Radioactive Waste Streams Research Articles

Roadmap of Graphite Moderator and Graphite-Matrix TRISO Fuel Management Options

Most high-temperature reactors use graphite as a moderator and structural material. This includes high-temperature gas-cooled reactors with helium cooling and TRi-structural ISOtropic (TRISO) fuel particles embedded in graphite, as well as fluoride salt–cooled high-temperature reactors with clean salt coolant and TRISO fuel particles embedded in graphite and thermal spectrum molten salt reactors with a graphite moderator and fuel dissolved in the salt. The largest volume radioactive waste stream from these reactors is the irradiated graphite. We describe herein a roadmap for management of these graphite wastes that contain radioactive 14C, tritium, and other radionuclides. There may be some graphite wastes with sufficiently low radioactivity levels that can be treated as nonradioactive waste and managed like other graphite waste. Management options for the graphite include (1) direct disposal, (2) recycled back to the reactor or other nuclear applications, and (3) oxidizing the graphite with release as an effluent or underground sequestration of the carbon dioxide. Cosequestration of this carbon dioxide with carbon dioxide from industrial, biological, and cement production processes can isotopically dilute the 14C before sequestration to eliminate the possibility of exceeding individual radiation exposure limits. We also describe options for processing graphite-matrix TRISO fuel, including separating the bulk graphite to reduce the volumes of used fuel for disposal or processing to recover fissile materials. The inventories of radioactive isotopes in different carbon wastes vary by many orders of magnitude; thus, there is no single economic option for the management of all graphite waste.

The performance of iron-silicate-based biochar as a sorbent material towards 133Ba retention from radioactive liquid waste

Abstract The application of Phalaris seed peel (PSP) for the production of biochar involves the pyrolysis process in an N2 environment, resulting in the creation of a cost-effective sorbent. Two distinct modifications were conducted on the existing biochar (BC), employing just silicate (BC/SiO2) and in combination with iron-silicate (BC/SiO2/Fe). Several analytical methods were used to look at the modified biochar’s physical and chemical properties. These included scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDX), X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), thermogravimetric analysis-differential thermal analysis (TGA-DTA), and surface area analysis. Based on the initial investigations, it has been revealed that the use of silica and iron as the second modification is a more suitable approach for effectively retaining 133Ba from liquid radioactive waste streams. The investigation of sorption kinetics and isotherms was conducted to enhance our understanding of the process. The Langmuir isotherm model demonstrates the most optimal correlation for sorption, yielding a maximum sorption capacity (Q max) of 31 mg/g. Furthermore, an evaluation was performed on the BC/SiO2/Fe sorbent material by subjecting it to a mixture of simulated radioactive liquid waste, which included 133Ba, 60Co, and 137Cs.The experimental results indicate that BC/SiO2/Fe exhibits a comparatively higher sorption capacity for 133Ba when compared to 60Co and 137Cs as competing ions.

A prototype system for the removal of 137Cs from liquid radioactive waste using reverse osmosis membrane

Cesium removal from aqueous solutions of radioactive waste streams is a challenge in the field of radioactive waste management; this is due to the small atomic radii of Cs+ metal ions and their high migration ability. So, the development of a withstand system for the removal of Cs+ is crucial. In the current study, the removal of radioactive cesium from aqueous solutions using an RO-TLC membrane was studied. Two modifications were conducted; the first is to enlarge the cesium metal ion radii by interacting with mono- and dibasic acids, namely, stearic acid, tartaric acid, citric acid, and EDTA, and the second is the modification of the RO membrane pore size via reaction with the same acids. The modification was confirmed using SEM, FTIR, and EDX analysis techniques. The Cs+ and K+ rejection capacities and water permeability across the membrane at 1.5 bars were evaluated. Along with using the above-mentioned acids, the Cs+ metal ion retention index (RCs) was also obtained. It was found that employing EDTA as a chelating agent in an amount of 1.5 g/L in conjunction with the variation of feed content since it provided the highest value of RCs ~ 98% when used. Moreover, the elution of Cs+ using water, EDTA, ammonia, and HCl is also investigated. The optimal value of the eluent concentration was (0.25 M) HCl. Finally, Langmuir and Freundlich isotherm models were applied for a better understanding of the sorption process. The results of the present work more closely match the Langmuir isotherm model to determine the dominance of the chemical sorption mechanism.

