[en] In keeping with the 12-year history of this conference, GLOBAL 2007 focuses on future nuclear energy systems and fuel cycles. With the increasing public acceptance and political endorsement of nuclear energy, it is a pivotal time for nuclear energy research. Significant advances have been made in development of advanced nuclear fuels and materials, reactor designs, partitioning, transmutation and reprocessing technologies, and waste management strategies. In concert with the technological advances, it is more important than ever to develop sensible nuclear proliferation policies, to promote sustainability, and to continue to increase international collaboration. To further these aims, GLOBAL 2007 highlights recent developments in the following areas: advanced integrated fuel cycle concepts, spent nuclear fuel reprocessing, advanced reprocessing technology, advanced fuels and materials, advanced waste management technology, novel concepts for waste disposal and repository development, advanced reactors, partitioning and transmutation, developments in nuclear non-proliferation technology, policy, and implementation, sustainability and expanded global utilization of nuclear energy, and international collaboration on nuclear energy

[en] In 2005, experts from AREVA and RIAR performed a joint research work on the feasibility study of a plant reprocessing 1000 t/y of LWR spent nuclear fuel by the gas-fluoride and pyro-electrochemical techniques developed at RIAR. This work was based on the RIAR experience in development of pyrochemical processes and AREVA experience in designing UNF reprocessing plants. UNF reprocessing pyrochemical processes have been developed at RIAR at laboratory scale and technology for granulated MOX fuel fabrication and manufacturing of vibro-packed fuel rods is developed at pilot scale. The research work resulted in a preliminary feasibility assessment of the reprocessing plant according to the norms and standards applied in France. The study results interpretation must integrate the fact that the different technology steps are at very different stage of development. It appears clearly however that in its present state of development, pyro-electrochemical technology is not adapted to the treatment of an important material flow issuing from thermal reactors. There is probably an economic optimum to be studied for the choice of hydrometallurgical or pyro-electrochemical technology, depending on the area of application. This work is an example of successful and fruitful collaboration between French and Russian specialists. (authors)

[en] We need more oil energy to take out oil under the ground. Limit resources make us consider other candidates of energy source instead of oil. Electricity shall be the main role more and more like electric vehicles and air conditioners so we should consider electricity generation ways. When we consider what kind of electric power generation is the best or suitable, we should not only power generation plant but whole process from mining to power generation. It is good way to use EPR, Energy Profit Ratio, to analysis which type is more efficient and which part is to do research and development when you see the input breakdown analysis. Electricity by the light water nuclear power plant, the hydrogen power plant and the geothermal power plant are better candidates from EPR analysis. Forecasting the world primly energy supply in 2050, it is said that the demand will be double of the demand in 2000 and the supply will not be able to satisfy the demand in 2050. We should save 30% of the demand and increase nuclear power plants 3.5 times more and recyclable energy like hydropower plants 3 times more. When the nuclear power plants are 3.5 times more then uranium peak will come and we will need breed uranium. I will analysis the EPR of FBR. Conclusion: A) the EPR of NPS in Japan is 17.4 and it is the best of all. B) Many countries will introduce new nuclear power plants rapidly may be 3.5 times in 2050. C) Uranium peak will happen around 2050. (author)

[en] KAERI is developing the pyro-process technology to minimize the burden on permanent disposal of spent nuclear fuel. In addition, KAERI has developed the Korean Reference System for potential spent nuclear fuel disposal since 1997. The deep geologic disposal system is composed of a multi-barrier system in a crystalline rock to dispose of 36,000 MT of spent nuclear fuel (SNF) from a CANDU and a PWR. Quite recently, introduction of advanced nuclear fuel cycles such as pyro-processing is a big issue to solve the everlasting disposal problem and to assure the sustainable supply of fuel for reactors. To compare the effect of direct disposal of SNF with that of the high level waste disposal for waste generated from the advanced nuclear fuel cycles, the total system performance assessment for two different schemes is developed; one for direct disposal of SNF and the other for the introduction of the pyro-processing and direct disposal CANDU spent nuclear fuel. The safety indicators to assess the environmental friendliness of the disposal option are annual individual doses, toxicities and risks. Even though many scientists use the toxicity to understand the environmental friendliness of the disposal, scientifically the annual individual doses or risks are meaningful indicators for it. The major mechanisms to determine the doses and risks for direct disposal are as follows: (1) Dissolution mechanisms of uranium dioxides which control the dissolution of most nuclides such as TRU's and most parts of fission products. (2) Instant release fraction of highly soluble nuclides such as I-129, C-135, Tc-99, and others. (3) Retardation and dilution effect of natural and engineered barriers. (4) Dilution effect in the biosphere. The dominant nuclide is I-129 which follows both congruent and instantaneous release modes. Since its long half life associated with the instantaneous release I-129 is dominant well beyond one million. The impact of the TRU's is negligible until the significant decay of I-129. Also, Sr-90 and Cs-137 shares 89 % of the total decay heat. In this sense, the pyro-processing contributes significantly to reduce the risks and doses of disposal. Also, the separation of decay heat generating nuclides significantly reduces the disposal areas. Results indicate the followings: (1) To minimize the annual dose, it is important to properly manage the release mechanism of I-129 by pyro-processing. (2) To minimize the underground repository area, it is important to remove the decay heat sources, Cs-137 and Sr-90. (3) To minimize the burden to monitor a repository, it i s important to control long lived radionuclides such as I-129 and TRU's. If instantaneous release is gone, the source term controlled by congruent release will diminish quite significantly. Also, if heat sources are safety stored after removal above the ground for a certain period of time, then the underground repository area will be shrunken quite significantly also. These two factors are clear advantages of the pyro-processing to preserve environment and a future generation without significant worry over nuclear proliferation. The electro refining processes in combination with winning technologies will give additional benefits to the environment by removing and recycling long lived TRU's. To fully analyze the benefits of these two processes for environmental protection more detailed researches are needed: (1) To identify the accurate inventories from these processes. (2) To understand the dissolution mechanism of solidified wastes. Also, the new management concept development is recommended to effectively dispose of solidified wastes and possibly metal ingots and store heat generating waste above the ground. (authors)

