Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu, Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha spectrometry.     

A triple continuum one-dimensional transport model for colloid facilitated contaminant migration in sets of parallel fractures with fracture skin

A triple continuum one-dimensional transport model is developed to analyse colloid facilitated contaminant transport in fractured geological formations. The model accounts for contaminant transport in the fracture, reversible deposition onto fracture surfaces and onto the colloids, diffusion into the rock formation and irreversible deposition of colloids onto the fracture surfaces. Sorption of the contaminant onto the fracture surfaces and onto suspended and deposited colloids are assumed to follow the linear equilibrium assumption (LEA); whereas the irreversible deposition of colloids onto the fracture skin surface is assumed to be governed by the linear kinetic sorption isotherms. The resulting coupled contaminant transport equations are solved using a numerical model employing fully implicit finite difference method based formulation. Results clearly demonstrate that the presence of the fracture skin significantly influences colloid facilitated contaminant migration in fractured formations. Fracture skin porosity and fracture skin diffusion coefficient are demonstrated to be the critical fracture skin properties that affect colloid facilitated contaminant migration in fractures. The impact of different colloid parameters on contaminant transport is investigated. The distribution coefficient for contaminant sorption onto the suspended colloids is found to be the most significant colloid related parameter influencing contaminant migration in fractured formation with fracture skin.     

China's progress in radionuclide migration study over the past decade (2010–2021): Sorption,transport and radioactive colloid

The radionuclide (RN) migration study is not only helpful to understand environmental behavior of RNs, but also can establish the basis for the safety assessment of geological disposal of high-level radioactive waste (HLW). In the context of China's HLW disposal, this review briefly summaries the progress of China's RN migration studies over the past decade regarding three aspects, RN sorption, RN transport and radioactive colloid. Domestic studies from other disciplines (such as geology and environmental science) are also included in this review because they can provide references for the RN migration study. Overall, China has achieved clear progress in RN migration study over the past decade, although large-scaled field experiments are lacked and a gap still exists comparing with the international advanced level. Finally, several suggestions are proposed for future RN migration research in China.     

Recovery of actinides extracted by Truex solvent from high level waste using complexing agents

This work deals with the batch studies on stripping of actinides extracted by a mixture octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and tri-n-butyl phosphate (TBP) in n-dodecane (Truex solvent) from simulated high level waste (HLW) solution. The stripping of americium and plutonium from acid-bearing CMPO-TBP mixture is carried out using a mixture of weak acid, weak base and complexing agent as strippant. A mixture of formic acid, hydrazine hydrate and citric acid appeared to be best suited for efficient stripping of americium and plutonium. With appropriate modifications in the concentration of individual constitutents, this strippant can be used for the recovery of actinides from loaded Truex solvent with any acid content.     

Radiometric detector options to aid in DOE high activity waste tank in situ characterization efforts

Radioactive waste cleanup and subsequent closure of waste storage tanks is currently underway at the Savannah river site, prompting the need to characterize the residual contents (heels) of the tanks. Occasionally, results from laboratory analyses indicate alternative sub-sampling strategies are needed, resulting in repetitive efforts to sample and analyze tank bottoms. The development of a system for in situ tank analyses using a radiometric probe, which could be lowered into a waste tank, could aid in identifying waste structures on tank bottoms requiring further sampling and characterization. Ideally, the probe would provide information for determining which structures were higher in concentrations of actinides and fission products characteristic of DOE high level waste (HLW) heels. Although only a limited set of isotopes can be measured directly without extensive radiochemical separations, the low-energy photon spectra of HLW do offer some intriguing possibilities for characterization using a radiometric probe. One possibility for obtaining a low-energy photon spectrum in the presence of high levels of interfering radiation would be to design a probe primarily based upon recently developed technology from Amptek Inc. Such a detector would be relatively insensitive to the high photon background, which would paralyze conventional gamma probes (i.e. sodium iodide) subjected to the same radiological conditions. The prototype detector is capable of successfully obtaining high resolution measurements at very high count rates (in excess of 500,000 counts per second). An overview of measurements obtained from various HLW samples using the prototype Amptek detector, as well as some additional detector technologies, which could enhance this prototype, will be discussed.     