STUDIES TO ESTABLISH THE SCALING FACTOR METHODOLOGY FOR LIQUID RADIOACTIVE WASTE GENERATED BY TRIGA REACTOR

During TRIGA reactor operation significant amounts of liquid radioactive waste are generated requiring appropriate management to ensure their safe disposal. Radiological characterization is an important activity needed to accomplish the waste acceptance criteria for different management steps including disposal. The purpose of this paper is to assess the activity concentrations of the (_1^3)H , (_38^90)Sr, (_28^63)Ni (difficult-to-measure), and gamma emitters (easy-to-measure) radionuclides in liquid radioactive waste generated by the TRIGA reactor operation and to establish a correlation among them to determine the scaling factors for this waste category. The radiological protocols applied for (_28^63)Ni and (_38^90)Sr, separation and purification involve extraction chromatography using selective resins, while for (_1^3)H separation the distillation technique was applied. The assessment of gamma emitters was carried out using gamma-ray spectrometry and the detection of (_1^3)H , (_38^90)Sr, and (_28^63)Ni was performed by liquid scintillation counting. The chemical yield of each experiment was determined by measuring the carrier added in each test by inductively coupled plasma optical emission spectrometry. The activity concentrations of (_38^90)Sr, and (_28^63)Ni obtained were correlated with the activity concentration of (_27^60)Co for establishing the scaling factor for the liquid radioactive waste stream. There were obtained good correlations but more experimental data are needed to determine the relevant scaling factors for these difficult-to-measure radionuclides. The scaling factors, once established, will improve and optimize the radiological characterization methodology applied for liquid radioactive waste as an easy and rapid method for assessment of concentration activities of difficult-to-measure radionuclides. Until now, no scaling factors data were reported for this liquid radioactive waste generated by TRIGA reactor operated by Institute for Nuclear Research Pitesti. The results obtained in this work represent preliminary data for establishing correlations between difficult-to-measure and easy-to-measure radionuclides and will be used for developing a scaling factor methodology. This will further be used to determine the inventory of difficult-to-measure radionuclides in different waste streams generated by TRIGA operation and decommissioning (foreseen after 2035).

Immobilization of radioactive sulphate waste simulate in polymer–cement composite based on recycled expanded polystyrene foam: evaluation of the final waste form resistance for Cs-134 and Co-60 leachability

The present study investigates to incorporate spiked sulfate solution simulate into polymer–cement composite (PCC) based on recycled expanded polystyrene foam. The main aim is to convert the waste stream into leach resistance solid forms able to slow down or even to retard the back release of hazardous radionuclides to the surrounding. Effective parameters e.g. leachant medium, temperature, radioactivity contents, leachant volumes and radionuclides speciation versus the leaching time were studied. The incremental leach rate (Rn, cm/day) and the leach index (Lx) were evaluated for the final waste form after 145 days. The experimental results revealed that the Lx for the all variances can fulfil the waste acceptance criteria for the disposal facility and are above the threshold value of 6. Moreover, leach rate percentages for Cs-134 and Co-60 were not exceeding 2%. The acquired data, based on lab leaching experiments, can recommend the developed PCC under consideration for solidification of radioactive sulphate waste stream safely.

