Periodic Reporting for period 2 - MIMOSA (MultI-recycling strategies of LWR SNF focusing on MOlten SAlt technology)
Período documentado: 2023-12-01 hasta 2025-05-31
While the EU has the highest share of spent nuclear fuel (SNF) reprocessing worldwide, most spent nuclear fuel is still declared waste. MIMOSA develops an accessible, cost/risk optimized and sustainable multi-recycling strategy of light-water reactor (LWR) spent nuclear fuel, based primarily on multi-recycling of plutonium and reprocessed uranium in light water reactors, combined with one of the most promising advanced nuclear energy systems, the fast-spectrum chloride molten-salt reactors (Cl MSRs). The ambition is to provide a real option towards a closed nuclear fuel cycle in the EU while securing energy supply and using already available infrastructure in the EU such as the Orano reprocessing plant in La Hague (France). MIMOSA has identified six general scientific and technological objectives:
1. Perform a global analysis of tangible strategies for multi-recycling of spent nuclear fuel from LWR in the EU, based on multi-recycling in LWRs and Cl MSRs,
2. Assess chloride molten salt properties to support studies of reactor process,
3. Conduct simulations of chloride molten salt composition evolution in the reactor to support fuel cycle calculations, safety evaluation and the analysis of valuable isotopes,
4. Demonstrate several key aspects of technical feasibility and performance of Cl MSRs,
5. Study chloride molten salt recycling and back-end options in synergy with La Hague plant,
6. Evaluate the capacity of MSRs to produce valuable isotopes for other applications, and the way to extract, separate and purify them.
1. Perform a global analysis of tangible strategies for multi-recycling of spent nuclear fuel from LWR in the EU, based on multi-recycling in LWRs and Cl MSRs,
2. Assess chloride molten salt properties to support studies of reactor process,
3. Conduct simulations of chloride molten salt composition evolution in the reactor to support fuel cycle calculations, safety evaluation and the analysis of valuable isotopes,
4. Demonstrate several key aspects of technical feasibility and performance of Cl MSRs,
5. Study chloride molten salt recycling and back-end options in synergy with La Hague plant,
6. Evaluate the capacity of MSRs to produce valuable isotopes for other applications, and the way to extract, separate and purify them.
Spent nuclear fuel inventories, energy needs projections, nuclear fleet assumptions and potential evolutions have been assessed for all EU countries. In parallel, a Reduced-Order-Model of MSR has been developed and integrated in the scenario analysis tool. Several simulation rounds of scenarios have been performed, addressing the role of MSRs within EU27 for LWR spent fuel multi-recycling purpose. The (still evolving) results of the simulations are now available for further analysis and parametric studies.
The salt mixtures in NaCl-MgCl2-PuCl₃ and NaCl-ThCl₄-PuCl₃ that had been selected previously have been synthesized in order to perform characterization and further experimental work. Most of the salt properties measurement equipment has been set up and tested, and most measurements have been performed with the NaCl-MgCl2-PuCl₃ salt. They are still on-going for the NaCl-ThCl₄-PuCl₃ salt.
Three MSR configurations have been selected. Neutronic depletion simulations in these reactor configurations have been performed, to assess the evolution of salt mixtures composition, in particular fission products. The results have been used as inputs for the assessment of used salt treatment processes and valuable isotopes management processes. In addition, decay heat and doses calculations have also been performed for the salt mixtures.
A chloride salt irradiation experiment has been performed in the LR-0 reactor in the Czech Republic, during which the neutronic properties of a selection of chloride salts have been measured. Comparison with nuclear data libraries revealed differences in critical parameters, especially in systems with high chlorine concentrations, emphasizing the need for experimental validation of nuclear data.
In order to undergo the test program for structural materials corrosion tests that had been set up at the beginning of the project, the testing methods, the corrosion products measurement methods and the corrosion test set ups and loops have been developed. Selected material samples have been fabricated and tested for their resistance to corrosion when in contact with chloride salt, in variable configurations and exposure durations. Dynamic corrosion tests are now under preparation.
A corrosion / irradiation test bench has been constructed for combined phenomena investigation.
Eight scenarios have been defined, for which the used fuel salt compositions have being calculated. These compositions have been used as input parameters for the technical study of used fuel salt treatment in La Hague. No major obstacles have been identified for the treatment of salt from a demonstrator reactor (1 m3), whether or not chlorine enrichment is used. However, significant dilution is necessary to meet the reprocessing plant criteria, particularly to meet criticality safety limits and to avoid disruption of existing process streams or final waste conditioning.
Dissolution studies have started. The experimental setup design for dissolution is complete, and the work is advancing steadily toward meeting task objectives.
The pyrochemical salt treatment alternatives have also been studied. After defining an experimental plan, tests have been completed. Two extraction methods—reductive extraction on Al-Mg alloy and electrolytic extraction on Al/Cu alloy—achieved over 99% efficiency. Results show promise for separating lanthanides and possibly plutonium in MgCl2-based salts, though validation with Pu data is still needed.
