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Effect of nitric acid concentration and temperature upon the dissolution rate of MOx pellets at 30 % dissolution
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The focus of this paper is the chemistry of mixed uranium plutonium oxide (MOx,) in nitric acid. An overview of dissolution chemistry is discussed by comparing the differences in the dissolution characteristics of uranium and plutonium oxides. An overview of batch dissolution experiments, studying the dissolution chemistry of high surface area MOx...
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... experiments with 5.0 %Pu MOx pellets, each with a mass of ca. 10 g, have been used to study the effect of: nitric acid in the absence and presence of added nitrous acid at 80 C temperature nitric acid concentration. An example of the effect of temperature and nitric acid concentration upon the dissolution rate is shown in Figure 4. In contrast to the results of the powder dissolution experiments at high temperature and high nitric acid concentrations, the dissolution rate becomes less sensitive to the effect of nitric and nitrous acid concentrations. ...
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... Essentially, after overcoming the initial sluggish period, known as the induction period, the dissolution rate experiences a substantial increase. Researchers have delved into the effects of nitrogen oxides and nitric acid concentration on this dissolution process, revealing that the concentrations of nitric acid and nitrogen oxides exert a notable influence [48]. Notably, under typical dissolution conditions, the oxidation process involving nitrous acid outpaces that involving nitric acid [49], providing a mechanistic explanation for the inherently self-catalytic nature of the nitric acid dissolution process. ...
... Experiments conducted by Vollath [48] to investigate the dissolution of MOX fuel in nitric acid with varying plutonium contents yielded insightful results, as illustrated in Fig. 3. The findings distinctly reveal a substantial decrease in the solubility of MOX fuel in nitric acid when the plutonium content surpasses 35%. ...
... This observation underscores the significant impact of temperature on the dissolution of MOX fuel. Fig. 3. Solubility of MOX fuel at different plutonium contents in 5/10 mol.L-1 nitric acid solution [48]. ...
The fast neutron reactor is an internationally promising fourth-generation reactor. The main fuel for this reactor is a mixed oxide fuel, and its reprocessing is currently one of the technical challenges being tackled by various countries. One of the difficulties in the reprocessing of mixed oxide (MOX) fuel lies in the nitric acid dissolution process. The high Pu content in MOX fuel can lead to issues such as solvent radiolysis, nuclear criticality, increased insoluble residues, and slow dissolution rates during the nitric acid dissolution process. These challenges have yet to be effectively addressed. This article discusses the chemical aspects of nitric acid dissolution of MOX fuel and investigates the impact of fuel manufacturing processes, the addition of metal catalyst ions, hydrofluoric acid addition, fuel plutonium content, dissolution temperature, and ultrasonic assistance on the nitric acid dissolution of MOX fuel. A review of various countries' engineering practices related to MOX fuel dissolution is presented. Based on the research findings and experiences, a potentially feasible future industrial processing route for MOX fuel is proposed, and future research priorities are outlined.
... Considering conventional wet-reprocessing, the dissolution time of Pu in an acid chemical would not be a critical factor to determine the time of in-situ mining (Benedict et al., 1981;Carrott et al., 2012;Gupta et al., 2000;Marc et al., 2017;Mineo et al., 2012;Patil et al., 1980;Rudisill et al., 2008). The operation time for this scenario would be highly dependent on the time of drilling. ...
Spent nuclear fuel in a geological repository could be an easily accessible source of Pu for state and non-state proliferators after the level of radioactivity significantly decreases in the long term. Such concerns have been continuously raised, but little systematic analysis has been conducted to compare costs and benefits of long-term safeguards to protect Pu in spent nuclear fuel in a final repository. A system thinking model combined with game theory was developed to evaluate the costs and benefits of safeguards of the final repository. In this analysis, the safeguards system is considered as a strategic game between two competing players – a proliferation player and a safeguards player. Detailed data, such as drilling rates and costs, are carefully obtained from real world industrial applications. The costs of safeguards are predicted from the data of commercial satellite monitoring. The model is applied to two potential proliferation scenarios. One is the construction of a clandestine access tunnel using a tunnel boring machine. The other one is chemical extraction through drilling of a vertical deep borehole. The calculated costs of long-term safeguards are mapped for various decisions of the proliferation player and the safeguards player to suggest the optimized approach for the long-term safeguards of the final repository.
