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In the epi-cadmium neutron-induced fission of 239Pu, cumulative and independent yields of various fission products within the mass ranges of 83 to 117 and 123 to 156 have been measured by using an off-line gamma-ray spectrometric technique. The spectrum average neutron energy is 1.9 MeV. Charge distribution correction on the cumulative yields was applied to obtain their post-neutron mass yields. Mass yield distribution parameters such as the peak-to-valley (P/V) ratio, the full-width at tenth-maximum of light and heavy mass wings, the average light mass and heavy mass , and the average neutron number <ν> were obtained. The mass yields in the epi-cadmium and thermal neutron–induced fission of 239Pu were compared to examine the role of excitation energy on the P/V ratio and the nuclear structure effect. [ABSTRACT FROM AUTHOR]

Rabbit facilities are systems that pneumatically transport a sample carrier (known as a rabbit) to and from a radiation source. However, trace impurities in the polyethylene (PE) rabbit may interfere with measurements of short-lived isotopes. Activated impurities with short half-lives as 28Al interfere with the fission product γ-ray, therefore the rabbit cannot be measured immediately. Neutron activation analysis was performed on various PE samples obtained from different manufacturers and the highest quality PE was used to fabricate a high-purity PE rabbit. The impurities identified in the PE study were 27Al, 48Ca, 37Cl, 50Cr, 65Cu, 127I, 26Mg, 55Mn, and 30Si. [ABSTRACT FROM AUTHOR]

In this study, the novelly designed Sr 0.5 Zr 2 (PO 4) 3 –Ce 0.5 Sm 0.5 PO 4 composite ceramics were utilized as nuclear waste form to simultaneously immobilize simulated fission products Sr and actinide nuclides. The phase transition mechanism, thermokinetics and densification behavior of Sr 0.5 Zr 2 (PO 4) 3 –Ce 0.5 Sm 0.5 PO 4 composite ceramics based on in-situ synthesis were systemically analyzed by XRD, in-situ XPS, TG-DSC, SEM and density measurements. The results demonstrate that the formation of Ce 0.5 Sm 0.5 PO 4 solid solution phase preceded that of Sr 0.5 Zr 2 (PO 4) 3 phase, of which there is a valence change of Ce from Ce4+ to Ce3+ before 1023 K. Additionally, as for the exothermic crystallization enthalpy, change of exothermic crystallization enthalpy and activation energy, the values of Sr 0.5 Zr 2 (PO 4) 3 and Ce 0.5 Sm 0.5 PO 4 phases in composite ceramics are 286.17 – 515.61 mJ, 19.11 – 35.38 J/g, 702.45 kJ/mol and 145.21 – 232.44 mJ, 9.26 – 15.44 J/g, 450.04 kJ/mol, respectively. It is then indicated that the formation of Ce 0.5 Sm 0.5 PO 4 phase is definitely more favorable than that of Sr 0.5 Zr 2 (PO 4) 3 phase. Simultaneously, the relative densities of Sr 0.5 Zr 2 (PO 4) 3 –Ce 0.5 Sm 0.5 PO 4 composite ceramics are more than 95 % when the sintering temperature exceeds 1273 K. [ABSTRACT FROM AUTHOR]

The ceramic products derived from cesium-strontium co-doped geopolymers ((Cs, Sr)-GP) have been synthesized by microwave sintering to realize the co-immobilization of Cs and Sr. Results show that (Cs, Sr)-GP with various Cs/Sr ratios (2–4) form Sr-feldspar-pollucite composite ceramic, the formation temperatures of monoclinic Sr-feldspar decrease from 1000 ℃ (for Sr-GP) to 800 ℃ (for (Cs, Sr)-GP) with Cs addition. The ceramic product heat treated from the (4Cs, Sr)-GP at 1000 ℃ has a high crystalline and optimal leaching resistance, is considered as an ideal host to immobilization the real fission products containing Cs and Sr. [ABSTRACT FROM AUTHOR]

The spent alum sludge was employed for the removal of specific fission products including 137Cs, 85Sr, 152+154Eu, 99Mo, and 99Tc radionuclides from radioactive waste. The physicochemical features of spent alum sludge were investigated. The results illustrated that the uptake percentages were 94.2, 90.5, 87.6, 87.1, and 76.6%, correspondingly, for 99Mo, 152+154Eu, 99Tc, 85Sr, and 137Cs. The kinetic studies illustrated that the adsorption process obeyed Langmuir kinetics which is a combination of first-order and second-order expressions. The findings demonstrated that the spent alum sludge can be applied successfully as a low-cost and effective adsorbent for the treatment of radioactive waste. [ABSTRACT FROM AUTHOR]

Online monitoring of lanthanide fission product concentrations has been a great challenge in the pyrochemical reprocessing of spent nuclear fuels. Electrochemical techniques, with the advantages of accuracy and timeliness, provide a feasible strategy. In this work, the relationship between lanthanum ion (La(III)) concentration and reaction current was investigated by various electrochemical methods in LiCl–KCl eutectic salt. Tafel test results show that there is a linear relationship between the exchange current density (i0) and ion concentration (i0 = 0.06618x + 0.04409). In addition, according to the comparative analysis of electrochemical test results, it is found that anodic stripping voltammetry is a potential ion concentration detection technology. The peak current (ipc) and ion concentration satisfy the relationship: ipc = 0.08503x − 0.02245. According to the equation, the calculated La(III) concentration is almost consistent with that analyzed by inductively coupled plasma atomic spectrometer (ICP-AES), which provides a plausible way for ion concentration monitoring in high-temperature molten salt. [ABSTRACT FROM AUTHOR]