[en] In the frame of the French radioactive waste management acts of December 1991 and June 2006, separation processes for minor actinide selective recovery have been developed to significantly decrease the radiotoxicity of the ultimate waste produced by the nuclear industry. Several routes are envisaged for actinide/lanthanide separation, either in two cycles, using two different solvents (generally a DIAMEX followed by a SANEX process), or in one single cycle involving the same solvent during the whole process. The DIAMEX-SANEX concept described in this paper is new concept: the organic phase, which consists of a cationic exchanger (phosphorus acid) and the malonamide developed for the DIAMEX process (N,N'-dimethyl-N,N'- di-octyl-hexyl-ethoxy-malonamide named DMDOHEMA), is split at a given stage of the process to avoid interactions between these two extractants during the co-extraction step of actinides and lanthanides. This paper describes some results obtained with di-n-hexyl phosphoric acid (HDHP), which fulfills the required criteria for process development. For instance, this reagent can easily extract lanthanides from a weak acidic aqueous solution, and it can be selectively separated from DMDOHEMA, thanks to a basic solution. (authors)

[en] In March 2007, the first reprocessing of the 'Type B' MOX spent fuels of the Prototype Advanced Thermal Reactor FUGEN was initiated at Tokai Reprocessing Plant as a plant-scale demonstration of MOX fuel reprocessing. The operation was advanced satisfactorily and it has been confirmed that the MOX fuels as well as UO₂ fuels can be reprocessed safely. Some characteristics of MOX fuels on reprocessing, such as properties of undissolved residue affecting the clarification process, are becoming visible. Reprocessing of the 'Type B' MOX fuels will be continued for several more years from now on, further investigations on solubility of fuels, characteristics of undissolved residues, progress of solvent degradation and so on will be continued. (authors)

[en] According to a planning building of NPP in Russia the accumulating of wastes will increase, so the foundation of reprocessing plant of new generation becomes required. The most prepared place for the new plant is the area of MCC (Krasnoyarskij region) that lies in the heart of Russian territory and far from oceans. Because of this the demands to the volume of liquid LLW are very stringent. On the base of acquired Russian and international experience the development of the technology for the plant of the next (third) generation conditionally named Simplified PUREX process has begun. This technology must provide the reduction of liquid LLW formation and reduce the total value of reprocessing in comparison with the project of Russian plant RT-2. The building of Experimental-demonstration center (EDC) on MCC is proposed for testing and working through the main aims of new technology (both Simplified Purex and some other technologies).The Simplified Purex technology in comparison with classical liquid-extraction technology contains the substantial amount of thermo-chemical (dry) operations in the head of the process and it allows simplifying the next hydrometallurgical operations to reduce the SNF reprocessing value in general. (authors)

