Simulated Fission Products Research Articles

ICP-OES and NAA analysis of aerosols generated from molten iron–concrete reactions in the presence of simulated fission products

ICP-OES and NAA analysis of aerosols generated from molten iron–concrete reactions in the presence of simulated fission products

Manufacturing and characterization of simulated fission product (FP)-doped uranium–plutonium mixed oxide solid solutions (FP = Zr, Nd, Sm, and Gd)

Seven uranium–plutonium mixed oxides (MOX) comprising 10 and 20 mol% of solid simulated fission products (FPs) were prepared via powder metallurgy. Zirconium, neodymium, samarium, and gadolinium were selected as the FPs owing to their high fission yields and solubility within the uranium–plutonium fluorite structure. A laboratory-scale manufacturing process was developed to prepare solid solutions that are prerequisites for thermodynamical investigations. The microstructure and crystal structure of the specimens were analyzed. Furthermore, a mechanistic approach toward cation interdiffusion is proposed. Finally, our hybrid crystallographic model was employed to re-evaluate the ionic radii of the species constituting the FP-MOX samples.

Transition mechanism and thermokinetic analysis of Sr0.5Zr2(PO4)3–Ce0.5Sm0.5PO4 composite ceramics for nuclear waste forms based on in-situ synthesis

In this study, the novelly designed Sr0.5Zr2(PO4)3–Ce0.5Sm0.5PO4 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 Sr0.5Zr2(PO4)3–Ce0.5Sm0.5PO4 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 Ce0.5Sm0.5PO4 solid solution phase preceded that of Sr0.5Zr2(PO4)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 Sr0.5Zr2(PO4)3 and Ce0.5Sm0.5PO4 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 Ce0.5Sm0.5PO4 phase is definitely more favorable than that of Sr0.5Zr2(PO4)3 phase. Simultaneously, the relative densities of Sr0.5Zr2(PO4)3–Ce0.5Sm0.5PO4 composite ceramics are more than 95 % when the sintering temperature exceeds 1273 K.

Structural stability and proton beam irradiation effects on simulated metal fluoride waste–loaded iron phosphate glass

AbstractNext‐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.

Characterization of Cs deposits formed by the interaction of simulated fission product CsOH in the gas phase and concrete at 200°C

Characterization of Cs deposits formed by the interaction of simulated fission product CsOH in the gas phase and concrete at 200°C

Thermal conductivity measurement of uranium–plutonium mixed oxide doped with Nd/Sm as simulated fission products

The thermal conductivities of near-stoichiometric (U,Pu,Am)O2 doped with Nd and/or Sm, which are major fission products (FPs) generated during irradiation of uranium–plutonium mixed oxides (MOX) fuels, as simulated fission products are evaluated from 1073 to 1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneities are prepared using ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the phonon transport model λ=(A+BT)−1. The dependences of the coefficients A and B on the Nd/Sm content (CNd and CSm, respectively) are evaluated as: A(mK/W)=1.70×10−2+0.93CNd+1.20CSm, B(m/W)=2.39×10−4.

Catalyzed oxidation of IG-110 nuclear graphite by simulated fission products Ag and Pd nanoparticles

To evaluate the stability of nuclear materials in high temperature gas reactors under air ingress conditions, catalytic oxidation of IG-110 graphite by two simulated fission products, metallic Pd and Ag, was studied in oxidative atmosphere and at temperatures up to 1000 °C using an integrated furnace, mass spectroscopy and infrared spectroscopy system. Transmission electron microscopy and X-ray diffraction studies show that Pd and Ag nanoparticles were successfully introduced onto powdery IG-110 graphite through an impregnation and subsequent heat-treatment process. The combined mass spectroscopy and infrared spectroscopy methods allow simultaneous analysis of two gaseous products, CO and CO2, and separate measurements of activation energy for their formation reactions. It was found that the introduction of Pd or Ag to IG-110 graphite substantially catalyzed the oxidation of graphite, characteristic of decreased onset temperatures for the oxidation of graphite. Moreover, the catalytic effects by Pd and Ag are considerably different based on measured concentration ratios of CO2 to CO as a function of oxidation temperatures. Ag makes the graphite oxidation commence at approximately 400 °C with CO2 being the dominant product. In contrast, Pd significantly increases the concentration ratio of CO2 to CO at temperatures higher than approximately 690 °C, although it decreases the onset temperature for the oxidation reaction to around 525 °C. To understand the catalytic difference, the mechanism of the graphite oxidation is discussed based on the changes of surface oxygen species on Ag and Pd.

