Cemented Radioactive Waste Research Articles

Study the Effect of Rice Husk Ash on Durability of Cemented Radioactive Waste Block

Study the Effect of Rice Husk Ash on Durability of Cemented Radioactive Waste Block

Photofission delayed gamma-ray measurements in a large cemented radioactive waste drum during LINAC irradiation

Nuclear materials characterization in 870 L cemented waste drums is challenging due to significant neutron and gamma ray attenuation. Among the existing non-destructive approaches, photon active interrogation is of particular interest. A LINAC accelerates electrons to energies greater than 10 MeV, leading to highly penetrating interrogating bremsstrahlung photons that can reach nuclear materials in the depth of the 870 L waste drum and allow the detection of delayed gamma rays emitted by photofission products. Because the interrogating photon and neutron fluxes are extremely high during irradiation, measurements with high purity germanium detectors are usually performed after a long irradiation time, for example by moving the detector or the waste drum. Therefore, such measurements are only sensitive to delayed gamma rays associated to photofission products having sufficient long half-life of at least tens of seconds, which cuts a valuable signal coming from fission products with a shorter half-life. In this paper, we report gamma rays measurements of very short lived photofission products using a 3 in × 3 in cylindrical LaBr3(Ce) fast scintillation detector. A depleted uranium sample was irradiated by a pulsed bremsstrahlung source with maximum energy 17.5 MeV in a macro pulsing mode (1 s irradiation, 2 s cooling). This allowed to acquire a strong delayed gamma signal within a time window of 2 s.A clear signature of nuclear materials was observed even in the depth of a concrete mock-up, with an excellent signal-to-noise ratio beyond 3.5 MeV where the background of nonnuclear activated materials is negligible. These results are compared to a semi analytical model that qualitatively agrees but underestimates the measurements by a constant factor 1.7, probably caused by a wrong beam intensity normalization. Indeed, in this work the LINAC pulse time structure (macro pulsing) was different from its nominal working point used to characterize the beam (usual pulsing mode). Since it allows direct calculations of the detector response to delayed gamma-ray photofission measurements, this model is a fast calculation alternative to time-consuming Monte Carlo simulations, in view of further studies that will follow this feasibility demonstration.

Co-Extraction of Uranium and Mercury Using Ion Exchange from Cemented Radioactive Waste Sulfuric Leachate in Iodide Media

The production of medical isotopes in Chalk River Laboratories facilities (Chalk River, ON, Canada) has resulted in a large quantity of cemented radioactive waste (CRW) containing valuable elements such as uranium. From the perspective of recovering and ultimately recycling valuable elements from CRW, the solubilization of key constituents such as uranium, mercury, and cesium has been previously investigated using H2SO4/KI. However, to achieve recycling of these elements, separation must be performed as they are co-solubilized. In this study, the extraction of uranium and mercury by chelating resin Lewatit TP260 from surrogate cemented radioactive waste (SCRW) leaching solution in sulfuric media and in the presence of iodide was investigated. Extraction of U and Hg was assessed as a function of the concentration of KI (0.12 M to 0.24 M) used during the SCRW dissolution process. Continuous experiments showed that the Lewatit TP260 functional group, aminomethylphosphonic acid, had a high affinity for U. Mercury was also extracted onto the Lewatit TP260. However, the presence of iodide in the SCRW leaching solution increased the competition between the adsorbed mercury of the stationary phase and the iodide–mercury complexes of the mobile phase. Additionally, the reusability of the resin was tested through extraction and desorption cycles. Due to the presence of trivalent cation, the capacity of Lewatit TP260 for U and Hg decreases with the number of cycles.

Pilot Scale Validation of a Chemical Process for Uranium, Cesium, and Mercury Recovery from Cemented Radioactive Wastes