... Dissolution kinetics for uranium oxides in HNO 3 are well documented, with additives such as metal nitrites being reported as catalysts to the dissociation mechanism[33]. The dissolution process in HNO 3 has also been described as autocatalytic, which stems from the by-product formation of NO 2 gas and, under oxygen atmosphere, H 2 O 2[34]. For this reason, agitation of the acid solution is often avoided to promote the uptake of NO 2 gas as it forms, further catalysing the dissolution of uranium. ...
Environmental sampling to detect trace nuclear signatures is key component of international nuclear treaty enforcement. Herein, we explored rapid chemical extraction methods coordinated with measurement systems to provide faster, simpler assay of low level uranium from environmental samples. A key problem with the existing analytical method for processing environmental surface samples is the requirement for complete digestion of sample and sampling material. This is a time-consuming and labor-intensive process that limits laboratory throughput, elevates analytical costs, and increases background levels. Promising extraction methods were competitively evaluated for their potential to quickly and efficiently remove different chemical species of uranium from standard surface sampling material. A preferred combination of carbonate and peroxide solutions is shown to give rapid and complete form of uranyl compound extraction and dissolution. This simplified and accelerated extraction process is demonstrated with standard sampling material to be compatible with standard inductive coupled plasma mass spectrometry methods for uranium isotopic assay as well as rapid screening techniques such as X-ray fluorescence (XRF). Rapid extraction of the entire swipe is shown to allow efficient XRF assay of all collected material for simple, fast, nanogram-level XRF assay of the sample. The new methods have direct application in the support of nuclear safeguards treaty enforcement efforts as well as health and safety monitoring. The general approach described may have applications beyond uranium to other trace analytes of nuclear forensic interest (e.g., rare earth elements and plutonium) as well as heavy metals for environmental and industrial hygiene monitoring.
... Once the exercise and methodology to study the dissolution behaviour of UO 2 system is completed, similar approach added with necessary safety precautions would be adopted to study the dissolution behaviour of urania plutonia mixed oxide fuel in nitric acid medium. Literature only says that MOX fuel is expected to have sluggish dissolution as the plutonium content increases in it (Carrott et al., 2012;Uriarte and Rainey, 1965), but it doesn't provide any account of systematic approach about the intrinsic kinetics of MOX fuel dissolution in nitric acid. Hence if dissolution studies similar to those reported in this article are carried out for MOX system, it should be possible unravel its intrinsic dissolution kinetics. ...
Plutonium multi-recycling is one of the main aims to gradually achieve a fully closed and sustainable nuclear fuel cycle in the future, particularly once fast reactors will be deployed into the fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, development of innovative concepts are required to meet the scientific, technical and economical challenges that are involved in multi-recycling of spent nuclear fuels with high plutonium contents. Specifically, the issues inherent in the hydrometallurgical treatment of spent mixed oxide (MOX) fuels containing increasing plutonium content must be addressed through new efficient processes for fuel dissolution, plutonium separation, MOX fabrication and waste conditioning. Some of the leading concepts being developed internationally are described in this chapter.
In order to gain insights into the kinetics of dissolution of Gen(IV) MOX fuels, sintered pellets of Th0.25U0.75O2 were prepared as surrogates. The study of the kinetics of dissolution was achieved to determine the impact of the incorporation of 25 mol % of thorium in UO2 on the rate and on the mechanism of dissolution. Especially, various acidic media were used to evidence changes occurring in the rate of the limiting step and to establish specific rate laws. For most of the investigated dissolution media the dissolution reaction was found to be congruent. The homogeneity of the Th0.25U0.75O2 solid/solutions allowed both elements to be recovered at the same rate. The way of synthesis by oxalate co-precipitation, then conversion in mixed oxide, allowed the quantitative dissolution of the elements. In aerated hydrochloric acid solutions, the kinetics of the overall dissolution reaction appeared to be controlled by adsorption and desorption of protons on activated surface sites. The dissolution rate determined in aerated sulfuric acid solution was enhanced compared to hydrochloric acid solution of the same acidity. This result was interpreted in terms of formation of sulfate ions surface complexes that favor the detachment of the actinides from the mixed oxide solid solution. In nitric acid solutions, the dissolution of the Th0.25U0.75O2 pellets followed several successive steps. During this first steady state period, the kinetics of the overall dissolution reaction appeared to be controlled by the oxidation of U(IV) by HNO3 at the solid/solution interface. A rate law was established by adding a redox contribution to the proton-promoted surface contribution defined for aerated hydrochloric acid solutions. Then, an increase of the normalized dissolution rate was observed, which was attributed to the simultaneous increase of the specific surface area of the pellet and of the HNO2 concentration in solution.
