Actinide partitioning - Application to waste management

    Abonneau, E.; Baron, P.; Berthon, C.; Berthon, L.; Beziat, A.; Bisel, I.; Bonin, L.; Bosse, E.; Boullis, B.; Broudic, J.C.; Charbonnel, M.C.; Chauvin, N.; Den Auwer, C.; Dinh, B.; Duhamet, J.; Escleine, J.M.; Grandjean, S.; Guilbaud, P.; Guillaneux, D.; Guillaumont, D.; Hill, C.; Lacquement, J.; Masson, M.; Miguirditchian, M.; Moisy, P.; Pelletier, M.; Ravenet, A.; Rostaing, C.; Royet, V.; Ruas, A.; Simoni, E.; Sorel, C.; Vaudano, A.; Venault, L.; Warin, D.; Zaetta, A.; Pradel, P.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Forestier, A.; Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Sollogoob, P.; Vernaz, E.; Bazile, F.; Parisot, J.P.; Finot, P.; Roberts, J.F

    2008-07-01

    subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)

  17. Actinides in liquid waste formed in the regeneration of nuclear fuel from a VVER-1000 reactor

    In the radiochemical reprocessing of spent fuel from nuclear reactors, a considerable amount of liquid, solid, and gaseous waste is formed; this waste is potentially dangerous to humans and requires the development of special and complex technological techniques for its localization and reliable long-term storage. The most hazardous are liquid wastes of high specific activity - water-tailings solutions obtained in the first cycle of extraction after the removal of uranium and plutonium. These solutions contain more than 99.9% of all the other transuranic elements - isotopes of neptunium, americium, and curium. Where necessary, some fission products and actinides may be removed from wastes of high specific activity for subsequent use. The quantity, composition, and activity of these wastes varies within broad limits, depending on the type and power of the reactor, the initial nuclide composition of the fuel, and its specific energy yield, the regeneration technology, and other factors. The aim of the present work is to study some radiation characteristics of the wastes of high specific activity formed in the radiochemical reprocessing of the fuel from a VVER-1000 water-cooled, water-moderated reactor

  18. Fabrication of inert-matrix nitride fuel pins for the irradiation test at JMTR

    Nitride fuel pins containing inert matrix such as ZrN and TiN were fabricated for the irradiation test at JMTR, aiming at understanding irradiation behavior of nitride fuel for transmutation of minor actinides. Minor actinides are surrogated by plutonium in the present fuel pin. This report describes the preparation and characterization of fuel pellets, and fabrication of fuel pins. The irradiation for 11 cycles from May 2002 to November 2004 at JMTR was completed without any failure of fuel pins. (author)

  19. Effects of burn-up, recovery efficiency and waste form on the environmental impact of fusion-fission transmutation systems

    The effects of fuel cycle parameters on nuclear waste environmental impact are analyzed for an advanced system that includes a Fusion-Fission Hybrid reactor. The system aims at reduction of the long-term radiotoxicity of waste by transmuting highly radiotoxic transuranics. However, the radiological risk of the system is measured by annual doses to the public, which are controlled by reactor operations, fuel cycle processes, waste treatment processes, and design of geological repositories. In this study, the waste environmental impact for a fuel cycle with a Fusion-Fission transmutation is analyzed as a function of three different parameters: burn-up, recovery efficiency and waste form durability for two different geological repositories, one with low actinide solubility and the other with high solubility. It is found that burn-up and recovery efficiency effects on environmental impact strongly depend on repository conditions, while the most influential parameter is found to be the durability of the waste form. (author)

  20. Electrochemistry of actinide and lanthanide in molten salt system

    In the partition and transmutation processes of reprocessing of spent fuel or radioactive waste in nuclear power plant, the dry type reprocessing method using molten salt and liquid metal as a solvent is studied. Most especially researches on the electrolysis of the actinide nitride in the molten salts corresponding to reprocessing of nitride fuel cannot be found. This report is a research result about the electro-chemical behavior of actinide and lanthanide on the electrode in molten LiCL-KCL eutectic system. When anode potential was less than -0.4V in recovery of U metal by the molten salt electrolysis of UN, the electrolysis efficiency of the recovery is not influenced by the generation of UNCL and the oxidation-reduction reaction of U4+/U3+. Moreover, generation of a chlorination nitride was not seen in the case where PuN and NpN are used. (H. Katsuta)

  1. Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

    Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance

  2. Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  3. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

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

    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)

  5. The actinide waste problem in perspective

    The long lived alpha emitting actinide waste nuclides of transplutonium elements such as Np, Am, Cm etc (also called Byproduct Actinides or BPA for short) which are proposed to be disposed of as part of High Active Waste (HAW) in deep underground geological repositories has been a persistent source of concern to opponents and critics of nuclear fission energy. In this context the recent finding of the authors that each and every transuranium nuclide, without exception, can independently support a self sustaining chain reaction raises the important philosophical question: Is it justified to continue to refer to these nuclides as nuclear waste ? Our computations have revealed that the Ksub(eff) of an assembly of each of these nuclides increases linearly with the fissility parameter (Z2/A), its threshold value for Ksub(eff) to exceed unity being 34.1 for fissile (odd neutron) nuclides and 34.9 for fissible (even neutron) nuclides. In other words higher the (Z2/A) better is its performance as a fission reactor fuel. This finding suggests that the long lived actinide waste problem can be solved by separating all the actinide nuclides from the High Active Waste stream and recycling them back into any hard spectrum fission reactor. The studies strongly support the concept of partitioning-transmutation (p-t) revived with great enthusiasm in Japan under the banner of the OMEGA proposal. However it is found that there is no need to resort to any exotic devices such as proton accelerators or fusion reactor blankets for nuclear incineration. In the context of the 232Th/233U fuel cycle it is worth noting that the quantum of transuranium nuclides generated per se is smaller by several orders of magnitude as compared to that arising from 235U/238U bearing fuels. Thus on the whole it appears that in the thorium fuel cycle partitioning and recycle of byproduct nuclides would be a less cumbersome undertaking. (author). 26 refs., 6 figs., 3 tabs

  6. Fusion transmutation of waste: design and analysis of the in-zinerator concept.

    Durbin, S. M.; Cipiti, Benjamin B.; Olson, Craig Lee; Guild-Bingham, Avery (Texas A& M University, College Station, TX); Venneri, Francesco (General Atomics, San Diego, CA); Meier, Wayne (LLNL, Livermore, CA); Alajo, A.B. (Texas A& M University, College Station, TX); Johnson, T. R. (Argonne Mational Laboratory, Argonne, IL); El-Guebaly, L. A. (University of Wisconsin, Madison, WI); Youssef, M. E. (University of California, Los Angeles, CA); Young, Michael F.; Drennen, Thomas E. (Hobart & William Smith College, Geneva, NY); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Morrow, Charles W.; Turgeon, Matthew C.; Wilson, Paul (University of Wisconsin, Madison, WI); Phruksarojanakun, Phiphat (University of Wisconsin, Madison, WI); Grady, Ryan (University of Wisconsin, Madison, WI); Keith, Rodney L.; Smith, James Dean; Cook, Jason T.; Sviatoslavsky, Igor N. (University of Wisconsin, Madison, WI); Willit, J. L. (Argonne Mational Laboratory, Argonne, IL); Cleary, Virginia D.; Kamery, William (Hobart & William Smith College, Geneva, NY); Mehlhorn, Thomas Alan; Rochau, Gary Eugene

    2006-11-01

    Due to increasing concerns over the buildup of long-lived transuranic isotopes in spent nuclear fuel waste, attention has been given in recent years to technologies that can burn up these species. The separation and transmutation of transuranics is part of a solution to decreasing the volume and heat load of nuclear waste significantly to increase the repository capacity. A fusion neutron source can be used for transmutation as an alternative to fast reactor systems. Sandia National Laboratories is investigating the use of a Z-Pinch fusion driver for this application. This report summarizes the initial design and engineering issues of this ''In-Zinerator'' concept. Relatively modest fusion requirements on the order of 20 MW can be used to drive a sub-critical, actinide-bearing, fluid blanket. The fluid fuel eliminates the need for expensive fuel fabrication and allows for continuous refueling and removal of fission products. This reactor has the capability of burning up 1,280 kg of actinides per year while at the same time producing 3,000 MWth. The report discusses the baseline design, engineering issues, modeling results, safety issues, and fuel cycle impact.

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

    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

  8. Advanced fuels with reduced actinide generation. Proceedings of a technical committee meeting

    Nuclear energy can play an important future role in supplying the world population with energy. However, this form of energy will be successful only under certain conditions: it must meet very strict safety requirements, it must be economically competitive, and it must be acceptable to the public. Nuclear power produces radioactive wastes and in several countries the public raises concern about safety. Much development work on advanced nuclear power systems is going on in several countries, with participation of both governmental and private industries to meet these conditions. In the framework of this IAEA activity the Technical Committee Meeting on Advanced Fuels with Reduced Actinide Generation was organized. The aim of the meeting was to highlight current research activities and to identify new research areas and fields of possible co-operation. The scope of the meeting included advanced fuels for all types of nuclear reactors: light water reactors, heavy water reactors, high temperature reactors, fast reactors, molten salt reactors and for accelerator driven systems. Other topics covered a wide range of investigations made, or to be made in the Member States. Refs, figs, tabs

  9. Synthesis and structural characterization of actinide oxalate compounds

    Oxalic acid is a well-known reagent to recover actinides thanks to the very low solubility of An(IV) and An(III) oxalate compounds in acidic solution. Therefore, considering mixed-oxide fuel or considering minor actinides incorporation in ceramic fuel materials for transmutation, oxalic co-conversion is convenient to synthesize mixed oxalate compounds, precursors of oxide solid solutions. As the existing oxalate single crystal syntheses are not adaptable to the actinide-oxalate chemistry or to their manipulation constrains in gloves box, several original crystal growth methods were developed. They were first validate and optimized on lanthanides and uranium before the application to transuranium elements. The advanced investigations allow to better understand the syntheses and to define optimized chemical conditions to promote crystal growth. These new crystal growth methods were then applied to a large number of mixed An1(IV)-An2(III) or An1(IV)-An2(IV) systems and lead to the formation of the first original mixed An1(IV)-An2(III) and An1(IV)-An2(IV) oxalate single crystals. Finally thanks to the first thorough structural characterizations of these compounds, single crystal X-ray diffraction, EXAFS or micro-RAMAN, the particularly weak oxalate-actinide compounds structural database is enriched, which is essential for future studied nuclear fuel cycles. (author)

  10. Bulk separation of actinides and lanthanides from actual high level liquid waste of PUREX origin using tetra-(2-ethylhexyl) diglycolamide

    Partitioning and transmutation of minor actinides is emerging as one of the preferred options for the management of high level liquid waste (HLLW) generated during the reprocessing of spent nuclear fuel. Various processes viz., DIAMEX process, TRUEX process etc. have been developed and are being tested for their use in actual application. Of late, a new class of extractant, diglycolamides, with good radiation and chemical stability and complete incinerability has emerged as the front runner for partitioning. This paper deals with the testing of indigenous and bulk synthesized N,N,N',N' tetra-(2-ethylhexyl) diglycolamide (TEHDGA) for the bulk separation of actinides and lanthanides (An and Ln) from actual HLLW

  11. A fusion transmutation of waste reactor

    A design concept and the performance characteristics for a fusion transmutation of waste reactor (FTWR)--a sub-critical fast reactor driven by a tokamak fusion neutron source--are presented. The present design concept is based on nuclear, processing and fusion technologies that either exist or are at an advanced stage of development and on the existing tokamak plasma physics database. A FTWR, operating with keff≤0.95 at a thermal power output of about 3 GW and with a fusion neutron source operating at Qp=1.5-2, could fission the transuranic content of about a hundred metric tons of spent nuclear fuel per full-power-year and would be self-sufficient in both electricity and tritium production. In equilibrium, a nuclear fleet consisting of Light Water Reactors (LWRs) and FTWRs in the electrical power ratio of 3/1 would reduce the actinides discharged from the LWRs in a once-through fuel cycle by 99.4% in the waste stream that must be stored in high-level waste repositories

