Flexible Waste Management to Increase the Effectiveness of Minor Actinide P&T Technology

Partitioning and transmutation (P&T) technologies have been developed for minor actinides (MA) to reduce the high level waste (HLW) volume and long-term radiotoxicity. Although the MA P&T can reduce the potential radiotoxicity effectively by 1-3 orders of magnitude, the actual operation of P&T requires several tens of years for developing elemental technologies of nuclide separation, MA containing fuel fabrication, transmutation and their practical systematization. The high level liquid waste (HLLW) containing MA is presently vitrified immediately after spent fuel reprocessing, stored about 50 years at surface facility and will be disposed of at deep geological repository. Vitrified HLW form works as an excellent artificial barrier against nuclides release during storage and disposal. On the other hand, it is difficult to recover MA from the form. So the present waste management scheme has an issue of MA P&T technology application until its deployment, which will produce much amount of vitrified HLW including long-lived MA without P&T application. Thus the authors proposed the flexible waste management method to increase the effectiveness of the MA P&T. The system adopts the HLLW calcination instead of the vitrification to produce granule for its dry storage of about 50 years until the MA P&T technology will be applicable. The granule should be easily dissolved by the nitric acid solution to apply the typical aqueous MA partitioning technologies to be developed. This paper reports the purpose of the study, the feasibility evaluation results for the calcined granule storage and the evaluation results for the environmental burden reduction effect.     

Hydrometallurgical Reprocessing of BREST-OD-300 Mixed Uranium-plutonium Nuclear Fuel

The duration of external fuel cycle of BREST-OD-300 reactor with mixed U-Pu nitride fuel (MNIT) including hydrometallurgical reprocessing should not exceed 3 years. An average burnup of the fuel should be 6% of heavy metal (HM) with the potential increase up to 10% HM. Therefore, the technology should provide the reprocessing of spent nuclear fuel (SNF) after less than 2 years cooling time and with fissile materials (FM) content of 10 – 15%. Pellets technology has been chosen for the MNIT fuel production. That means necessity to receive the recycled actinides oxides of high purification coefficient (∼ 10⁶). Currently on a laboratory scale, the following process stages have been tested on the real products: actinide oxides production and rare-earth and trans-plutonium elements separation. Moreover, on a pilot scale the process of high level radioactive waste (HLW) and intermediate level radioactive waste (ILW) concentration by evaporation has been tested, as well as the Am-Cm separation. In 2015, the design of the MNIT SNF reprocessing facility has been started, placed at the JSC Siberian Chemical Plant site as a part of the pilot demonstration power complex (PDPC) with BREST-OD-300 reactor. MNIT SNF reprocessing plant (RP) should be put in operation after 2020.     

The role sol–gel process for nuclear fuels-an overview

The paper reviews the sol–gel methods used for the preparation of nuclear fuel materials in the form of microspheres. It also discusses how these microspheres can be fabricated into nuclear fuels for reactors such as High Temperature Gas Cooled Reactors and Fast Reactors. The performance of these microsphere-based fuels is reviewed. More recent applications, such as the transmutation of minor actinides, (Np, Am and Cm) and hydrogen production, are also briefly covered.     

Potentialities of fluoride-based salts for specific nuclear reprocessing: Overview of the R&D program at CEA

Initially studied in the frame of the first French act on radioactive waste management (December 1991), the pyrotechnology is currently assessed by the Nuclear Energy Direction of the Commissariat à l’Energie Atomique (CEA) within the succeeding act (June 2006) as a potential alternative to hydrometallurgy for the reprocessing of targets or dedicated fuels (coming from accelerator-driven systems or ADS) considered for the minor actinides transmutation.The R&D program is mainly focused on the evaluation of the fluoride melts as interesting media for operating separation between the actinides and the fission products. Two separation techniques are currently evaluated; the first one uses the liquid-liquid extraction technique between molten fluoride and liquid metal at high temperature, the second one is based on an electrolytic separation in a molten fluoride melt. Both are promising in terms of separation efficiency. This paper gives an overview of the current studies and presents the last main experimental results.     

Current Progress in R&D of Fluoride Volatility Method

Fluoride Volatility Method is regarded to be a promising advanced pyrochemical reprocessing technology, which can be used for reprocessing mainly of oxide spent fuels coming from current LWRs or future GEN IV fast reactors. The technology should be chiefly suitable for the reprocessing of advanced oxide fuels with inert matrixes of very high burn-up and short cooling time, which can be hardly reprocessed by hydrometallurgical technologies. Fluoride Volatility Method is based on direct fluorination of powdered spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (mostly UF6) are separated from the non-volatile ones (trivalent minor actinides and majority of fission products). The subsequent operations necessary for partitioning of volatile fluorides are condensation and evaporation of volatile fluorides, thermal decomposition of PuF6 and finally distillation and sorption used for the purification of uranium product.     