Competitive Reduction of Pertechnetate and Chromate in Radioactive Waste Streams via Zero-Valent Iron

Competitive Reduction of Pertechnetate and Chromate in Radioactive Waste Streams via Zero-Valent Iron

Ниобий в проблеме обеспечения долговременной радиационной безопасности

The article provides a general overview discussing niobium and partially zirconium isotopes and their behavior in experimental and natural environments. In terms of weight, zirconium and niobium isotopes account for a quite tangible fraction of radioactive waste streams generated by nuclear industry. As it comes to natural environments, the behavior of substances containing these isotopes has been poorly studied even over the initial period of their one-time intake. Moreover, their chronic long-term intake over hundreds and thousands of years hasn’t been basically studied at all, which raises certain difficulties when demonstrating the long-term safety of radioactive waste disposal. This overview also evaluates the experimental studies focused on niobium behavior in natural environments, as well as some individual hypotheses about its potential behavior. The paper evidences that new experimental studies should be launched to support the forecasts regarding niobium isotope impacts on various types of biota, including human.

Elaboration and characterization of molybdenum titanium tungsto-phosphate towards the decontamination of radioactive liquid waste from 137 Cs and 85Sr

The crystalline phase of molybdenum titanium tungsto-phosphate (MoTiWPO4) as an inorganic sorbent material was synthesized via the sol–gel method. The physicochemical characteristics of MoTiWPO4 were evaluated by using Fourier transform infrared (FT-IR), scanning electron microscope (SEM), energy dispersive X-ray (EDX), thermal analysis (TGA-DTA), and X-ray diffraction (XRD). MoTiWPO4 sorbent material exhibits a high chemical resistance to HNO3, HCl, and alkaline media. MoTiWPO4 has good thermal stability as it retained about 75.63% of its saturation capacity upon heating at 500 °C. The sorption studies for several metal ions revealed marked high sorption efficiency of MoTiWPO4 towards Cs+ and Sr2+ ions which reached 99% and 95%, respectively. The saturation capacity of MoTiWPO4 for Cs+ and Sr2+ is 113 and 109 mg/g, respectively. MoTiWPO4 is approved to be successfully eliminating both 137Cs and 85Sr from liquid radioactive waste streams by %eff. of 92.5 and 90.3, respectively, in the presence of competing ions from 60Co(divalent) and 152Eu (trivalent), confirming the batch experiment results for the removal of Cs+ and Sr2+ metal ions. Furthermore, the decontamination factor exceeds 13.3 in the case of 137Cs and 10.3 for 85Sr.

Adsorption behavior of trace elements of 90Sr on MnO2–ZrO2 loaded with polyacrylonitrile polymer from aqueous solutions

A MnO2–ZrO2-polyacrylonitrile (MnO2–ZrO2-PAN) composite ion exchanger was produced and its properties were examined by Fourier-transformed infrared spectroscopy, scanning electron microscopy, The BET (Brunauer, Emmett and Teller) surface area, X-Ray diffraction analysis and thermogravimetric analysis. The adsorption of Strontium (Sr) from solutions by MnO2–ZrO2-PAN composite was studied thru batch experiments. The distribution Coefficient of Sr (II) on the composite sorbent was investigated against pH, interaction time, and primary concentration ion. To study the kinetics of adsorption, Pseudo-first-order and Pseudo second-order adsorption kinetics were studied and the results revealed that adsorption kinetics better fit to the pseudo-second-order model. Three iso-temperature models, Langmuir, Freundlich, and Temkin were applied to fit the experimental results. Among those models, Langmuir revealed the most suitable one with minimum deviation. The created composite exhibited strong compatibility to the elimination of Y (III), Ni (II), Pb (II), and Co (II) from radioactive waste streams. On the other, it is evident from the data that the quantifiable extraction of Sr (II) ions from Zr (IV), Mo (VI), and La (III) is feasible. MnO2–ZrO2 Loaded with (PAN) Polymer was figured out to have high ion exchange capacity and thermal stability and selectivity for strontium.