  12. Assessment of the transmutation capability an accelerator driven system cooled by lead bismuth eutectic alloy

    1. PURPOSE The reduction of long-lived fission products (LLFP) and minor actinides (MA) is a key point for the public acceptability and economy of nuclear energy. In principle, any nuclear fast reactor is able to burn and transmute MA, but the amount of MA content has to be limited a few percent, having unfavourable consequences on the coolant void reactivity, Doppler effect, and delayed neutron fraction, and therefore on the dynamic behaviour and control. Accelerator Driven Systems (ADS) are instead able to safely burn and/or transmute a large quantity of actinides and LLFP, as they do not rely on delayed neutrons for control or power change and the reactivity feedbacks have very little importance during accidents. Such systems are very innovative being based on the coupling of an accelerator with a subcritical system by means of a target system, where the neutronic source needed to maintain the neutron reaction chain is produced by spallation reactions. To this end the PDS-XADS (Preliminary Design Studies on an experimental Accelerator Driven System) project was funded by the European Community in the 5th Framework Program in order both to demonstrate the feasibility of the coupling between an accelerator and a sub-critical core loaded with standard MOX fuel and to investigate the transmutation capability in order to achieve values suitable for an Industrial Scale Transmuter. This paper summarizes and compares the results of neutronic calculations aimed at evaluating the transmutation capability of cores cooled by Lead-Bismuth Eutectic alloy and loaded with assemblies based on (Pu, Am, Cm) oxide dispersed in a molybdenum metal (CERMET) or magnesia (CERCER) matrices. It also describes the constraints considered in the design of such cores and describes the thermo-mechanical behaviour of these innovative fuels along the cycle. 2. DESCRIPTION OF THE WORK: The U-free composite fuels (CERMET and CERCER) were selected for this study, being considered at European level

  13. Transmutation and the Global Nuclear Energy Partnership

    , somewhat, the uranium ore and enrichment requirements at a given level of power production, but has the disadvantage of producing non-fissile plutonium isotopes and the so-called minor actinides (neptunium, americium and curium), some of which act as neutron poisons, and thus, require increasing uranium enrichment, eventually raising fuel costs beyond practical limits. The French only use one recycle of plutonium in their power reactors. The future 'burning' (transmutation by fission) of used plutonium (and the other transuranics) could, if put into large-scale practice, eliminate one of the more serious proliferation problems in the world today, the accumulation of large quantities of separated civilian plutonium. It is generally accepted by the world's technical community that the effective way to transmute transuranics is by fissioning them in a fast reactor (i.e., reactors not containing light materials used to slow down, by collision fission, neutrons in LWRs to velocities equal to thermal velocities or the media temperature). (author)

  14. Proposed future R+D activities on advanced fuel cycles at PSI

    This paper outlines proposed PSI activities for the future under the following headings: - reactor physics R+D (plutonium recycling in LWRs, Pu-burning fast reactors, actinide transmutation in accelerator-driven systems), - materials technology R+D (Pu-fuels in LWRs, materials for advanced systems). (author) 12 refs

  15. On fusion driven systems (FDS) for transmutation

    Aagren, O (Uppsala Univ., Aangstroem laboratory, div. of electricity, Uppsala (Sweden)); Moiseenko, V.E. (Inst. of Plasma Physics, National Science Center, Kharkov Inst. of Physics and Technology, Kharkov (Ukraine)); Noack, K. (Forschungszentrum Dresden-Rossendorf (Germany))

    2008-10-15

    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented

  16. Reactor physics experiments related to transmutation in the KUCA

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  17. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    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)

  18. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    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)

  19. Studies on separation, conversion and transmutation of long-living radionuclides. A contribution to advanced disposal of high-level radioactive wastes; Untersuchungen zur Abtrennung, Konversion und Transmutation von langlebigen Radionukliden. Ein Beitrag zur fortschrittlichen Entsorgung von hochradioaktiven Abfaellen

    Modolo, Giuseppe

    2014-07-01

    The future role and acceptance of nuclear energy will be decisively determined by the safe operation of existing and future facilities and by convincing solutions for nuclear waste management. With respect to the long half-lives of some radionuclides (actinides and fission products) and the related question as to whether the release of radionuclides from a repository can be prevented over very long periods of time, alternatives to the direct disposal of spent nuclear fuels are discussed internationally. As a potential complementary solution, the technological option with partitioning and transmutation (P and T) is considered. This method separates and converts the long-lived radionuclides into stable, short-lived nuclides via neutron reactions in dedicated facilities. Against this background, the first main chapter of the present work looks at the chemical separation of actinides from high-level reprocessing wastes. In order to achieve a better understanding of the processes at the molecular level, basic investigations were also performed on separating actinides(III) via liquid-liquid or liquid-solid extraction. At the same time, reversible processes were developed and tested on the laboratory scale with the aid of mixer-settlers and centrifugal extractors. The subsequent chapter focuses on separating the long-lived fission product iodine-129 from radioactive wastes as well as from process effluents arising from reprocessing. As part of this work, different simple chemical and physical techniques were developed for complete recovery with respect to transmutation or conditioning in host matrices that are sufficiently stable for final storage. Its high mobility and radiological properties make iodine-129 relevant for the long-term safety assessment of final repositories. In addition, transmutation experiments on iodine-127/129 targets were performed using high-energy protons (145-2600 MeV). Due to the expected low cross sections (<100 mb), transmutation with protons

  20. Restructuring and redistribution of actinides in Am-MOX fuel during the first 24 h of irradiation

    In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95

  1. AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report

    William Carmack; Randy D. Lee; Pavel Medvedev; Mitch Meyer; Michael Todosow; Holly B. Hamilton; Juan Nino; Simon Philpot; James Tulenko

    2005-06-01

    The U.S. Advanced Fuel Cycle Program and the Atomic Energy Canada Ltd (AECL) seek to develop and demonstrate the technologies needed to minimize the overall Pu and minor actinides present in the light water reactor (LWR) nuclear fuel cycles. It is proposed to reuse the Pu from LWR spent fuel both for the energy it contains and to decrease the hazard and proliferation impact resulting from storage of the Pu and minor actinides. The use of fuel compositions with a combination of U and Pu oxide (MOX) has been proposed as a way to recycle Pu and/or minor actinides in LWRs. It has also been proposed to replace the fertile U{sup 238} matrix of MOX with a fertile-free matrix (IMF) to reduce the production of Pu{sup 239} in the fuel system. It is important to demonstrate the performance of these fuels with the appropriate mixture of isotopes and determine what impact there might be from trace elements or contaminants. Previous work has already been done to look at weapons-grade (WG) Pu in the MOX configuration [1][2] and the reactor-grade (RG) Pu in a MOX configuration including small (4000 ppm additions of Neptunium). This program will add to the existing database by developing a wide variety of MOX fuel compositions along with new fuel compositions called inert-matrix fuel (IMF). The goal of this program is to determine the general fabrication and irradiation behavior of the proposed IMF fuel compositions. Successful performance of these compositions will lead to further selection and development of IMF for use in LWRs. This experiment will also test various inert matrix material compositions with and without quantities of the minor actinides Americium and Neptunium to determine feasibility of incorporation into the fuel matrices for destruction. There is interest in the U.S. and world-wide in the investigation of IMF (inert matrix fuels) for scenarios involving stabilization or burn down of plutonium in the fleet of existing commercial power reactors. IMF offer the

  2. Recovery of minor actinides from spent fuel using TPEN-immobilized gels

    Koyama, S.; Suto, M.; Ohbayashi, H. [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai (Japan); Oaki, H. [Solutions Research Organization, Tokyo Institute of Technology, Tokyo (Japan); Takeshita, K. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2013-07-01

    A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (so called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained

  3. Fusion/transmutation reactor studies based on the spherical torus concept

    The paper presents a conceptual design for a compact fusion/transmutation experimental reactor based on the spherical torus concept, CFER-ST. A set of plasma parameters suitable for the nuclear waste transmutation blanket are given. The transmutation neutronics, integer structure, thermo-hydraulics, liquid curtain wall and magnet shield design, etc., for two types of minor actinide transmutation blankets, namely the lead-bismuth eutectic cooled blanket and the FLiBe eutectic self-cooled blanket, along with the relevant calculation results, are presented. The preliminary results show that the proposed fusion/transmutation system and the relevant parameters can meet the design goals

  4. State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems

    Development of innovative fuels such as homogeneous and heterogeneous fuels, ADS fuels, and oxide, metal, nitride and carbide fuels is an important stage in the implementation process of advanced nuclear systems. Several national and international R and D programmes are investigating minor actinide-bearing fuels due to their ability to help reduce the radiotoxicity of spent fuel and therefore decrease the burden on geological repositories. Minor actinides can be converted into a suitable fuel form for irradiation in reactor systems where they are transmuted into fission products with a significantly shorter half-life. This report compares recent studies of fuels containing minor actinides for use in advanced nuclear systems. The studies review different fuels for several types of advanced reactors by examining various technical issues associated with fabrication, characterisation, irradiation performance, design and safety criteria, as well as technical maturity. (authors)

  5. Conjugates of Actinide Chelator-Magnetic Nanoparticles for Used Fuel Separation Technology

    Qiang, You; Paszczynski, Andrzej; Rao, Linfeng

    2011-10-30

    The actinide separation method using magnetic nanoparticles (MNPs) functionalized with actinide specific chelators utilizes the separation capability of ligand and the ease of magnetic separation. This separation method eliminated the need of large quantity organic solutions used in the liquid-liquid extraction process. The MNPs could also be recycled for repeated separation, thus this separation method greatly reduces the generation of secondary waste compared to traditional liquid extraction technology. The high diffusivity of MNPs and the large surface area also facilitate high efficiency of actinide sorption by the ligands. This method could help in solving the nuclear waste remediation problem.

  6. ACTINET: a European Network for Actinide Sciences

    Full text of publication follows: The research in Actinide sciences appear as a strategic issue for the future of nuclear systems. Sustainability issues are clearly in connection with the way actinide elements are managed (either addressing saving natural resource, or decreasing the radiotoxicity of the waste). The recent developments in the field of minor actinide P and T offer convincing indications of what could be possible options, possible future processes for the selective recovery of minor actinides. But they point out, too, some lacks in the basic understanding of key-issues (such as for instance the control An versus Ln selectivity, or solvation phenomena in organic phases). Such lacks could be real obstacles for an optimization of future processes, with new fuel compounds and facing new recycling strategies. This is why a large and sustainable work appears necessary, here in the field of basic actinide separative chemistry. And similar examples could be taken from other aspects of An science, for various applications (nuclear fuel or transmutation targets design, or migration issues,): future developments need a strong, enlarged, scientific basis. The Network ACTINET, established with the support of the European Commission, has the following objectives: - significantly improve the accessibility of the major actinide facilities to the European scientific community, and form a set of pooled facilities, as the corner-stone of a progressive integration process, - improve mobility between the member organisations, in particular between Academic Institutions and National Laboratories holding the pooled facilities, - merge part of the research programs conducted by the member institutions, and optimise the research programs and infrastructure policy via joint management procedures, - strengthen European excellence through a selection process of joint proposals, and reduce the fragmentation of the community by putting critical mass of resources and expertise on

  7. Utilization of fast reactor excess neutrons for burning minor actinides and long lived FPs

    An evaluation is made on a large MOX fuel fast reactor's capability of burning minor actinides and long lived fission products (FPs) without imposing penalties on core nuclear and safety characteristics. The excess neutrons generated in the fast reactor core are fully utilized not only to generate the fissile material but also to transmute the minor actinides and long lived FPs. The FP target assemblies which consist of Tc-99 and I-129 are loaded into the selected blanket positions whereas the minor actinides are loaded to the rest of the blanket. A long term FP accumulation scenario is also considered in the mix of FP burner fast reactor and non-burner LWRs. (author)