Solidification of glass-ceramics of simulated fission products RE3+, Sr2+, Ba2+ in chlorine-containing waste molten salt

Embedding nuclear waste in glass-ceramic and immobilizing nuclides in ceramic lattice is an effective way for the disposal of high-level radioactive waste. In this paper, a method of solidification of simulated various nuclides was proposed, i.e., RE3+(RE = La, Sm, Nd, Dy), Sr2+ and Ba2+ precipitated from waste molten salt in the form of REPO4, SrCO3 and BaCO3 were solidified in glass-ceramics. To avoid the decomposition of SrCO3 and BaCO3 at high temperature, SrCO3/BaCO3 containing Cl− salt was further sintered with NH4H2PO4 to form Sr5(PO4)3Cl/Ba5(PO4)3Cl ceramics. It was found that the prepared REPO4 belonged to monoclinic or tetragonal crystal system, while Sr5(PO4)3Cl and Ba5(PO4)3Cl belonged to hexagonal crystal system. REPO4, Sr5(PO4)3Cl and Ba5(PO4)3Cl ceramics were co-solidified in iron phosphate glass. BET results showed that the ceramics had a dense structure without any pore inside. XRD, TEM and HRTEM results showed all ceramics had high crystallinity, and nuclides could enter the lattice structure of ceramics through isomorphic replacement, which made the nuclides stable in the crystal structure. The effects of embedding rate on the volume density, Vickers hardness and wettability of glass-ceramics were explored. It was found that the density of the glass-ceramics gradually increased with the increase of ceramic embedding rate, however, the Vickers hardness firstly increased and then decreased. When the embedding rate reached 20 wt%, the Vickers hardness of the glass-ceramics could reach 583.90 GPa. The water contact angles of glass-ceramics with an embedding rate 0–40 wt% were measured to be 70.45°–84.05°, indicating glass-ceramics having a good water leaching resistance. Furthermore, the normalized leaching rate NRi of La3+, Sm3+, Nd3+, Dy3+, Sr2+, Ba2+, Cl− on the 28th day were estimated to be 7.53 × 10−7, 5.02 × 10−7, 5.12 × 10−7, 4.04 × 10−7, 1.22 × 10−3, 1.59 × 10−4, 1.07 × 10−4 g‧m−2‧d−1, which indicating that all elements remained good leaching resistance.

Catalyzed oxidation of nuclear graphite by simulated fission products Sr, Eu, and I

The influence of three fission products Sr, Eu, and I on the oxidation of IG-110 nuclear graphite was studied in the temperature range of 400 to 1000 °C. Sr and Eu were introduced as chlorides, and I was introduced as NaI. The temperature dependence of both CO2 and CO production during the graphite oxidation measured with mass spectroscopy and infrared spectrometry shows that the introduction of these three compounds to graphite significantly decreases the onset temperature for the oxidation of graphite. Among the three compounds, NaI is the most active towards the oxidation reaction, characterized by a significant decrease of the onset temperature from approximately 650 to 400 °C before and after its introduction to graphite. Separate measurements of CO2 and CO concentration at varying temperatures enable the calculation of the activation energy for the formation of CO2 and CO. The activation energies for the oxidation of pure and fission product-impregnated graphite samples decrease in the following order: standard IG-110 graphite, EuCl3-impregnated IG-110, SrCl2-impregnated IG-110, and NaI-impregnated IG-110. This trend indicates that the three compounds catalyze the oxidation of graphite at temperatures relevant to the operation of high-temperature gas-cooled reactors. Furthermore, it is found that the three compounds can also affect the molar ratio of reaction products CO2 and CO, and the rates of the graphite oxidation. At temperatures higher than about 850 °C, the impregnated samples exhibit lower CO2: CO ratios than the pure graphite. Different from EuCl3 and NaI, the introduction of SrCl2 decreases the graphite oxidation rates at temperatures higher than about 770 °C. Their catalytic mechanism can be understood based on a redox cycle of the intermediate active species, promoting the dissociation of molecular oxygen and transfer to the carbon.

