Uranyl Nitrate Solution Research Articles

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O 2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th, 233U)O 2 mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with 232U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO 2 is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O 2 pellets. In this study, fabrication of (Th,U)O 2 mixed oxide pellets containing 3–5 wt.% UO 2 was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.

ChemInform Abstract: Rapid Self‐Assembly of Uranyl Polyhedra into Crown Clusters.

AbstractThree chemically distinct clusters built from 32 uranyl peroxide polyhedra, (NH4)40 [(UO2)32(O2)40(OH)24] (I), (NH4)40 [(UO2)32(O2)44(OH)26] (II), and a third one with a composition consistent with coprecipitation of (I) and (II), self‐assemble and crystallize within 15 min from aqueous solutions of uranyl nitrate, ammonium hydroxide, and hydrogen peroxide in different molar ratios under ambient conditions.

Heterogeneous reduction of U 6+ by structural Fe 2+ from theory and experiment

Computational and experimental studies were performed to explore heterogeneous reduction of U 6+ by structural Fe 2+ at magnetite (Fe 3O 4) surfaces. Molecular Fe–Fe–U models representing a uranyl species adsorbed in a biatomic bidentate fashion to an iron surface group were constructed. Various possible charge distributions in this model surface complex were evaluated in terms of their relative stabilities and electron exchange rates using ab initio molecular orbital methods. Freshly-cleaved, single crystals of magnetite with different initial Fe 2+/Fe 3+ ratios were exposed to uranyl-nitrate solution (pH ∼ 4) for 90 h. X-ray photoelectron spectroscopy and electron microscopy indicated the presence of a mixed U 6+/U 5+ precipitate heterogeneously nucleated and grown on stoichiometric magnetite surfaces, but only the presence of sorbed U 6+ and no precipitate on sub-stoichiometric magnetite surfaces. Calculated electron transfer rates indicate that sequential multi-electron uranium reduction is not kinetically limited by conductive electron resupply to the adsorption site. Both theory and experiment point to structural Fe 2+ density, taken as a measure of thermodynamic reducing potential, and sterically accessible uranium coordination environments as key controls on uranium reduction extent and rate. Uranium incorporation in solid phases where its coordination is constrained to the uranate type should widen the stability field of U 5+ relative to U 6+. If uranium cannot acquire 8-fold coordination then reduction may proceed to U 5+ but not necessarily U 4+.

Isolation and analyses of uranium tolerant Serratia marcescens strains and their utilization for aerobic uranium U(VI) bioadsorption

Enrichment-based methods targeted at uranium-tolerant populations among the culturable, aerobic, chemo-heterotrophic bacteria from the subsurface soils of Domiasiat (India's largest sandstone-type uranium deposits, containing an average ore grade of 0.1 % U(3)O(8)), indicated a wide occurrence of Serratia marcescens. Five representative S. marcescens isolates were characterized by a polyphasic taxonomic approach. The phylogenetic analyses of 16S rRNA gene sequences showed their relatedness to S. marcescens ATCC 13880 (≥99.4% similarity). Biochemical characteristics and random amplified polymorphic DNA profiles revealed significant differences among the representative isolates and the type strain as well. The minimum inhibitory concentration for uranium U(VI) exhibited by these natural isolates was found to range from 3.5-4.0 mM. On evaluation for their uranyl adsorption properties, it was found that all these isolates were able to remove nearly 90-92% (21-22 mg/L) and 60-70% (285-335 mg/L) of U(VI) on being challenged with 100 μM (23.8 mg/L) and 2 mM (476 mg/L) uranyl nitrate solutions, respectively, at pH 3.5 within 10 min of exposure. his capacity was retained by the isolates even after 24 h of incubation. Viability tests confirmed the tolerance of these isolates to toxic concentrations of soluble uranium U(VI) at pH 3.5. This is among the first studies to report uranium-tolerant aerobic chemoheterotrophs obtained from the pristine uranium ore-bearing site of Domiasiat.