  8. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    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)

  9. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    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 (AnF3) 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 7Li,Na,Be/F and 7Li,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)

  10. Towards an assessment of partition and transmutation

    In France, after 15 years of research launched by the 1991 Law on High Level Waste, l'Office Parlementaire de l'Evaluation des Choix Scientifiques et Technologiques has published its report due to the Chambers after hearings of the actors of the research and of different stakeholders. The first conclusions are that Partition/Transmutation (P and T), Disposal and Interim Storage have to be considered as complementary ways of dealing with the High Level Waste. The technical feasibility of the Partition process is about to be proven, including a small scale demonstration on active solutions in the Atalante facility. The feasibility of Transmutation will be comforted by the present experiments in the Phenix facility. There are several recommendations in the report: the timing of implementation of a partition transmutation cycle in France will have to coincide with the need for a renewal of the present reprocessing plant and the need for a fuel cycle serving a new generation of Power Reactors. At the same time, P and T may induce disposal cost reduction that may cover a part of the cost in P and T. As a matter of fact, present reprocessing and recycling reduces the waste volume and toxicity and therefore is beneficial to the disposal capacity and cost. However, at the present time, disposal costs reduction due to P and T are difficult to estimate, since optimization of disposal is still to be done. P and T costs are difficult to estimate as well, since they may include not only separative shops, targets fabrication and reprocessing units but actinide incinerators. This paper provides some reflections on the possible benefits in implementing P and T. The benefits will come either from toxicity reduction, thermal load reduction, or from other less tangible reasons as it may increase the positive perception of a disposal by reducing uncertainties or time perspective. Our proposition is to focus more the next R and D programs on the assessments of P and T benefits and

  11. Methods For The Calculation Of Pebble Bed High Temperature Reactors With High Burnup Plutonium And Minor Actinide Based Fuel

    The graphite moderated Modular High Temperature Pebble Bed Reactor enables very flexible loading strategies and is one candidate of the Generation IV reactors. For this reactor fuel cycles with high burnup (about 600 MWd/kg HM) based on plutonium (Pu) and minor actinides (MA) fuel will be investigated. The composition of this fuel is defined in the EU-PuMA-project which aims the reduction of high level waste. There exist nearly no neutronic full core calculations for this fuel composition with high burnup. Two methods (deterministic and Monte Carlo) will be used to determine the neutronics in a full core. The detailed results will be compared with respect to the influence on criticality and safety related parameters. (authors)

  12. Partitioning and Transmutation - Physics, Technology and Politics

    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

  13. Partitioning and Transmutation. Annual Report 2002

    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

  14. 次锕系元素在加速器驱动的次临界快堆中嬗变的研究%Study of Transmutation of Minor Actinides in Accelerator-Driven Sub-critical Fast Reactor

    杨永伟; 古玉祥

    2001-01-01

    选取加速器驱动次临界快堆(ADSFR),进行嬗变来自于PWR(U)乏燃料 中次锕系元素 的研究。在堆芯内,燃料为NpAmCm的氧化物,选取液态钠为冷却剂。利用下列程序对所选方 案进行物理计算和分析:LAHET -模拟质子与靶核的相互作用;MCNP4A-模拟次临界包层内 20MeV以下的中子与材料核的相互作用;ORIGEN2-利用MCNP4A的输出提供的一群等效截面对 堆芯进行燃耗计算。计算分析的结果表明:考虑临界安全、功率密度和燃耗等因素,利用所 选方案进行次锕系元素嬗变是可行的。%Accelerator-Driven Sub-critical Fast Reactor (ADSFR)is chosenfor transmu ta tion of minor actinides from the spent fuel of PWR(U). In the core, the fuel type is (PuNpAmCm)Ox. Liquid sodium is chosen as coolant The neutronics calcul ation and analysis of the selected scheme have been done by using the following codes: LAHET, for the simulation of the interaction between the protons and the nuclei of the target; MCNP4A, for the simulation of interaction between neutron s with energy below 20MeV and the nuclei of materials in the sub-critical blank e t; ORIGEN2, for the multi-region burnup calculation of the blanket by using the one-group effective cross-section provided in the output of MCNP4A. The neutro ni cs calculation and analysis show that the proposed scheme is feasible for trans mutation of minor actinides, considering the factors such as the criticality s afety, power density, burnup, etc.

  15. A study of the transmutation performance of externally driven sub-critical assemblies

    For transmutation systems based on externally driven sub-critical assemblies with a fast neutron spectrum, there is an incentive to expose the actinides directly to the source neutrons, since these neutrons have higher energies than the fission neutrons. To evaluate the transmutation effectiveness of such systems, a parameter study based on the PHOENIX system, i.e. a sodium-cooled system with a minor actinide (MA) oxide fuelled target was performed. An interesting result is that the high-energy source neutrons give rise to a 20-25% increase in the fission-to-capture ratio of the important (fissionable) nuclides. Moreover, the performance of such a system can be further improved by substituting the oxide fuel by metal fuel and by reducing the volume fraction of steel in the target. Replacing the liquid sodium coolant by liquid lead has only a small effect on the fission-to-capture ratio, however, for a given proton current, the neutron production in the target increases. 17 refs., 5 tabs., 2 figs

  16. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SFLn/Am obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as a Zr

  17. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    Sypula, Michal

    2013-07-01

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SF{sub Ln/Am} obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as

  18. 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'

    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)

  19. Preliminary neutronics analysis of a spallation target for transmutation

    Accelerator Driven subcritical System (ADS) was recognized as an effective nuclear waste transmutation device. Target in liquid or solid in an independent loop bombarded by the charged particle beam was considered as the neutron source. Heavy metal was chosen as target material or coolant. The present work was to discuss the possibility of taking Minor Actinides as part of spallation target material, for a better transmutation performance of entire ADS. According to the thermal cooling and irradiation time limitation, a conceptual design of target for transmutation was proposed. And preliminary neutronics analysis for target performance assessment including neutron flux, neutron yield as well neutron spectrum is shown in this work. (author)

  20. Comparison of accelerator-based with reactor-based nuclear waste transmutation schemes

    An overview of the most significant studies in the last 35 years of partitioning and transmutation of commercial light water reactor spent fuel is given. Recent Accelerator-based Transmutation of Waste (ATW) systems are compared with liquid-fuel thermal reactor systems that accomplish the same objectives. If no long-lived fission products (e.g. 99Tc and 129I) are to be burned, under ideal circumstances the neutron balance in an ATW systems becomes identical to that for a thermal reactor system. However, such a reactor would need extraordinarily rapid removal of internally-generated fission products to remain critical at equilibrium without enriched feed. The accelerator beam thus has two main purposes (1) the burning of long-lived fission products that could not be burned in a comparable reactor's margin (2) a relaxing of on-line chemical processing requirements without which a reactor-based system cannot maintain criticality. Fast systems would require a parallel, thermal ATW system for long-lived fission product transmutation. The actinide-burning part of a thermal ATW system is compared with the Advanced Liquid Metal Reactor (ALMR) using the well-known Pigford-Choi model. It is shown that the ATW produces superior inventory reduction factors for any near-term time scale. (author)

  1. Preliminary neutronics design analysis on accelerator driven subcritical reactor for nuclear waste transmutation

    By taking minor actinides (MA) transmutation performance as evaluation index, preliminary neutronics design analyses were performed on ADS-NWT which is a lead-alloy cooled accelerator driven subcritical reactor for nuclear waste transmutation. In the specific design, liquid lead-bismuth eutectic (LBE) and transuranic metallic dispersion fuel were used as coolant and a fuel of ADS-NWT, respectively. The neutronics calculations and analyses were performed by using CAD-based multi-functional 4D neutronics and radiation simulation system named VisualBUS and the nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). The preliminary results showed that based on specific deign of MA/Pu volume ratio of 7 : 3, the transmutation rate of MA was approximately 650 kg/a, the high thermal reactor power output was ∼1000 MW when energy self-sustaining was satisfied and relatively deep subcriticality and negative reactivity coefficients made sure of good inherent safety of ADS-NWT. (authors)

  2. The U.S. accelerator transmutation of waste program

    A national project to develop a future capability to separate actinides and long-lived fission products from spent fuel, to transmute them, and to dispose off the remaining waste in optimal waste forms has begun in the United States. This project is based on the Accelerator-driven Transmutation of Waste (ATW) program developed during the 1990s at Los Alamos National Laboratory, and has its technological roots in several technologies that have been developed by the multi-mission laboratories of the U.S. Department of Energy (DOE). In the Fiscal Year 1999 Energy and Water Appropriation Act, the U.S. Congress directed the DOE to study ATW and by the end of FY99 to prepare a 'roadmap' for developing this technology. DOE convened a steering committee, assembled four technical working groups consisting of members from many national laboratories, and consulted with several individual international and national experts. The finished product, 'A Roadmap for Developing ATW Technology - A Report to Congress', recommends a five-year, $281 M, science-based, technical-risk-reduction program. This paper provides an overview of the U.S. Roadmap for developing ATW technology, the organization of the national ATW Project, the critical issues in subsystems and technological options, deployment scenarios, institutional challenges, and academic and international collaboration

  3. Neutronic features of pebble-bed reactors for transmutation applications

    Pebble-bed reactors offer very appealing characteristics for radioactivity confinement and for withstanding thermal transients. Besides that, pebble-bed reactors have a peculiar degree of freedom in the radius of the active core of the pebble (where the fuel is located) as compared to the outer radius of the pebble, which has a coating of pure graphite. By varying the aforementioned radius, very different types of neutron spectra can be formed, which in turn gives very different values of the average cross sections that govern the isotopic composition evolution, and particularly the elimination of the most relevant transuranics. Preliminary conclusions of this work show that there is a very broad design window for exploiting the transmutation capabilities of pebble-bed reactors in a scenario of inherent safety features. A 99,9% elimination of Pu-239 associated to a 99% elimination of Pu-240 and Pu-241 can be reached, with some increment of the Pu-242 contents (which is extremely long-lived, less radio-toxic and decays into the natural nuclide U-238). Am and Cm are also transmuted to a significant level, although some residual higher A actinides will remain. (authors)

  4. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  5. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  6. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading

  7. Advanced Aqueous Separation Systems for Actinide Partitioning

    Nash, Ken [Washington State Univ., Pullman, WA (United States); Martin, Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lumetta, Gregg [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-04-02

    One of the most challenging aspects of advanced processing of used nuclear fuel is the separation of transplutonium actinides from fission product lanthanides. This separation is essential if actinide transmutation options are to be pursued in advanced fuel cycles, as lanthanides compete with actinides for neutrons in both thermal and fast reactors, thus limiting efficiency. The separation is difficult because the chemistry of Am3+ and Cm3+ is nearly identical to that of the trivalent lanthanides (Ln3+). The prior literature teaches that two approaches offer the greatest probability of devising a successful group separation process based on aqueous processes: 1) the application of complexing agents containing ligand donor atoms that are softer than oxygen (N, S, Cl-) or 2) changing the oxidation state of Am to the IV, V, or VI state to increase the essential differences between Am and lanthanide chemistry (an approach utilized in the PUREX process to selectively remove Pu4+ and UO22+ from fission products). The latter approach offers the additional benefit of enabling a separation of Am from Cm, as Cm(III) is resistant to oxidation and so can easily be made to follow the lanthanides. The fundamental limitations of these approaches are that 1) the soft(er) donor atoms that interact more strongly with actinide cations than lanthanides form substantially weaker bonds than oxygen atoms, thus necessitating modification of extraction conditions for adequate phase transfer efficiency, 2) soft donor reagents have been seen to suffer slow phase transfer kinetics and hydro-/radiolytic stability limitations and 3) the upper oxidation states of Am are all moderately strong oxidants, hence of only transient stability in media representative of conventional aqueous separations systems. There are examples in the literature of both approaches having been described. However, it is not clear at present that any extant process is sufficiently robust for application at the scale

  8. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

  9. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

    1990-08-01

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs.

