High-level Waste Management Research Articles

Phosphate-Based Dechlorination of Electrorefiner Salt Waste using a Phosphoric Acid Precursor.

Electrochemical processing of spent nuclear fuel in molten chloride salts results in radioactive salt waste. Chlorine removal from the salt has been identified as an effective and efficient first step in the management of high-level waste. In this work, a simple salt was dechlorinated with a phosphoric acid phosphate precursor, resulting in a glassy dechlorinated product. The dechlorination efficacy was evaluated in air and argon environments. This work serves as an initial step to advance the Technological Readiness Level of H3PO4-based dechlorination step toward implementation of iron phosphate waste forms to immobilize electrochemical fuel reprocessing salt waste streams.

Thermal Effects and Glass Crystallization in Composite Matrices for Immobilization of the Rare-Earth Element–Minor Actinide Fraction of High-Level Radioactive Waste

The current policy of managing high-level waste (HLW) derived in the closed nuclear fuel cycle consists in their vitrification into B-Si or Al-P vitreous forms. These compounds have rather limited capacity with respect to the HLW (5–20 wt%), and their properties change over time due to devitrification of the glasses. Cardinal improvement in the management of HLW can be achieved by their separation onto groups of elements with similar properties, followed by their immobilization in robust waste forms (matrices) and emplacement in deep disposal facilities. One of the possible fractions contains trivalent rare-earth elements (REEs) and minor actinides (MAs = Am and Cm). REEs are the fission products of actinides, which are mainly represented by stable isotopes of elements from La to Gd as well as Y. This group also contains small amounts of short-lived radionuclides with half-lives (T1/2) from 284 days (144Ce) to 90 years (151Sm), including 147Pm (T1/2 = 2.6 years), 154Eu (T1/2 = 8.8 years), and 155Eu (T1/2 = 5 years). However, the main long-term environmental hazard of the REE–MA fraction is associated with Am and Cm, with half-lives from 18 years (244Cm) to 8500 years (245Cm), and their daughter products: 237Np (T1/2 = 2.14 × 106 years), 239Pu (T1/2 = 2.41 × 104 years), 240Pu (T1/2 = 6537 years), and 242Pu (T1/2 = 3.76 × 105 years), which should be immobilized into a durable waste form that prevents their release into the environment. Due to the heat generated by decaying radionuclides, the temperature of matrices with an REE–MA fraction will be increased by hundreds of centigrade above ambient. This process can be utilized by selecting a vitreous waste form that will crystallize to form durable crystalline phases with long-lived radionuclides. We estimated the thermal effects in a potential REE–MA glass composite material based on the size of the block, the content of waste, the time of storage before immobilization and after disposal, and showed that it is possible to select the waste loading, size of blocks, and storage time so that the temperature of the matrix during the first decades will reach 500–700 °C, which corresponds to the optimal range of glass crystallization. As a result, a glass–ceramic composite will be produced that contains monazite ((REE,MA)PO4) in phosphate glasses; britholite (Cax(REE,MA)10-x(SiO4)6O2) or zirconolite ((Ca,REE,MA)(Zr,REE,MA)(Ti,Al,Fe)2O7), in silicate systems. This possibility is confirmed by experimental data on the crystallization of glasses with REEs and actinides (Pu, Am). The prospect for the disposal of glasses with the REE–MA fraction in deep boreholes is briefly considered.

Nuclear Waste Management

Nuclear energy production generates nuclear waste. Nuclear and radioactive waste, especially high level waste (HLW) and intermediate level waste (ILW), require special long-term safe management solutions. One part of the solution involves recycling through reprocessing of spent nuclear fuel. Recycling reduces the volume of nuclear waste and allows for the reuse of some of its components. This can be achieved through methods such as adsorption, ion exchange, coagulation, flotation, filtration, chemical precipitation, reverse osmosis, and solvent extraction, like the PUREX process. Another possible future solution involves partitioning and transmutation technology, which can reduce the production of nuclear waste. The best long-term solution is the immobilisation of HLW and ILW in a solid matrix. Materials used for this purpose include glasses, cements, bitumen, geopolymers, concrete, and ceramics as a promising material. While cementation is still the most commonly used immobilisation method due to its low cost and simplicity, vitrification is a more permanent long-term solution. Deep geological disposal in combination with vitrification and a robust multi-barrier system is considered the most acceptable solution for safe nuclear waste isolation. This review provides insight into the mostly commonly used and promising immobilisation materials, as well as the most effective methods and technologies currently in use and under development for the management of HLW and ILW in nuclear waste management.