This paper reports the studies carried out on the development of a process for the recovery of 1,3-dioctyloxycalix[4]arene-18-crown-6 (CC-6) from the spent solvent generated during industrial scale separation of 137Cs from High Level Liquid Waste (HLLW). Initial works involved the validation of a conceptual process design and optimization of the process parameters with a simulated solvent. The optimized process was then tested with actual spent solvent to recover calix-crown-6. Such a process to recover undegraded and potentially re-usable calix-crown-6 from the spent solvent has been attempted for the first time with the aim of value recovery and reduced spent solvent management burden. [ABSTRACT FROM AUTHOR]

AbstractAn innovative design is introduced for neutron transmutation employing a proton accelerator in conjunction with a compact subcritical system. The transmutation converter comprises a spherical target enveloped by a subcritical assembly. The subcritical assembly consists of a moderator and low-enriched uranium in shell plates. The subcritical assembly has an inner radius of 10 cm and a thickness of 40 or 55 cm. The material used for the target is lead, and beryllium or beryllium oxide is used as a moderator. Low-enriched uranium in the subcritical assembly contains 5% 235U. The transmutation half-life is inversely proportional to the integral of epithermal 99Tc capture rates. The MCNP6 simulation demonstrates that the transmutation half-life is less than 1 year when exposed to 1-GeV protons at 5 mA. Additionally, it is notable that this half-life can be further reduced with increased proton energies and currents. Previous studies have reported that the 99Tc transmutation half-life using fast reactors and an accelerator-driven system ranges from tens to hundred years; this design concept represents a substantial advancement to previous research efforts. [ABSTRACT FROM AUTHOR]

True ternary fission, the fission of a nucleus into three fragments of nearly equal mass, is an elusive and poorly known process influenced by shell effects. An increase in the probability of this process with respect to binary fission, which is very low in spontaneous and neutron-induced fission, has been envisaged. Heavy-ion-induced reactions are adopted due to the possibility of an increase in the fissility parameter and the excitation energy of the compound nuclei. Nuclei with mass number around A = 250, accessible in heavy-ion-induced reactions, are favorable and should be investigated. It is still debated if the process takes place in a single step, direct ternary fission, or in a two step, sequential ternary fission. The purpose of this work is to define experimental conditions and observables that allow the disentangling of the products from the direct and sequential ternary fission, as well as from the usual most probable binary fission. This step is essential for gaining insights into the ternary fission dynamics and the binary to ternary fission competition. The method proposed here is for simulating the kinematics of the ternary and binary fission processes to compute the energy distributions and angular correlations of direct and sequential ternary fission products, as well as those of binary fission. The reaction taken as a benchmark is 40Ar + 208Pb at 230 MeV and is supposed to form the 248Fm* compound nucleus. The simulation results have been filtered by considering the response function of a multi-coincidence detection system virtually constructed using the Geant4 simulation toolkit. The simulations support the possibility of separating the products of different multimodal fission decays with the proposed setup that consequently represents an effective tool to obtain insights into ternary fission from the observables selected. [ABSTRACT FROM AUTHOR]

This paper focuses on the highly radioactive, long-lasting nuclear waste produced by the currently operating fission reactors and on the sensitive issue of spent fuel reprocessing. Also included is a short description of the fission process and a detailed analysis of the more hazardous radioisotopes produced either by secondary reactions occurring in the nuclear installations or by decay of the fission fragments. The review provides an overview of the strategies presently adopted to minimize the harmfulness of the nuclear waste to be disposed, with a focus on the development and implementation of methodologies for the spent fuel treatments. The partitioning-conditioning and partitioning-transmutation options are analyzed as possible solutions to decrease the presence of long-lived highly radioactive isotopes. Also discussed are the chemical/physical approaches proposed for the recycling of the spent fuel and for the reusing of some technologically relevant isotopes in industrial and pharmaceutical areas. A brief indication is given of the opportunities offered by innovative types of reactors and/or of new fuel cycles to solve the issues presently associated with radioactive waste. [ABSTRACT FROM AUTHOR]

65Zn, a fission product found in cooling water reactors, poses significant environmental risks due to its toxicity. This study explores the use of Ca–Fe2+ phosphate (SB1) and Ca–Fe3+ phosphate (SB2) as sorbents for Zn(II), prepared via the wet chemical method. The Zn(II) solution, spiked with 65Zn radionuclides, was analyzed radiometrically. Optimal sorption conditions were determined to be pH 4.5, a contact time of 24 h, and a sorbate volume to sorbent mass ratio of 1:10 at 20 °C. The Langmuir isotherm model best fit the adsorption data, indicating monolayer adsorption capacities of 0.574 mmol g−1 for SB1 and 0.621 mmol g−1 for SB2. Sorption kinetics followed a quasi-nth-order model. Furthermore, 0.1 M FeCl3 effectively desorbed 99% of Zn(II) from both sorbents. The sorption process was found to be spontaneous and endothermic. These findings suggest that SB1 and SB2 have potential applications in recycling Zn(II) from the black mass of expired batteries. [ABSTRACT FROM AUTHOR]