  10. Advanced fuel developments for an industrial accelerator driven system prototype

    Delage, Fabienne; Ottaviani, Jean Pierre [Commissariat a l' Energie Atomique CEA (France); Fernandez-Carretero, Asuncion; Staicu, Dragos [JRC-ITU (Germany); Boccaccini, Claudia-Matzerath; Chen, Xue-Nong; Mascheck, Werner; Rineiski, Andrei [Forschungszentrum Karlsruhe - FZK (Germany); D' Agata, Elio [JRC-IE (Netherlands); Klaassen, Frodo [NRG, PO Box 25, NL-1755 ZG Petten (Netherlands); Sobolev, Vitaly [SCK-CEN (Belgium); Wallenius, Janne [KTH Royal Institute of Technology (Sweden); Abram, T. [National Nuclear Laboratory - NNL (United Kingdom)

    2009-06-15

    Fuel to be used in an Accelerator Driven System (ADS) for transmutation in a fast spectrum, can be described as a highly innovative concept in comparison with fuels used in critical cores. ADS fuel is not fertile, so as to improve the transmutation performance. It necessarily contains a high concentration ({approx}50%) of minor actinides and plutonium. This unusual fuel composition results in high gamma and neutron emissions during its fabrication, as well as degraded core performance. So, an optimal ADS fuel is based on finding the best compromise between thermal, mechanical, chemical, neutronic and technological constraints. CERCER and CERMET composite fuels consisting of particles of (Pu,MA)O{sub 2} phases dispersed in a magnesia or molybdenum matrix are under investigation within the frame of the ongoing European Integrated Project EUROTRANS (European Research programme for Transmutation) which aims at performing a conceptual design of a 400 MWth transmuter: the European Facility for Industrial Transmutation (EFIT). Performances and safety of EFIT cores loaded with CERCER and CERMET fuels have been evaluated. Out-of-pile and in-pile experiments are carried out to gain knowledge on the properties and the behaviour of these fuels. The current paper gives an overview of the work progress. (authors)

  11. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  12. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    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

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

    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

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

    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

  15. Nuclear transmutation characteristics of reduced moderation BWR (Thesis)

    In the present thesis, the nuclear transmutation characteristics of reduced moderation BWR, which decides the spent fuel characteristics and its safety in its nuclear fuel cycle, were investigated and compared with other types of reactors. The major conclusions were obtained as follows: The decay heat and radioactivity from FPs increases in fuel burn-up. However, they which normalized with burn-up are small for the reactor with low specific power and long operation period due to the decay during the long operation period. Breeder type of reduced moderation BWR shows low decay heat and radioactivity from FPs because of the long operation period approximately 3000 days which realized by the high conversion ratio. That also shows low decay heat and radioactivity from actinide nuclides due to the hard spectrum. MA recycling reactor of high conversion type of reduced moderation BWR was designed. The neptunium, which has large impact for environmental burden from the viewpoint of nuclide transport analysis, can be incinerated approximately 40% of loaded inventory which corresponds to 22 units of LWR per year. LLFP (99Tc, 129I, 135Cs) transmutation by breeder type of reduced moderation BWR was estimated. As a result, the support factor cannot be lower than unity for each LLFP nuclides. In other words, the reduced moderation BWR cannot reduce LLFP because the LLFP target cannot be loaded inner of the reactor core due to the small margin of core specification. It is expected that these results and the characteristics of other types of reactor shown in the present study benefit the discussion for various nuclear fuel cycle options. (author)

  16. Sphere-Pac Evaluation for Transmutation

    Icenhour, A.S.

    2005-05-19

    The U.S. Department of Energy Advanced Fuel Cycle Initiative (AFCI) is sponsoring a project at Oak Ridge National Laboratory with the objective of conducting the research and development necessary to evaluate the use of sphere-pac transmutation fuel. Sphere-pac fuels were studied extensively in the 1960s and 1970s. More recently, this fuel form is being studied internationally as a potential plutonium-burning fuel. For transmutation fuel, sphere-pac fuels have potential advantages over traditional pellet-type fuels. This report provides a review of development efforts related to the preparation of sphere-pac fuels and their irradiation tests. Based on the results of these tests, comparisons with pellet-type fuels are summarized, the advantages and disadvantages of using sphere-pac fuels are highlighted, and sphere-pac options for the AFCI are recommended. The Oak Ridge National Laboratory development activities are also outlined.

  17. Transmutation of neptunium, americium, technetium, and iodine in fast spectrum cores driven by accelerated protons

    A neutronic analysis is presented of three incinerator subcritical lattices, driven by accelerated protons and designed to transmute the minor actinides, the 99Tc and the 129I, of light water reactor (LWR) waste. A calculational methodology must first be established to enable a neutronic burnup analysis of fission cores driven by high-energy protons. The methodology is based on the following codes: HERMES, the Forschungszentrum Juelich adaptation of HETC, for high-energy interactions; MCNP3, for neutron interactions below 20 MeV of neutron energy; and KORIGEN, the Forschungszentrum Karlsruhe adaptation of ORIGEN, for burnup analysis. A result of applying the methodology is that the minor actinides, the 99Tc, and the 129I, of LWR waste may be transmuted in subcritical cores, driven by the spallation neutrons emanating from the bombardment of the cores with 1600-MeV protons. Three cores types are required. Core type 1 is fueled by the minor actinides and is a modification of the Brookhaven National Laboratory PHOENIX. With a proton current of 20 mA, the core incinerates the minor actinide waste of 14 LWRs. Core type 2 contains the 99Tc, 129I, and plutonium waste of 19 LWRs. With a proton beam of 130 mA, the core incinerates the technetium and 60% of the iodine. With a faction of the plutonium coming out of this core, the remaining 40% of 129I is incinerated in core type 3. All three cores run to 100,000 MWd/tonne or slightly higher; on the average, no core is a net consumer of grid electricity; all are cooled by sodium but remain subcritical with the loss of coolant

  18. Hydrophilic Clicked 2,6-Bis-triazolyl-pyridines Endowed with High Actinide Selectivity and Radiochemical Stability: Toward a Closed Nuclear Fuel Cycle.

    Macerata, Elena; Mossini, Eros; Scaravaggi, Stefano; Mariani, Mario; Mele, Andrea; Panzeri, Walter; Boubals, Nathalie; Berthon, Laurence; Charbonnel, Marie-Christine; Sansone, Francesco; Arduini, Arturo; Casnati, Alessandro

    2016-06-15

    There is still an evident need for selective and stable ligands able to separate actinide(III) from lanthanide(III) metal ions in view of the treatment of the accumulated radioactive waste and of the recycling of minor actinides. We have herein demonstrated that hydrophilic 2,6-bis-triazolyl-pyridines are able to strip all actinides in all the different oxidation states from a diglycolamide-containing kerosene solution into an acidic aqueous phase. The ascertained high actinide selectivity, efficiency, extraction kinetics, and chemical/radiolytic stability spotlight this hydrophilic class of ligands as exceptional candidates for advanced separation processes fundamental for closing the nuclear fuel cycle and solving the environmental issues related to the management of existing nuclear waste. PMID:27203357

  19. Decay calculations on medium-level and actinide-containing wastes from the LWR fuel cycle. Pt. 1

    A number of basic data on medium-level and actinide-containing waste streams from the LWR fuel cycle were evaluated and the activity and thermal decay power were calculated for the nuclide inventories of cladding hulls and fuel assembly structural materials, for feed clarification sludge, medium-level aqueous process waste, low-level solid transuranium waste and for medium-level reactor operating waste. The activity as a function of decay time of the medium-level wastes decreases within 500 to 600 years by 1 to 3 orders of magnitude and is at the same time about 1 to 2 orders of magnitude lower than the activity of the high-level waste. The thermal decay power of the medium-level wastes decreases after 10 to 100 years by about 3 orders of magnitude and is about a factor of 10 to 100 less than that of high-level waste. In the very long term the residual activity (and thermal power) decreases only slowly due to the long halflives of the dominant actinides. The activity after more that 1000 years is about 1 to 2 orders of magnitude lower than that of high-level waste, the low-level transuranium waste by a factor 10 to 4, respectively. The activity per unit volume of the packaged waste of the medium-level and actinide-containing wastes because of the bigger volume of the conditioned wastes is lower by 2 to 4 orders of magnitude up to about 500 years. After more than 1000 years the activities per unit volume are lower by a factor of 20 to 200 than that of high-level waste. (orig.)

  20. Advanced processes for minor actinides recycling: studies towards potential industrialization

    In June 2006, a new act on sustainable management of radioactive waste was voted by the French parliament with a national plan on radioactive materials and radioactive waste management (PNG-MDR). Concerning partitioning and transmutation, the program is connected to 4. generation reactors, in which transmutation of minor actinides could be operated. In this frame, the next important milestone is 2012, with the assessment of the possible transmutation roads, which are either homogeneous recycling of the minor actinides in the whole reactor fleet, with a low content of M.A (∼3%) in all fuel assemblies, or heterogeneous recycling of the minor actinides in about one third of the reactor park, with a higher content of M.A. (∼20%) in dedicated targets dispatched in the periphery of the reactor. Advanced processes for the recycling of minor actinides are being developed to address the challenges of these various management options. An important part of the program consists in getting closer to process implementation conditions. The processes based on liquid-liquid extraction benefit from the experience gained by operating the PUREX process at the La Hague plant. In the field of extracting apparatus, a large experience is available. In the field of extracting apparatus, a large experience is already available. Nevertheless, the processes present specificities which have to be considered more precisely. They have been classified in the following fields: - Evolution of the simulation codes, including phenomenological representations: with such a simulation tool, it will be possible to assess operating tolerances, lead sensitivity studies and calculate transient states; - Definition of the implementation conditions in continuous contactors (such as pulse columns), according to the extractant physico-chemical characteristics; - Scale-up of new extractants, such as malonamides used in the DIAMEX process, facing purity specifications and costs estimation; - Solvent clean

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

    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)

  2. Study of kinetics of extraction of actinides in processes of reprocessing of nuclear fuels

    This research thesis reports a bibliographical study on extraction kinetics. After some generalities on solvent-based extraction, and on the chemistry of actinides in solution, on the methods of kinetics study which are generally used and their mathematical treatments, the author compares the published results for the extraction kinetics of nitric acid, uranium VI, uranium IV, neptunium IV and plutonium IV

  3. Subcritical set coupled to accelerator (ADS) for transmutation of radioactive wastes: an approach of computational modelling

    Nuclear fission devices coupled to particle accelerators ADS are being widely studied. These devices have several applications, including nuclear waste transmutation and producing hydrogen, both applications with strong social and environmental impact. The essence of this work was to model an ADS geometry composed of small TRISO fuel loaded with a mixture of MOX uranium and thorium target material spallation of uranium, using methods of computational modeling probabilistic, in particular the MCNPX 2.6e program to evaluate the physical characteristics of the device and their ability to transmutation. As a result of the characterization of the spallation target, it can be concluded that production of neutrons per incident proton increases with increasing dimensions of the spallation target (thickness and radius), until it reached the maximum production of neutrons per incident proton or call the region saturation. The results obtained in modeling the ADS device bed kind of balls with respect to isotopic variation in the isotopes of plutonium and minor actinides considered in the analysis revealed that accumulation of mass of the isotopes of plutonium and minor actinides increase for subcritical configuration considered. In the particular case of the isotope 239Pu, it is observed a reduction of the mass from the time of burning of 99 days. The increase of power in the core, whereas tungsten spallation targets and Lead is among the key future developments of this work

  4. SIMULATION OF THE NUCLEAR-REACTOR ACTIVE-ZONE ELEMENTS WITH THICK ROTATING LAYER OF MICRO-PARTICLE FUEL FOR RADIOACTIVE WASTE TRANSMUTATION

    V. V. Sorokin

    2015-06-01

    Full Text Available The effective transmutation of radioactive isotopes into the stable ones with the use of neutrons requires the neutron high-flux and the spectra with significant part of fast and resonance neutrons. It is advisable to alternate a range of specified-duration irradiation sessions with revamping the composition of waste. The depleted fuel of the commercial reactor comprises near 1 % of such isotopes of their individual mass in the batch loading which amounts to several tens of kilograms. The article considers a perspective nuclear reactor for radioactive waste transmutation as regards its design, thermal physics and hydrodynamics. Mobile microparticles of the fuel build up the active zone of the reactor and form a steady dense ringshaped layer. The layer rotates within immovable vortex chamber using the energy of the coolant, i.e. water. The micro particles cool down with the coolant unmediated.The  formulaic  valuation  of  the  device  capacity  with  water  under  pressure  comes to 1–5 MW per 1 liter of the layer. The condition of avoided boiling sets the most restrictive limitations  to  the  capacity.  The  bulk  of  the  layer  constricts  to  tens  of  liters  inasmuch as enlarging the chamber dimensions reduces the rotary acceleration and the force confining the fuel micro-particles on the free surface of the layer. The author offers and substantiates with calculations the active zone composed of several layers or a layer with a large ratio of the volume to the surface area for achieving criticality of nuclear fuel load with limitations on enrichment. The vortex chambers in case of the active zone of several layers can have the joint coolant exscapes along the axis. Implementation of the chambers with reverse vortices in composite active zones with joint escapes allows reducing the flow rotation below the vortex reactor along the coolant course.