Benchmark Critical Experiments of a Heterogeneous System of Uranium Fuel Rods and Uranium Solution Poisoned with Gadolinium, and Application of Their Results to JACS Validation

A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of the Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark criticality data concerning the dissolving process in reprocessing plants. Focusing on the use of gadolinium as a soluble poison, critical mass was measured for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with gadolinium (Gd). Fuel rods were arrayed with a 1.5 cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 330 gU/L, and the Gd concentrations were varied up to 0.1 gGd/L. The other series of experiments were also conducted, as reference cases, varying uranium concentration in the fuel solution without Gd. The results provided benchmark criticality data for validation of neutron multiplication factor calculation on heterogeneous systems such as a dissolver. Validation calculation of JACS based on the newly obtained benchmarks supports the justification of its utilization for the criticality safety analysis.

Behavior of Actinide Elements and Fission Products in Recovery of Uranyl Nitrate Hexahydrate Crystal by Cooling Crystallization Method

To elucidate various kinds of actinide element and fission product behavior, U crystallization experiments were carried out with a uranyl nitrate solution and with a solution in which irradiated fast reactor core fuel was dissolved. Insoluble residue simulating that found in actual reactor operation was not incorporated into the uranyl nitrate hexahydrate (UNH) crystal in the course of the U crystallization. However, the decontamination factors (DFs) were below 10 even when the UNH crystal was washed because the mother liquor containing the simulated insoluble residue occupied the interspaces of the agglutinated UNH crystal. In the U crystallization process, the DF of Pu was >40 when the UNH crystal was washed. But, Np was not removed from the UNH crystal because Np was oxidized to Np(VI) in the feed solution and thus was co-crystallized with U(VI). Cesium exhibited different behavior depending on whether Pu was present. Although a high DF of Cs was obtained in the case of uranyl nitrate solution without Pu, Cs was hardly separated at all from the UNH crystal formed from the dissolver solution of irradiated fast reactor core fuel. It is likely that crystals of a mixed salt of Pu and Cs, Cs2Pu(NO3)6, precipitated from the dissolver solution. Since Ba precipitated as Ba(NO3)2 during the crystallization process, its DF was low after the UNH crystal was washed. On the other hand, Am, Cm, Rb, Sr, Zr, Nb, Ru, Sb, and rare earth elements remained in the mother liquor at the time of U crystallization. Therefore, portions of these elements in the mother liquor that was attached to the surface of the UNH crystal were washed away with HNO3 solution.

Investigation of Uranyl Nitrate Ion Pairs Complexed with Amide Ligands Using Electrospray Ionization Ion Trap Mass Spectrometry and Density Functional Theory

Ion populations formed from electrospray of uranyl nitrate solutions containing different amides vary depending on ligand nucleophilicity and steric crowding at the metal center. The most abundant species were ion pair complexes having the general formula [UO(2)(NO(3))(amide)(n=2,3)](+); however, singly charged complexes containing the amide conjugate base and reduced uranyl UO(2)(+) were also formed as were several doubly charged species. The formamide experiment produced the greatest diversity of species resulting from weaker amide binding, leading to dissociation and subsequent solvent coordination or metal reduction. Experiments using methyl formamide, dimethyl formamide, acetamide, and methyl acetamide produced ion pair and doubly charged complexes that were more abundant and less abundant complexes containing solvent or reduced uranyl. This pattern is reversed in the dimethylacetamide experiment, which displayed lower abundance doubly charged complexes, but augmented reduced uranyl complexes. DFT investigations of the tris-amide ion pair complexes showed that interligand repulsion distorts the amide ligands out of the uranyl equatorial plane and that complex stabilities do not increase with increasing amide nucleophilicity. Elimination of an amide ligand largely relieves the interligand repulsion, and the remaining amide ligands become closely aligned with the equatorial plane in the structures of the bis-amide ligands. The studies show that the phenomenological distribution of coordination complexes in a metal-ligand electrospray experiment is a function of both ligand nucleophilicity and interligand repulsion and that the latter factor begins exerting influence even in the case of relatively small ligands like the substituted methyl-formamide and methyl-acetamide ligands.