This study provides and investigates the fabrication of graphene oxide (GO) sheets decorated with chitosan nanoparticles (CSNP) through the hybridizing of GO and CS, by the addition of tetraethyl orthosilicate (TEOS) as a cross-linker agent. The fabricated GO-CSNP composite was characterized using several advanced techniques. Furthermore, various parameters affect the sorption of Sr(II), such as contact time, pH, initial concentration, dosage, temperature, and coexisting ions. The experimental results were in accordance with the pseudo-second-order reaction. The interaction mechanism between Sr(II) and GO-CSNP composite was accurately described by the Langmuir isotherm model with a maximum adsorption capacity of 473.93 mg/g. The GO-CSNP composite demonstrated exceptional selectivity for the sorption of Sr(II) over Y(III) at a high concentration ratio of 10:1 for Sr2+ to Y3+, respectively. Furthermore, the GO-CSNP adsorbent demonstrated considerable potential as a highly effective sorbent for the adsorption of Sr(II), Mo(VI), Cd(II), and Cs(I) ions. The results revealed that the prepared composite was effectively capable of removing various fission products. [ABSTRACT FROM AUTHOR]

This paper focuses on the highly radioactive, long-lasting nuclear waste produced by the currently operating fission reactors and on the sensitive issue of spent fuel reprocessing. Also included is a short description of the fission process and a detailed analysis of the more hazardous radioisotopes produced either by secondary reactions occurring in the nuclear installations or by decay of the fission fragments. The review provides an overview of the strategies presently adopted to minimize the harmfulness of the nuclear waste to be disposed, with a focus on the development and implementation of methodologies for the spent fuel treatments. The partitioning-conditioning and partitioning-transmutation options are analyzed as possible solutions to decrease the presence of long-lived highly radioactive isotopes. Also discussed are the chemical/physical approaches proposed for the recycling of the spent fuel and for the reusing of some technologically relevant isotopes in industrial and pharmaceutical areas. A brief indication is given of the opportunities offered by innovative types of reactors and/or of new fuel cycles to solve the issues presently associated with radioactive waste.

Experimental and theoretical studies of β electrons (electrons emitted in β-decay transitions) and their β-electron spectra have recently experienced a rapid expansion. These β spectral shapes have been used to study total β spectra of fission-product nuclei in the quest for explanation of the reactor-flux anomalies, and individual β transitions in search for β spectral shapes sensitive to the effective value of the weak axial coupling gA. In the former case the TAGS (total absorption gamma-ray spectroscopy) can be efficiently used to measure the total β spectral shapes and in the latter case dedicated measurements of the involved forbidden nonunique β transitions have been deployed. The fourth-forbidden nonunique decay transitions 113Cd(1/2+g.s → 113In(9/2+g.s and 115In(9/2+g.s 115Sn(1/2+g.s represent theoretically and experimentally much-studied cases where the total β spectra consist of these single transitions. In these particular cases the TAGS method could be used to assess the effective value of gA. In the present work we have identified five more interesting cases where a total β spectrum consists of a single transition. These spectra correspond to second-forbidden nonunique transitions and are gA and/or sNME dependent, where sNME denotes the so-called small relativistic vector nuclear matrix element. These studies have been performed using the nuclear shell model with well established effective Hamiltonians. With this we target to β transitions that would potentially be of high interest for the TAGS and present and future dedicated β -spectrum experiments. [ABSTRACT FROM AUTHOR]

Post-neutron mass yield distribution in the epi-cadmium neutron-induced fission of 233U has been carried out by measuring the cumulative yields of various fission products within the mass ranges of 77 to 117 and 123 to 153 using an off-line gamma-ray spectrometric technique. Independent yields of a few fission products were also measured by using the same technique. Charge distribution correction has been applied on cumulative yields to obtain the post-neutron mass yields. The mass yield distribution parameters such as full-width at tenth-maximum of light and heavy mass wings, the average light mass and heavy mass , and the average number of neutrons <ν> were obtained. The spectrum average neutron energy is 1.9 MeV. Thus, the role of excitation energy on the nuclear structure effect was examined by comparing the mass yield data in between the epi-cadmium and thermal neutron–induced fission of 233U. [ABSTRACT FROM AUTHOR]

Metallic alloy fuels from fast reactors will be reprocessed by a non‐aqueous electrochemical technique known as electrorefining. Electrorefining of spent metal fuel results in the accumulation of heat‐generating fission products, especially 137Cs in the eutectic salt. These fission products need to be removed from the salt so as to reduce the decay heat load and contamination. The melt crystallization technique is a simple separation process being pursued for the separation of fission products. This is a single‐step process that utilizes the difference in solubility of fission products in the solid and liquid phase. This process is also less energy intensive. Crystallization involves the purification of a substance from a liquid mixture by solidification of the desired component. Since separation is based on selective solidification, which involves phase change, it is essential to evaluate the transient solidification pattern, liquid fraction distribution, solute distribution, and separation time and efficiency. Numerical modeling of the separation of CsCl from a LiCl‐KCl eutectic mixture has been carried out using Ansys Fluent (19.2). The effect of crystallizer geometry and cooling rates on the separation of CsCl has been numerically simulated. [ABSTRACT FROM AUTHOR]

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Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit. [ABSTRACT FROM AUTHOR]