Small Cyclic Diglycolamides: Tautomerism, Solvent Extraction and Coordination with f-Elements: One Strain to Rule Them All.

The search for new effective extractants is an important task for the management of high-level liquid waste (HLW) generated during the reprocessing of spent nuclear fuel. Here, we synthesized a series of diglycolamides with cyclic substituents for the first time. We disclosed their coordination with f-element nitrates [La(NO3)3 and UO2(NO3)2] by SC-XRD study and complexation properties toward Am(III), Ln(III), and U(VI) during solvent extraction from nitric acid solutions. Using dynamic nuclear magnetic resonance (NMR) and density functional theory (DFT) calculations, the importance of tautomerism in the extraction properties of diglycolamides was shown.

MINERAL RESOURCES OF THE NUCLEAR INDUSTRY OF RUSSIA AND ISOLATION OF RADIOACTIVE WASTE

The tasks of ensuring the sustainable development of the nuclear energy industry in Russia are considered, which are aimed at solving the problems of the initial and final stages of the nuclear fuel cycle from the reproduction of the mineral resource base of uranium to the management of high-level liquid waste from the reprocessing of spent nuclear fuel. An analysis was made of the trends in the development of the mineral resource base of uranium in connection with the predicted growth in electricity generation at nuclear power plants. Based on the example of the territory of southeastern Transbaikalia, within the framework of topical mineralogical studies, new approaches to the allocation and justification of prospecting areas are proposed based on models of mineral-forming systems of uranium deposits using Earth remote sensing data, GIS modeling and neural network technologies. A detailed analysis of trends in solving the problems of the final stages of the nuclear fuel cycle has been carried out. For isolation and subsequent disposal of the REE-actinide fraction of high-level waste, it is proposed to use phosphate glass ceramics with monazite

Tuning the Alkyl Chain of Nitrilotriacetamide for Selectively Extracting Trivalent Am over Eu Ions.

The successful management and safe disposal of high-level nuclear waste necessitate the efficient separation of actinides (An) from lanthanides (Ln), which has emerged as a crucial prerequisite. Mixed donor ligands incorporating both soft and hard donor atoms have garnered interest in the field of An/Ln separation and purification. One such example is nitrilotriacetamide (NTAamide) derivatives, which have demonstrated selectivity in extracting minor actinide Am(III) ions over Eu(III) ions. Nevertheless, the Am/Eu complexation behavior and selectivity remain underexplored. In the work, a comprehensive and systematic investigation has been conducted for [M(RL)(NO3)3] complexes (M = Am and Eu) utilizing relativistic density functional theory. The NTAamide ligand (RL) is substituted with various alkyl groups, namely, methyl, ethyl, propyl, n-butyl, n-pentyl, n-hexyl, n-heptyl, and n-octyl. Thermodynamic calculations show that the alkyl chain length in NTAamide is capable of tuning the separation selectivity of Am and Eu. Moreover, the differences in calculated free energies between Am and Eu complexes are more negative for R = Bu-Oct than Me-Pr. This indicates that elongation of the alkyl chain can increase the efficiency of selective separation of Am(III) from Eu(III). Based on the quantum theory of atoms in molecules and charge decomposition analyses, it has been observed that the strength of Am-RL bonds is higher than that of Eu-RL bonds. This disparity is attributed to a greater degree of covalency in Am-RL bonds and a higher level of charge transfer from ligands to Am within complexes containing these bonds. Energies of occupied orbitals with the central N character are recognized overall lower for [Am(OctL)(NO3)3] than for [Eu(OctL)(NO3)3], indicative of stronger complexation stability of the former. These results offer valuable insights into the separation mechanism of NTAamide ligands, which can help guide the development of more powerful agents for An/Ln separation in future applications.