Calculation of the (α, n) Emission from Plutonium Nitrate and Plutonium Uranyl Nitrate Solutions

Neutron coincidence measurements of plutonium samples with uncertainties <0.5% could reduce the amount of costly destructive analysis required for nuclear material accountancy in plutonium handling plants. The ratio of (α, n) emission to spontaneous fission neutron emission, α, of plutonium samples is important to the interpretation of neutron coincidence measurements. When the “known alpha” analysis method is used, an error on the α value propagates to approximately the same percentage error on the measured plutonium mass. Molality data of Charrin and the SOURCES code have been used to update the calculation of α for both pure plutonium nitrate solutions and plutonium/uranyl nitrate solutions of different concentrations and acidity. This paper gives equations for the density of the solution as a function of heavy metal concentration and for the α weight factors that can be used in the analysis of neutron coincidence measurements.

Determination of uranium isotopic ratio (235U/238U) using extractive electrospray ionization tandem mass spectrometry

Extractive electrospray ionization (EESI) mass spectrometry (MS), a typical MS platform for detection of organic compounds, has been used for quantitative detection of uranium isotopic ratio (235U/238U) in uranyl nitrate solution samples, which were prepared from various samples including natural water samples, uranium ore samples and soil samples using nitric acid. This method requires minimal sample pretreatment and no high resolution instruments. The typical time for analysis of a single liquid sample was about 5 min since the prepared sample solutions were directly infused into the EESI source without separation or preconcentration. The relative errors for uranium isotope measurements were in the range of 0.21%–0.25% and the corresponding relative standard deviation (RSD) values were 1.54%–1.81% for the ore samples. The results obtained by EESI-MS for direct analysis of 5 soil samples were validated by using ICP-MS. These findings confirmed that EESI-MS can be employed for quantitative measurement inorganic compounds present in a complex matrix. The fast detection of uranium isotopes shows potential applications of EESI-MS in nuclear research laboratories and the nuclear energy industry.

A NEW FORMULATION CONTAINING CALIXARENE MOLECULES AS AN EMERGENCY TREATMENT OF URANIUM SKIN CONTAMINATION

Cutaneous contamination represents the second highest contamination pathway in the nuclear industry. Despite that the entry of actinides such as uranium into the body through intact or wounded skin can induce a high internal exposure, no specific emergency treatment for cutaneous contamination exists. In the present work, an innovative formulation dedicated to uranium skin decontamination was developed. The galenic form consists in an oil-in-water nanoemulsion, which contains a tricarboxylic calixarene known for its high uranium affinity and selectivity. The physicochemical characterization of this topical form revealed that calixarene molecules are located at the surface of the dispersed oil droplets of the nanoemulsion, being thus potentially available for uranium chelation. It was demonstrated in preliminary in vitro experiments by using an adapted ultrafiltration method that the calixarene nanoemulsion was able to extract and retain more than 80% of uranium from an aqueous uranyl nitrate contamination solution. First ex vivo experiments carried out in Franz diffusion cells on pig ear skin explants during 24 h showed that the immediate application of the calixarene nanoemulsion on a skin contaminated by a uranyl nitrate solution allowed a uranium transcutaneous diffusion decrease of about 98% through intact and excoriated skins. The calixarene nanoemulsion developed in this study thus seems to be an efficient emergency system for uranium skin decontamination.