  5. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR → reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k∞) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k∞ using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO2 (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% 237Np for transmutation purposes (T). Results: k∞ based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k∞ decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k∞ decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth ∼-20%Δk/k), hence improving the effectiveness of packaging. (author)

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

    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

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

    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

  8. Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems

    H.R. Trellue

    2003-06-01

    Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium ({approx}40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation.

  9. Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems

    Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium (∼40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation

  10. Studies on separation, conversion and transmutation of long-living radionuclides. A contribution to advanced disposal of high-level radioactive wastes

    The future role and acceptance of nuclear energy will be decisively determined by the safe operation of existing and future facilities and by convincing solutions for nuclear waste management. With respect to the long half-lives of some radionuclides (actinides and fission products) and the related question as to whether the release of radionuclides from a repository can be prevented over very long periods of time, alternatives to the direct disposal of spent nuclear fuels are discussed internationally. As a potential complementary solution, the technological option with partitioning and transmutation (P and T) is considered. This method separates and converts the long-lived radionuclides into stable, short-lived nuclides via neutron reactions in dedicated facilities. Against this background, the first main chapter of the present work looks at the chemical separation of actinides from high-level reprocessing wastes. In order to achieve a better understanding of the processes at the molecular level, basic investigations were also performed on separating actinides(III) via liquid-liquid or liquid-solid extraction. At the same time, reversible processes were developed and tested on the laboratory scale with the aid of mixer-settlers and centrifugal extractors. The subsequent chapter focuses on separating the long-lived fission product iodine-129 from radioactive wastes as well as from process effluents arising from reprocessing. As part of this work, different simple chemical and physical techniques were developed for complete recovery with respect to transmutation or conditioning in host matrices that are sufficiently stable for final storage. Its high mobility and radiological properties make iodine-129 relevant for the long-term safety assessment of final repositories. In addition, transmutation experiments on iodine-127/129 targets were performed using high-energy protons (145-2600 MeV). Due to the expected low cross sections (<100 mb), transmutation with protons

  11. Separation of Nuclear Fuel Surrogates from Silicon Carbide Inert Matrix

    The objective of this project has been to identify a process for separating transuranic species from silicon carbide (SiC). Silicon carbide has become one of the prime candidates for the matrix in inert matrix fuels, (IMF) being designed to reduce plutonium inventories and the long half-lives actinides through transmutation since complete reaction is not practical it become necessary to separate the non-transmuted materials from the silicon carbide matrix for ultimate reprocessing. This work reports a method for that required process

  12. Radiotoxicity and Risk Reduction of TRU Elements from Spent Fuel by Transmutation in the Light Water Reactor

    A conventional PWR of type VVER-440 operating in a sustainable advanced fuel cycle mode with complete recycling of TRU elements in an Inert Matrix Combined Fuel Assembly (IMC-FA) in the same reactor was investigated. A preliminary assessment with the differences between various nuclear fuel cycles in terms of the risk analysis and its indicators has been conducted. The results indicate that the sustainable advanced fuel cycle option can, for the same amount of energy generation, significantly reduces both the amounts and radiotoxicity of the spent nuclear fuel in comparison with the conventional once-through UO2 or MOX fuel cycles

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

    García, C.; Rosales, J.; García, L.; Pérez-Navarro, A.; Escrivá, A.; Abánades Velasco, Alberto

    2011-01-01

    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, cooled by helium which enables high temperatures (...

  14. Gamma ray beam transmutation

    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 197Au and 129Iodine. 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 129Iodine 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

  15. Neutron measurements for innovative fuel cycle and transmutation performed at the CEN Bordeaux-Gradignan : transfer techniques applied to the protactinium case

    Transfer reaction techniques have been used to determine neutron induced fission cross section (σn,f) of the short lived 233Pa nucleus, which is of importance for the Th-U fuel cycle for innovative reactors. The σn,f of 233Pa has been determined from the product of the fission probability of 234Pa measured in transfer reaction 232Th(3He,p) with the calculated compound nucleus formation cross section in the 233Pa+n reaction. The validity of this method has been tested with the existing data for direct neutron experiments on long-lived target nuclei 231Pa and 230Th. Transfer reaction techniques have been used too for the determination of capture cross section (σn,y) of 233Pa. This method will be extended to other highly radioactive actinides (such as 242-245Cm isotopes). (author)

  16. Comparison of radiotoxicity of uranium, plutonium, and thorium spent nuclear fuel at long-term storage

    Time dependence of radio-toxicity of actinides from spent uranium, MOX-plutonium, and thorium fuel calculated for storage during 1000 years is discussed in the paper. Calculations are based on the nuclear fuel of the VVER-1000 type reactor. Recommendations for uranium and plutonium spent fuel could be done to perform chemical separation of plutonium, americium, curium before long-term controllable storage. Americium should be separated after 50-70 years of storage for sufficient conversion of Pu-241 in Am-241. Cm-244 decays almost completely after 100 years. Extracted americium (possibly, with long-lived curium isotopes) should be directed to transmutation and plutonium should be reused. The separation of actinides is also effective to reduce decay heat power. In thorium spent fuel, the overwhelming share of radio-toxicity is determined by U-232. It is obvious that the repeated use of thorium fuel will be accompanied by accumulation of radio-toxicity. For a one-fold use of thorium fuel with deep U-233 burnup, it is necessary to perform additional deep burn-out (transmutation) of uranium fraction containing both U-233 and U-232. The further reduction of radio-toxicity by several orders can be obtained by extraction and transmutation of plutonium fraction (Pu-238). The transmutation of Th-228 - daughter nuclide of U-232 - is not necessary because Th-228 decays practically completely in 10 years together with its short-lived daughter nuclides

  17. Nuclear waste transmutation

    A deep repository for safe long-term storage of long-lived radioactive materials (waste) arising from nuclear fuel irradiation in reactors is a need generally accepted, whatever the strategy envisaged for further use of the irradiated fuel (e.g.: reprocessing and re-use of uranium and plutonium; no reprocessing and final disposal). To assess the impact on the environment of a waste repository, one is lead naturally to consider the impact of radiation on man and to define the radiotoxicity of the different isotopes. The toxicity of the materials stored in a repository is function of time and at a given time is the sum of the activities of each radionuclide multiplied by appropriate danger coefficients. This time dependent sum R, is a source of 'potential' radiotoxicity. It has been pointed out (in reference 1), that R does not measure 'risk', which has to take into account 'actual pathways and probability of radioactive release to the biosphere'. It is well understood that (e.g. in the case of spent PWR fuel) the main contributor to R are actinides, Pu being the main component (see table I). In the case of risk, the situation is by far more complex and dependent on the modeling of different geological environments. In the analysis made in reference 1 the predominant role of Tc-99, I-129 and Cs-135 has been pointed out. The same analysis also stresses that actinides will be by far less relevant with respect to the highly soluble and mobile fission products. (authors). 13 refs., 2 tabs., 2 figs

  18. Concept of the heavy water MA-burner with the neutral fuel matrix

    The concept of the heavy water moderated and cooled critical MA-burner with the solid neutral fuel matrix is proposed. The distinguishing feature of the system is the high thermal neutron flux level. This leads to the high neutron reaction rates on the actinides and, consequently, to the low values of MA transmutation time. The concept of MA stage-transmutation strategy is proposed for this system. The transmutation process is divided into several time-stages of different duration and each of them includes a proper number of the burner's identical fuel cycles with the stage-peculiar feed and discharge fuel compositions. Some basic design features of the proposed MA burner are given. Results of one MA stage-transmutation strategy are presented. It is concluded that the proposed concept promises to be an efficient one and may be realized based on the current technologies, regarding both system design and fuel reprocessing ones. Some possible ways of the stage-transmutation strategy efficiency further increasing are proposed, in particular, reasonable distribution of transmutation stages between the fast systems and the thermal ones. (author)

  19. EU strategy in partitioning and transmutation and its implementation within the EURATOM Framework Programmes

    A robust strategy is followed in the European Union (EU) in the area of partitioning and transmutation (P and T) for sustainability of nuclear energy by promoting collaborative research and training among member states within the EURATOM multi-annual Framework Programmes (FP) and Sustainable Nuclear Energy Technology Platform (SNE-TP). Once-through cycle ('disposal of spent fuel as is') does not appear to be sustainable for nuclear energy production. Reprocessing of the spent fuel and transmutation of minor actinides in dedicated devices would reduce considerably the radio-toxic inventory of the disposed waste in geological repositories. This is of significant importance in non-proliferation strategy and radiological terrorism and reduces risks in case of an inadvertent human intrusion. The separation of main heavy metals (uranium and plutonium) reduces the volume and thermal output of the waste to be disposed of, which increases effectively the capacity of the repository. Furthermore, extraction of heat-bearing components (Sr and Cs) from the waste 'can reduce thermal output' of the disposed waste. Concerns of the public related to long-life of the waste could largely be overcome by P and T as it would reduce half-life of most of the waste to be disposed of to a couple of hundred years and it could thus come to the aid of geological disposal community in securing a broadly agreed political solution of waste disposal in geological repositories. Nevertheless, additional cost, additional secondary waste, activation products, intermediate-level waste and dose to workers in the process of P and T itself will contribute to defining an optimal transmutation scheme. A double-strata approach with subcritical accelerator-driven systems (ADS) and/or critical fast reactors (Generation IV systems) is being considered. A decision on the choice is planned in a couple of years. Geological disposal of the remaining waste (separation/transmutation losses) will nevertheless be

  20. Advanced Fuel Cycle Initiative AFC-1D, AFC-1G, and AFC-1H End of FY-07 Irradiation Report

    Debra J Utterbeck; Gray S Chang; Misit A Lillo

    2007-09-01

    The purpose of the U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), is to develop and demonstrate the technologies needed to transmute the long-lived transuranic isotopes contained in spent nuclear fuel into shorter-lived fission products. Success in this undertaking could potentially dramatically decrease the volume of material requiring disposal with attendant reductions in long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is investigation of irradiation/transmutation effects on actinide-bearing metallic fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Goals of this initiative include addressing the limited irradiation performance data available on metallic fuels with high concentrations of Pu, Np and Am, as are envisioned for use as actinide transmutation fuels. The AFC-1 irradiation experiments of transmutation fuels are expected to provide irradiation performance data on non-fertile and low-fertile fuel forms specifically, irradiation growth and swelling, helium production, fission gas release, fission product and fuel constituent migration, fuel phase equilibria, and fuel-cladding chemical interaction. Contained in this report are the to-date physics evaluations performed on three of the AFC-1 experiments; AFC-1D, AFC-1G and AFC-1H. The AFC-1D irradiation experiment consists of metallic non-fertile fuel compositions with minor actinides for potential use in accelerator driven systems and AFC-1G and AFC-1H irradiation experiments are part of the fast neutron reactor fuel development effort. The metallic fuel experiments and nitride experiment are high burnup analogs to previously irradiated experiments and are to be irradiated to = 40 at.% burnup.