Study Analysis of Multiplication Factor on ADS Transmutation System

New research technology for partition and transmutation (P-T) of minor actinide (MA) contained in high-level waste (HLW) has been carried out to design accelerator-based transmutation system or commonly known as Accelerator Driven Subcritical (ADS) device. The objective of this system is to eliminate or minimize high-level radioactive waste that has long half-life and to develop long-term safety assurance in HLW management. Analysis done by using Origen2.1 and mCnP6 codes. As matrix, Th-232 with optimum weight 75% is used.. Comparing with , analysis results using MCNP shows that the use of PWR nuclear spent fuel for ADS device can be done by reprocessing to eliminate U-238 nuclide, which is the source of the formation of plutonium nuclide and minor actinides. From the results and discussion, it can be concluded that observation on transmutation of transuranic nuclides using ORIGEN2.1 Code on ADS device has been conducted successfully with LMFBR library. 15%-Th232 fuel has been justified as the most optimum ADS fuel based on K-inf reached at EOC on various fuel types. Mass reduction occurs at EOC for U-236, U-235, U-234, Th-232, Np-237, Pu-238, Am-241, Pu-242, and Cm 244. Radionuclides that experience mass changes during cooling time are Pu-238, Pu-240, Am-241, Pa-233, and Cm-244.

Na-montmorillonite to Fe(II)-Mt using ferrous citrate/ascorbate obtained by dissolving iron powder

Bentonite containing montmorillonite (Mt) as the major clay mineral is the potential buffer material of engineered barrier system in the deep geological repository for high-level waste management. The corrosion products of canister/or overpack material (carbon steel) may alter the basic characteristics of bentonite through Fe/clay interaction and formation of FeMt to certain extent. A convenient method is required for quantitative preparation of Fe(II)-Mt because the only few direct methods reported have some limitations. For this, weak ferrous complexes having good solubility to facilitate the cation-exchange with bentonite is essential. Therefore, Fe(II)-citrate, Fe(II)-ascorbate and Fe(II)-citrate/ascorbate were obtained here by dissolving the iron powder (Fe0) in citric acid (CA), ascorbic acid (AA) and their equimolar mixtures from 25 mM to 300 mM at 70 °C under N2 atmosphere. The redox potential was measured to monitor the dissolution of Fe0 in these mild organic acids. Higher dissolution of Fe0 occurred in CA + AA mixture. The dissolved iron in the form of ferrous complex was reasonably stable in inert atmosphere, Fe2+/Fetotal ≈ 1. The parent clay mineral (bentonite, NaMt) was treated with Fe(II) − cit/asc to form Fe(II) − Mt. through cation exchange process. The CEC and ferrous to ferric ratio of the Fe(II)-Mt was close to the stoichiometric amount. The XRF, XRD, ac impedance, FT-IR, TGA/DSC, XPS and FE-SEM are used for further characterization. In-situ complexation reaction of interlayer Fe2+ ions in Fe(II)-Mt with o-phenanthroline was also observed.

Dissolution Behavior of Simulated Spent Nuclear Fuel in LiCl-KCl-UCl3 Molten Salt

Considering the necessity of the development of methods to reduce the burden of storage and disposal of high-level radioactive waste, in this study, we propose a nuclide separation technique using molten salt immersion. The dissolution behavior of simulated spent nuclear fuels (SSFs) immersed in a LiCl-KCl-UCl3 (LKU) molten salt at 500°C was investigated using a combination of thermodynamic and experimental studies. Surrogates of transuranic elements (TRUs), that is, rare earth elements (REs), in the SSFs were dissolved into the molten salt without any damage to the UO2 structure of the SSFs. The results suggest that the LKU salt technique can be used to separate REs and, potentially, TRUs from actual spent nuclear fuels (SNFs). It is thought that this technique is advantageous over the conventional TRU recovery techniques because the majority of the SNFs (i.e., UO2) remained stable, thus reducing the process burden. Several SNF treatment process options using this technique were suggested. This study will serve as a guide for future studies on the management of high-level waste discharged from nuclear reactors.

Political Mediation in Nuclear Waste Management: a Foucauldian Perspective

This paper aims to open up high-level waste management practices to a political philosophical questioning, beyond the enclosure implied by the normative ethics approaches that prevail in the literature. Building on previous insights derived from mediation theory (in particular the work of Verbeek and Dorrestijn), Foucault and science and technology studies (in particular Jasanoff’s work on socio-technical imaginaries), mediation theory’s appropriation of Foucauldian insights is shown to be in need of modification and further extension. In particular, we modify Dorrestijn’s figure of “technical determination of power relations” to better take into account the (literal and figurative) aspects of imagination, and complement Dorrestijn’s work with the figures of techno-scientific mediation, and the inherently political figures of socio-technical and state-technical mediation, both based on Foucault’s notion of governmentality. Our analysis implies that the practical implementation of a high-level nuclear waste (HLW) management strategy will require the “stitching together” of these different mediations, which is an inherently political task.