Benefits of using multiple Raman laser wavelengths for characterizing defects in a UO2 matrix

AbstractRaman spectroscopy is one of the most useful techniques for studying the structure of UO2 and changes due to specific defects caused by doping, changes in stoichiometry, irradiation, or heating under oxidizing conditions. In this paper, we illustrate several aspects of the application of Raman techniques to the study of UO2, including the use of wavelength‐dependent excitation (455, 532, and 785 nm) to assess the effects of doping (Nd, Th, and Zr), ion irradiation, and in situ heating and oxidation (UO2 to U3O8). Additionally, we show examples of how correlative microscopy is possible using electron backscatter diffraction combined with Raman maps of specific vibration bands or of laser‐induced luminescence generated by rare‐earth dopants in the matrix. For each of these applications, we suggest optimal excitation wavelengths that vary depending on the desired data. Blue (455 nm) excitation tends to promote oxidation even at low powers, but because Raman spectra change little with doping, irradiation‐induced changes are easier to observe. Green (532 nm) excitation is optimal for observing electron–phonon resonance effects in UO2 and offers a good compromise for high‐temperature oxidation experiments, delivering high‐quality spectra for both UO2 and U3O8. Infrared (785 nm) excitation is best for observing “defect” bands associated with doping in UO2, as changes with irradiation are small. Raman spectroscopy is particularly suited for studying the stability of UO2 towards oxidation in the presence of dopants simulating fission products, where electron–phonon resonant effects, dopant ion luminescence, and mapping can be used together to investigate structural rearrangement as a function of temperature. These techniques can offer insight into microstructural changes in UO2 fuels at higher burnups envisioned in future reactors.

Microwave-sintering preparation, phase evolution and chemical stability of Na1-2xSrxZr2(PO4)3 ceramics for immobilizing simulated radionuclides

Sodium zirconium phosphate (NaZr2(PO4)3, NZP) family ceramics were considered as a potential host matrix for the immobilization of nuclear waste. In this study, the dense Sr-incorporated NZP ceramics with the formula of Na1-2xSrxZr2(PO4)3 (x = 0, 0.1, 0.2, 0.25, 0.3, 0.4, and 0.5) were fabricated by microwave-assisted solid-state reaction method. The phase evolution, densification behavior and chemical stability of the as-prepared samples were systematically investigated by XRD, Rietveld refinement, Raman, SEM-EDS and PCT method. The results demonstrated that Na in the NZP lattice could be substituted by the simulated fission product Sr, and the phase composition gradually changed from NZP to Sr0.5Zr2(PO4)3 (SrZP) with the increase of Sr content. The samples presented high densification and uniform elements distribution after microwave sintering at 1100 °C for 2 h. Moreover, the as-prepared samples exhibited the superior chemical stability. The normalized elemental leaching rate of Sr, Na, P, and Zr were 10−3, 10−4, 10−3, and <10−7 g⋅m−2⋅d−1, respectively.

Phase evolution and microstructure of new Sr0.5Zr2(PO4)3-NdPO4 composite ceramics prepared by one-step microwave sintering

Sodium Zirconium Phosphate (NaZr2(PO4)3, hereinafter NZP) and monazite are both potential materials for immobilization of nuclear waste. In this work, novel (1-x)Sr0.5Zr2(PO4)3-xNdPO4 composite ceramics (x = 0, 0.2, 0.4, 0.6, 0.8, and 1.0) for simultaneously immobilizing the simulated fission product (FP) Sr and trivalent minor actinide (MA) Nd were prepared by one-step microwave sintering technique, in which Sr and Nd were immobilized into NZP and monazite type structures, respectively. The phase evolution and microstructure of the samples were investigated by X-ray diffraction (XRD), Raman, and backscattering scanning electron microscopy (BSE). The results showed that the expected composite ceramics were successfully obtained by one-step microwave sintering at 1050 °C for 2 h. The as-prepared samples consisted of Sr0.5Zr2(PO4)3 and NdPO4 phases, and the content of the two phases varied regularly as x changed, generally conforming to the designed nominal chemical composition. Importantly, the composite ceramics presented the homogenous and dense microstructure. The relative density of the composite ceramics was more than 95%, meanwhile, the Vickers-hardness of the samples was higher than 600 MPa. It was indicated that NZP-monazite type composite ceramics could be a potential matrix for the simultaneous immobilization of actinide and fission product.