MIMOSA is investigating vitrification of representative chloride salt compositions (NaCl-MgCl2-ThCl4-PuCl3) including key fission products into borosilicate and phosphate glass. High-quality vitrified samples have been obtained. Vitrification experiments with iron phosphate glass have begun.
Valuable isotopes produced in MSRs among other fuel salt mixtures isotopes have been assessed and quantified. The applications for these isotopes and the associated market have been described.
A thermodynamic database on the platinoids chloride phases has been developed, and some experiments have been performed to validate the metallic state of the platinoids and noble metals.
A device to test the ability and efficiency of gas bubbling to extract solid particles in molten salt has been constructed and tests have started.
The salt mixtures in NaCl-MgCl2-PuCl₃ and NaCl-ThCl₄-PuCl₃ that had been selected previously have been synthesized in order to perform characterization and further experimental work. Most of the salt properties measurement equipment has been set up and tested, and most measurements have been performed with the NaCl-MgCl2-PuCl₃ salt. They are still on-going for the NaCl-ThCl₄-PuCl₃ salt.
Three MSR configurations have been selected. Neutronic depletion simulations in these reactor configurations have been performed, to assess the evolution of salt mixtures composition, in particular fission products. The results have been used as inputs for the assessment of used salt treatment processes and valuable isotopes management processes. In addition, decay heat and doses calculations have also been performed for the salt mixtures.
A chloride salt irradiation experiment has been performed in the LR-0 reactor in the Czech Republic, during which the neutronic properties of a selection of chloride salts have been measured. Comparison with nuclear data libraries revealed differences in critical parameters, especially in systems with high chlorine concentrations, emphasizing the need for experimental validation of nuclear data.
In order to undergo the test program for structural materials corrosion tests that had been set up at the beginning of the project, the testing methods, the corrosion products measurement methods and the corrosion test set ups and loops have been developed. Selected material samples have been fabricated and tested for their resistance to corrosion when in contact with chloride salt, in variable configurations and exposure durations. Dynamic corrosion tests are now under preparation.
A corrosion / irradiation test bench has been constructed for combined phenomena investigation.
Eight scenarios have been defined, for which the used fuel salt compositions have being calculated. These compositions have been used as input parameters for the technical study of used fuel salt treatment in La Hague. No major obstacles have been identified for the treatment of salt from a demonstrator reactor (1 m3), whether or not chlorine enrichment is used. However, significant dilution is necessary to meet the reprocessing plant criteria, particularly to meet criticality safety limits and to avoid disruption of existing process streams or final waste conditioning.
Dissolution studies have started. The experimental setup design for dissolution is complete, and the work is advancing steadily toward meeting task objectives.
The pyrochemical salt treatment alternatives have also been studied. After defining an experimental plan, tests have been completed. Two extraction methods—reductive extraction on Al-Mg alloy and electrolytic extraction on Al/Cu alloy—achieved over 99% efficiency. Results show promise for separating lanthanides and possibly plutonium in MgCl2-based salts, though validation with Pu data is still needed.
MIMOSA is investigating vitrification of representative chloride salt compositions (NaCl-MgCl2-ThCl4-PuCl3) including key fission products into borosilicate and phosphate glass. High-quality vitrified samples have been obtained. Vitrification experiments with iron phosphate glass have begun.
Valuable isotopes produced in MSRs among other fuel salt mixtures isotopes have been assessed and quantified. The applications for these isotopes and the associated market have been described.
A thermodynamic database on the platinoids chloride phases has been developed, and some experiments have been performed to validate the metallic state of the platinoids and noble metals.
A device to test the ability and efficiency of gas bubbling to extract solid particles in molten salt has been constructed and tests have started.
MIMOSA’s choice has been to focus on the demonstration of several key aspects of technical feasibility and performance, which, linked to the use of this specific type of salt, will push this advanced technology beyond the current state of the art and contribute to acceleration of its deployment.
The technology readiness level (TRL) of the whole concept based on fast-spectrum Cl MSRs can be assessed at 1-2. There are mainly three parts to be considered in the concept: the reactor itself and its neutronics aspects, the materials and salt interreacting between each other, and the fuel cycle aspects (recycling and waste management). The first part, mainly physics, is not covered in MIMOSA. Concerning the materials and salt cycle aspects, the objective of MIMOSA is to raise the chemistry TRL from 1 or 2 to 3 or 4. This will be a great leap forward.
The technology readiness level (TRL) of the whole concept based on fast-spectrum Cl MSRs can be assessed at 1-2. There are mainly three parts to be considered in the concept: the reactor itself and its neutronics aspects, the materials and salt interreacting between each other, and the fuel cycle aspects (recycling and waste management). The first part, mainly physics, is not covered in MIMOSA. Concerning the materials and salt cycle aspects, the objective of MIMOSA is to raise the chemistry TRL from 1 or 2 to 3 or 4. This will be a great leap forward.