Monte-Carlo based investigation of individual dosimetry in deep geological repository for high-level nuclear waste with consideration of realistic body postures

Deep geological disposal with provisions for retrieval is one of the most promising strategies for the definite management of high-level nuclear waste. However, retrieval of disposed waste might require certain extensive work of personnel near the waste package that might also enhance the level of radiation exposure. Hence, a precise estimation of the personal dose with consideration of realistic body postures is highly desired for optimization of individual working scenarios. In this study, the near field of a generic geological disposal facility was modeled with a horizontal emplacement drift in saline host rock. Inside the drift, a shielded disposal cask loaded with spent nuclear fuel was placed on the ground. A whole-body stylized phantom with moveable limbs was applied to represent a reference worker. Several relevant realistic body postures that might be encountered in a disposal facility were investigated. The corresponding radiation exposure was calculated and compared with a general-purpose Monte Carlo code.

Sorption and surface complexation modeling of 137Cs on Fe(II)-montmorillonite clay mineral relevant to nuclear waste disposal

Abstract Bentonite, a natural clay, is a promising candidate to be used as a buffer/backfill material in the geological disposal systems for management of high level waste (HLW). Fe released due to corrosion of canister/overpack may result in alteration of Na-montmorillonite (Na-Mt), present in bentonite, to Fe-montmorillonite (Fe-Mt) in a span of several years after the emplacement of vitrified HLW in deep geological repositories. For realistic performance assessment, it is essential to understand the sorption behavior of altered Mt with regard to the radionuclides present in HLW. Cs is one of the high-yield (137Cs t1/2 = 30.1 y, 6%) and long-lived (135Cs t1/2 = 2 × 106 y) fission products in spent fuels. The objective of present study is to understand the effect of various parameters, viz., time (0–48 h), pH (3.0–9.0), ionic strength (0.001–1 M) [Cs(I)] (10−10–10−3 M) and Fe dissolution on sorption behavior of Cs(I) on Fe(II)-Mt through batch sorption experiments. Fe(II)-Mt was synthesized by reducing Fe(III)-Mt using ascorbic acid as reducing agent in N2 atmosphere. The near-constancy in Cs(I) sorption on Fe(II)-Mt with pH (≥4), and decrease with increasing ionic strength, illustrate the ion exchange as dominant mode of Cs(I) sorption. Further, Cs(I) sorption isotherm on Fe(II)-Mt is found to be linear. The estimation of dissolved iron in the supernatant of Fe(II)-Mt suspensions demonstrated that dissolved Fe decreased with increase in pH and increased with increase in ionic strength. Moreover, the Fe2+/Fetotal ratio determined in all experiments was close to unity, thereby depicting that Fe(II) did not oxidize to Fe(III), except when suspension pH was ≥ 5.5. For comparison, Cs(I) sorption was also studied on Na(I)-Mt and compared with that on Fe(III)-Mt. Surface complexation modeling of Cs(I) sorption on the three clay minerals, viz., Na-Mt, Fe(II)-Mt and Fe(III)-Mt, has been successfully carried out.

Structure of NaFeSiO4, NaFeSi2O6, and NaFeSi3O8 glasses and glass-ceramics

AbstractThe crystallization of iron-containing sodium silicate phases holds particular importance, both in the management of high-level nuclear wastes and in geosciences. Here, we study three as-quenched glasses and their heat-treated chemical analogs, NaFeSiO4, NaFeSi2O6, and NaFeSi3O8 (with nominal stoichiometries from feldspathoid, pyroxene, and feldspar mineral groups, i.e., Si/Fe = 1, 2, and 3, respectively) using various techniques. Phase analyses revealed that as-quenched NaFeSiO4 could not accommodate all Fe in the glass phase (some Fe crystallizes as Fe3O4), whereas as-quenched NaFeSi2O6 and NaFeSi3O8 form amorphous glasses. NaFeSi2O6 glass is the only composition that crystallizes into its respective isochemical crystalline polymorph, i.e., aegirine, upon isothermal heat-treatment. As revealed by Mössbauer spectroscopy, iron is predominantly present as fourfold-coordinated Fe3+ in all glasses, though it is present as sixfold-coordinated Fe3+ in the aegirine crystals (NaFeSi2O6), as expected from crystallography. Thus, Na-Fe silicate can form a crystalline phase in which it is octahedrally coordinated, even though it is mostly tetrahedrally coordinated in the parent glasses. Thermal behavior, magnetic properties, iron redox state (including Fe K-edge X-ray absorption), and vibrational properties (Raman spectra) of the above compositions are discussed.