As a fission product in high‐level radioactive nuclear waste, Mo has low solubility in borosilicate glass. Fe2O3 is not only a prevalent transition metal element but also a major corrosion product in high‐level radioactive nuclear waste. Against this backdrop, the effect of Fe2O3 content on the structure and chemical durability of typical molybdenum‐containing sodium borosilicate glasses for nuclear waste immobilization are studied. The results show that the samples containing more than 3.85 mol% Fe2O3, a completely homogenous amorphous glass sample is obtained. Moreover, the mechanism of the effect of Fe2O3 on the solubility of Mo is discussed in detail. In this work, a portion of Fe3+ is reduced to Fe2+ and enters into the glasses as a charge compensation ion as Fe2+O6. Concurrently, Fe3+ ions contribute to the formation of the glass networks as Fe3+O4. Iron incorporation can improve the chemical durability of the sample. [ABSTRACT FROM AUTHOR]

Sea-to-land transfer of irradiated uranium in soil samples was measured along a transect from the Cumbrian coast south of the Windscale Sellafield nuclear site in a NE direction across the Cumbrian coastal plain. Soluble uranium components were extracted by ion exchange chromatography and ratios 236U:238U and 235U:238U were measured using multi-collector ICP-MS. The method employed demonstrated the utility of 236U ratio measurements for the determination of irradiated uranium in environmental samples. The results produced for uranium were like those that have been reported for other fission product and nuclear fuel components that have been shown to be marine in origin. Measured 236U levels were highest close to the seashore and decreased rapidly as an exponential function of distance. The results indicated the absence of depleted uranium, such as might result from the deposition of aerosols generated by the testing of DU munitions at the nearby Eskmeals firing range in all but one sample. [ABSTRACT FROM AUTHOR]

To evaluate the dosimetric performance of indigenously developed 106Ru-106Rh ophthalmic plaques for treatment of eye cancer, utilizing Monte Carlo simulation, a custom-built phantom, radiochromic films, and diode detectors. 106Ru/106Rh round type eye plaques has been developed from recovered fission product 106Ru, by Bhabha Atomic Research Centre (BARC), India, to treat retinoblastoma and choroidal melanoma. The plaque is made up of three silver layers of purity 99.9%, namely, a 15.8 mm diameter and 0.9 mm thick backing plate, a 13.3 mm diameter and 0.2 mm thick silver substrate that contains 106Ru, and a 0.1 mm thick silver window. The assembly is spherically shaped with a radius of 12 mm and sealed in a controlled environment during the production process. For dosimetry of the plaques, sources with less than 20% non-uniformity in activity distribution were considered. The dosimetry was carried out using Monte Carlo simulation, a custom-built phantom, radiochromic films, and diode detector. In order to validate the accuracy of the dosimetry methodology, dosimetry parameters for a BEBIG CCA type eye plaque (Eckert & Ziegler BEBIG, Berlin, Germany), were generated and compared. The non-uniformity of the activity distribution within the plaque was determined to be within 14%. The absorbed dose rates to water on the central axis at depths of 1 mm and 2 mm were measured as 138.32 mGy/min and 106.26 mGy/min, respectively, with an extrapolated dose at zero depth of 170.70 mGy/min. For BEBIG CCA, these values were found to be 6.8%, 135.32 mGy/min, 103.74 mGy/min, and 172.50 mGy/min, respectively. Monte Carlo simulated and measured percentage depth dose values exhibit good agreement with each other. Additionally, the measured dose rate for the BARC eye plaque closely aligns with the recommended values outlined in the GEC-ESTRO guidelines. The dosimetry of the indigenously developed 106Ru-106Rh eye plaque was carried out using indigenously designed and developed eye phantom. A robust dosimetry methodology was formulated and rigorously tested against data obtained from a commercially available eye plaque. This cost-effective ophthalmic plaque can be a promising asset in the field of ocular oncology, offering the potential for highly efficient and targeted treatments while minimizing potential harm to healthy ocular structures. [ABSTRACT FROM AUTHOR]

Next‐generation nuclear reactor technologies such as the molten salt reactor utilize alkali metal fluoride salts as both fuel and coolant. In the present study, the suitability of iron phosphate glass (IPG) as a vitrification matrix for alkali metal fluoride (NaF, CaF2) and simulated fission product loaded fluoride (NdF3, CeF3, SmF3) waste has been explored. The structural change in the metal fluoride–loaded IPG has been analyzed thoroughly using Raman and fourier transform infrared (FTIR) spectroscopy. Thermal analysis showed that the stability and glass forming ability of IPG improved upon loading the same with various mixed metal fluorides. Mössbauer data and X‐ray absorption spectroscopy at Fe K‐edge explored the minute changes in the local structure. The effect of radiation emanating from radioactive wastes in the fluoride‐loaded IPG has been scrutinized via 4.5 MeV proton beam irradiation. Our study firmly establishes the applicability of IPG as suitable vitrification matrix for radioactive metal fluoride–loaded nuclear wastes. [ABSTRACT FROM AUTHOR]

90Sr is one of the most monitored fission products. New efficient methods for strontium separation are therefore sought. Over the past 20 years, crown ethers solutions in ionic liquids (ILs) have been confirmed to efficiently extract strontium from acidic solutions. Despite all the work done in this field, the extraction mechanism is still not completely clear. Depending on experimental conditions, the extraction is believed to proceed either in the form of an ion-association complex [Sr·CE·(NO3)2] or via exchange between [Sr·CE]2+ and two IL cations. This work aims to study the influence of several extraction parameters and compare them with the presumed mechanisms. [ABSTRACT FROM AUTHOR]