U1-xAmxO2±δ MABB Fabrication in the Frame of the DIAMINO Irradiation Experiment

In the frame of the heterogeneous transmutation of minor actinides in MABB (Minor Actinides Bearing-Blanket) fuels, the CEA program DIAMINO aims to assess the influence of americium content and microstructure on helium release and fuel swelling. For DIAMINO experiment, four sets of U1-xAmxO2±δ ceramic fuels, namely two compositions (x = 0.075, 0.15) and two microstructures (highly dense: > 95%TD, and highly porous: < 85%TD), were produced in the ATALANTE facility following innovative preparation methods. Pellets were further characterized by several techniques and were all found to be in full compliance with the strict specifications required for such irradiation programs.     

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.     

Effect of the Structure of Amido-polynitrogen Molecules on the Complexation of Actinides

The complexation and solvent extraction of Eu(III) and actinides in different oxidation states (Am(III), Pu(IV), Np(V)) by bitopic molecules with a dipyridyl-phenanthroline cycle as nitrogen unit and one or two amido functions are described. The complexation has been studied in methanol-water solution with hydrophilic molecules to enhance knowledge about this new family of ligands and to identify the most interesting structural effect. Some extraction tests have been performed with lipophilic molecules of the family to check the possible utility of the new class of ligands under representative fuel reprocessing conditions. These first studies have demonstrated that the presence of a preorganized N-donors unit like dipyridyl-phenanthroline improves the ligand's affinity for actinides and its An/Ln selectivity.     

Installation of a Shielded SIMS for the Analysis of Irradiated Nuclear Fuels

A CAMECA IMS 6F secondary ion mass spectrometer (SIMS) for the analysis of irradiated nuclear fuel has been installed in the Microbeam Analysis Laboratory of the Institute for Transuranium Elements (ITU). This device is specially equipped with heavy metal shielding to enable the safe examination of irradiated nuclear fuel samples with activities up to 75 GBq. At ITU the shielded SIMS will be used in conjunction with EPMA taking advantage of the complementary nature of the two techniques and will make important contributions to ongoing research programmes such as the safety of nuclear fuels, the partitioning and transmutation programme and the characterisation of spent fuels. The paper describes the shielded SIMS installation and presents a selection of results from the commissioning tests.     

The EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

Application of ICP-MS and HR-ICP-MS for the characterization of solutions generated from corrosion testing of spent nuclear fuel

Summary The corrosion behavior of spent nuclear fuels under simulated geologically unsaturated and oxidizing conditions are being studied by subjecting both unirradiated and irradiated nuclear fuels to dripping groundwater. Solutions and solid materials are periodically sampled and subsequently analyzed to determine concentrations of groundwater and fuel components in these materials to elucidate corrosion mechanisms. The analyses are performed primarily by inductively coupled plasma mass spectrometry (ICP-MS). For ICP-MS we use the method of internal standardization with direct external calibration with multi-elemental standards possessing natural isotopic abundances for the determination of concentrations groundwater components and indirect instrumental response calibration for the determination of fuel components. Additionally, we are utilizing high resolution inductively coupled plasma mass spectrometry (HR-ICP-MS) to enhance our ability to determine concentrations of low-solubility actinides at ultratrace concentrations.     

Main Results of the French Program on Partitioning of Minor Actinides,a Significant Improvement Towards Nuclear Waste Reduction

In the framework of the successive 1991 and 2006 Waste Management Act, French government supported a very significant R&D program on partitioning and transmutation of minor actinides (MA). This program aims to study potential solutions for still minimizing the quantity and the hazardousness of final waste, by MA recycling. Indeed, MA 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. Within this framework, this paper aims to present the most recent progress obtained in CEA on the development of innovative actinide partitioning hydrometallurgical processes in support of their recycling, either in an homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in an heterogeneous mode (MA are recycled at higher concentration in specific targets, at the periphery of the reactor core). Recovery performances obtained on recent tests in high active conditions of the so-called GANEX and DIAMEX-SANEX process will be presented and discussed in light of the potential P&T scenarios. Finally, recent developments regarding the recycling of the sole Am will be presented as well as the results obtained on highly active solutions for this so-called EXAM process. This set of results gives to the French government a portfolio of potential recycling processes which could be separately and progressively implemented if decided.     

Chemistry of nuclear fuel reprocessing: Current status

Current status on the chemical aspects of nuclear fuel reprocessing is presented with special emphasis on the Purex process which continues to be the process of choice for the last four decades. Better decontamination from fission products, new methods for uraniumplutonium partitioning and removal of actinides from high active waste are challenging areas in process chemistry. The development work on TRUEX and DIAMEX process for treating high active waste is briefly described. An overview of pyrochemical processes, which are important for Integral Fast Reactor Concept, is presented.     