Reevaluating the efficacy of moderate annealing in nuclear waste vitrification for sustainable high-level waste management

The reliable immobilization of nuclear waste, particularly high-level waste (HLW), is a key issue for sustainable utilization of nuclear energy. Annealing process is adopted in nuclear waste vitrification to improve mechanical properties of products; however, its efficacy on chemical stability of treated HLW glass remain uncertain. This study aims to reevaluate the effects of moderate annealing on chemical durability of sodium borosilicate glass matrix incorporated with simulated HLW. Two types of leaching tests, i.e. acid leaching and the modified Product Consistency Test (PCT), were performed to examine the leaching behavior of HLW glassy products controlled by corrosion mechanisms of ion exchange and hydrolysis, respectively. The acid leaching results showed that annealed glass samples released lanthanum (La) species at a faster rate than that of quenched samples, but the effects of annealing on the release behavior of alkaline and alkaline-earth elements were limited. In contrast, the PCT results showed minor difference between quenched and annealed samples, which suggests that the small structure modification caused by moderate annealing was not comparable with the dominant hydrolysis reaction occurred in the PCT process. According to the free-volume theory, as well as evidences derived from Raman characterization and thermodynamic measurements, La was inferred to exist in the free-volume region in the HLW glass, and the annihilation of free volume during annealing could mobilize La from its stable surrounding environment, and thus reducing the chemical stability of nuclear waste glass.

Removal of Cesium from Simulated Spent Fuel Dissolver Liquor

With a resurgence of the nuclear industry’s fortunes, waste management will be an even greater consideration; reducing wastes, better segregation and treatment that lower the impact on waste storage facilities and repositories, in particular geological repositories will help the sustainability of the industry. We at the University of Central Lancashire (UCLan) proposed in a previous publication a sequential chromatographic separation process, Alternative Reprocessing Technology (ART) for Fission Products (FPs) and Minor Actinides (MAs) separation from spent fuel dissolver liquor, as an improvement to/replacement for PUREX (Plutonium Uranium Redox Extraction). This publication addresses the removal of one particular fission product, cesium, and its impact on waste management, and down-stream PUREX operations. Although our proposed process is still in its infancy, its impact on the PUREX process could be significant, with major gains in the separation circuit, waste management of High Level Waste (HLW) and subsequently waste disposal to and design of a geological repository. This paper briefly describes; our concept, some preliminary experimental data and why re-classification of the bulk of HLW (High Level Waste) to Intermediate Level Waste (ILW) would be possible.

Modelling the multiscale behaviour of claystone: deformation, rupture, and hydro-mechanical phenomena around underground galleries

In the context of underground exploitation, the behaviour of rocks near galleries and tunnels conditions their stability. Underground drilling generates deformations, damage, fracturing, and significant modification of flow characteristics in the surrounding rock. However, the influence of small-scale characteristics and behaviour on the rock deformations and damage at engineering scale remains a complex issue. Consequently, the multiscale behaviour of a clay rock is modelled starting from the large scale of the excavation damaged zone around galleries and then enriching the approach by considering microstructural characteristics from the scale of mineral inclusions. Lastly, a double-scale numerical framework is considered. It allows to relate small- to large-scale rock behaviour in terms of deformations and material rupture. In fact, the development of damage and cracking at microscale allows to predict large-scale fracturing. The developed method focuses on a claystone in the particular context of long-term management of high-level nuclear wastes by deep geological repository. The results highlight the possibilities of double-scale computing in the prediction of the behaviour of underground engineering structures.