In situ TEM study of heavy-ion irradiation-induced amorphisation and electron beam-induced recrystallisation in powellite (CaMoO4)

Molybdenum-rich radioactive waste streams in the UK will be vitrified into glass composite high-level waste (HLW) products containing powellite (CaMoO4) crystals. These materials will be exposed to self-irradiation in a geological disposal facility for thousands of years. Our heavy-ion and electron irradiation results, together with a comprehensive review of previous studies, reveal new details on the radiation tolerance of powellite. In this study, powellite crystals were exposed to 500 keV Ar2+ and 600 keV Xe2+ ion irradiations between -160 and -80 °C and were analysed in situ via transmission electron microscopy (TEM). At low temperatures (between -160 and -105 °C), the critical amorphisation dose (Dc) is higher for the Xe irradiation, most likely due to the significant defect recovery in the cascade core. Above -105 °C, the Dc is higher for the Ar irradiation. The dynamic annealing of radiation-induced defects becomes more efficient as the temperature increases; our results suggest that point defects formed during Ar irradiation are more susceptible to dynamic annealing compared with defects in the cascades generated during Xe irradiation. We also present the first evidence of 300 keV electron beam-induced recrystallisation in amorphous powellite by studying its response from room temperature to cryogenic temperatures inside the TEM using a wide range of electron fluxes (0.6–5.2 × 1016 e/cm2/s). In contrast to previous research, our new data suggest that powellite is susceptible to amorphisation by alpha recoil nuclei, and that beta irradiation may cause defect recombination.

Post-Modification of a Robust Covalent Organic Framework for Efficient Sequestration of 99 TcO4 - /ReO4.

Nuclear industry spent fuel reprocessing and some radioactive contamination sites often involve high acidity and salinity environments. Currently developed and reported sorbents in 99 TcO4 - sequestration from the nuclear waste are unstable and show low adsorption efficiency in harsh conditions. To address this issue, we developed a post-synthetic modification strategy by grafting imidazole-based ionic liquids (ILs) onto the backbone of covalent organic framework (COF) via vinyl polymerization. The resultant COF-polyILs sorbent exhibits fast adsorption kinetics (<5 min) and good sorption capacity (388 mg g-1 ) for ReO4 - (a nonradioactive surrogate of 99 TcO4 - ). Outstandingly, COF-polyILs composite shows superior ReO4 - removal even under highly acidic conditions and in the presence of excess competing ions of Hanford low-level radioactive waste stream, benefiting from the stable covalent bonds between the COF and polyILs, mass of imidazole rings, and hydrophobic pores in COF.

PH-dependent Cs leaching from nitrate bearing geopolymers: A combined experimental and model analysis

Geopolymers are considered as promising wasteforms for the immobilization of low-level radioactive waste streams containing 137Cs. This work presents an experimental study and modelling of the pH-dependent leaching of Cs from nitrate bearing geopolymers. Longer curing periods and higher curing temperatures induce the formation of nitrate bearing zeolitic phases (NO3-sodalite and NO3-cancrinite), that exhibit higher Cs/Na selectivity relative to nitrate-free zeolitic phases (zeolite A and zeolite X). Such phase formation was found to have a significant effect on the pH-dependent leaching of Cs, as well as on the leaching of other major elements in the geopolymeric system. An innovative, semi-empirical geochemical thermodynamic model was developed for the sodium aluminum silicate hydrate (NASH) solid solution, based on the pH-dependent leaching experimental data. The model succeeded to describe most of the studied system, with emphasis on their interaction with Cs.

Strontium interaction with unblended and Low-pH cementitious wasteforms: pH-dependent leaching experiments and geochemical modelling

