Concentrations Of Fission Products Research Articles

Rare Earth Silicide Formation on Si Electrode in LiCl-KCl Melt to Establish a Novel Used Salt Treatment Process

Pyroprocessing of spent nuclear fuels which uses chemical and electrochemical reactions in molten salts has been developed as one of the promising technologies for advanced nuclear fuel cycle. A main step of the pyroprocessing is electrorefining process where a recovery of actinides from spent nuclear fuels is accomplished in an electrolyte, LiCl-KCl melt. After the electrorefining process, fission products (rare earths (REs), alkalis and alkaline earths) chlorides accumulate in the melt, the used LiCl-KCl melt.The accumulated fission products must be removed from the used melt to be stabilized in a waste form suitable for a geological disposal. For this purpose, a zeolite is utilized in the conventional used salt treatment process: The zeolite is contacted with the used melt to absorb and occlude the fission products in its structure and then the zeolite is converted to a stable chlorine-containing waste form, a glass-bonded sodalite. However, the achievable concentration of fission products in the glass-bonded sodalite is limited to lower values (1.6 ~ 4.3 wt%) compared with that in the conventional oxide-base glass waste form (6.4 ~ 15.2 wt%).In order to pursue a reduction of an environmental burden, this paper proposes a novel used salt treatment process utilizing the glass to decrease the volume of the waste form. REs, dominant among fission products, are electrochemically recovered from the used melt in a form of their silicides (REn+ + xSi + ne- → RESix). Then, the silicides are oxidized to be dissolved in the glass matrix. Si is selected as the alloying material, meaning that REs are not required to be separated from the Si before stabilizing in the glass waste form which is composed mainly of Si. After the electrochemical recovery of REs, the remaining fission products, alkalis and alkaline earths, are removed from the used melt in a similar manner to the conventional one; absorption and occlusion in the zeolite to be converted to the glass-bonded sodalite.To investigate a basic behavior of a series of an individual RE silicide (La-Si, Ce-Si, Pr-Si, Nd-Si, Sm-Si, Eu-Si and Gd-Si) formation, cyclic voltammetry and potentiostatic electrolysis were performed using a Si wafer (n-type, <110>, 0.02 Ωcm) as the working electrode in LiCl-KCl melt containing the corresponding RE chloride at 723 K. It was confirmed that all of the above REs were recovered from the melt as their silicides. Silicides with various morphologies were found to be formed on the Si electrode; a pillar shaped silicide growing perpendicularly to the Si substrate for Sm, a dense and thin silicide layer covering the Si substrate for Eu and a porous silicide layer between a dense silicide layer and the remaining Si substrate for the rest of the REs. The obtained results confirmed a feasibility of the key step composing of the proposed used salt treatment process; REs recovery as silicides from the used LiCl-KCl melt. In addition, a preliminary test of the proposed used salt treatment process, Ce recovery as silicide from LiCl-KCl-CeCl3 melt and the oxidation of the Ce silicide to be dissolved in a glass matrix, was demonstrated and an optimization of the processing condition is under way.Acknowledgement Parts of the presented results were obtained in the project “Development of highly flexible technology for recovery and transmutation of minor actinide” entrusted to CRIEPI by the Ministry of Education, Culture, Sports, Science and Technology (MEXT).

Development of a robust, accurate and efficient coupling between PLEIADES/ALCYONE [formula omitted] fuel performance code and the OpenCalphad thermo-chemical solver