DECSERVIS: a tool for radioactive decay series visualisation

We have developed an interactive visualisation tool, decay series visualisation (DECSERVIS), for exploring the three natural radioactive decay chains. Through DECSERVIS, one can investigate the full decay scheme of any natural decay chain radionuclide to obtain the number of nuclides, their masses, activities, and activity ratios, accounting for all the daughters, starting from initial conditions freely chosen by the user. The tool has been developed particularly for user friendly and flexible operation. Chain decay in closed systems can be explored as a function of time with various graphical presentations such as solid curve and column diagrams or animation. We present several exploration examples related to geological dating. DECSERVIS will be freely available on request.     

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.     

High Sorption and Selective Extraction of Actinides from Aqueous Solutions

The selective elimination of long-lived radioactive actinides from complicated solutions is crucial for pollution management of the environment. Knowledge about the species, structures and interaction mechanism of actinides at solid–water interfaces is helpful to understand and to evaluate physicochemical behavior in the natural environment. In this review, we summarize recent works about the sorption and interaction mechanism of actinides (using U, Np, Pu, Cm and Am as representative actinides) on natural clay minerals and man-made nanomaterials. The species and microstructures of actinides on solid particles were investigated by advanced spectroscopy techniques and computational theoretical calculations. The reduction and solidification of actinides on solid particles is the most effective way to immobilize actinides in the natural environment. The contents of this review may be helpful in evaluating the migration of actinides in near-field nuclear waste repositories and the mobilization properties of radionuclides in the environment.     

Diffusion of iodide, cesium and strontium in charnockite rock mass

Experiments have been conducted to study the diffusion of iodide, cesium and strontium ions through intact and fractured rock samples of charnockite rock formation present at Kalpakkam, India. The diffusion coefficient (D _i ) of these ions for the intact and fractured rock samples has been evaluated and is found to be ranging from 0.76·10^–14 to 6.25·10^–14 m²/s and 0.67·10^–9 to 84.6·10^–9 m²/s, respectively. The study reveals that despite the sorbing nature of Cs and Sr, these ions diffuse comparatively faster than the iodide ion (non-sorbing) through the intact and fractured rock mass. The results have been validated vis-à-vis those reported in the literature and a good agreement has been noticed. This paper deals with details of the testing methodology developed to assess the radionuclide migration as well as the rock mass suitability for safe disposal of the radioactive waste.     

Synthesis and application of a macroporous silica-based polymeric octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide composite for the chromatographic separation of high level liquid waste

Summary The Minor Actinides Recovery from HLW by Extraction Chromatography (MAREC) process was used mainly for the separation of minor actinides (MAs) and some specific fission products (FPs) from highly active liquid waste (HLW) by the composite CMPO/SiO₂-P of the macroporous silica based polymeric octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide (CMPO) and others. In this study a cascade of chromatographic separation was performed on a 3.0M HNO₃ solution containing 5.0 ^. 10^-3M of 13 elements, at 323 K. The cascade consisted of three columns the first and second ones were packed with CMPO/SiO₂-P and the third with SiO₂-P particles. The first column was employed to prepare various eluents containing saturated CMPO. The second column was used for separation into groups. The CMPO of CMPO/SiO₂-P was recovered from the effluent by the third column and a CMPO-free effluent containing minor actinides was obtained. The elements contained in the simulated HLW of 3.0M HNO₃ were separated into (1) a non-adsorption group (Sr, Cs, and Ru etc.), (2) a MA-hRE (heavy rare earth)-Mo-Zr group, and (3) a lRE (light rare earth) group by eluting with 3.0M HNO₃, 0.05M DTPA (diethylenetriaminepentaacetic acid) (pH 2.0) and HNO₃ (pH 3.5), respectively. The resultant MA-hRE-Mo-Zr mixture containing minor actinides was then separated into the groups (1) Pd-Ru, (2) MA-hRE, and (3) Mo-Zr by utilizing 3.0M HNO₃, distilled water, and 0.05M DTPA (pH 2.0) as eluents. More than 92% of CMPO in the MA-hRE containing effluent was adsorbed by SiO₂-P particles. The effectivity and technical feasibility of MAREC process were demonstrated.     