  1. Effects of actinide burning on waste disposal at Yucca Mountain

    Partitioning the actinides in spent fuel and transmuting them in actinide-burning liquid-metal reactors (ALMRs) is a potential method of reducing public risks from the geologic disposal of nuclear waste. In this paper, the authors present a comparison of radionuclide releases from burial at Yucca Mountain of spent fuel and of ALMR wastes. Two waste disposal schemes are considered. In each, the heat generation of the wastes at emplacement is 9.88 x 107 W, the maximum for the repository. In the first scheme, the repository contains 86,700 tonnes of initial heavy metal (IHM) of light water reactor (LWR) spent fuel. In the second scheme, all current LWRs operate for a 40-yr lifetime, producing a total of 84,000 tonnes IHM of spent fuel. This spent fuel is treated using a pyrochemical process in which 98.4% of the uranium and 99.8% of the neptunium, plutonium, americium, and curium are extracted and fabricated into ALMR fuel, with the reprocessing wastes destined for the repository. The ALMR requires this fuel for its startup and first two reloads; thereafter, it is self-sufficient. Spent ALMR fuel is also pyrochemically reprocessed: 99.9% of the transuranics is recovered and recycled into ALMR fuel, and the wastes are placed in the repository. Thus, in the second scheme, the repository contains the wastes from reprocessing all of the LWR spent fuel plus the maximum amount of ALMR reprocessing wastes allowed in the repository based on its heat generation limit

  2. Actinides recycling assessment in a thermal reactor

    Highlights: • Actinides recycling is assessed using BWR fuel assemblies. • Four fuel rods are substituted by minor actinides rods in a UO2 and in a MOX fuel assembly. • Performance of standard fuel assemblies and the ones with the substitution is compared. • Reduction of actinides is measured for the fuel assemblies containing minor actinides rods. • Thermal reactors can be used for actinides recycling. - Abstract: Actinides recycling have the potential to reduce the geological repository burden of the high-level radioactive waste that is produced in a nuclear power reactor. The core of a standard light water reactor is composed only by fuel assemblies and there are no specific positions to allocate any actinides blanket, in this assessment it is proposed to replace several fuel rods by actinides blankets inside some of the reactor core fuel assemblies. In the first part of this study, a single uranium standard fuel assembly is modeled and the amount of actinides generated during irradiation is quantified for use it as reference. Later, in the same fuel assembly four rods containing 6 w/o of minor actinides and using depleted uranium as matrix were replaced and depletion was simulated to obtain the net reduction of minor actinides. Other calculations were performed using MOX fuel lattices instead of uranium standard fuel to find out how much reduction is possible to obtain. Results show that a reduction of minor actinides is possible using thermal reactors and a higher reduction is obtained when the minor actinides are embedded in uranium fuel assemblies instead of MOX fuel assemblies

  3. Recycling and transmutation of nuclear waste. ECN-Petten and Belgonucleaire contributions in the framework of 'Partitioning and transmutation studies of the 4th CEC programme on rad waste management and disposal'

    A 'Strategy study on nuclear waste transmutation' by Netherlands Energy Research Foundation (ECN) and Belgonucleaire (BN) in the frame of the EU R and D Programme 1990/1994 on management and storage of radioactive waste has been executed in collaboration with AEA Technology, CEA and Siemens. First of all the motivation for transmuting long-lived radioactive products has been formulated, next transmutation of Tc-99 and I-129 in fission reactors has been studied for the PWR, HFR, Superphenix, and the CANDU reactor. Cross section libraries have been improved for ORIGEN-S on the basis of JEF2.2 and EAF3. This study has been amended by a graphical representation of important reactions for activation of cladding and inert matrix materials. By means of the derived new data libraries, some sample calculations on transmutation of americium in thermal reactors have been performed. Implications of recycling plutonium and americium in the form of MOX fuel in light water reactors have been investigated. It became clear from the present study that trasmutation of the existing plutonium has the highest priority and that reduction of minor-actinides is next on the priority list. Thirdly, the (difficult) large-scale transmutation of Tc-99 and of I-129 could reduce the leakage dose risks. It also seems most worthwhile to be careful with naturally occurring U-234 in the waste, as this will in the long run lead to a substantial increase of the 'natural' radon dose in the neighbourhood of the storage facility. (orig.)

  4. Recycling and transmutation of nuclear waste. ECN-Petten and Belgonucleaire contributions in the framework of `Partitioning and transmutation studies of the 4th CEC programme on rad waste management and disposal`

    Abrahams, K. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Brusselaers, P. [Belgonucleaire S.A., Brussels (Belgium); Evrard, G. [Belgonucleaire S.A., Brussels (Belgium); La Fuente, A. [Belgonucleaire S.A., Brussels (Belgium); Maldague, T. [Belgonucleaire S.A., Brussels (Belgium); Pilate, S. [Belgonucleaire S.A., Brussels (Belgium); Renard, A. [Belgonucleaire S.A., Brussels (Belgium)

    1995-12-01

    A `Strategy study on nuclear waste transmutation` by Netherlands Energy Research Foundation (ECN) and Belgonucleaire (BN) in the frame of the EU R and D Programme 1990/1994 on management and storage of radioactive waste has been executed in collaboration with AEA Technology, CEA and Siemens. First of all the motivation for transmuting long-lived radioactive products has been formulated, next transmutation of Tc-99 and I-129 in fission reactors has been studied for the PWR, HFR, Superphenix, and the CANDU reactor. Cross section libraries have been improved for ORIGEN-S on the basis of JEF2.2 and EAF3. This study has been amended by a graphical representation of important reactions for activation of cladding and inert matrix materials. By means of the derived new data libraries, some sample calculations on transmutation of americium in thermal reactors have been performed. Implications of recycling plutonium and americium in the form of MOX fuel in light water reactors have been investigated. It became clear from the present study that trasmutation of the existing plutonium has the highest priority and that reduction of minor-actinides is next on the priority list. Thirdly, the (difficult) large-scale transmutation of Tc-99 and of I-129 could reduce the leakage dose risks. It also seems most worthwhile to be careful with naturally occurring U-234 in the waste, as this will in the long run lead to a substantial increase of the `natural` radon dose in the neighbourhood of the storage facility. (orig.).

  5. Layer thickness evaluation for transuranic transmutation in a fusion–fission system

    Highlights: • Layer thickness for transmutation in a fusion–fission system was evaluated. • The calculations were performed using MONTEBURNS code. • The results indicate the best thickness and volume ratio to induce transmutation. - Abstract: Layer thickness for transuranic transmutation in a fusion–fission system was evaluated using two different ways. In the first one, transmutation layer thicknesses were designed maintaining the fuel rod radius constant; in the second part, while the transmutation layer thickness increases, the fuel rod radius decreases maintaining ks (source-multiplication factor) ≈0.95. Spent fuel reprocessed by UREX+ method and then spiked with thorium and uranium composes the transmutation layer. The calculations were performed using MONTEBURNS code (MCNP5 and ORIGEN 2.1). The results indicate the best thickness and the volume ratio between the coolant and the fuel composition to induce transmutation

  6. Layer thickness evaluation for transuranic transmutation in a fusion–fission system

    Velasquez, Carlos E., E-mail: carlosvelcab@eng-nucl.mest.ufmg.br [Departamento de Engenharia Nuclear—Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627 Campus UFMG, 31.270-90, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq, Rio de Janeiro, RJ (Brazil); Rede Nacional de Fusão (FINEP/CNPq), Rio de Janeiro, RJ (Brazil); Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Departamento de Engenharia Nuclear—Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627 Campus UFMG, 31.270-90, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq, Rio de Janeiro, RJ (Brazil); Rede Nacional de Fusão (FINEP/CNPq), Rio de Janeiro, RJ (Brazil); Veloso, Maria Auxiliadora F., E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear—Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627 Campus UFMG, 31.270-90, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq, Rio de Janeiro, RJ (Brazil); Rede Nacional de Fusão (FINEP/CNPq), Rio de Janeiro, RJ (Brazil); Costa, Antonella L., E-mail: antonella@nuclear.ufmg.br [Departamento de Engenharia Nuclear—Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627 Campus UFMG, 31.270-90, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq, Rio de Janeiro, RJ (Brazil); Rede Nacional de Fusão (FINEP/CNPq), Rio de Janeiro, RJ (Brazil)

    2015-05-15

    Highlights: • Layer thickness for transmutation in a fusion–fission system was evaluated. • The calculations were performed using MONTEBURNS code. • The results indicate the best thickness and volume ratio to induce transmutation. - Abstract: Layer thickness for transuranic transmutation in a fusion–fission system was evaluated using two different ways. In the first one, transmutation layer thicknesses were designed maintaining the fuel rod radius constant; in the second part, while the transmutation layer thickness increases, the fuel rod radius decreases maintaining k{sub s} (source-multiplication factor) ≈0.95. Spent fuel reprocessed by UREX+ method and then spiked with thorium and uranium composes the transmutation layer. The calculations were performed using MONTEBURNS code (MCNP5 and ORIGEN 2.1). The results indicate the best thickness and the volume ratio between the coolant and the fuel composition to induce transmutation.

  7. Modifying SCALE-4.1 for transmutation calculations

    A special sequence of selected modules and data libraries of the SCALE-41. computer code package was developed for the study of the feasibility of transmuting the transuranium isotopes and long-lived fission products accumulated, in spent fuel from LWRs. The specific reactor type considered for the transmutation is a molten salt reactor that is fueled by, and only by, the transuranium isotopes and long-lived fission products from the light water reactors spent fuel. (authors). 2 refs., 1 fig

  8. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  9. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale

  10. Property database of TRU nitride fuel

    西 剛史; 荒井 康夫; 高野 公秀; 倉田 正輝

    2014-01-01

    The purpose of this study is to prepare a property database of nitride fuel needed for the fuel design of accelerator-driven system (ADS) for transmutation of minor actinide (MA). Nitride fuel of ADS is characterized by high content of Pu and MA as principal components, and addition of a diluent material such as ZrN. Experimental data or evaluated values from the raw data on properties Pu and MA nitrides, and nitride solid solutions containing ZrN are collected and summarized, which cover the...