In this paper, a robust, accurate and efficient coupling between the PLEIADES/ALCYONE 2.1 fuel performance code and the OpenCalphad thermo-chemical solver is presented. A challenging 3.5D simulation of a power ramp, consisting of the 3D simulations of all the fuel pellets within a complete rodlet, serves as a base case to illustrate the difficulty inherent to extensive thermochemical equilibrium calculations (more than 1.5 million in this example) within fuel performance simulations, related both to the wide temperature and pressure range encountered in irradiated nuclear fuels. In the first part of this paper, a detailed analysis of the main quantities of interest (i.e. temperature, hydrostatic pressure, fission products concentration) is provided to illustrate the couplings at hand during a power ramp and, above all, to demonstrate the good agreement of the results with those already published in the literature and with available measurements of xenon, cæsium and iodine releases. In the core of this paper, several numerical strategies are proposed to reduce the overall time spent in the thermochemical solver by defining at each node and for each time step a good initial estimate of the set of stable phases that are likely to form. The first one, referred to as the spatial strategy, relies on an intelligent reorganization of the mesh nodes as a function of temperature or hydrostatic pressure with an appropriate initialization of the thermochemical equilibrium calculation at a node by the phase composition calculated at the node with the closest thermo-mechanical conditions. The second strategy, referred to as the hybrid strategy, combines the spatial strategy during the power transient (where large power variations take place at each time step) with an initialization of the nodal equilibria from the solution obtained at the previous time step in case of small power variations. The 3.5D simulation has been run several times in order to assess and compare the spatial and hybrid strategies. The simulations show that the spatial strategy is the most efficient. It leads to an overall increase of the calculation time related to the incorporation of thermochemistry in the 3.5D fuel performance simulation of less than 2.25%, showing the feasibility of large thermal-chemical–mechanical simulations within the PLEIADES/ALCYONE 2.1 fuel performance code. Finally, as a first step towards more complicated simulations (i.e. including a greater number of phases and/or performed with a finer spatial discretization), a primitive parallelization algorithm based on a multiprocess approach is proposed. Combined with the most efficient spatial strategy, it leads to a substantial reduction of the extra (real) calculation time related to thermochemistry which become less than 0.65% of the total computation time when the thermochemical equilibrium calculations are distributed over 4 cores.

Rare Earth Silicide Formation on Si Electrode in LiCl-KCl Melt to Establish a Novel Used Salt Treatment Process

After repeating the electrorefining of spent nuclear fuels, rare earth, alkaline earth, and alkali fission products accumulate in the used LiCl-KCl melt. This paper proposes a novel used salt treatment process where dominant fission products, rare earths, are recovered electrochemically from the used melt in a form of their silicides and the silicides are then oxidized to be dissolved in a glass matrix as the waste form. The glass waste form stabilizing a higher concentration of fission products than the conventional waste form provides a reduction in volume of the waste form. The results of cyclic voltammetry using a Si electrode in LiCl-KCl melt containing a rare earth chloride (LaCl3, CeCl3, PrCl3, NdCl3, SmCl2, EuCl2, or GdCl3) at 723 K indicated the corresponding lanthanide silicide formation. The rare earth silicides formed by potentiostatic electrolysis were analyzed by XRD and SEM-EDX. The obtained results confirmed a feasibility of rare earths recovery as silicides from the used LiCl-KCl melt, which was a key step of the proposed used salt treatment process.

A first principle based study on the mechanical and thermal properties of UO2: Effect of La and Dy fission product concentrations

The amount of fission products substitution in the uranium dioxide (UO2) fuel matrix and subsequently their interaction with the host is important in the perspective of safe and efficient operation of nuclear reactors as fission products can change the structural, mechanical and thermo-physical properties of the host fuel significantly. Therefore, structural, elastic, mechanical and thermo-physical properties of lanthanide (Ln) fission products doped UO2 with different doping concentrations of Ln atoms are studied by employing density functional theory with Hubbard U correction. In this study, two Ln atoms (Ln = La and Dy) and different doping concentrations in UO2e.g. 6.25%, 12.5%, 25%, and 50% are considered. The mechanical property such as bulk modulus of Ln doped UO2 are directly related to the Ln doping concentration. Expressions are generated to extract various properties at very low and high Ln doping concentrations. The thermo-physical properties like heat capacity and coefficient of thermal expansion are also evaluated based on quasi-harmonic approximation to the phonons frequencies which are obtained from density functional perturbation theory. The UO2 lattices with higher Ln atom concentrations exhibit new electronic energy bands near the band gap region, which are dominated by Ln atomic orbitals. From the present study, we derived correlations between La and Dy doping concentration and various thermo-physical properties of UO2. Moreover, the properties obtained from our first principle calculations show a good agreement with the available experimental and theoretical values. Thus, our systematic study with varying fission product concentrations will provide insight to the UO2 fuel properties under reactor conditions.