The radioactive fission product 90Sr has a sufficient half-life (28.8 years) to be detected long after its appearance in the environment. After its uptake into the soil-edible plant system, it enters the food chain and represents a potential source of contamination that threatens human health. Due to these facts, tracking the distribution of the artificial radionuclide 90Sr in the soil–edible plant system is a subject of intense research. The tracking of the 90Sr radionuclide distribution in the soil profile, as well as in the crops on the long-term experimental fields was carried out using beta radiation spectrometry. The radiochemical analytical method was used to analyze the 90Sr content in cultivated soil and crops. The conducted study focused on the experimental substantiation of the developed model for predicting the behavior of 90Sr in the cultivated soil–crop system. The results of using the applied radioecological model for the transfer of 90Sr from the soil to the above-ground part of crops showed a relatively good agreement with the experimentally determined values of the soil–crop transfer factor, which indicates that the used model can be successfully applied for the prediction of the behavior of 90Sr in the soil–soil solution–crop system. [ABSTRACT FROM AUTHOR]

The improved contribution function (ICF) method based on generalized perturbation theory (GPT) is applied to the compression of CINDER90 depletion library. A series of representative problems for PWR are defined with different fuel materials and operation conditions. According to the characteristics that the reactivity indicator is only meaningful in the neutron radiation calculation while the decay heat indicator is affected by both radiation and decay stage, the ICF method is adopted in the selection of candidate nuclides, then actinides and fission product nuclides of compressed system are obtained. In addition, the absorber nuclides of the compressed system are identified by analyzing the depletion process of the poison with a large absorption cross section separately. Finally, a compressed depletion library consisting of 489 nuclides is generated. The numerical results show that the compressed library maintains high precision in the calculation of k‐inf, nuclide concentration, decay heat power, and pin power. Meanwhile, the use of compressed depletion library leads to a significant reduction of computing resources. [ABSTRACT FROM AUTHOR]

Highly reactive fission product cesium may volatilize at high temperatures. PbO–B 2 O 3 –ZnO glass with low melting temperature and excellent chemical durability can be used as a potential matrix for immobilizing Cs. In this work, five PbO–B 2 O 3 –ZnO glasses with different contents of Pb were synthesized via a melt-quenching process. These specimens were characterized with XRD, Raman, DTA and viscosity measurements. The results showed that the melting temperature decreases, viscosity decreases and structural polymerization increases with increasing lead content. Cs-doped 35PbO–60B 2 O 3 –5ZnO glasses were synthesized at low temperatures. XRD results revealed the amorphous nature of the glass solidified bodies. EDS and ICP-MS studies confirmed that Cs is retained in the glass. Clear vitrified waste products containing up to 13.6 wt% Cs were obtained. The chemical durability of the samples was evaluated via PCT-A and MCC-1 method. For the glass solidified body comprising a Cs/matrix ratio of 0.136, the leaching rate of Cs was 1.6 × 10−1 g m−2 d−1 from the PCT test, which indicates that it has a good chemical stability. 35PbO–60B 2 O 3 –5ZnO glass was expected to be a suitable matrix for Cs immobilization. [ABSTRACT FROM AUTHOR]

A foreseen feature of the Molten Salt Fast Reactor is the adoption of a bubbling system for the removal of gaseous and metallic fission products (FPs). This mechanism injects helium bubbles into the core to remove FPs from the salt through floating and mass transfer mechanisms for metallic and gaseous FPs, respectively. The present work is aimed at analyzing this helium bubbling system, focusing on gaseous FPs. We investigate both operational and safety-related features in order to get information useful for the design and to assess the convenience of its adoption. Accordingly, our investigations split into two strands: (1) analyzing the characteristics of the bubbling system itself and (2) assessing the safety features of the reactor in its presence. In order to perform the above analyses, we add the capability to simulate production, transport, and mass transfer of an arbitrary number of gaseous FPs to a preexisting multiphysics solver, built with the OpenFOAM suite. In terms of operational characterization, our analyses quantify the removal efficiency through a characteristic removal time and estimate the poisoning effect of gaseous FPs. In addition, we evaluate the activity and decay heat of the removed gas, which is an aspect crucial for the design of the off-gas unit, and the effect of the bubbling system on the power versus the fuel mass flow rate curve, which is a possible control mechanism. Among our safety-related studies, we first evaluate the void coefficient, determining upper bounds on the helium flow rate in order to avoid prompt supercriticality in case of prompt loss of helium injection. The latter accidental scenario is also analyzed considering the thermal-hydraulic dynamics of the system. We also discuss another accident: complete loss of helium removal. [ABSTRACT FROM AUTHOR]

Derek Parfit argues that fission is prudentially better for you than ordinary death. But is having more fission products with good lives prudentially better for you than having just one? In this paper, we argue that it is. We argue that, if your brain is split and the halves are transplanted into two recipients (who both have good lives), then it is prudentially better for you if both transplants succeed than if only one of them does (other things being equal). This upshot rules out, among other things, that the prudential value of standing in the relation that matters in survival to multiple people is equal to their average well-being. [ABSTRACT FROM AUTHOR]