Cementitious materials are widely-applied for the solidification and stabilization of low and intermediate level radioactive waste streams, that often contain strontium-90. Low-pH cement pastes, based on ordinary Portland cement (OPC) supplemented with pozzolanic additives, were found to have better Sr retention capabilities than unblended OPC pastes. However, precise definition of the phases and processes that control Sr retention in unblended and low-pH OPC pastes is still lacking. The present study combines experimental work with thermodynamic modelling to investigate the retention of Sr in unblended pastes compared to low-pH pastes. The unblended and low-pH pastes were prepared each in two formulations: one based on reagent grade alite (tri-calcium silicate) and the other based on OPC. The chemical and mineralogical behavior of the paste were studied by pH-dependent leaching test. The pastes characteristics were compared in order to assess the impact of: (1) the addition of amorphous silica to the pastes and (2) the presence of minor OPC phases (which are absent in the alite-based pastes) on Sr binding. The low-pH systems showed one order of magnitude lower Sr leachability than the unblended systems at the natural pH. Moreover, the leaching curves indicated that the low-pH systems bind Sr effectively at pH > 7 while for the unblended systems binding is effective only at pH > 10. These results are attributed to the difference in Sr binding to low Ca/Si C–S–H, dominant in low-pH systems, versus the high Ca/Si C–S–H, dominant in unblended systems. Additionally, it was shown that Sr binds mainly to the C–S–H phase rather than to minor phases. The leaching results were thermodynamically modelled and correlated successfully with the experimental findings. A novel modelling approach is offered for the consideration of partial pozzolanic reaction and SiO2-gel formation. The effect of these processes on strontium retention has been successfully demonstrated.

Solidification/Stabilization Technology for Radioactive Wastes Using Cement: An Appraisal.

Across the world, any activity associated with the nuclear fuel cycle such as nuclear facility operation and decommissioning that produces radioactive materials generates ultramodern civilian radioactive waste, which is quite hazardous to human health and the ecosystem. Therefore, the development of effectual and commanding management is the need of the hour to make certain the sustainability of the nuclear industries. During the management process of waste, its immobilization is one of the key activities conducted with a view to producing a durable waste form which can perform with sustainability for longer time frames. The cementation of radioactive waste is a widespread move towards its encapsulation, solidification, and finally disposal. Conventionally, Portland cement (PC) is expansively employed as an encapsulant material for storage, transportation and, more significantly, as a radiation safeguard to vigorous several radioactive waste streams. Cement solidification/stabilization (S/S) is the most widely employed treatment technique for radioactive wastes due to its superb structural strength and shielding effects. On the other hand, the eye-catching pros of cement such as the higher mechanical strength of the resulting solidified waste form, trouble-free operation and cost-effectiveness have attracted researchers to employ it most commonly for the immobilization of radionuclides. In the interest to boost the solidified waste performances, such as their mechanical properties, durability, and reduction in the leaching of radionuclides, vast attempts have been made in the past to enhance the cementation technology. Additionally, special types of cement were developed based on Portland cement to solidify these perilous radioactive wastes. The present paper reviews not only the solidification/stabilization technology of radioactive wastes using cement but also addresses the challenges that stand in the path of the design of durable cementitious waste forms for these problematical functioning wastes. In addition, the manuscript presents a review of modern cement technologies for the S/S of radioactive waste, taking into consideration the engineering attributes and chemistry of pure cement, cement incorporated with SCM, calcium sulpho-aluminate-based cement, magnesium-based cement, along with their applications in the S/S of hazardous radioactive wastes.

On the Sustainable Utilization of Geopolymers for Safe Management of Radioactive Waste: A Review

The application of geopolymers for the safe management of radioactive waste has not been implemented on a large scale, where they are tirelessly examined with the purpose of facilitating the practicality and feasibility of the actual application towards the sustainable performance of these materials. This review therefore compiles the findings of the utilization of geopolymers as sorbents for removal of radio-contaminants from aqueous waste streams and as immobilization matrices for the containment of different radioactive wastes. The investigated geopolymer base materials encompass a wide range of reactive aluminosilicate precursor sources that include natural materials, industrial wastes, and chemicals. This work introduces to the reader the scientific interest in the field of geopolymer studies, their sustainability analysis, and their application in the nuclear industry, in particular in radioactive waste treatment and immobilization. The geopolymer classification, radiation stability, and structural characterizations were summarized with special reference to the characterization of the structure alteration due to the inclusion of functional materials or radioactive wastes. The effect of the application of metakaolin-based materials, fly ash-based materials and other base materials, and their blend on radio-contaminant removal from aqueous solutions and the immobilization of different problematic radioactive waste streams were reviewed and analyzed to identify the gaps in the sustainable performance of these materials. Finally, perspectives on geopolymer sustainability are presented, and the identified gaps in sustainable application included the need to investigate new areas of application, e.g., in pretreatment and membrane separation. The reusability and the regeneration of the geopolymer sorbents/exchangers need to be addressed to reduce the material footprints of this application. Moreover, there is a need to develop durability tests and standards based on the record of the application of the geopolymers.