PHYSICAL CHARACTERISTICS OF FLY ASH NANO PARTICLES AS BACKFILL ON THE RADIOACTIVE WASTE REPOSITORY

Characterization of fly ash nano particles as backfilled material candidate in the radioactive waste repository has been done. The objective of this research is to determine the permeability (K) and migration rate (Vr) of uranium in the backfilled material of fly ash, zeolite, and zeolite + fly ash mixtures. The experiment was carried out by the fixed bed method in a column contains fly ash, or zeolite, or zeolite+fly ash mixtures. It was filled with the saturated water and was flown by uranyl nitrate solution of 500 ppm as the simulated uranium. The uranium effluents was sampled in every 15 minutes and it was analyzed using spectrometer. The concentration of which represented as Ct and by using concentration profile of Co/Ct, then Vr of uranium in the backfilled material can be determined. The experiment result showed that ≤ 38 mm of fly ash particles sizes could improve the characteristic feature of 196 mm of zeolite sizes as backfilled material with the decreasing permeability values from Kzeolite = 4.06x10-3 cm/s to Kzeolite+fly ash = 5.00x10-5 cm/s and the decreasing of the migration rate from Vr zeolite = 1.65x10-4 cm/s to Vr zeolite+fly ash = 2.91x10-6 cm/second. Keywords: fly ash, zeolite, backfill materials, uranium migration, radioactive waste repository

THE USE OF FLY ASH AND ZEOLITE MIXTURE AS BACKFILL MATERIAL IN THE RADIOACTIVE WASTE REPOSITORY

An investigation of the contribution of fly ash in the fly ash-zeolite mixture as the backfill material on the shallow land burial of radioactive waste has been done. The experiment objective is to know the effect of zeolite particle size and fly ash-zeolite weight ratio on physical properties such as permeability (K) and dispersion characteristic such as effective dispersion coefficient (De) in the fly ash-zeolite form as backfill material. The experiment was carried out by the fixed bed method in the column filled by the fly ash-zeolite mixture with a fly ash-to-zeolite weight percent ratio of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 in the water saturated condition flown by uranyl nitrate solution at concentration (Co) of 500 ppm. The concentration of uranium in the effluents in interval 15 minutes represented as Ct was analyzed by spectrophotometer, then using Co and Ct, data effective dispersion coefficient (De) in the backfill material were determined. The experiment data showed that -400 mesh fly ash and -70+80 mesh zeolite on fly ash-to-zeolite with weight percent ratio of 40/60 with K = 5.00x10-5cm/second and De = 1.11.10-5 cm2/second can be used as backfill material. Keywords: backfill material, fly ash, radioactive waste, zeolite

DISPERSION AND SORPTION CHARACTERISTICS OF URANIUM IN THE ZEOLITE-QUARTZ MIXTURE AS BACKFILL MATERIAL IN THE RADIOACTIVE WASTE REPOSITORY

The experiment of sorption and dispersion characteristics of uranium in the zeolite-quartz mixture as candidate of raw material of backfill material in the radioactive waste repository has been performed. The objective is to know the effect of zeolite and quartz grain size on the zeolite-to-quartz weight ratio that gives porosity (ε), permeability (K), and dispersivity (α) of uranium in the zeolite-quartz mixture as backfill material. The experiment was carried out by fixed bed method in the column filled by the zeolite-quartz mixture with zeolite-to-quartz weight percent ratio of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 wt. % in the water saturated condition flowed by uranyl nitrate solution of 500 ppm concentration (Co) as uranium simulation which was leached from immobilized radioactive waste in the repository. The concentration of uranium in the effluents represented as Ct were analyzed by spectrophotometer Corning Colorimeter 253 every 15 minutes, then using Co and Ct uranium dispersivity (α) in the backfill material was determined. The experiment data shown that 0.196 mm particle size of zeolite and 0.116 mm particle size of quartz on the zeolite-to-quartz weight ratio of 60/40 wt. % with ε = 0.678, K = 3.345x10-4 cm/second, and α = 0.759 cm can be proposed as candidate of raw material of backfill material in the radioactive waste repository. Keywords: backfill material, quartz, radioactive waste, zeolite