Since xenon (Xe) production is always an unavoidable part of the fission products in fuel pellets, the challenges of its presence have always been the subject of many papers. In line with these goals, in the present paper, the effect of the presence of grain boundaries (GBs) with misorientations (θ G B ) of 36 ° , 39 ° , 52 ° , 53 ° , 71 ° , and 129 ° in 10%Xe-UO2 (uranium dioxide in which 10% of its uranium atoms have been replaced by xenon atoms) on the behavior of xenon (diffusion, nucleation, and formation of clusters) was investigated by performing molecular dynamics (MD) simulations. The results showed that xenon atoms aggregate in the GB with misorientations of 36 ° and 53 ° and form larger clusters relative to other GBs. This was interpreted due to the low formation energy of these two GBs in comparison with other misorientations. A decrease in the number of xenon atoms was also observed at a slight distance from these two GB regions, indicating their sink efficiency. The calculation of diffusion coefficients also indicated that the presence of these two GBs increases the coefficients (xenon, oxygen, and uranium). All of these demonstrate the effective role of θ G B = 36 ° and θ G B = 53 ° in swelling. To reduce the destructive effect of xenon atoms on the fuel pellet, a conceptual design in the form of nanoporous was proposed. The results of the MD simulation of such a design showed that the presence of nano-porosity significantly reduces xenon clusters. [ABSTRACT FROM AUTHOR]

Since xenon (Xe) production is always an unavoidable part of the fission products in fuel pellets, the challenges of its presence have always been the subject of many papers. In line with these goals, in the present paper, the effect of the presence of grain boundaries (GBs) with misorientations (θ G B ) of 36 ° , 39 ° , 52 ° , 53 ° , 71 ° , and 129 ° in 10%Xe-UO2 (uranium dioxide in which 10% of its uranium atoms have been replaced by xenon atoms) on the behavior of xenon (diffusion, nucleation, and formation of clusters) was investigated by performing molecular dynamics (MD) simulations. The results showed that xenon atoms aggregate in the GB with misorientations of 36 ° and 53 ° and form larger clusters relative to other GBs. This was interpreted due to the low formation energy of these two GBs in comparison with other misorientations. A decrease in the number of xenon atoms was also observed at a slight distance from these two GB regions, indicating their sink efficiency. The calculation of diffusion coefficients also indicated that the presence of these two GBs increases the coefficients (xenon, oxygen, and uranium). All of these demonstrate the effective role of θ G B = 36 ° and θ G B = 53 ° in swelling. To reduce the destructive effect of xenon atoms on the fuel pellet, a conceptual design in the form of nanoporous was proposed. The results of the MD simulation of such a design showed that the presence of nano-porosity significantly reduces xenon clusters. [ABSTRACT FROM AUTHOR]

The efficacy of filters to trap volatile radiotoxic nuclear fission products depends on the thermodynamic stability of these species within the filter material. Using atomic scale modeling based on density functional theory together with a dispersion correction, we predict the structures and energies of volatile fission product atoms and molecules trapped by a crystalline organic electride Cs+(15C5)2•e−. Endothermic encapsulation energies indicate that Kr and Xe are not captured by this electride. Conversely, encapsulation is very strong for Br, I, and Te, with respect to atoms and dimers as reference states, leading to the formation of trapped Br−, I−, and Te− ions. While both Rb and Cs are encapsulated exothermically (without significant charge transfer), their encapsulation is markedly weaker than that calculated for Br, I, and Te. Encapsulation of homonuclear dimers (Br2, I2, and Te2) as anionic molecular species is thermodynamically favorable, though they will disproportionate if sufficient encapsulation sites are available. Conversely, encapsulation of heteronuclear dimers (Rb–Br, Rb–I, Cs–Br, and Cs–I) is unfavorable with respect to their bulk solids as reference states. [ABSTRACT FROM AUTHOR]

The efficacy of filters to trap volatile radiotoxic nuclear fission products depends on the thermodynamic stability of these species within the filter material. Using atomic scale modeling based on density functional theory together with a dispersion correction, we predict the structures and energies of volatile fission product atoms and molecules trapped by a crystalline organic electride Cs+(15C5)2•e−. Endothermic encapsulation energies indicate that Kr and Xe are not captured by this electride. Conversely, encapsulation is very strong for Br, I, and Te, with respect to atoms and dimers as reference states, leading to the formation of trapped Br−, I−, and Te− ions. While both Rb and Cs are encapsulated exothermically (without significant charge transfer), their encapsulation is markedly weaker than that calculated for Br, I, and Te. Encapsulation of homonuclear dimers (Br2, I2, and Te2) as anionic molecular species is thermodynamically favorable, though they will disproportionate if sufficient encapsulation sites are available. Conversely, encapsulation of heteronuclear dimers (Rb–Br, Rb–I, Cs–Br, and Cs–I) is unfavorable with respect to their bulk solids as reference states. [ABSTRACT FROM AUTHOR]

In response to the low separation efficiency of actinide nuclide (An) and fission product (FP) in the separation process of high‐level radioactive waste, as well as the large amount of secondary waste. A novel Ln(1‐x)SrxZrO(3.5‐0.5x) (0 ≤ x ≤1) multiphase ceramic waste form (Nd3+ simulating An3+), which can cure both An and Sr simultaneously, was in‐situ synthesized by sol‐gel spray pyrolysis method. The series of multiphase ceramics are composed of cubic pyrochlore phase Nd2Zr2O7 (NZO) and orthorhombic perovskite phase SrZrO3 (SZO), without any accompanying impurities. The two phases exhibit excellent compatibility, and their content can change regularly with the change of x. Nd and Sr occupy their most stable crystal lattice positions, with low transposition doping levels. The density of the composite ceramics is high, and the measured density is as high as 88 % of the theoretical density. The above experimental results confirm that the new multiphase ceramics can cure An and Sr simultaneously and separately, showing strong adaptability to waste components and high chemical stability. [ABSTRACT FROM AUTHOR]

