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The fission product release and transport during the accident of Fukushima Daiichi Unit 3 was analysed in the present paper as a complement to the thermal-hydraulic analysis presented in the first part of the present study, where three sequences obtained with MELCOR 2.1. were evaluated. The calculation results were compared against the available data for fission products: i) dose rate measurements in the containment drywell and suppression chamber, ii) the environmental release data by reverse calculations performed with WSPEEDI-II, and iii) estimations of water releases to the turbine and reactor buildings. From the fission product analysis perspective, the best results were obtained with case 2 sequence, further confirming the results obtained in the first part of the analysis. However, the comparison of the analysis results with the dose rate measurements in the drywell indicates that the reactor pressure vessel failure took place 2 h earlier than predicted in the case 2, i.e., at ca. 62 h after scram.

A Filtered Containment Venting System (FCVS) is a safety system incorporated in Nuclear Power Plants (NPPs) to filter the steam and noncondensable gases released from the containment during a severe accident scenario to avoid overpressure in the containment structures while isolating most radioactive products inside. This paper presents the experimental and numerical simulation results from confirmatory tests performed in the VEFITA (Venting Filter Assessment) test facility at the Paul Scherrer Institut (PSI) in Switzerland. This facility is designed to evaluate the performance of an FCVS by assessing key phenomena and parameters such as internal recirculation patterns, changes in inventory, level swell, and the Decontamination Factor (DF) of aerosol and iodine species in simulated venting processes. The current study focuses on the level swell (gas holdup) phenomenon. The level swell during gas injection is particularly of interest since it is associated with the heights of gases and liquids in the venting vessel and thus affects the design and operation of an FCVS. The experiment is performed by injecting the noncondensable gas and steam at various mass flow rates through a sparger nozzle to the vessel, which is partially filled with washing fluid as a bubble column. The level swell and other physical variables are recorded and analyzed. System codes (RELAP5 and TRACE) input decks are developed for this facility, and experimental transients are simulated. The code simulation results, as well as previous physical gas holdup models based on homogeneous-heterogeneous regime transition, are compared with current experimental data to study the level swell in this FCVS.

In this study, the aerosol release from the silver-indium-cadmium control rods during the air-ingress phase of the QUENCH-18 bundle test was investigated both experimentally and by using simulation tools. During the QUENCH-18 test, the aerosol mass size distribution was measured using two Berner low pressure impactors, aerosol samples were collected on filters, and the particle number size distribution was monitored online with an Electrical Low Pressure Impactor (ELPI). After the experiment, the elemental composition of the aerosol samples was determined using inductively coupled plasma optical emission spectrometry. The first released aerosol particles contained mostly cadmium and afterwards, also silver was released with cadmium in significant amounts, with the indium concentration staying at lower level. The MELCOR 2.2 code was used to simulate the aerosol deposition in the QUENCH facility in order to assess the total release of the control rod materials during the test. Based on the measurement and simulation results, the total releases of cadmium, silver and indium during the test were determined to be 9.0 g, 6.5 g, and 1.2 g, respectively. A thermodynamic equilibrium model was used to investigate the speciation of the control rod material in the test conditions. The calculations indicate that the aerosol particles measured during the first aerosol release consisted mostly of cadmium oxide and during the main release phase, the main aerosol release consisted mainly of Cd and Ag compounds with smaller amounts of indium.

The accident sequence of Fukushima Unit 3 was analysed with MELCOR 2.1. Hundreds of calculations have been performed and three plausible scenarios which predicted remarkably well the main signatures (i.e. pressure in RPV and containment and containment water level) have been selected and studied in the present analysis. The main signatures, namely pressure in the RPV and drywell and water levels were satisfactorily reproduced by the three proposed plausible scenarios. Major uncertainties concerning the time of RPV failure, possibility of MCCI as well as the transport of hydrogen responsible for the explosion of unit 3 and unit 4 were addressed in this study. The study was complemented with plant observations such as containment inspections, muon measurements and photographs taken during the accident. The results point out that the plausible sequence lie between case 1 and case 2 presented in this study. The state of the core after 350 h seems to be in agreement with our case 2, where ca. 64 tons of debris and stainless steel structures remained in the lower head and ca. 70 tons of debris were ejected to the pedestal. Taking into account the limitation in the modelling of the failure modes of a BWR, this estimation is very uncertain. Further discussion will be provided in the second part of the paper where the fission product release and transport will be studied and compared against available source term calculations.