Sorption behavior of Np (IV), Np (V) and Am (III) in the disturbed zone between engineered and natural barriers

In order to be more confident of the performance assessment of high-level radioactive waste disposal, radionuclide transport must be investigated in more detail in the disturbed host rock region adjacent to the engineered barriers where disturbance has been introduced during the construction and waste-emplacement period. Geochemical, hydrological, and rock-mechanical properties should be quite different from those of undisturbed host rock. We have to elucidate the effect of bentonite intrusion into intersecting fractures from the standpoint of radionuclide confinement. In the present work, sorption distribution ratios (Kd's) of Np and Am are measured experimentally for various values or redox potential (Eh) in a simulated rock fracture filled with bentonite. The Kd of Am is approximately 6.5×10³ ml/g and found to be insensitive to the redox potential. Under anaerobic conditions, the Kd of Np is approximately 6×10⁴ ml/g. Under aerobic conditions, Kd is as small as 30 to 100 ml/g. This is the first report to measure the sorption behavior of Np and Am in a simulated rock fracture filled with bentonite (namely, in a disturbed zone) under pH, Eh and ionic strength control. We aan make use of these Kd data for numerically evaluating the mass transfer from bentonite filled fractures into the water-flowing fracture network¹.     

Diffusion of iodide, cesium and strontium in charnockite rock mass

Experiments have been conducted to study the diffusion of iodide, cesium and strontium ions through intact and fractured rock samples of charnockite rock formation present at Kalpakkam, India. The diffusion coefficient (D _i ) of these ions for the intact and fractured rock samples has been evaluated and is found to be ranging from 0.76·10^–14 to 6.25·10^–14 m²/s and 0.67·10^–9 to 84.6·10^–9 m²/s, respectively. The study reveals that despite the sorbing nature of Cs and Sr, these ions diffuse comparatively faster than the iodide ion (non-sorbing) through the intact and fractured rock mass. The results have been validated vis-à-vis those reported in the literature and a good agreement has been noticed. This paper deals with details of the testing methodology developed to assess the radionuclide migration as well as the rock mass suitability for safe disposal of the radioactive waste.     

Sorption speciation of lanthanides/actinides on minerals by TRLFS, EXAFS and DFT studies: a review

Lanthanides/actinides sorption speciation on minerals and oxides by means of time resolved laser fluorescence spectroscopy (TRLFS), extended X-ray absorption fine structure spectroscopy (EXAFS) and density functional theory (DFT) is reviewed in the field of nuclear disposal safety research. The theoretical aspects of the methods are concisely presented. Examples of recent research results of lanthanide/actinide speciation and local atomic structures using TRLFS, EXAFS and DFT are discussed. The interaction of lanthanides/actinides with oxides and minerals as well as their uptake are also of common interest in radionuclide chemistry. Especially the sorption and inclusion of radionuclides into several minerals lead to an improvement in knowledge of minor components in solids. In the solid-liquid interface, the speciation and local atomic structures of Eu(III), Cm(III), U(VI), and Np(IV/VI) in several natural and synthetic minerals and oxides are also reviewed and discussed. The review is important to understand the physicochemical behavior of lanthanides/actinides at a molecular level in the natural environment.     

Removal of Cs from simulated high-level waste solutions by extraction using chlorinated cobaltdicarbollide in a mixture of nitrobenzene and xylene

Efficacy of chlorinated cobaltdicarbollide in a modified diluent, 20% nitrobenzene in xylene was tested for the extraction and recovery of Cs from simulated high-level waste (HLW) solutions generated from PHWR-fuel reprocessing. Concentration of the reagent, composition of the diluent, numbers of contacts, the nature of stripping agents are some of the parameters optimized for the complete removal of Cs from such waste solutions. The above solvent extraction procedure can be applied to genuine HLW solutions for effective reduction of the dose due to Cs so that HLW can be handled in fume hoods for its characterization.     