High strontium adsorption performance of layered zirconium phosphate intercalated with a crown ether.

Effective removal of strontium isotopes in radioactive waste streams has important implications for the environment and the sustainable development of nuclear energy. In this work, a zirconium phosphate/18-crown-ether-6 (ZrP/18C6) composite was prepared using the intercalation method by loading crown ether into zirconium phosphate. The composite was structurally and morphologically characterized by XRD, FT-IR, XPS, and SEM. The adsorption experiments of Sr2+ onto the ZrP/18C6 composite were conducted as a function of temperature, pH, Sr2+ concentration and competing ions. The results indicate ZrP/18C6 can adsorb 98.6% of Sr2+ within 30 minutes at an Sr2+ concentration of 100 mg L-1 and maintain a high removal rate with a distribution coefficient of 7 × 105 mL g-1 when Sr2+ is at a low level of 4.28 mg L-1. The ZrP/18C6 composite reached a maximum adsorption capacity of 195.74 mg g-1 at an Sr2+ concentration of 380 mg L-1, which is significantly higher than the 43.03 mg g-1 of α-ZrP. The adsorption performance of Sr2+ onto ZrP/18C6 is not significantly affected by temperature, pH and competing ions. Furthermore, the adsorption kinetics and thermodynamics were analyzed based on the adsorption data obtained in the present work. It is shown that the adsorption of Sr2+ onto ZrP/18C6 follows the pseudo-second-order model and the Langmuir monolayer model, respectively. Additionally, the adsorption mechanism of Sr2+ by ZrP/18C6 is discussed.

Leaching resistance study on immobilization of MnO2/HMSS with sorbed simulated nuclide strontium by metakaolin-based geopolymers

Radioactive waste generated in nuclear energy utilization poses great environmental risks, and the safe disposal technology has become a hot research topic in the field of nuclear environmental security. Immobilization is a common radioactive waste disposal technique, and the leaching resistance of the solidified body is the key to safe disposal. In the study, metakaolin was used as a raw material to solidify strontium, a common nuclide in radioactive waste streams, and then conducted leaching resistance tests. Using control variables, the experiments were used to investigate the effects of adsorption pretreatment prior to solidifying, solidifying with and without the addition of adsorbent on the effect of immobilization. The leaching resistance of the solidified bodies were investigated in three leaching environments: 25℃ deionized water, pH=1 sulfuric acid solution, and 5%wt magnesium sulfate solution during the 42d test cycle. The results showed that both leaching rate and cumulative leaching fraction of the 42d samples for the leaching resistance of radioactive waste immobilization under the environment of deionized water at 25℃. For the three leaching conditions, the adsorption pretreatment with (4:1)MnO2/HMSS before solidifying or the addition of (4:1)MnO2/HMSS while solidifying could effectively improve the leaching resistance of solidified body, and the former was more effective. The results of this study can provide reference for the disposal of strontium-contanining radioactive waste.

Removal of Cs-137 Radionuclide by Resorcinol-Formaldehyde Ion-Exchange Resins from Solutions Simulating Real Liquid Radioactive Waste.