Data assimilation of post irradiation examination experiments to adjust fission yields

Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using data from post-irradiation examination experiments. The adjusted fission yields improve the bias and reduce the uncertainty of predicted fission product concentrations in spent fuel. Our approach modifies fission yields by adjusting the model parameters of the code GEF with post-irradiation examination experiments. We used the BFMC data assimilation method to account for the non-normality of GEF's fission yields. In the application that we present, the assimilation decreased the average bias of the predicted fission product concentrations from 26% to 15%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in ENDF/B-VIII.O. For Pu-239 thermal fission, the average relative difference from ENDF/B-VIII.O was 16% before data assimilation and 11% after. For the standard deviations of the fission yields, GEF's were, on average, 16% larger than those from ENDF/B-VIII.O before data assimilation and 15% smaller after.

Data assimilation of post-irradiation examination data for fission yields from GEF

Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged.

Analysis of Some Computer Codes Used for Evaluation of Spent Nuclear Fuel Radiation Parameters

The radiation parameters such as radionuclide content and activities, fluxes and energy spectrum of gamma and neutrons of spent nuclear fuel are essential when planning further spent fuel management options – interim wet or dry storage or disposal into a geological repository. Radiation parameters determine the design of a storage or disposal facility, what materials, structures and thicknesses of structures should be used to provide adequate biological shielding.&#x0D; Experimental measurements of spent fuel radiation parameters are rather complicated and expensive, therefore numerical methods are widely used. Various computer codes (APOLLO, BOXER, CASMO, FISPACT, ORIGEN-S, WIMS, etc.) have been developed to simulate the irradiation processes of nuclear fuel and to obtain resulting radiation parameters. Irrespective of the used computer code, the input data firstly must be entered into that code. When simulating nuclear fuel irradiation and burn-up in the reactor core, the geometrical parameters of the fuel assembly, materials’ data (chemical compositions, densities), the operating parameters of the reactor (power, operation time, coolant parameters, etc.) shall be entered into the program as initial data. Fairly often approximations of the input data are performed, for example, fuel rods in a fuel assembly are homogenized and geometrically described as a solid cylinder, the reactor operation time is assumed as continuous and at constant power. The particularity of the input data and accepted assumptions depend on what initial information is available and on the capabilities of the computer code. The modelled spent fuel radiation parameters depend not only on the input data and assumptions, but also on the cross-section databases that are used in computer codes. Computer codes TRITON, ORIGEN-S and FISPACT have been used to model the concentration of actinides and fission products in the spent fuel from the RBMK-1000 reactor. The obtained results are compared and possible reasons for the differences in the modelling results are discussed.

Rapid concentration and isotopic measurements of ultra-trace 235U fission products with comparison to an ORIGEN isotope depletion model

This article presents the application of an automated online separation–direct analysis method, RAPID (Rapid Analysis of Post-Irradiation Debris), for the simultaneous measurement of both radioactive and stable fission isotopes from an irradiated highly-enriched uranium target. Developed for the measurement of the concentration and isotopic composition of over 40 elements down to the femtogram level, the RAPID method possesses the sensitivity, stability, and precision required to achieve accurate, low-level analyses of elements of non-natural origin. The isotopic compositions and concentrations of key fission elements cesium, strontium, yttrium, lanthanum, cerium, praseodymium, neodymium, promethium, and samarium have been measured repeatedly over a six-week period. The validity of these measurements was confirmed using ORIGEN (Oak Ridge Isotope GENeration), an isotope depletion and decay modeling software, to within 1–3%.

A Method to Estimate Fission Product Concentration Uncertainty in a Multi-Time-Step MCNP6 Code Nuclear Fuel Burnup Calculation

The Monte Carlo N-Particle (MCNP6) radiation transport code is widely used to perform material transmutation and depletion calculations using the embedded module CINDER90. CINDER90 is capable of obtaining fission product and transuranic nuclide concentrations with a high level of accuracy in irradiated nuclear fuel. This information is very useful for many nuclear applications including reactor design and analysis, nuclear safeguards, nuclear security, and nuclear forensics, to name a few. However, at present the MCNP6 code does not estimate the overall statistical uncertainty in the nuclide concentrations reported at the end of a depletion calculation. We report our approach using a random sampling method to estimate stochastic uncertainty in fission product nuclide concentration using various parameters reported in MCNP6 output and how these uncertainties are affected by the calculation parameters.