A significant amount of radioactive iodine was release to the environment during the accident in the Fukushima Daiichi nuclear power plant. In this numerical study, we used the IMPAIR 2.2 code to simulate the iodine release during containment venting from the suppression chamber gas space and drywell leakage from unit 3 of the power plant. The IMPAIR code was coupled to the Dakota software and uncertainty quantification study and sensitivity analysis of the simulations was performed. The results indicated that both molecular iodine and organic methyl iodide were formed in the water volumes of the plant, where it was transferred to the gas phase of the modeled structures. The sensitivity analysis indicated that the water pH was the most sensitive parameter on the gaseous iodine formation and release during the venting or leakage periods. Additionally, the water pH was shown to affect the speciation of the gas phase iodine. Overall, the simulations performed in this work showed that iodine volatilized from the contaminated water phase in the suppression chamber and the drywell of Fukushima Daiichi unit 3 have an important contribution to the iodine source term to the environment.

International audience; Spent fuel pools (SFPs) are large structures equipped with storage racks designed to temporarily storeirradiated nuclear fuel removed from the reactor. SFP severe accidents have long been considered ashighly improbable since the accident progression is slow (in comparison with reactor core accidents)and let time to corrective operator actions. However, the accident at the Fukushima Dai-ichi NuclearPower Plants has highlighted the vulnerability of nuclear fuels that are stored in SFPs in case of prolongedloss-of-cooling accidents and consequently renewed international interest in the safety of SFPs. In thiscontext, the AIR-SFP project, funded by the Euratom 7th FP in the frame of the NUGENIA+ project, waslaunched in May 2015 with 15 participants. One of the objectives was to assess the applicability ofSevere Accident (SA) codes, which were initially developed for reactor applications, to the calculationof transients in SFPs. To reach this objective, a benchmark, including a criticality risk assessment, was carriedout. The degradation progression was computed by 14 participants with 6 different SA codes and 5have participated to the criticality risk assessment. Main results are presented as well as conclusions thathave been drawn concerning SA codes readiness to address these ‘‘beyond-scope” scenarios. 2018 Elsevier Ltd. All rights reserved.

In order to develop an iodine scrubbing model for nuclear severe accidents, experimental data on filtration efficiency of a wet scrubber for gaseous molecular iodine are collected to improve understanding of the basic mechanisms involved in iodine chemical scrubbing. A bubble column reactor 1.5 m high and 0.2 m in diameter equipped with an injection nozzle and a bubble breaker is used to investigate retention under relevant flow regimes. Scrubber is loaded with a scrubbing solution containing sodium hydroxide and sodium thiosulphate, no irradiation is provided. Retention efficiency is found to be strongly dependent on flow regime and on residence time as models predict. It was found that jet-like injection has a strong effect on filtration efficiency introducing a significant dependence on iodine initial concentration in gas phase believed to be due to nozzle hydrodynamics.

The treatment of chemistry and thermodynamics in the integral severe accident codes is typically limited. A more accurate treatment of the chemistry during the severe accident modelling is, therefore, of great interest. For this purpose, the work is focused on the development of coupling of the MELCOR code with chemical thermodynamic calculations using GEMS codes and HERACLES database. Developed coupling between the two codes, called cGEMS, allowed for the improved thermodynamic description of the fission product release from the nuclear fuel under severe accident conditions. VERDON-1 test was selected as an experimental reference for the simulations. Experimental release behaviour of Mo, Cs and Ba observed in VERDON-1 experiment was reproduced by the developed coupled code. Performed simulations provided detailed information about the fission product speciation at different redox conditions. The obtained information and the developed code provides a more accurate description of the fission product behaviour and release during severe accidents.