The efficiency of the removal of Cs-137 radionuclides with porous and non-porous resorcinol-formaldehyde resins from alkaline solutions simulating the composition of real liquid radioactive waste (LRW) streams has been evaluated. Resins were synthesized through the polycondensation of resorcinol and formaldehyde in an alkaline medium at a molar ratio of 1.8/2.2 and a temperature of 210 °C. The Cs-137 distribution coefficients on RFRs in alkaline solutions simulating LRW were above 103 mL/g under static sorption conditions. In a model solution with pH 11, the full dynamic sorption capacity of non-porous RFR was 0.178 mmol/g. The values of the full dynamic sorption capacities of porous RFRs were 0.274 and 1.035 mmol/g for resins obtained with calcium carbonate and toluene as templates, respectively. When the sizes of RFR beads increased two-fold, the volume until 5% cesium breakthrough decreased by 20-40%. The most pronounced beneficial effect of the RFR's porosity was observed at flow rates from 25 to 50 BV/h. It was shown that the negative effect of metal cations on Cs-137 uptake increases in the following order: Na+ &lt; Mg2+ &lt; Ca2+ &lt; K+. The number of bed volumes of LRW-simulating solution decontaminated with RFRs until 5% cesium breakthrough was above 450; that is higher than the value of known commercially available analogs. The latter shows that the developed RFRs are promising for application in technological schemes of alkaline LRW processing.

In Situ Synthesis of Zeolites in Solidified Fly Ash/Cement Matrices with High Na Content: Products, Pore Structure, and Leaching Behavior

Due to its high retention capacity and well durability, fly ash-based geopolymers rich of zeolites are considered to be potential solidified materials for low- and intermediate-level radioactive waste stream. This study mainly focused on the phase transformation, pore structure, and leaching characteristic of Sr2+ and Cs+ in solidified fly ash/cement matrices under different sodium hydroxide contents or sodium nitrate concentrations after thermal curing of 28 d. The results obtained from the experiments indicated that the geopolymers prepared from pure fly ash had the potential to transform into ordered structures at high temperatures of 90°C instead of room temperature, such as Na-P1 zeolite and chabazite. With the increase in sodium hydroxide contents, more sodium nitrate participated in the reaction to form a zeolite phase in solidified fly ash/cement matrices, while the coexistence phenomenon of chabazite and Na-P1 zeolite appeared under the condition of high sodium hydroxide contents of more than 5%. Just a proper number of zeolites can effectively inhibit the leaching of radionuclides Sr2+, as too many zeolites would lead to the increase in porosity, which was harmful to the retention of radionuclides. Solidified fly ash/cement matrices at a sodium hydroxide content of 5% or at a sodium nitrate concentration of 300 g/L obtained the lowest cumulative leaching fraction of Sr2+. Moreover, sodium hydroxide increased the cumulative leaching fraction of Cs+ which reversely decreased significantly with the increase in sodium nitrate concentrations.

Precise radiocarbon determination in radioactive waste by a laser-based spectroscopic technique

The precise and accurate determination of the radionuclide inventory in radioactive waste streams, including those generated during nuclear decommissioning, is a key aspect in establishing the best-suited nuclear waste management and disposal options. Radiocarbon ([Formula: see text]) is playing a crucial role in this scenario because it is one of the so-called difficult to measure isotopes; currently, [Formula: see text] analysis requires complex systems, such as accelerator mass spectrometry (AMS) or liquid scintillation counting (LSC). AMS has an outstanding limit of detection, but only a few facilities are available worldwide; LSC, which can have similar performance, is more widespread, but sample preparation can be nontrivial. In this paper, we demonstrate that the laser-based saturated-absorption cavity ring-down (SCAR) spectroscopic technique has several distinct advantages and represents a mature and accurate alternative for [Formula: see text] content determination in nuclear waste. As a proof-of-principle experiment, we show consistent results of AMS and SCAR for samples of concrete and graphite originating from nuclear installations. In particular, we determined mole fractions of 1.312(9) F[Formula: see text] and 30.951(7) F[Formula: see text] corresponding to ∼1.5 and 36.2 parts per trillion (ppt), respectively, for two different graphite samples originating from different regions of the Adiabatic Resonance Crossing activator prototype installed on one irradiation line of an MC40 Scanditronix cyclotron. Moreover, we measure a mole fraction of 0.593(8) F[Formula: see text] ([Formula: see text] ppt) from a concrete sample originating from an external wall of the Ispra-1 nuclear research reactor currently in the decommissioning phase.