  11. Emerging nuclear energy and transmutation systems: Core physics and engineering aspects

    The Technical Committee Meeting (TCM) on Core Physics and Engineering Aspects of Emerging Nuclear Energy Systems for Energy Generation and Transmutation held in December 2000, was convened by the IAEA on the recommendation of its Technical Working Group on Fast Reactors (TWG-FR). The objectives of this TCM were threefold: to review the status of Research and Development activities in the area of hybrid systems for energy generation and transmutation, to discuss specific scientific and technical issues covering the different R and D topics of these systems; and to recommend to the IAEA activities that would be specifically targeted to the needs of the Member States performing R and D in this field. The TCM had not called for broad overview papers of the various R and D fields. Apart from a rather brief presentation by each delegation of the general issues and the status of the R and D in the respective country, the IAEA had called for in-depth technical papers addressing one or more of the following topics: accelerator driven systems (ADS) concepts, requirements and features of ADS accelerators, target development, experiments and validation, sub-critical core studies, technology of heavy liquid metals, fuel and fuel processes development, and fuel cycle studies. Forty-five participants from eleven countries and one international organization attended the TCM, and thirty papers were presented. The status information presented in the delegates' general statements and in some of the papers is as of the time of the TCM. Thus, other later material should also be referenced for more current information. One such source of information is the Web Site of IAEA's project on Technology Advances in Fast Reactors and Accelerator Driven Systems for Actinide and Long lived Fission Product Transmutation (http://www.iaea.org/inis/aws/fnss/). However, the technical information provided in the papers, representing the bulk of the information presented, remains valid

  12. Long-Lived Fission Product Transmutation Studies

    A systematic study on long-lived fission products (LLFPs) transmutation has been performed with the aim of devising an optimal strategy for their transmutation in critical or subcritical reactor systems and evaluating impacts on the geologic repository. First, 99Tc and 129I were confirmed to have highest transmutation priorities in terms of transmutability and long-term radiological risk reduction. Then, the transmutation potentials of thermal and fast systems for 99Tc and 129I were evaluated by considering a typical pressurized water reactor (PWR) core and a sodium-cooled accelerator transmutation of waste system. To determine the best transmutation capabilities, various target design and loading optimization studies were performed. It was found that both 99Tc and 129I can be stabilized (i.e., zero net production) in the same PWR core under current design constraints by mixing 99Tc with fuel and by loading CaI2 target pins mixed with ZrH2 in guide tubes, but the PWR option appears to have a limited applicability as a burner of legacy LLFP. In fast systems, loading of moderated LLFP target assemblies in the core periphery (reflector region) was found to be preferable from the viewpoint of neutron economy and safety. By a simultaneous loading of 99Tc and 129I target assemblies in the reflector region, the self-generated 99Tc and 129I as well as the amount produced by several PWR cores could be consumed at a cost of ∼10% increased fuel inventory. Discharge burnups of ∼29 and ∼37% are achieved for 99Tc and 129I target assemblies with an ∼5-yr irradiation period.Based on these results, the impacts of 99Tc and 129I transmutation on the Yucca mountain repository were assessed in terms of the dose rate. The current Yucca Mountain release evaluations do not indicate a compelling need to transmute 99Tc and 129I because the resulting dose rates fall well below current regulatory limits. However, elimination of the LLFP inventory could allow significant relaxation of

  13. Transmutation of Americium in Fast Neutron Facilities

    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...

  14. Advanced fuels for fast reactors

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  15. Accelerator-driven transmutation technologies

    The basic principles of accelerator-driven transmutation technologies (ADTT) are outlined and their assets highlighted. Current designs of ADTT facilities pursue 3 basic objectives: (i) Systems designed to generate power and convert nuclear wastes produced by conventional nuclear reactors into long-lived radioisotopes by transmutation. Such isotopes will be separated from molten salts by centrifugal separation. A single subcritical assembly will 'burn' wastes produced by several conventional NPPs. (ii) Systems for power generation using thorium fuel. Such systems are not designed for transmutation of nuclear wastes. The amount of transuranium elements produced by the thorium cycle is minimal, whereby the problem of storage of very long lived isotopes is virtually eliminated. (iii) Systems for transmutation of plutonium reclaimed from nuclear weapons. As to the future of ADTT in comparison with nuclear fusion, an asset of the former is that there remain no unsolved principal physical problems that would preclude its implementation. What has to be solved is materials and technological problems and, in particular, the financial problem. Implementation of ADTT is impossible in any way other than on the basis of a wide international cooperation. There exists a group of people dealing with ADTT in the Czech Republic, joining academic and industrial experts; this group is fostering contacts with the Los Alamos National Laboratory, U.S.A. The Institute of Nuclear Physics, Academy of Sciences of the Czech Republic, has set up an ADTT Documentation Center, which is accessible to any person interested in this promising field of science and technology. (P.A.). 3 figs

  16. Actinides draw down process for pyrochemical reprocessing of spent metal fuel

    Perumal, Suyambu Vannia; Reddy, Bandi Prabhakara; Ravisankar, Guruswamy; Nagarajan, Krishnamurthy [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Chemistry Group

    2015-06-01

    An experimental system has been setup inside an argon atmosphere glove box to carry out studies on the actinide draw down process, an important step of the pyroprocess flow sheet. Studies have been carried out at 723 K on the extraction of uranium from LiCl-KCl eutectic salt containing 0.5% UCl{sub 3} using lithium-0.03% cadmium alloy. In these experiments, 5 L of the molten salt and 5 L of the molten alloy are fed to a single stage extractor where 250 mL of the salt and 250 mL of the alloy come into contact. A three stage stirrer having straight as well as pitched blade turbine impellers was used to mix the molten alloy and salt phases having high interfacial tension of 450 dyne/cm. The results of the studies indicate that a U extraction efficiency of 99% could be achieved.

  17. A concept of self-completed fuel cycle based on lead-cooled nitride-fuel fast reactors

    A concept of nuclear energy total system was studied based on the nitride fuel cycle and inherent safety lead-cooled fast reactors. In the nitride fuel reprocessing, a new concept for pyrochemical method was proposed due to reducing fuel cycle cost. The present designed lead-cooled fast reactors have higher safety, economics and minor actinide transmutation efficiency than those of MOX-fuel fast reactors. The construction of 1500 MWt plant is feasible as a result for technology studies for aseismic, steam-generator and reactor configuration systems. (author)

  18. Current status and future plan of research and development on partitioning and transmutation technology in Japan

    After a first check and review on partitioning and transmutation (P and T) technology by the Japan Atomic Energy Commission (JAEC) in 2000, significant progress was made in respective research areas of partitioning, fuel fabrication, transmutation and fuel recycling in Japan. The second check and review on P and T technology was made by the JAEC in 2008-2009 to illustrate the benefit and significance of P and T, to review the current state of P and T technology in and outside Japan, and to discuss how to conduct future research and development. The final report, issued in April 2009, mentions that the significance of P and T technology can be reduced to three points: reduction of the potential hazard, mitigation of the requirement for geological repository site, and enhancement of the options in the design of the whole system of waste disposal. The current technology levels of P and T for both FBR and ADS were evaluated. Although the technology levels of some parts of the FBR cycle system are between basic research and engineering demonstration, P and T technology in general is still in the basic research stage because of the lack of experimental data on minor actinides (MA). It was, therefore, strongly recommended to accumulate experimental data for MA as a common basis for both FBR and ADS. (authors)

  19. Separation of actinides from irradiated An–Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl–KCl

    Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Murakami, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Claux, B.; Meier, R.; Malmbeck, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tsukada, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany)

    2015-04-15

    Highlights: • Electrorefining process in molten LiCl-KCl using solid Al electrodes was demonstrated. • High separation factors of actinides over lanthanides were achieved. • Efficient recovery of actinides from irradiated nuclear fuel was achieved. • Uniform, dense and well adhered deposits were obtained and characterised. • Kinetic parameters of actinide–aluminium alloy formation were evaluated. - Abstract: An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl–KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An–Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U{sub 67}–Pu{sub 19}–Zr{sub 10}–MA{sub 2}–RE{sub 2} (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide–aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes.

  20. Separation of actinides from irradiated An–Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl–KCl

    Highlights: • Electrorefining process in molten LiCl-KCl using solid Al electrodes was demonstrated. • High separation factors of actinides over lanthanides were achieved. • Efficient recovery of actinides from irradiated nuclear fuel was achieved. • Uniform, dense and well adhered deposits were obtained and characterised. • Kinetic parameters of actinide–aluminium alloy formation were evaluated. - Abstract: An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl–KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An–Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U67–Pu19–Zr10–MA2–RE2 (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide–aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes

  1. Fabrication of nuclear fuel by powder injection moulding: Study of the binders systems and the de-binding of feedstock containing actinide powder

    Powder Injection Moulding (PIM) is identified as an innovative process for the nuclear fuel fabrication. Technological breakthrough compared to the current process of powder metallurgy, the impact of actinide powder's specificities on the different steps of PIM is performed. Alumina powders simulating actinide powder have been implemented with a reference binders system. Thermal and rheological studies show the injectability and the de-binding of feedstocks with adequate solid loading (≥50 %vol), thanks to the de-agglomeration during the mixing step, which allow to obtain net shape fuel pellet. Specific surface area of powders, acting as a key role in behaviour's feedstocks, has been integrated in analysis models of viscosity prediction according to the shear rate. Also conducted studies on uranium oxide powder show that the selected binders systems, which have a compatible rheological behaviour with PIM process, impact the de-agglomeration of powder and final microstructure of the fuel pellet, consistent with the results obtained on alumina powders. Independent behaviour of binders and uranium oxide powder, showing no adverse chemical reaction against the PIM process, show a residual mass of carbon of about 150 ppm after sintering. Binders system using polystyrene, resistant to radiolysis phenomena and loadable more than 50 %(vol) of actinide powder, shows the promising potential of PIM process for the fuel fabrication. (author)

  2. Deep-Burn: making nuclear waste transmutation practical

    In the Deep-Burn concept, destruction of the transuranic component of light water reactor (LWR) waste is carried out in one burn-up cycle, accomplishing the virtually complete destruction of weapons-usable materials (Plutonium-239), and up to 90% of all transuranic waste, including the near totality of Neptunium-237 (the most mobile actinide in the repository environment) and its precursor, Americium-241. Waste destruction would be performed rapidly, without multiple reprocessing of plutonium, thus eliminating the risks of repeated handling of weapons-usable material and limiting the generation of secondary waste. There appears to be no incentive in continuing the destruction of waste beyond this level. An essential feature of the Deep-Burn Transmuter is the use of ceramic-coated fuel particles that provide very strong containment and are highly resistant to irradiation, thereby allowing very extensive destruction levels ('Deep Burn') in the one pass, using gas-cooled modular helium reactor (MHR) technology developed for high-efficiency energy production. The fixed moderator (graphite) and neutronically transparent coolant (helium) provide a unique neutron energy spectrum to cause Deep-Burn, and inherent safety features, specific to the destruction of nuclear waste, that are not found in any other design. Deep-Burn technology could be available for deployment in a relatively short time, thus contributing effectively to waste problem solutions. Extensive modeling effort has led to conceptual Deep-Burn designs that can achieve high waste destruction levels (70% in critical mode, 90% in with a supplemental subcritical step) within the operational envelope of commercial MHR operation, including long refueling intervals and the highly efficient production of energy (approximately 50%). To the plant operator, a Deep-Burn Transmuter will be identical to its commercial reactor counterpart

  3. Fuel depletion analyses for the HEU core of GHARR-1: Part I: Actinide inventory

    The Ghana Research Reactor-1 (GHARR-1), a 30kW, 90.2% HEU fueled (U-Al) MNSR type reactor went critical on December 17, 1994. Under operating conditions of 2.5 hours per day for five days a week at a peak thermal neutron flux of 1.0x1012n/cm2.s, the estimated core life is ten years. After the fuel is depleted, the entire spent fuel assembly will be replaced with a fresh LEU core in the spirit of the RERTR program and in accordance with current trend. This paper presents the results of multigroup fuel burnup and depletion analysis of the GHARR-1 fuel lattice using the WIMSD/4 transport lattice code. The results contained in this paper would be used as microscopic database in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the GHARR-1 core. (author)

  4. Recent Advances in Fuel for Fast Reactors: Synthesis, Properties, Safety Performance

    This presentation provides an overview of advances made in fast reactor fuel safety research in the last decades. MOX fuels were selected for many fast reactors. The need to minimize waste through partitioning and transmutation strategies has seen new fuel forms evolve, including fertile and non-fertile targets to host the minor actinides. Despite the proven safety performance of MOX fuels and the progress made in nitride and carbide driver fuels and also in minor actinide oxide fuel research programmes, improvements in knowledge and understanding of the safety performance of these fuels can be made. Above all, breakthroughs in simulation and modelling need to be harnessed for dedicated experiments, leading to even more reliable and robust engineering codes for the qualification of all fast reactor fuels. (author)

  5. Solvent extraction studies with substituted malonamides and oligopyridines. Influence of structure and chemical properties on the extraction ability of trivalent actinides and lanthanides