In this work, the separation of Co(II) ions from some fission products using zirconium molybdo silicate (ZrMoSi) sorbent was studied. ZrMoSi sorbent was prepared by the co-precipitation method and characterized using different analytical tools such as FT-IR, SEM, XRF, XRD, TGA, and DTA. ZrMoSi sorbent was found to have the molecular formula ZrO2.MoO3(0.1).SiO2(10.2).4.3 H2O. The sorption behavior of Co(II) involves the effect of shaking time, pH, initial Co(II) concentrations, desorption, and recycling. The sorption data is dependent on pH and ZrMoSi has high separation factors for Co(II) from Cs(I) and Sr(II). Reaction kinetics follow the pseudo-2nd-order model with an equilibrium time of 60 min and sorption isotherms are more applicable to a Langmuir isotherm. Desorption of Co(II) from the loaded sorbent was studied using different eluents and the best eluant is HCl (93.39%). The recycling results of Co(II) from aqueous solutions are excellent and revealed that ZrMoSi sorbent can be used as a promising sorbent to remove Co(II) from liquid waste. [ABSTRACT FROM AUTHOR]

The Saba thyroid shield is a new X-ray dose reduction tool for radiosensitive organs that can reduce the X-ray dose without degrading the image quality. This study investigated its effectiveness in reducing the X-ray dose on the thyroid surface during diagnostic X-ray. This study was conducted in the lab of the Medical Physics Department at Umm-Alqura University. The experiments were performed in the year 2022. A mixed fission product phantom and a solid-state detector were used to measure the radiation dose on the thyroid surface during diagnostic X-ray using a tube voltage within 40–120 kVp. The effectiveness of the Saba thyroid shield was examined in two settings: in-field and out-of-field. Measurements were conducted for these two settings without and with the Saba thyroid shield, and the attenuation percentages were computed. Analysis of covariance was used to test the tool's effectiveness in X-ray dose reduction. The average in- and out-of-field dose attenuation percentages using the Saba thyroid shield were 80.12 ± 0.12% and 81.13 ± 0.18%, respectively. A significant difference was shown between the measurements done without and with the Saba thyroid shield (p = 0.000) for both the in-field and out-of-field doses. The in-field and out-of-field dose reduction levels using the shield were approximately the same. This study demonstrated that the newly developed Saba thyroid shield made of copper and bismuth effectively reduces both the in-field and out-of-field radiation doses. [ABSTRACT FROM AUTHOR]

To minimize the use of organic solvents and achieve a selective adsorption separation of minor actinides (MA(III)) and Ln(III), adsorbents consisting of SiO2–P particles impregnated with hexaoctylnitrilotriacetamide (HONTA) were prepared. Batch experiments were performed to evaluate the adsorption properties of the HONTA/SiO2–P adsorbent for MA(III) and simulated fission products in HNO3 solution. The uniform distribution of Eu(III) on the adsorbent was confirmed via particle-induced X-ray emission analysis, and X-ray photoelectron spectroscopy revealed that EuNO3 was the adsorbed species. This study provides a simplified process for the separation of MA(III) using a HONTA/SiO2–P-packed column. [ABSTRACT FROM AUTHOR]

A rapid method to separate Ta, Po, Au, Pt, and W from a sample containing mixed fission and activation products was developed using three commercially available extraction chromatography resins stacked in series. When the separation was tested using all the target activation products, even those with short half-lives were measurable. Combining multiple extraction chromatography resin cartridges in tandem allows for multiple short-lived activation products to be separated from one sample addition without the complication of multiple steps. [ABSTRACT FROM AUTHOR]

Mastery with regard to the evolution of the chemical form of fission products in the primary circuit is essential for the development and improvement of models, which can predict the release of fission products into the primary loop under normal or accidental operating conditions and analyze the consequences of severe accidents. Here, the chemical forms of fission products in the primary loop of the HTR-PM are predicted by the thermodynamic database method. Furthermore, the microscopic mechanisms of the reactions between radionuclides Cs and I with steam are systematically investigated using density functional theory. The predominant reaction products of Cs + H2O and I + H2O include CsO, CsOH, IO, IOH, and HI. The bond evolution properties are analyzed using the electron localization function, quantum theory of atoms in molecules, bond order density, and mayer bond order. The reaction rate constants are determined utilizing the variational transition state theory. Within the temperature range of the HTR-PM primary circuit (250–750 °C), the reaction rate constant of OIH2 is effectively zero, and the reaction rate constant of HI is higher than that of IOH. The results of the above microscopic calculations are consistent with the macroscopic thermodynamic calculations, and these data serve as valuable ESI for the experiments. [ABSTRACT FROM AUTHOR]

Experiments on high-temperature reprocessing of VVER-1000 SNF with a burnout depth of 47 GW day/t U were carried out in order to determine the dependence of the extent of semi-volatile fission product removal on the time of SNF reprocessing in an oxidizing atmosphere for subsequent optimization of operating modes of this process. It was found that, during high-temperature reprocessing at 1200°С for 8 h, more than 30% of Mo, 90% of Cs, and 100% of Ru and Tc are removed. [ABSTRACT FROM AUTHOR]