Evaluation of atmospheric dispersion of radioactive materials in a severe accident of the BNPP based on Gaussian model

Regarding the importance of radioactive materials dispersion in the environment and their effect on living organism's well-being, this article examines the atmospheric dispersion of these materials around the Bushehr nuclear power plant (BNPP) in a hypothetical severe accident. One of practical dispersion models is AERMOD which is based on Gaussian distribution. This model which is an extension of Gaussian model is normally used for predicting the gaseous pollution. The aim of this study is to investigate the dispersion of radioactive materials in BNPP's surroundings due to a specific type of Large Break LOCA (LBLOCA) accident. In this study, the wind direction, wind speed, and the topological conditions of the power plant's environment are considered. In order to execute this model, the meteorology and topological data of the region have been provided. Besides, using the CONTAIN code, the data about released radioactive materials from the containment have been provided. The double ended cold leg break (DECL) accident in containment has been modeled and its validity has been checked with FSAR. Extracting all required data, the simulation is done in AERMOD and the volumetric concentration and activity of fission products are calculated. The volumetric activity of fission products in Tangestan, Bandar-E-Bushehr and Delvar cities are evaluated as 7.5658E+07, 7.5136E+07 and 2.9655E+06 (Bq/m3), respectively. From results it could be seen that the most dispersion of the radioactive materials are in direction of the non-residential regions, although the pollution of the cities around the power plant is inevitable.

Computational and experimental forensics characterization of weapons-grade plutonium produced in a thermal neutron environment

Abstract The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural UO2 fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately 20 μg of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural UO2 fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural UO2 fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%.

Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes

This paper investigates how concentrations of certain fission products and their related gamma-ray emissions can be used to discriminate between uranium oxide (UOX) and mixed oxide (MOX) type fuel. Discrimination of irradiated MOX fuel from irradiated UOX fuel is important in nuclear facilities and for transport of nuclear fuel, for purposes of both criticality safety and nuclear safeguards. Although facility operators keep records on the identity and properties of each fuel, tools for nuclear safeguards inspectors that enable independent verification of the fuel are critical in the recovery of continuity of knowledge, should it be lost.A discrimination methodology for classification of UOX and MOX fuel, based on passive gamma-ray spectroscopy data and multivariate analysis methods, is presented. Nuclear fuels and their gamma-ray emissions were simulated in the Monte Carlo code Serpent, and the resulting data was used as input to train seven different multivariate classification techniques. The trained classifiers were subsequently implemented and evaluated with respect to their capabilities to correctly predict the classes of unknown fuel items.The best results concerning successful discrimination of UOX and MOX-fuel were acquired when using non-linear classification techniques, such as the k nearest neighbors method and the Gaussian kernel support vector machine. For fuel with cooling times up to 20 years, when it is considered that gamma-rays from the isotope 134Cs can still be efficiently measured, success rates of 100% were obtained. A sensitivity analysis indicated that these methods were also robust.

Determination of concentrations of fission products by ICP–AES in solutions from spent nuclear fuel reprocessing

A procedure was developed for determining the gravimetric concentration of U and fission products (Ba, Mo, Zr, Sr, Rb, Pd, and Cd) in solutions from spent nuclear fuel (SNF) reprocessing, using inductively coupled plasma atomic emission spectrometry (ICP–AES). The optimum analysis parameters were found. The analytical lines that are not significantly affected by the matrix element were chosen for the elements being determined. The total margin of error in measurements of the metal concentrations (0.1–2500 mg L–1) by the suggested procedure is in the interval 1–20% (for the minimal concentration). The detection thresholds of uranium fission products in model solutions at optimum parameters vary in the interval from 0.020 to 0.1 mg L–1.