Segregation behavior of Fe and Gd in molten corium during solidification progress

For a criticality assessment of the fuel debris generated by the Fukushima–Daiichi Nuclear Power Plant accident, knowing the segregation of neutron absorber materials, i.e., Gd, B, and Fe, in the fuel debris is necessary. Although B may mostly evaporate during melting, Fe and Gd are expected to remain in the molten corium. To understand the redistribution behavior of Gd and Fe during the solidification of the molten corium, solidification experiments with simulated corium (containing UO2, ZrO2, FeO, and Gd2O3 with a small amount of simulated fission products such as MoO3, Nd2O3, SrO, and RuO2) were performed using a cold crucible induction heating method. The simulated corium was slowly cooled from 2500 °C and solidified from the bottom to the top of the melt. An elemental analysis of the solidified material showed that the Fe-concentration in the inner region increased up to approximately 3.4 times that in the bottom region. This suggested that FeO might be concentrated in the residual melt and that, consequently, the concentration of Fe increased in the later solidification region. On the contrary, the Gd concentration in the periphery region was found to be approximately 2.0 times higher than that in the inner region, suggesting the segregation of Gd in the early solidified phase and to remain with the residual, UO2-rich phase even after oxidative losses. No significant segregation was observed for the simulated fission products.

U-10Zr and U-5Fs: Fuel/cladding chemical interaction behavior differences

A fuel performance phenomenon that has previously limited the peak cladding temperature of metallic fast reactor fuel is fuel/cladding chemical interaction (FCCI). This study examined the differences in the phenomenon exhibited by the more recent fuel alloy, U-10Zr and a previously used fuel alloy, U-5Fs (Fs – fissium). Fs is comprised of Mo, Ru, Pd, Zr and Rh simulated fission products representing reprocessing carryover. Studies on the U-Zr alloy had shown the often controlling influence on FCCI of lanthanide (Ln) fission products (Ln: Ce, Nd, Sm, etc.) while a review of the U-Fs alloy behavior indicated a dominance of basic fuel/cladding interdiffusion (Fe, Ni with U). The recent studies on U-Zr fuel showed that migration of lanthanide fission products to the fuel/cladding interface facilitated the Ln dominance of FCCI behavior. This study was focused on understanding why Ln’s do not migrate in the U-5Fs alloy to the extent that they do in U-10Zr, or why they do not interact with the cladding like in U-10Zr fuel pins. A small fuel/cladding interaction layer was found containing the cladding and lanthanide components. However, it was also found that the lanthanide fission products are at least partially bound to agglomerates that also contained Pd, Rh and Zr, all minor components of Fs.

Study on reaction of 15-15Ti with fission products under various oxygen potential

ABSTRACT The effect of oxygen potential on the Fuel-Cladding Chemical Interaction (FCCI) of 15-15Ti cladding material was studied in this paper. In the test, the materials have been exposed to simulated fission products at 600◦C for 120 h under the oxygen potential in equilibrium with Mo/MoO2 and Cr/Cr2O3.The general corroded depth was roughly 30 µm. Furthermore, the results also show that corroded depth on the cladding surface increase with the increase of oxygen potential. No intergranular attacks were observed, which might be caused by the absence of the high-temperature gradient. A single layer of CrTe is assumed to be formed under low oxygen potential, whereas cesium and oxygen may play an important role under high oxygen potential. Moreover, FCCI may impose a great threat to the integrity of China designed 15-15Ti material.

Murataite: a matrix for immobilizing waste generated in radiochemical reprocessing of spent nuclear fuel

Incorporation of waste from spent nuclear fuel pyrochemical reprocessing and calcines from the decontamination of glove box and hot cell equipment into murataite-based ceramics was studied. The phase and chemical compositions of the ceramics containing precipitates simulating fission products in molten chlorides were examined. The radiation stability of murataite matrices was studied by incorporating 244Cm isotope (1.8 wt%). In the ceramics produced by melting at 1325 and 1350 °C the murataite phases was rendered to be X-ray amorphous at doses of 2.46 × 1018 and 2.53 × 1018 α-decay/g (0.21 dpa), while for the sample sintered at 1250 °C the amorphization dose was found to be 2.73 × 1018 α-decay/g (0.21 dpa). The murataite structure was recovered after the annealing at 1250 °C for 5 h in air. Both the pristine and amorphized samples had very low leachability of Cm and major elements. Production of highly durable murataite-based ceramics containing waste surrogate after spent nuclear fuel (SNF) pyrochemical reprocessing and calcine after the evaporation of decontamination solutions were demonstrated.