BENTONITE-QUARTZ SAND AS THE BACKFILL MATERIALS ON THE RADIOACTIVE WASTE REPOSITORY

An investigation of the contribution of quartz sand in the bentonite mixture as the backfill materials on the shallow land burial of radioactive waste has been done. The experiment objective is to determine the effect of quartz sand in a bentonite mixture with bentonite particle sizes of -20+40, -40+60, and -60+80 mesh on the retardation factor and the uranium dispersion in the simulation of uranium migration in the backfill materials. The experiment was carried out by the fixed bed method in the column filled by the bentonite mixture with a bentonite-to-quartz sand weight percent ratio of 0/100, 25/75, 50/50, 75/25, and 100/0 on the water saturated condition flown by uranyl nitrate solution at concentration (Co) of 500 ppm. The concentration of uranium in the effluents in interval 15 minutes represented as Ct was analyzed by spectrophotometer, then using Co and Ct, retardation factor (R) and dispersivity () were determined. The experiment data showed that the bentonite of -60+80 mesh and the quartz sand of -20+40 mesh on bentonite-to-quartz sand with weight percent ratio of 50/50 gave the highest retardation factor and dispersivity of 18.37 and 0.0363 cm, respectively. Keywords: bentonite, quartz sand, backfill materials, radioactive waste

Decomposition of uranyl nitrate in the silica gel matrix under the action of microwave radiation

Decomposition of aqueous solutions of uranyl nitrate in a matrix of granulated silica gel of KSKG grade under the action of microwave radiation (MWR) was studied. Microwave irradiation leads not only to formation of solid decomposition products UO3, UO2(OH)NO3, and their hydrates in pores of KSKG granules, but also to accumulation of gaseous NOx and H2O. The presence of NOx in KSKG pores leads to HNO3 formation in the course of washing of sorbent granules with water. This prevents hydrolysis of uranyl nitrate and formation of UO2(OH)2·H2O in KSKG pores. Washout of uranium with water and HClO4 solutions from the KSKG fraction containing products of decomposition of 2 and 10 g of the initial UO2(NO3)2·6H2O under the action of MWR (hereinafter denoted as KSKG-P-I) was studied. Upon ∼7-day contact of the solid and liquid phases at the total ratio S : L = 1 : 20, from 5 to 14% of U passes into the aqueous phase from KSKG-P-I samples obtained in experiments with 10 and 2 g of UO2(NO3)2·6H2O, respectively. In the course of repeated treatments of KSKG-P-I with water, pH of the wash water increased from 3 to 6, owing to removal of NOx from KSKG pores. Then an insoluble phase of uranyl hydroxide UO2(OH)2·H2O, which can also be presented as hydroxylated uranium trioxide UO3·2H2O, is gradually formed from the solution obtained by treatment of KSKG-P-I with water. On treatment of KSKG-P-I with HClO4 solutions (pH 1–2), virtually all uranium species formed by MWR treatment of aqueous uranyl nitrate solutions in KSKG matrix dissolve (at a contact time of the solid and liquid phases of ∼21 days, the amount of U that passed into HClO4 solutions is ∼90%). The amount of the U form that is not extracted with HClO4 solutions and remains in KSKG granules is ∼12% of its initial amount. X-ray phase analysis suggests that the uranium species remaining in KSKG are silicate compounds formed by sorbent saturation with a uranyl nitrate solution and subsequent MWR treatment.

Formation of molybdenum and zirconium precipitates in concentrated uranyl nitrate solutions

The influence of the keeping time, temperature, and concentrations of U and HNO3 on the residual content of Mo and Zr in the solution was examined. An increase in the preliminary keeping time favors more complete precipitation. An increase in the uranium concentration to 800 g l−1 and/or in the temperature to 105°C leads to growing formation of precipitates of molybdic acid (in the absence of Zr) or zirconium molybdate. The presence of zirconium promotes, and an increase in the acidity impedes the process.