Needs of countries with longer timescale for deep geological repository implementation

Countries operating nuclear power plants have to deal with the tasks connected to spent fuel and high-level radioactive waste management. There is international consensus that, at this time, deep geological disposal represents the safest and most sustainable option as the end point of the management of high-level waste and spent fuel considered as waste. There are countries with longer timescale for deep geological repository (DGR) implementation, meaning that the planned date of commissioning of their respective DGRs is around 2060. For these countries cooperation, knowledge transfer, participation in RD&D programmes (like EURAD) and adaptation of good international practice could help in implementing their own programmes. In the paper the challenges and needs of a country with longer implementation timescale for DGR will be introduced through the example of Hungary.

The Influence of Zeolite (Sokyrnytsya Deposit) on the Physical and Chemical Resistance of a Magnesium Potassium Phosphate Compound for the Immobilization of High-Level Waste.

The manuscript presents the results of the development of new material for high-level waste (HLW) management: the magnesium potassium phosphate (MKP) compound. The possibility of using zeolite (Sokyrnytsya deposit) to increase the mechanical, thermal, and hydrolytic resistance of this compound with immobilized HLW was studied. The main component of the used natural zeolite is a mineral of the clinoptilolite–heulandite series, and quartz, microcline, and clay minerals (illite, sepiolite, and smectite) are present as impurities. The compressive strength of the compound, containing at least 4.2 wt % zeolite, is about 25 MPa. Compound containing 28.6 wt % zeolite retains high compressive strength (at least 9.0 MPa), even after heat treatment at 450 °C. The adding of zeolite to the composition of the compound increases its hydrolytic stability, while the leaching rate of the mobile nuclides 137Cs and 90Sr decreases up to one order of values. Differential leaching rate of radionuclides from the compound containing 28.6 wt % zeolite is 2.6 × 10−7 for 137Cs, 2.9 × 10−6 for 90Sr, 1.7 × 10−9 for 239Pu, and 2.9 × 10−9 g/(cm2∙day) for 241Am. Thus, the properties of the resulting compound correspond to the requirements for solidified HLW in Russia.

NCRP Vision for the Future and Program Area Committee Activities in 2018.

The National Council on Radiation Protection and Measurements' (NCRP) congressional charter aligns with our vision for the future: to improve radiation protection for the public and workers. This vision is embodied within NCRP's ongoing initiatives: preparedness for nuclear terrorism, increasing the number of radiation professionals critically needed for the nation, providing new guidance for comprehensive radiation protection in the United States, addressing the protection issues surrounding the ever-increasing use of ionizing radiation in medicine (the focus of this year's annual meeting), assessing radiation doses to aircrew related to higher altitude and longer flights, providing guidance on emerging radiation issues such as the radioactive waste from hydraulic fracturing, focusing on difficult issues such as high-level waste management, and providing better estimates of radiation risks at low doses within the framework of the Million Person Study of Low-Dose Radiation Health Effects. Cutting-edge initiatives included a reevaluation of the science behind recommendations for lens of the eye dose, recommendations for emergency responders on dosimetry after a major radiological incident, guidance to the National Aeronautics and Space Administration with regard to possible central nervous system effects from galactic cosmic rays (the high-energy, high-mass ions bounding through space), reevaluating the population exposure to medical radiation, and addressing whether the linear no-threshold model is still the best available for purposes of radiation protection (not for risk assessment). To address these initiatives and goals, NCRP has seven program area committees on biology and epidemiology, operational concerns, emergency response and preparedness, medicine, environmental issues and waste management, dosimetry, and communications. The NCRP vision for the future will continue and increase under the leadership of President-Elect Dr. Kathryn D. Held (Massachusetts General Hospital and Harvard Medical School, and current NCRP executive director and chief science officer). The NCRP quest to improve radiation protection for the public is hindered only by limited resources, both human capital and financial.

Economic assessment of the political and technical management of large-scale public projects. The deployment schedule of France’s deep geological repository of radioactive waste

In France, the 2006 act on radioactive waste management stipulated the rapid deployment of a deep geological repository (DGR) for high-level waste (HLW) management. This paper analyzes the impact of different DGR deployment schedules on project costs in order to discern an optimal schedule for the disposal of radioactive waste. It appears more economically favorable to extend the duration of storage rather than to dispose of the waste immediately. However, the French parliament has decided to act quickly, accepting a 1.5% excess cost every year, to avoid leaving the burden to future generations and to continue nuclear development. Secondly, the analysis reveals the economic benefits of shortening the duration of the disposal process.JEL classification: Q01, Q58, Q48.