Analysis of high-level waste for tetraphenylborate using HPLC

Summary One waste remediation process used at the Savannah River Site was the in-tank precipitation of the beta-emitting ¹³⁷Cs from high-level waste (HLW) using sodium tetraphenylborate (NaTPB) followed by processing the resulting decontaminated filtrate into grout at the Saltstone Production Facility (SPF). A simple method was developed for the monitoring of tetraphenylborate (TPB) in high-level waste (HLW) containing up to 0.38 Ci/gal of ¹³⁷Cs. Separation was achieved by extraction of the high sodium-bearing waste with acetonitrile followed by analysis using reversed-phase high performance liquid chromatography (HPLC). The sample preparation method allowed for the handling of an organic extraction layer that had 94% less acitivity than the HLW sample. The subsequent HPLC analysis of the extraction layer determined the TPB concentration in HLW waste to 0.8 mg/l with a %rsd of 8.     

Europium (III) and Uranium (VI) complexation by natural organic matter (NOM): Effect of source

For the safe long‐term storage of high‐level radioactive waste (HLW), detailed information about geo‐chemical behavior of radioactive and toxic metal ions under environmental conditions is important. Natural organic matter (NOM) can play a crucial role in the immobilization or mobilization of these metal ions due to its complexation and colloid formation tendency. In this study, the complexation of europium (as chemical homologue of trivalent actinides such as americium) and uranium (as main component of HLW) by ten humic acids (HA) from different sources and Suwannee NOM river extract has been analyzed. Capillary electrophoresis in combination with inductively coupled plasma mass spectrometry has been used for the evaluation of complex stability constants log β. In order to determine the complex stability constants a conservative single site model was used in this study. In dependence of their source and thus of NOM structure the log β values for the analyzed humic acids are in the range of 6.1–7.0 for Eu(III) and 5.2–6.4 for U(VI) (UO₂²⁺), respectively. In contrast to the results for HA the used Suwannee river NOM reveals log β values in the range of nearly two orders of magnitude lower (4.6 for Eu³⁺ and 4.5 for UO₂²⁺) under the geochemical conditions applied in this study.     

Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO₃) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.     

A facile and novel process for the preparation of <Superscript>147</Superscript>Pm source

Aiming the selective recovery of palladium from high level radioactive liquid waste (HLW), a chelating thiamide type sorbent, CWP–TU, was prepared by the modification of Japanese cedar wood powder (CWP). Convection oven and microwave heating were separately used for modification purpose and found that microwave heating is more effective over oven heating. CWP–TU was extensively studied for the adsorption of Pd(II) from nitric acid medium. The batch test showed that nitric acid concentration of 3 M or higher is favorable for Pd(II) loading. Consistent adsorption of Pd(II) under gamma irradiation condition demonstrated the feasibility of using CWP–TU in real HLW. Also, Pd(II) only adsorption from simulated HLW solution verified the palladium only selectivity of the sorbent as well as the lack of influence of coexisting metal ions on its affinity toward Pd(II). CWP–TU holds maximum Pd(II) loading capacities of 0.98 mol/kg at 30 °C and 1.04 mol/kg under gamma irradiation. A comparative study using some ion exchange resins revealed that the resins are either ineffective in nitrate medium or lack stability under irradiation.     

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.     

Drop calorimetric measurements on a versatile monazite phase loaded with simulated radioactive waste

Monazite is one of the candidate ceramic matrices for the immobilization of high level radioactive waste (HLW) from the reprocessing of spent nuclear fuel. The monazite phase, Ce_0.8Ca_0.2PO₄, can accommodate cations of different valences due to the mixed valence state (+3 and +4) of Ce in this compound, by facilitating the oxidation and reduction of the Ce³⁺ and Ce⁴⁺ as required by the in-coming cation. This will assist in accommodating HLW of different compositions in the monazite crystal structure even if the average valence of the HLW elements is other than 3. Therefore, the monazite phase, Ce_0.8Ca_0.2PO₄, can be a versatile host for the immobilization of HLW. The enthalpy increment and heat capacity of this versatile monazite phase and a simulated waste form based on it with 20 mass% HLW oxides were measured by drop calorimetry in the temperature range from 373 to 873 K, and the results are compared with those measured for CePO₄.