The removal of rare earth fission products and thorium from LiCl–KCl system is of great significance to purification and recovery LiCl–KCl during pyroprocessing flow. The removal efficiency of rare earth elements or thorium was investigated by adding precipitant Li3PO4 in LiCl–KCl molten salt. The characteristics of precipitation products were determined by XRD, SEM, and EDS analysis and the removal ratios of rare earth elements were compared under different conditions. The removal ratio of Th and rare earth elements (except Eu) can reach 95% when adding a slightly excess of Li3PO4 at 823 K. About 50% LiCl–KCl was entrained after precipitation reaction in mixture salt containing 20wt%ThF4, which showed the potential challenges to achieve the removal of Th and rare earth and the recovery of LiCl–KCl salt in the actual process. It is expected that the combination of precipitation and distillation method can improve to efficiently recover the LiCl–KCl mixed salt. [ABSTRACT FROM AUTHOR]

The study investigated the sorption behaviour of 99Tc and 125I on bottom sediments from the Japan Sea, Khanka Lake and commercially available sapropel. It was determined that organic matter can have an impact on the iodine sorption process. After a 20-day period, sapropel reached an equilibrium state with a sorption value of up to 100%. In real systems of bottom sediment–water in the Far East region, iodine sorption was found to be superior on marine bottom sediments with a sorption value reaching 85% within 20 days, as opposed to freshwater sediments which only reached up to 35% within the same period. Tc(VII) in the form of pertechnetate anion exhibits no sorption under either aerobic or anaerobic conditions in the pH range of 4–11. Under aerobic conditions at pH 11 Tc(IV) is sorbed up to 15% only on sapropel; under anaerobic conditions at pH 7 the sorption reaches 25% for lake sediments, 10% for marine sediments and when the pH is 11, the sorption value of Tc by sapropel is as high as 100%. [ABSTRACT FROM AUTHOR]

As a quick and affordable experimental technique, the heavy ion irradiation method may quickly learn about the correlation between the degree of amorphization and structural modification of samples with varying ion implantation. Using the heavy ion irradiation technique, the irradiation stability of multiphase ceramic waste made of Ln 2 Zr 2 O 7 /SrZrO 3 was assessed. Here, we found that multiphase ceramic waste forms still exhibit excellent radiation resistance while achieving high capacity and weak selectivity for immobilization of actinide nuclides (simulated by lanthanide elements) and fission products 90Sr (replaced by 89Sr). The analysis of grazing incidence and Raman spectroscopy provides evidence for explaining the radiation damage mechanism caused by 84Kr15+ heavy ions. Analysis of grazing incidence offers proof of the manner and extent of radiation damage brought on by 84Kr15+ ions. The impact and mechanism of crystal structure on radiation resistance in multiphase ceramics are discussed. [ABSTRACT FROM AUTHOR]

The fission products brought about by the growth of nuclear energy is increasing, and their radioactivity will seriously jeopardize human health and pollute the environment. The recycling of radioactive materials has become a problem that needs to be solved nowadays. In this paper, we simulate the adsorption behaviors of typical fission products Cs, Sr, and Co on the surface of WS2 based on first-principle study. 3 × 3 supercell is selected by convergence test and calculate and compare the parameters of adsorption sites, adsorption energy, and charge transfer. At the microelectronic level, we analyze the interactions of WS2 with the three nuclides in detail. In addition, the effect of temperature on the adsorption rate of each nuclide on the WS2 surface is further evaluated by empirical equations. The results show that fissionable metal nuclides tend to be located at the top of the metal atoms of two-dimensional transition metal sulfides (top site of the W atom of WS2), and Co, moreover, has a much larger adsorption energy than that of Cs and Sr due to its binding to W in a form similar to covalent bonds. Moreover, under high temperature conditions, WS2 is more favorable for selecting Co and separating it from Cs and Sr. WS2 is expected to be an excellent material for the separation and recovery of radionuclide Co. [ABSTRACT FROM AUTHOR]

The paper presents the results of validating the severe accident block of the EVKLID/V2 integral code used to calculate the processes of fission product release from the oxide fuel melt and dissociation of nitride fuel observed during the destruction of the core in a fast neutron reactor with liquid metal cooling. Based on the obtained results, the uncertainty of calculating individual parameters, including the fraction of released fission products and the loss of the fuel mass during dissociation, is presented. [ABSTRACT FROM AUTHOR]

Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting inability to resolve individual pulses. Low-power reactor facilities offer the possibility to study reactor gamma-ray fields, a domain of experiments hitherto poorly explored. In this work, we present gamma-ray spectroscopy experiments performed with various detectors in two reactors: The EPFL zero-power research reactor CROCUS, and the neutron beam facility at the Ohio State University Research Reactor (OSURR). We employed inorganic scintillators (CeBr3), organic scintillators (trans-stilbene and organic glass), and high-purity germanium semiconductors (HPGe) to cover a range of typical—and new—instruments used in gamma-ray spectroscopy. The aim of this study is to provide a guideline for reactor users regarding detector performance, observed responses, and therefore available information in the reactor photon fields up to 2 MeV. The results indicate several future prospects, such as the online (at criticality) monitoring of fission products (like Xe, I, and La), dual-particle sensitive experiments, and code validation opportunities. [ABSTRACT FROM AUTHOR]