    Separation and transmutation of spent nuclear fuel has been considered as a complement to direct disposal in a deep geological repository. The time needed for the waste to decay to natural background levels can be drastically decreased if the long-lived radionuclides, mainly the actinides, in the spent nuclear fuel are separated and transmuted to short-lived or stable nuclides. Critical reactors or subcritical accelerator-driven systems have been considered for the transmutation. An efficient chemical separation of long-lived actinides from fission products is necessary to achieve an efficient transmutation process. Solvent extraction techniques have been suggested as the separation method used prior to each transmutation cycle. Malonamides have been suggested as co-extracting agents for trivalent actinides and lanthanides from high acidity in a first step in such a separation process and nitrogen-donor extractants, such as oligopyridines in synergy with carboxylic acids, have shown potential to be able to separate the trivalent actinides from the lanthanides from low acidity in a second step. The extractive behaviour and chemical properties of several substituted malonamides and oligopyridines have been studied in this work and have been related to the structure and the basicity of the ligands. It was found that the basicity of the malonamide was strongly related to its molecular structure, and a malonamide with lower basicity was shown to give a better metal extraction, owing to the less severe competition between protons and metal cations for the binding site in the malonamides. Malonamides with aromatic groups attached on the nitrogens were shown to have the lowest basicity of all studied ligands. The basicity of the oligopyridines was also shown to be dependent on the molecular structure and extractants with low basicity generally resulted in a higher metal extraction. A synergistic mixtures with a carboxylic acid and terpyridine showed good selectivity for

  6. Neutronics of LBE target-cooled ADS for MA transmutation: Japan

    Purpose and goal: JAEA's reference design of ADS is a tank type 800 MWth subcritical reactor to transmute about 250 kg of minor actinides annually. A lead-bismuth eutectic (LBE) is used as both the primary coolant and the spallation target. A superconducting linear accelerator (SC-LINAC), whose proton energy and maximum current are 1.5 GeV and 20 mA (30 MW), is connected to produce spallation neutrons. The (MA, Pu) N fuel diluted by ZrN is used in the subcritical core. Because the relatively high power peaking factor will be observed at the burnup stage of low HII value, where the influence of the spallation neutrons is strong, Pu is added at the beginning of the first burnup cycle to mitigate the rapid increase of the burnup reactivity

  7. Effectiveness of LLFP (long-lived fission product) transmutation and the simplified recycling method

    Takaki, Naoyuki [Tokyo Electric Power Co., Yokohama (Japan). Nuclear Power R and D Center; Fujita, Reiko; Takagi, Ryuzo; Sekimoto, Hiroshi

    1997-12-31

    Feasibility of long-lived FP (LLFP) transmutation adopting simplified isotope separation process has been discussed in terms of neutron economy of a fast reactor in the equilibrium cycle. For minimization of nuclear waste, recycling of selected LLFP in addition to all of actinides is required. The recycle process, however, must be technically complicated due to the necessity of isotope separation of the LLFP with high separation factor. The authors proposed a simplified separation process of LLFP by using laser isotope separation technology. The concept is based on rough separation only of easily recoverable isotopes. Reactor and fuel cycle coupled analysis showed that rough removing a few stable isotopes from LLFP elements to the waste stream effectively reduced the parasitic neutron absorption for keeping the reactor critical, and also could be beneficial to restrict the boundless accumulation of radioactive waste. (author)

  8. Effectiveness of LLFP (long-lived fission product) transmutation and the simplified recycling method

    Feasibility of long-lived FP (LLFP) transmutation adopting simplified isotope separation process has been discussed in terms of neutron economy of a fast reactor in the equilibrium cycle. For minimization of nuclear waste, recycling of selected LLFP in addition to all of actinides is required. The recycle process, however, must be technically complicated due to the necessity of isotope separation of the LLFP with high separation factor. The authors proposed a simplified separation process of LLFP by using laser isotope separation technology. The concept is based on rough separation only of easily recoverable isotopes. Reactor and fuel cycle coupled analysis showed that rough removing a few stable isotopes from LLFP elements to the waste stream effectively reduced the parasitic neutron absorption for keeping the reactor critical, and also could be beneficial to restrict the boundless accumulation of radioactive waste. (author)

  9. Investigations into the behavior of high-burnup spent fuel and its further uses. Initial literature research

    The core of the text consists of 2 chapters: Description of advanced fuel cycles aimed at a further use of spent nuclear fuel (principles of scenarios of advanced fuel cycles (AFC); Mass-balance schemes of AFC scenarios; Technical and economic characteristics of AFC projects; A brief overview of technological processes of spent nuclear fuel processing; Transmutation; New aspects of radioactive waste management; Overview of nuclear fuel types for the advanced fuel cycle), and Strategy of the approach to advanced fuel cycles in the US. The Annex is devoted to advanced separation processes for application to nuclear fuel reprocessing (PUREX process; Extraction of the sum of trivalent lanthanides and actinides - TRUEX, DIAMEX, TRPO, DIDPA; Separation of trivalent actinides from trivalent lanthanides - SANEX, TALSPEAK; Separation of Am(III) from Cm(III); Separation of Cs and Sr; Dicarbollides; Complete spent fuel reprocessing processes - UREX+, ARTIST). (P.A.)

  10. Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

    Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

    2009-05-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

  11. Transmutation efficiency in the prismatic deep burner HTR concept by a 3D Monte Carlo depletion analysis

    This paper summarizes studies performed on the Deep-Burner Modular Helium Reactor (DB-MHR) concept-design. Feasibility and sensitivity studies as well as fuel-cycle studies with probabilistic methodology are presented. Current investigations on design strategies in one and two pass scenarios, and the computational tools are also presented. Computations on the prismatic concept-design were performed on a full-core 3D model basis. The probabilistic MCNP-MONTEBURNS-ORIGEN chain, with either JEF2.2 or BVI libraries, was used. One or two independently depleting media per assembly were accounted. Due to the calculation time necessary to perform MCNP5 calculations with sufficient accuracy, the different parameters of the depletion calculations have to be optimized according to the desired accuracy of the results. Three strategies were compared: the two pass with driver and transmuter fuel loading in three rings, the one pass with driver fuel only in three rings geometry and finally the one pass in four rings. The 'two pass' scenario is the best deep burner with about 70% mass reduction of actinides for the PWR discharged fuel. However the small difference obtained for incineration (∼5%) raises the question of the interest of this scenario given the difficulty of the process for TF fuel. Finally the advantage of the 'two pass' scenario is mainly the reduction of actinide activity. (author)

  12. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Nuttin A.; Siem S.; Ivanov E.; Méplan O.; David S; Guillemin P.; Wilson J.N.; Capellan N.; Rose S.J.

    2012-01-01

    The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor ac...

  13. Research on the chemical speciation of actinides

    A demand for the safe and effective management of spent nuclear fuel and radioactive waste generated from nuclear power plant draws increasing attention with the growth of nuclear power industry. The objective of this project is to establish the basis of research on the actinide chemistry by using advanced laser-based highly sensitive spectroscopic systems. Researches on the chemical speciation of actinides are prerequisite for the development of technologies related to nuclear fuel cycles, especially, such as the safe management of high level radioactive wastes and the chemical examination of irradiated nuclear fuels. For supporting these technologies, laser-based spectroscopies have been performed for the chemical speciation of actinide in an aqueous solutions and the quantitative analysis of actinide isotopes in spent nuclear fuels. In this report, results on the following subjects have been summarized. (1) Development of TRLFS technology for chemical speciation of actinides, (2) Development of LIBD technology for measuring solubility of actinides, (3) Chemical speciation of plutonium complexes by using a LWCC system, (4) Development of LIBS technology for the quantitative analysis of actinides, (5) Development of technology for the chemical speciation of actinides by CE, (6) Evaluation on the chemical reactions between actinides and humic substances, (7) Chemical speciation of actinides adsorbed on metal oxides surfaces, (8) Determination of actinide source terms of spent nuclear fuel

  14. Analysis of irradiated Magnox fuel for fission product and actinide elements

    This report describes the methods used for, and the results obtained from, chemical and isotopic analysis of four Magnox fuel samples of known irradiation history. The analysis of this material was carried out in conjunction with CEGB, Berkeley Nuclear Laboratories and BNFL. The data obtained was used to evaluate the accuracy of fuel composition as predicted by computer codes such as RICE (Reactor Inventory Code). (author)

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

    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

  16. Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste

    A spallation neutron source was modeled using a high energy proton accelerator. The aim of the core design is to optimise the core parameters for maximizing the minor actinides and fission products transmutation rates, which is created from the operation of nuclear power reactors for the production of electricity, while maintaining the structural material damage and decay heat as low as possible. The transmutation system is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1 GeV proton beam incident on a natural lead cylindrical target, 20 cm radius, 70 cm height , and entering the target through a 5.3 cm radius hole. The protons were uniformly distributed across the beam of radius 2 cm. The core is cylindrical assembly, 2.3 m radius, 4.6 m high. The wall thickness of the main vessel is 2 cm. The main vessel surrounded by a reflector made of graphite, 40 cm thick. The axes of proton beam and target are concentric with the main vessel axis. The structural walls and beam window are made of the same material, stainless steel, HT9. All dimensions of systems are results of target and core optimization that keeps most of the spallation neutrons within the lead target and transmutes the largest fraction of the long-lived waste. We investigated the following neutronics parameters with presence and absence of fissile materials: o spallation neutron and other particles such as proton, pions and muons yields (per one incident proton) from the spallation target, - spatial and energy distribution of the spallation neutrons, and protons in target, - heat deposition distribution in the spallation target, - heat deposition in beam window, core, reflector and structural

  17. TRU transmutation using ThO{sub 2}-UO{sub 2} and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    Bae, Gonghoon; Hong, Sergi [Dep. of Nuclear Engineering, KyungHee Univ., Yongin (Korea, Republic of)

    2012-10-15

    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{sub 2}-UO{sub 2} fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO{sub 2} in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO{sub 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 {approx}40% and {approx}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.

  18. Subsurface Biogeochemistry of Actinides

    Kersting, Annie B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Univ. Relations and Science Education; Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Glenn T. Seaborg Inst.

    2016-06-29

    A major scientific challenge in environmental sciences is to identify the dominant processes controlling actinide transport in the environment. It is estimated that currently, over 2200 metric tons of plutonium (Pu) have been deposited in the subsurface worldwide, a number that increases yearly with additional spent nuclear fuel (Ewing et al., 2010). Plutonium has been shown to migrate on the scale of kilometers, giving way to a critical concern that the fundamental biogeochemical processes that control its behavior in the subsurface are not well understood (Kersting et al., 1999; Novikov et al., 2006; Santschi et al., 2002). Neptunium (Np) is less prevalent in the environment; however, it is predicted to be a significant long-term dose contributor in high-level nuclear waste. Our focus on Np chemistry in this Science Plan is intended to help formulate a better understanding of Pu redox transformations in the environment and clarify the differences between the two long-lived actinides. The research approach of our Science Plan combines (1) Fundamental Mechanistic Studies that identify and quantify biogeochemical processes that control actinide behavior in solution and on solids, (2) Field Integration Studies that investigate the transport characteristics of Pu and test our conceptual understanding of actinide transport, and (3) Actinide Research Capabilities that allow us to achieve the objectives of this Scientific Focus Area (SFA and provide new opportunities for advancing actinide environmental chemistry. These three Research Thrusts form the basis of our SFA Science Program (Figure 1).

  19. Actinide Sciences at ITN - Basic Studies in Chemistry with Potential Interest for Partitioning, Fuel Fabrication and More

    The current activities in the area of actinide chemistry at ITN, comprising basic research studies in inorganic and organometallic chemistry, catalysis, gas-phase ion chemistry, thermochemistry, and solid state chemistry, are briefly described. Actinide (and lanthanide) chemistry studies at ITN will be pursued connecting basic research with potential applications in nuclear and non-nuclear areas. (authors)

  20. Product Conversion: The Link between Separations and Fuel Fabrication

    Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D. [Oak Ridge National Laboratory: P.O. Box 2008, Oak Ridge, Tennessee 37831-6384 (United States)

    2008-07-01

    Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)