The medical isotope (Mo-99) production at Chalk River Laboratory involves the dissolution of irradiated isotope targets prior to the extraction of Mo-99. This process generates a waste that is cemented in 5-gallon containers and transferred to a waste-management facility for intermediate storage. Over the past decades, a large number of five-gallon containers of cemented radioactive waste (CRW) were produced, and Canadian Nuclear Laboratories (CNL) must develop a process to convert this material to a permanent waste form. Research has been undertaken to develop an innovative method for the recovery of U, Hg, and Cs from surrogate cemented radioactive waste (SCRW). This paper presents the pilot scale validation of the operating parameters prior to the demonstration scale testing. Leaching tests at the pilot scale were conducted with 5 kg of SCRW to validate the main operating parameters and evaluate the reuse of the leaching solution. The mean solubilization yields obtained at the pilot scale were 81.9 ± 8.3% for Cs, 99.0 ± 1% for U, and 94.9 ± 4.5% for Hg. Columns with 100 g of KNiFC-PAN and 250 g of Lewatit TP214 allow for the separation of Cs and Hg from 60 L of leaching solution without U loss. Flow rates of 12.5 BV/h and 25 BV/h were suitable to achieve 99% separation of Hg and Cs, respectively. For the Hg resin, the capacity reached 23.4 mg/g, and the capacity for the Cs resin reached 0.79 mg/g. The pilot scale U extraction results showed that the U adsorption is selective, with a breakthrough at 36 BV (capacity for U of 3.70 mg/g). Uranium elution with 1 M Na2CO3 exceeded 99%, and subsequent precipitation with NaOH achieved 99% recovery. SEM data confirmed the high purity of the U solids produced as sodium di-uranate.

Обеспечение соответствия цементированных РАО в контейнерах НЗК критериям приемлемости для захоронения

In 2018–2020, work was carried out on experimental confirmation of compliance of cemented radioactive waste from blocks 1, 2 of Novovoronezh NPP-2 with the criteria of acceptability for final disposal. The requirements for the volume of technological control of Radwastes are justified. It is shown that compliance with the requirements of regulatory documents is ensured when the salt content of cemented Radwastes is up to 600 g/l. Corrective actions for deviations (the presence of boric acid in cemented Radwastes, complexing substances) are justified.

Immobilization of NPP evaporator bottom high salt-bearing liquid radioactive waste into struvite-based phosphate matrices

The high salt-bearing liquid radioactive waste (evaporator bottoms, EB) makes up the most voluminous NPP waste and needs solidification. In the paper presented, we introduce a novel formation process study of the struvite-based phosphate matrices ((K, NH4)MgPO4·6H2O) and the developed phosphate matrix compositions for the solidification of high salt-bearing solutions. The solutions simulate the EB of nuclear power plants with pressurized water reactors (NPP PWR). The effect of the EB's composition and salt content on the matrices' mechanical strength was investigated. The cesium-selective nickel-potassium ferrocyanide sorbent or 10–20% of MgO over the reaction stoichiometry, introduced at the matrix synthesis stage, allowed the production of matrices with the average 137Сs leach rate of less than 10-3 g cm-2 day-1 and the mechanical strength over 5 MPa. The matrices obtained completely satisfied the cemented radioactive waste requirements and contained up to 17–17.5 wt% of salts, which was 1.7–2.5 times higher compared to the Portland cement-based matrices.

Composite Binders for Solidification of Spent Ion-Exchange Resins

New cement compositions for solidifying mixtures of spent ion-exchange resins in sodium, nitrate, and tetraboric-acid form were developed. Depending on the composition of the cement and the ionic form of the anionite, the proposed cements make it possible to incorporate into the matrix 22–83 wt.% more ionexchange resins compared with using conventional portland cement. The matrices with heightened content of ion-exchange resins meet the requirements of cemented radioactive waste in terms of strength, water resistance, and leaching of 137Cs.

Plutonium Migration during the Leaching of Cemented Radioactive Waste Sludges

One of the most challenging components of the UK nuclear legacy is Magnox sludge, arising from the corrosion of Mg alloy-clad irradiated metallic U fuel that has been stored in high pH ponds. The sludges mainly comprise Mg hydroxide and carbonate phases, contaminated with fission products and actinides, including Pu. Cementation and deep geological disposal is one option for the long-term management of this material, but there is a need to understand how Pu may be leached from the waste, if it is exposed to groundwater. Here, we show that cemented Mg(OH)2 powder prepared with Pu(IV)aq is altered on contact with water to produce a visibly altered ‘leached zone’, which penetrates several hundred microns into the sample. In turn, this zone shows slow leaching of Pu, with long-term leaching rates between 1.8–4.4 × 10−5% of total Pu per day. Synchrotron micro-focus X-ray fluorescence mapping identified decreased Pu concentration within the ‘leached zone’. A comparison of micro-focus X-ray absorption spectroscopy (µ-XAS) spectra collected across both leached and unleached samples showed little variation, and indicated that Pu was present in a similar oxidation state and coordination environment. Fitting of the XANES spectra between single oxidation state standards and EXAFS modeling showed that Pu was present as a mixture of Pu(IV) and Pu(V). The change in Pu oxidation from the stock solution suggests that partial Pu oxidation occurred during sample ageing. Similarity in the XAS spectra from all samples, with different local chemistries, indicated that the Pu oxidation state was not perturbed by macro-scale variations in cement chemistry, surface oxidation, sample aging, or the leaching treatment. These experiments have demonstrated the potential for leaching of Pu from cementitious waste forms, and its underlying significance requires further investigation.