Resourceability on nuclear fuel cycle by transmutation approach

A resourceability on nuclear fuel cycle by transmutation of fission products in the spent fuel of nuclear reactors is discussed in this paper to investigate the feasibility of "creation and utilization" of Après ORIENT from Adv.-ORIENT cycle,in which chemical "separation and utilization" of nuclear rare metals(platinum group metals,Mo,Tc,rare earth,etc.) has been proposed since FY2006.Après ORIENT research program was newly initiated in FY2011 for nuclear transmutation of fission products into stable or short-lived highly-valuable elements.In the resourceability of rare earth metals from fission products,non-radioactive Nd and Dy can be created from Pr and Tb,respectively,by transmutation.Especially,the Dy creation has a relatively high feasibility of about 10-20 %/y in creation rate.A proper moderation of neutrons in blanket of fast reactors may be required to provide a high creation rate of La from Ba.In light platinum group metals,non-radioactive Ru can be created from Tc by transmutation,of which creation rate is about 4-5 %/y in blanket of fast reactors.Pd created from Rh is almost non-radioactively depending on the isotope fraction of 107 Pd.Rh creation from Ru is not feasible under the neutron irradiation of typical nuclear reactors.     

Status of sol–gel process for nuclear fuels

Sol–gel process provides an alternate route for fabrication of ceramic nuclear fuel. The sol–gel process provides several advantages over the conventional powder pellet fabrication process by eliminating handling of radioactive powders. The sol–gel process uses only fluids or fluid like materials, thus become amenable to remote handling. The sol–gel process has been developed for the production of coated particle fuels for High Temperature Gas Cooled Reactors (HTGRs), as sphere-pac fuel for Fast Breeder Reactors (FBRs) and as SGMP fuel for Thermal Reactors. Internal Gelation Process is one of the most important routes of the sol–gel process and has been accepted as the most promising process route globally. Several countries having plutonium or ²³³U based fuel program have developed sol–gel process for nuclear fuels. In India there is special interest for the development of the sol–gel process for the thorium–uranium fuels keeping in view the large resources of thorium in India. Sol–gel process for fuel fabrication is also very attractive route for closing the nuclear fuel cycle efficiently. Author is BRNS Raja Ramanna Fellow.     

Comparison of the effect of alpha and gamma radiolysis on the extraction of americium by C5-BTBP in cyclohexanone

Solvent extraction is a separation technique suitable for the treatment of used nuclear fuel. Two immiscible phases are contacted and the metals of interest are extracted from one phase into the other, most often using so called extractants. One group of extractants is the bis(triazine)-bipyridine (BTBP) type molecules. These molecules have been developed within EU research programs for the separation of actinides from lanthanides. During such an extraction process, the components of the two phases will be exposed to ionizing radiation, since the used fuel contains many highly radioactive species. Radiolytic reactions can alter the chemistry of the extracting system, and affect the metal extraction by degradation of the extractant and the formation of degradation products. In this paper the effect of irradiation with alpha particles and gamma rays, respectively, has been studied for one of the BTBP type molecules, C5-BTBP.     

An analysis of the potassium concentrations of soft drinks by HPGe gamma spectrometry

A pyrochemical processing has become one of the potential technologies for a future nuclear fuel cycle. An integrated multi-physics simulation and electrotransport model of a molten-salt electrolytic process are proposed and discussed with respect to the recovery of pure uranium when using thermochemical data. This study has been performed to provide information for diffusion boundary layers between the molten salt (KCl-LiCl) and electrode. The diffusion-controlled electrochemical model demonstrate a prediction of the electrotransport behaviors of LWR spent fuel as a function of the time up to the corresponding electrotransport satisfying a given applied current based on a galvanostatic electrolysis.     

Reactive Magnesium Nitride Additive: A Drop-in Solution for Lithium/Garnet Wetting in All-Solid-State Batteries

Garnet oxides such as Li_6.4La₃Zr_1.4Ta_0.6O₁₂ (LLZTO) are promising solid electrolyte materials for all-solid-state lithium-metal batteries because of high ionic conductivity, low electronic leakage, and wide electrochemical stability window. While LLZTO has been frequently discussed to be stable against lithium metal anode, it is challenging to achieve and maintain good solid-on-solid wetting at the metal/ceramic interface in both processing and extended electrochemical cycling. Here we address the challenge by a powder-form magnesium nitride additive, which reacts with the lithium metal anode to produce well-dispersed lithium nitride. The in situ formed lithium nitride promotes reactive wetting at the Li/LLZTO interface, which lowers interfacial resistance, increases critical current density (CCD), and improves cycling stability of the electrochemical cells. The additive recipe has been diversified to titanium nitride, zirconium nitride, tantalum nitride, and niobium nitride, thus supporting the general concept of reactive dispersion-plus-wetting. Such a design can be extended to other solid-state devices for better functioning and extended cycle life.     

Hot particles analysis originating from failed and damaged fuels

The increase of activities of fission products and transmutation products in the primary coolant of a nuclear power plant indicates the presence of fuel rod failures. The measurement of the activity concentration of the primary coolant was able to detect fuel failures in the reactor core. Microanalytical methods for examining individual hot particles have been developed and applied to fuel failure detection under normal operation conditions as well as during the severe fuel damage that occurred in the cleaning tank incident at Unit 2 of NPP Paks in April 2003. Several faulty fuel rods can be detected simultaneously by the characterization of individual hot particles originating from the primary water. The analysis of particles originating from the damaged fuels provides information relating to the dissolution process of the fuel debris.