Experimental and Computational Forensics Characterization of Weapons-Grade Plutonium Produced in a Fast Reactor Neutron Environment

A terrorist attack using an improvised nuclear device is one of the most serious dangers facing the United States. The work presented here is part of an effort to improve nuclear deterrence by developing a methodology to attribute weapons-grade plutonium to a source reactor by measuring the intrinsic physical characteristics of the interdicted plutonium. In order to demonstrate the developed methodology, plutonium samples were produced from depleted uranium dioxide (DUO2) surrogates irradiated in a fast-neutron environment. In order to replicate the neutron flux in a fast-neutron-spectrum reactor and obtain experimental samples emulating weapons-grade plutonium produced in the blanket of a fast breeder reactor, DUO2 samples were placed in a gadolinium sheath and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Previous computational work on this topic identified several fission products that could be used to distinguish between reactor types (fast and thermal reactors), specifically: 137Cs, 134Cs, 154Eu, 125Sb, 144Ce, 85Rb, 147Pm, and 150Sm along with the plutonium isotopes. Simulations of the fast neutron irradiation of the DUO2 fuel surrogates in the HFIR were carried out using the Monte Carlo radiation transport code MCNPX 2.7. Comparisons of the predicted values of plutonium and fission product concentrations to destructive and nondestructive assay measurements of neutron-irradiated DUO2 surrogates are presented here. The agreement between the predictions and gamma spectroscopic measurements in general were within 10% for 134Cs, 137Cs, 154Eu, and 144Ce. Additional experimental results (mass spectroscopy) agreed to within 5% for the following isotopes: 85Rb, 147Pm, 150Sm, 154Eu, 148Nd, 144Ce, and 239Pu. Two indicator isotopes previously suggested to differentiate between the reactor types were ruled out for use in the attribution methodology; 125Sb was ruled out due to the difficulty in accurately predicting its concentration, and 242Pu was ruled out because of its low content in weapons-grade plutonium.

Electrochemical Behavior of Samarium in Molten LiCl-KCl

The mass transport behavior of samarium was investigated in molten eutectic LiCl-KCl in order to develop real-time concentration monitoring capabilities for the pyrochemical reprocessing of used nuclear fuel. Pyrochemical reprocessing relies on electrolysis in molten LiCl-KCl to recycle fissile material from used nuclear fuel. Without the capabilities to monitor concentrations of various actinides and fission products, pyrochemical reprocessing cannot be commercialized due to strict material accountability requirements set by the International Atomic Energy Agency (IAEA) [1]. Therefore, physical electrochemical properties of these species in molten salts is extremely important. Cyclic voltammetry was utilized to investigate electrochemical behavior of samarium in LiCl-KCl. By establishing valid relationships between these parameters and the peak current, the diffusion coefficient of samarium in our system can be reported with confidence [2]. Experiments were conducted in a high-purity argon atmosphere glove box using a three-electrode cell contained in an alumina crucible inside a resistive furnace. A high-resolution translation stage equipped with ceramic electrode mounts was used to accurately control electrode positions inside the melt. Tungsten was used as the counter and working electrodes, and a silver wire was used as a quasi-reference (Ag|AgCl2). The variation in peak current that resulted from changing the concentration, scan rate and electrode surface area will be reported, along with calculated diffusion coefficient values for samarium in molten LiCl-KCl at 500°C. The effects of using molybdenum as a working electrode will also be reported and compared to results obtained using tungsten. A comparison between reported diffusion coefficient values obtained by separate research organizations and results obtained in these studies will also be discussed. Acknowledgements: This work was performed under the auspices of the Department of Energy (DOE) under contracts DE-NE0008262 and DE-NE0008236 as well as the US Nuclear Regulatory Commission (USNRC) under contracts NRCHQ-11-G-38-0039. Dr. Kenny Osborne serves as the program manager for the DOE award and Ms. Nancy Hebron-Isreal serves as the grants program officer for the NRC awards.

Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by “bottlenecks” in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

Investigation of different burnable absorbers effects on the neutronic characteristics of PWR assembly

There is an urgent need to find compact burnable BA materials as an alternative to the traditional materials (B4C and Ag–In–Cd). These materials must possess higher density, good thermophysical properties, radiation resistance, and swelling resistance. In this article, the most interesting rare earth nuclides for neutron absorption (gadolinium, erbium, dysprosium, hafanium and dysprosium) are studied and compared from the point of view of their possibilities for the control of core reactivity in PWRs to increase the cycle length. MCNPX code is used to design a three dimensional model for PWR assembly. The PWR assembly is loaded in different operation process with 2% Gd2O3 and 2% Er2O3 as integral burnable absorbers (IBAs). The effects of gadolinia and erbia on the thermal neutron flux and normalized power are investigated. The study presents that erbium is more effective IBA than gadolinium for a reactor with a very long cycle. The dysprosium titanate, dysprosium hafnate and hafnium diboride are investigated as material for a control rod in PWR assembly. The effects of these control rod materials on the multiplication factor and the concentrations of fission products are calculated. From the viewpoint of the achievable fuel discharge burnup, it was found that the suggested control rod materials achieve a satisfactory control rod worth (CRW), i.e., the CRW of the presented materials is comparable with the CRW of the standard AIC control rods.