Magnesium-Potassium Phosphate Matrix for Immobilization of 14C

Possibility of using a low-temperature magnesium-potassium phosphate matrix to solve the problem of immobilizing the radioactive wastes containing radioactive carbon (14C) in the form of calcium carbonate was examined. The physicochemical characteristics of the compounds obtained were determined. Large values of the ultimate compression strength (22 ± 5 MPa), which satisfy the technical requirements for cemented radioactive wastes (no less than 4.9 MPa), were obtained. The minimum carryover of carbon dioxide into the atmosphere in the course of synthesis and in keeping of samples for 14 days was noted: not more than 3 wt % relative to the starting CaCO3. The leaching rate of carbonate ions from magnesium-potassium compounds by 28th day of contact with air does not exceed 10‒9 g cm‒2 day‒1, with this value for the rest of the compound components not exceeding 10‒4 g cm‒2 day‒1. Thus, it was found that the magnesium‒potassium phosphate matrix is an alternative to the cementation for solidification of radioactive wastes containing 14C.

Uranium, Cesium, and Mercury Leaching and Recovery from Cemented Radioactive Wastes in Sulfuric Acid and Iodide Media

The Canadian Nuclear Laboratories (CNL) is developing a long-term management strategy for its existing inventory of solid radioactive cemented wastes, which contain uranium, mercury, fission products, and a number of minor elements. The composition of the cemented radioactive waste poses significant impediments to the extraction and recovery of uranium using conventional technology. The goal of this research was to develop an innovative method for uranium, mercury and cesium recovery from surrogate radioactive cemented waste (SRCW). Leaching using sulfuric acid and saline media significantly improves the solubilization of the key elements from the SRCW. Increasing the NaCl concentration from 0.5 to 4 M increases the mercury solubilization from 82% to 96%. The sodium chloride forms a soluble mercury complex when mercury is present as HgO or metallic mercury but not with HgS that is found in 60 °C cured SRCW. Several leaching experiments were done using a sulfuric acid solution with KI to leach SRCW cured at 60 °C and/or aged for 30 months. Solubilization yields are above 97% for Cs and 98% for U and Hg. Leaching using sulfuric acid and KI improves the solubilization of Hg by oxidation of Hg0, as well as HgS, and form a mercury tetraiodide complex. Hg and Cs were selectively removed from the leachate prior to uranium recovery. It was found that U recovery from sulfuric leachate in iodide media using the resin Lewatit TP260 is very efficient. Considering these results, a process including effluent recirculation was applied. Improvements of solubilization due to the recycling of chemical reagents were observed during effluent recirculation.

Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina) , it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

Microbiological evaluation of the condition of cement compounds with radioactive wastes after long-term storage in near-surface repositories

Analysis of the core material taken by check drilling of a monolith of cemented radioactive waste in near-surface repositories operated for 15–45 years revealed the presence of damaged areas in the cement matrix. An association of microorganisms capable to cause degradation of silicate materials was found in cement compounds with radioactive waste. The bacterial genera, the metabolite concentrations, and the character of the biogenic acid damage of the cement matrix were determined. It was concluded that the safety level of the longterm storage of conditioned radioactive waste forms can decrease because of disturbances of the immobilization properties of the cement matrix as a result of biogenic damage of its microstructure. Biocidal additives were suggested for preventing degradation processes in the modern stages of radioactive waste cementation.

Long-term field and laboratory leaching tests of cemented radioactive wastes

Experiments with real and simulated radioactive cementitious wasteforms were set up to compare the leaching behaviour of cementitious wasteforms containing nuclear power plant operational waste in field and laboratory test conditions. Experiments revealed that the average annual 137Cs leach rate in deionised water was about thirty-five times greater compared with the measured average value for the 1st year of the field test. Cumulative leached fraction of 137Cs for 1st year (3.74%) was close to values reported in literature for similar laboratory experiments in deionised water, however more than two orders of magnitude higher than the 1st year leached fraction of 137Cs in the repository test (0.01%). Therefore, to compare field and laboratory test results, a scaling factor is required in order to account for surface to volume factor difference, multiplied by a temperature factor and a leach rate decrease coefficient related to the ground water composition.