Physicochemical principles of preparation of U(VI) carbonate solutions for extraction reprocessing in the Carbex process

Physicochemical principles of preparation of U(VI) carbonate solutions in the step of oxidative dissolution of U3O8 and UO2 in the Carbex process are considered. Carbonate solutions with the U(VI) concentration higher than 100 g L–1, suitable for subsequent final purification of uranium by extraction, can be prepared under the conditions of formation of U(VI) carbonate–peroxide complexes in the course of dissolution with prevention of hydrolysis of U(VI) compounds. The behavior of impurities simulating some fission products in the course of oxidative dissolution was studied, and the decontamination factors of U(VI) from the chosen simulated fission products were determined.

Experimental and Monte Carlo investigation of mass attenuation coefficients of fission product isotopes in molecular precipitates

Mass attenuation coefficients for molecular precipitates have been measured and compared to results using the EGS5 Monte Carlo computer code and results from an empirical formula. This study assesses the mass attenuation coefficients of isotopes that can be readily produced by thermal neutron activation of elements in a reactor and precipitated in molecular compounds. Good agreement exists between measured results and EGS5 simulated results. Results are within reasonable agreement with the empirical mass attenuation formula. These results are further compared to simulated fission product isotopes. This study suggests mass attenuation coefficients of molecular precipitates can be approximated using EGS5, especially in the instance of radioisotopes produced predominantly through uranium fission.

Solubility of dialkylalkyl phosphonates in supercritical carbon dioxide: Experimental and modeling approach

Dialkylalkyl phosphonates have found renewed interest in the selective extraction of actinides from different matrices. In this context, two dialkylalkyl phosphonates, dibutylbutyl phosphonate (DBBP) and diamylamyl phosphonate (DAAP) were synthesised and their solubilities in supercritical carbon dioxide (SCCO2) medium were determined in the pressure range of 10–25 MPa at 313–333 K. Solubility studies were carried out to examine their utility as ligands during supercritical fluid extraction of actinides. Solubilities of dialkylalkyl phosphonates ranged from 0.06 to 0.12 mole fraction. Experimental solubilities were correlated using Mendez-Santiago, Chrastil, solution theory with Wilson activity coefficient model and association model based on van Laar activity coefficient model. Comparison among these models revealed that the association model with van Laar activity coefficient model provides better solubility predictions. SCCO2 containing DAAP was employed for the (i) selective extraction of uranium in the presence of simulated fission products and (ii) extraction of uranyl nitrate from acidic medium.

Multi-elemental Gd, Eu, Sm, Nd isotope ratio measurements by liquid chromatography coupled to MC-ICPMS with variable Faraday cup configurations during elution

The high-precision isotopic characterization of actinides and fission products in nuclear samples is fundamental for various applications such as the management of spent nuclear fuel or the validation of neutronic calculation codes. However multi-elemental isotope ratio measurements by mass spectrometric techniques are hampered by the presence of both spectral and non-spectral interferences as complex sample matrices are encountered in such topics, but also due to the lack of high precision mass spectrometers able to cover the entire mass spectrum. This work describes a new LC-MC-ICPMS approach allowing simultaneous high-precision and multi-elemental isotope ratio measurements of four fission products of interest for nuclear issues (Nd, Sm, Eu, Gd) within a single elution run. Variable motorized Faraday cup configurations were successively used during a specifically designed elution procedure in order to take into account the non-natural Nd, Sm, Eu, Gd isotopic compositions encountered in irradiated nuclear samples. This new method, involving the relevant isotopic reference standard injection timings for on-line mass bias corrections, was validated by the analysis of a simulated fission product fraction from a 235U-irradiated target. Reproducibilities better than 2‰ (k=2), comparable to those obtained by off-line measurements and the classic sample-standard bracketing mass bias correction approach, were obtained for all isotope ratios, except those involving isotopes with a transient signal peak apex lower than 100mV, for which the reproducibilities were comprised between 2‰ and 6‰.