[en] Dibutyl phosphoric acid (HDBP) formed by tributyl phosphate (TBP) destruction is soluble both in aqueous alkaline solutions and in organic solvents in acidic media. So, the solvent based on HDBP and its zirconium salt (ZS-HDBP) dissolved in polar diluent, e.g. in diluted TBP, is interesting for radwaste treatment, minimizing the amount of secondary organic wastes. Addition of Zr to 0.2 mol/L HDBP dissolved in 30% TBP results in successful extraction of lanthanides and actinides at the optimum ratio Zr:HDBP=1:(8-9) from 1.5 mol/L HNO₃, followed by their back-washing with 5 mol/L HNO₃. Partitioning of yttrium and cerium RE subgroups (the latter together with TPE) with the separation factor from 5 to 50 is also possible with purification from molybdenum. Strontium is extracted by 0.4 mol/L ZS HDBP from 0.3 mol/L HNO₃ and stripped with 1.5 mol/L HNO₃. ZS-HDBP solution in 30% TBP is also capable of extraction of residual Np, Pu and corrosion-born iron. Stripping of these elements requires salt-free complexants. Solvents containing ZS-HDBP have high capacity, while TBP presence doubles it because of synergetic effect. The maximum solvent loading of 0.2 mol/L ZSHDBP in xylene was found as 8.0 g/L Eu and ∼6 g/L Mo. The mixture of DTPA and formic acid is suitable for TPE/RE partitioning using ZS-HDBP as a solvent with separation factors for Ce/Am and Eu/Am of 67 and 9, respectively. Two variants of two-cycle flowsheet for TPE and RE partitioning after their joint recovery have been proposed, which differs by order of the cycles with acidic and buffer media at the partitioning. Both variants were successfully tested using simulate solutions on the centrifugal contactor rig with TPE/RE separation factor being ∼60, the major impurity being Nd. Correction of the solvent composition because of HDBP loss due to its solubility in aqueous phase, especially at acidity less than 0.2 mol/L HNO₃, was also taken into consideration. Further investigations on HLW partitioning were aimed at Cs, Sr, TPE and RE joint extraction with the mixture of chlorinated cobalt dicarbollide (CCD) and ZSHDBP (dissolved in fluorinated diluent F-3). Joint recovery of Cs, Sr, TPE with RE, Mo and Fe from 0.5-3.0 mol/L HNO₃ was demonstrated in the presence of polyethylene glycol (PEG). Extraction system is stable and has good hydrodynamic properties. A synergetic effect was observed in the system especially at the acidity 0.5-1 mol/L.TPE and RE back-washing with 5-8 mol/L HNO₃ is possible, while Cs and Sr back-washing needs cation replacement. Such a result can achieved also without Zr, but with 0.5-1.5 mol/L HDBP in F-3. (authors)

[en] At Rokkasho Reprocessing Plant (RRP), the first commercial reprocessing plant in Japan, the test operation has been carried out step by step with 'water and steam', 'chemical products', 'depleted uranium' and 'spent fuels' toward the planned start of the commercial operation. Water Test was performed as the final stage of plant construction work and functioning of each equipment was tested with water and steam. In Chemical Test the performance of each equipment and unit was verified with chemical products such as nitric acid. In Uranium Test with depleted uranium, function and performance of equipment such as the sharing machine and the dissolver was verified. All its tests were completed by 22 January 2006. Active Test has been performed with spent fuels for the verification of safety functions and performances of equipment and facilities related to the processing of fission products and of plutonium, which had not been tested previously. Active Test which has been in progress since 31 March 2006 is divided into 5 steps, and Step 1, Step 2 and Step 3 are already completed. (authors)

[en] This paper outlines the operations involved in cleaning the control rod for the complementary shutdown system in the Phenix Power Plant, the French sodium-cooled fast reactor. The Phenix reactor is controlled by six control rods and a complementary shutdown system. The latter comprises a control rod and a mechanism maintaining the rod in position by means of an electromagnet. The electromagnet is continuously supplied with power and holds the rod control assembly in position by magnetisation on a plane circular surface made from pure iron. The bearing capacity of the mechanism on the rod was initially 80 daN with a rod weight of 26.3 daN. This deteriorated progressively over time. The bearing surface of the rod and the electromagnet became contaminated with a deposit of sodium oxides and metallic particles, thus creating an air gap. This reached a figure of 36 daN in 2005 and was deemed not to be sufficient to prevent the rod from dropping at the wrong time during reactor operation. The Power Plant thus decided to replace the rod mechanism in the reactor in an initial phase, followed by the control rod itself. As the Phenix Power Plant had no spare control rods left, they initiated a 'salvage' plan, over two stages, for the rod removed from the reactor and placed in the fuel storage drum: - Inspection of the bearing surface of the rod by means of a borescope to check whether the rod could be salvaged, - A cleaning operation on the bearing face and checks on the bearing capacity of the rod. The operation is subject to very stringent requirements: the rod must not be taken out of the sodium to ensure that it can be reused in the reactor. The operation must thus take place in the fuel storage drum where there are no facilities for such an operation and where operating conditions are very hostile: high temperatures (the sodium in the fuel storage drum is at a temperature of 150 deg. C, high dose rate (3 mGy/h on the bearing surface) and the bearing surface is submerged under a metre of sodium. 4 poles will have been required for this operation: - a pole for draining the rod and uncovering the bearing surface, - a borescope pole, - a pole for cleaning the bearing surface, - a pole for inspecting the bearing surface. The operation resulted in a very satisfactory bearing capacity of approx. 80 daN. This paper also examines the history of contamination of the control rod and the rod mechanism. It outlines the hypotheses for the source of this contamination and the measures taken to counteract this problem. (authors)