Material characterization in cemented radioactive waste with the associated particle technique

The elemental characterization of materials constituting radioactive waste is of great importance for the management of storage and repository facilities. To complement the information brought by gamma or X-ray imaging, the performance of a fast neutron interrogation system based on the associated particle technique (APT) has been investigated by using MCNP simulations and by performing proof-of-principle experiments. APT provides a 3D localisation of the emission of fast neutron induced gamma rays, whose spectroscopic analysis allows to identify the elements present in specific volumes of interest in the waste package. Monte Carlo calculations show that it is possible to identify materials enclosed behind the thick outer envelop of a ≈1 m 3 cemented waste drum, provided the excited nuclei emit gamma rays with a sufficient energy to limit photon attenuation. Neutron attenuation and scattering are also predominant effects that reduce the sensitivity and spatial selectivity of APT, but it is still possible to localise items in the waste by neutron time-of-flight and gamma-ray spectroscopy. Experimental tests confirm that the elemental characterization is possible across thick mortar slabs.

Evaluation of the properties of cemented radioactive wastes with prolonged testing in mound type repository

Cemented wastes to be stored for a long time are investigated using physicochemical and microbiological methods of analysis. It is determined that the cement compound containing radioactive wastes conforms to the regulations pertaining to strength and radionuclide leach rate. No effects due to natural factors in the form of filtering waters and temperature differentials on the change in the properties of the cement compounds were found.

39-years Performance of Cemented Radioactive Waste in a Mound Type Repository

ABSTRACTLong-term field tests of cemented aqueous radioactive wastes in an experimental mound type surface repository were carried out at Moscow Scientific and Industrial Association “Radon” from 1965 to 2004. Aqueous radioactive wastes of different compositions containing short-lived radionuclides including 90Sr and 137Cs at concentrations from 0.34 to 1.8 MBq/L were immobilized using cementation technology. Water solution to cement ratio was 0.66, grout mixing time 10-15 minutes, and cement paste hardening time 7 days. 73 cement blocks with a volume of 0.027 m3 were disposed of for long-term tests in a simple mound type surface repository. The atmospheric precipitates, which contacted radioactive cement blocks, were collected and analyzed for the content of radionuclides. In August 2004 the experimental repository was opened, cemented blocks, underlying and covering materials were retrieved for analyses. XRD analyses showed that along with amorphous tobermorite gel the main crystal phases in cements are calcite and portlandite. Both visual inspection and radiometric analyses demonstrate that cemented blocks are in good condition and that the cement paste has retained radionuclides from the wastes. Thus after 39 years of storage in the mound type repository the cemented aqueous wastes are reliable immobilized.

The Effect of Cellulose Degradation Products on the Solubility and Sorption of Pu and Am in Alkaline-Plume-Affected and in Unaffected Boom Clay

ABSTRACTThe potential effect of water-soluble degradation products (DPs) formed from cellulose-containing cemented radioactive waste on the behaviour of Pu(IV) and Am(III) in Alkaline-Plume-Affected (APA) and in unaffected Boom clay was assessed. After solubility measurements in APA Real Clay Water, sorption experiments of Pu(IV) and Am(III) onto APA Boom clay were performed in the presence or not of DPs or pure α-isosaccharinic acid (α-ISA). Moreover, the sorption of α-ISA onto the unaffected clay was studied.

Experimental and Theoretical Investigations of Corrosion Mechanisms in Cemented Waste Forms

Evaluation of the long-term radionuclide release from cemented waste forms in contact with solutions like salt brine requires a detailed chemical analysis of the corrosion processes. The results can be integrated into a theoretical model. For this purpose, experiments were performed with simulated cemented radioactive waste forms, giving the concentration profiles of the elements involved in the corrosion process. The profiles are measured parallel to the direction of corrosion, mainly by electron microprobe analysis. The profiles are compared with profiles computed with the recently developed DIFMOD computer code. It computes both leaching of specimen transportation of the reactive ions, such as Mg/sup 2 +/ and SO/sub 4//sup 2 -/, contained in a quinary brine, into the cement product and the chemical reactions occurring in the waste form. On the basis of diffusion constants from the literature and fitted material constants, a good agreement between the calculated and measured concentration profiles has been achieved. The penetration of the corrosion front into waste forms is computed as an example of a practical application of the model.