Characterization of alumina-supported uranium oxide catalysts in methane oxidation

Synthesis of alumina-supported uranium oxide catalysts and their characterization in methane oxidation have been performed. The catalysts containing 5% U were prepared by impregnation, solid-phase synthesis and mechanical mixing techniques. Samples were studied using BET, XRD, HRTEM and XPS methods. It was shown that physicochemical properties and activity of the catalysts depend on the preparation procedure and calcination temperature and are determined by the extent of the interaction between active component and support. Most active are the catalysts prepared by impregnation of alumina with an aqueous solution of uranyl nitrate. It was found that the 5%U/Al 2O 3 catalyst calcined at 1000 °C has the highest catalytic activity, and it is explained by the formation of highly active nanodispersed state of uranium on the surface of the support.

Energy dispersive X-ray fluorescence determination of cadmium in uranium matrix using Cd Kα line excited by continuum

An energy dispersive X-ray fluorescence method for determination of cadmium (Cd) in uranium (U) matrix using continuum source of excitation was developed. Calibration and sample solutions of cadmium, with and without uranium were prepared by mixing different volumes of standard solutions of cadmium and uranyl nitrate, both prepared in suprapure nitric acid. The concentration of Cd in calibration solutions and samples was in the range of 6 to 90 µg/mL whereas the concentration of Cd with respect to U ranged from 90 to 700 µg/g of U. From the calibration solutions and samples containing uranium, the major matrix uranium was selectively extracted using 30% tri-n-butyl phosphate in dodecane. Fixed volumes (1.5 mL) of aqueous phases thus obtained were taken directly in specially designed in-house fabricated leak proof Perspex sample cells for the energy dispersive X-ray fluorescence measurements and calibration plots were made by plotting Cd Kα intensity against respective Cd concentration. For the calibration solutions not having uranium, the energy dispersive X-ray fluorescence spectra were measured without any extraction and Cd calibration plots were made accordingly. The results obtained showed a precision of 2% (1 σ) and the results deviated from the expected values by < 4% on average.

Amine-templated uranyl selenates with chiral [(UO2)2(SeO4)3(H2O)]2– layers: topology, isomerism, structural relationships

Abstract Eleven new amine-templated uranyl selenate hydrates have been prepared by evaporation from aqueous solution of uranyl nitrate, selenic acid and the respective amines. The structures of the compounds have been solved by direct methods and refined using least-squares techniques. Each structure is based upon [(UO2)2(SeO4)3(H2O)]2– layers of corner-sharing UO7 pentagonal bipyramids and SeO4 tetrahedra. The layers are based upon 4- and 6-membered rings arranged in different fashion. In topology I, 6-membered rings form edge-sharing chains, whereas, in topology II, they form corner-sharing chains. Layers with the topology II exist in two geometrical isomers that differ in the system of ‘up’ and ‘down’ orientations of tetrahedra relative to the plane of the layer. There are two isomers, one of which is chiral and the other is achiral. The layers with the topology II are chiral. Chirality is induced by the combination of orientations of tetrahedra and direction of the U → H2O bond. The analysis of the relationships between composition and shape of amine molecules and layer topology reveals two important regularities. 1. Aliphatic components of amine molecules tend to associate with 6-MRs of the inorganic layers. 2. Molecules with longer and spacious aliphatic components favor formation of the layers with topology II, whereas those with shorter aliphatic components prefer layers with the topology I.

Benchmark Critical Experiments and FP Worth Evaluation for a Heterogeneous System of Uranium Fuel Rods and Uranium Solution Poisoned with Pseudo-Fission-Product Elements

A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark data concerning the dissolving process in a reprocessing plant. Focusing on the introduction of the burn-up credit, critical mass measurement was conducted for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with pseudo-fission-product (FP) elements—samarium, cesium, rhodium, and europium. Fuel rods were arrayed at a 1.5-cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 320 gU/L, and the FP element concentrations were adjusted to be equivalent to that in a burn-up of about 30GWd/t. The result provides basic experimental data for validation of computational methods to evaluate the reactivity effect of each FP element, as well as benchmark criticality data for validation of neutron multiplication factor calculation for heterogeneous systems of spent fuel. In this report, the details of the experiments and benchmark models will be presented as well as the procedure and the result of separate reactivity worth evaluation for each FP element. The experimental results and the computational evaluation results will also be compared.