Suppression pool testing at the SIET laboratory: experimental investigation of critical phenomena expected in the Fukushima Daiichi suppression chamber

In the unlikely event of a nuclear power plant long duration station black-out, as in the Fukushima Daiichi (1F) severe accident (SA), it was recognized that the suppression chamber (S/C) functions of heat sink and fission product (FP) scrubbing will degrade, resulting in the S/C pressure increase, reduction of the scrubbing efficiency and subsequent necessity of venting operations. Consequently, a relatively large amount of FPs, in particular highly volatile elements (e.g. CH3I), is likely to be dispersed into the environment. As a method to evaluate the degradation of the pool characteristics under discharge of steam and non-condensable gases through vent pipes and steam through different quencher geometries of make-up systems, an experimental campaign was recently started at the SIET research laboratory in Italy. Two different quencher geometries, representing vent pipes and the reactor core isolation cooling (RCIC) exhaust pipes in 1F2 and 1F3, were adopted. Several combinations of steam and air mass flow rates were tested to scale down the main conditions occurred during the 1F SA. Measurements of pool water temperature in different locations and visualization with high-speed camera represent the main outcome of the experimental activity. The preliminary results have demonstrated that a relatively small concentration of air in the steam flow is able to suppress the occurrence of chugging of the steam, with reduced mixing in the pool. Both RCIC quenchers adopted induced large chugging at the bottom of the pool which are effective to avoid temperature stratification, thanks to the large water recirculation and vertical mixing within the pool. At decreased subcooling, mixing in the pool ceases and the quenchers with holes disposed in the vertical direction, as in the RCIC exhaust pipe of the 1F unit 3, introduce intense stratification that drastically reduces the condensation efficiency of the S/C pool. Quencher of 1F2 RCIC does not present stratification possibly dependent on the distance of the pipe outlet to the pool floor. Given the reduced size of the pool compared to the plant scale, the observed phenomena should not be extrapolated for the whole S/C. The objective of the ongoing experimental activity is to construct a database based on the high-speed filming, measurements of major quantities such as water temperature, steam pressure and FPs concentration to foster the development of physical models for both lumped parameter SA codes and detailed computational fluid dynamics software, in an effort to enhance the understanding of the complex phenomena following the 1F accident.

Lattice Parameter Behavior with Different Nd and O Concentrations in (U1−yNdy)O2±X Solid Solution

The solid solution of (U1−yFPy)O2±x has the same fluorite structure as UO2±x, and the lattice parameter is affected by dissolved fission product and oxygen concentrations. The relation between the lattice parameter and the concentrations of neodymium and oxygen in the fluorite structure of (U1−yNdy)O2±x was investigated using X-ray diffraction. The lattice parameter behavior in the (U1−yNdy)O2±x solid solution shows a linear change as a function of the oxygen-to-metal ratio and solubility of neodymium. The lattice parameter depends on the radii of ions forming the fluorite structure and also can be expressed by a particular rule (modified Vegard’s law). The numerical analyses of the lattice parameters for the stoichiometric and nonstoichiometric solid solutions were conducted, and the lattice parameter model for the (U1−yNdy)O2±x solid solution was assessed. A very linear relationship between the lattice parameter and the Nd and O concentration for the stoichiometry and nonstoichiometry of the (U1−yNdy)O2±x solid solution was verified.

Determination of rare-earth elements in nitric acid solutions with high concentrations of uranium and macrocomponents

Spectral and matrix effects of uranium and other macrocomponents on the determination of rare-earth elements (REE) are studied. The experimental data are updated and verified using a fractional factorial experiment. A procedure is developed for the direct determination of REE in solutions with high concentrations of uranium and fission products (FP) using inductively coupled plasma-atomic emission spectroscopy (ICP-AES).