Much is still not known about the end-state of core materials in each of the units at Fukushima Daiichi Nuclear Power Station (Daiichi) that were operating on March 11, 2011. The Nuclear Energy Agency of the Organization for Economic Development has launched the Preparatory Study on Analysis of Fuel Debris (PreADES) project as a first step to reduce some of these uncertainties. As part of the PreADES Task 1, relevant information was reviewed to confirm the accuracy of graphical depictions of the debris endstates at the damaged Daiichi units, which provides a basis for suggesting future debris examinations. Two activities have been completed within the PreADES Task 1. First, relevant knowledge from severe accidents at the Three Mile Island Unit 2 and the Chernobyl Nuclear Power Plant Unit 4 was reviewed, along with results from prototypic tests and hot cell examinations, to glean insights that may inform future decommissioning activities at Daiichi. Second, the current debris endstate diagrams for the damaged reactors at Daiichi were reviewed to confirm that they incorporate relevant knowledge from plant observations and from severe accident code analyses of the BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) 1 and 2 projects. This paper highlights Task 1 insights, which have the potential to not only inform future Decontamination and Decommissioning activities at Daiichi but also provide important perspectives for severe accident analyses and management, particularly regarding the long-term management of a damaged nuclear site following a severe accident.

Integral source-term analyses are performed using MELCOR for a PWR Station Blackout (SBO) sequence leading to induced steam generator tube rupture (SGTR). In the absence of any mitigation measures, such a sequence can result in a containment bypass where the radioactive materials can be released directly to the environment. In some SGTR scenarios flooding of the faulted SG secondary side with water can mitigate the accident escalation and also the release of aerosol-borne and volatile radioactive materials. Data on the efficiency of aerosol scrubbing in an SG tube bundle were obtained in the international ARTIST project. In this paper ARTIST data are used directly with parametric MELCOR analyses of a mitigated SGTR sequence to provide more realistic estimates of the releases to environment in such a type of scenario or similar. Comparison is made with predictions using the default scrubbing model in MELCOR, as a representative of the aerosol scrubbing models in current integral codes. Specifically, simulations are performed for an unmitigated sequence and 2 cases where the SG secondary was refilled at different times after the tube rupture. The results, reflecting the experimental observations from ARTIST, demonstrate enhanced aerosol retention in the highly turbulent two-phase flow conditions caused by the complex geometry of the SG secondary side. This effect is not captured by any of the models currently available. The underlying physics remains only partly understood, indicating need for further studies to support a more mechanistic treatment of the retention process.

In experimental prototypes of pebble-bed reactors, significant quantities of graphite dust have been observed due to rubbing between pebbles as they flow through the core. At the typical operating conditions in these reactors, which feature high temperatures, pressures, and a helium atmosphere, limited data is available on the frictional properties of the pebble surfaces, and as a result, a conceptual design of a scaled-down version of a pebble-bed reactor has been proposed to investigate this issue in detail. However, this raises general questions about how the granular flow in a scaled facility will emulate that in a full-size reactor. To address this, simulations of granular flow in pebble-bed reactors using the discrete-element method (DEM) have been carried out in a full-size geometry (using 440,000 pebbles) and compared to those in geometries scaled down by factors of 3:1 and 6:1. Differences in velocity profiles, pebble ordering, pebble wear, and stresses are examined, and the effect of friction is discussed. The results show complex behavior due to discrete pebble packing effects, although several simple scaling rules can be derived.

Spectroscopy and imaging of condensed matter have benefited greatly from the availability of intense X-ray beams from synchrotron sources, both in terms of spatial resolution and of elemental specificity. The advent of the X-ray free electron laser (X-ray FEL) provides the additional features of ultra-short pulses and high transverse coherence, which greatly expand possibilities to study dynamic processes and to image non-crystalline materials. The proposed SwissFEL facility at the Paul Scherrer Institute is one of at present four X-ray FEL projects worldwide and is scheduled to go into operation in the year 2017. This article describes a selection of problems in nuclear materials science and technology that would directly benefit from this and similar X-ray FEL sources. X-ray FEL-based experiments are proposed to be conducted on nuclear energy-related materials using single-shot X-ray spectroscopy, coherent X-ray scattering and/or X-ray photon correlation spectroscopy in order to address relevant scientific questions such as the evolution in time of the irradiation-induced damage processes, the deformation processes in nuclear materials, the ion diffusion processes in the barrier systems of geological repositories, the boiling heat transfer in nuclear reactors, as well as the structural characterization of graphite dust in advanced nuclear reactors and clay colloid aggregates in the groundwater near a radioactive waste repository.

A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet–bundle interactions close to the tube break. Droplets were entrained from the water surface with high gas flow rates carrying aerosol particles with them. However, compared to particle retention in the water close to the tube break, the effect of droplet entrainment on particle transport was small.

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO2 agglomerates showed that for high velocities in the range of 100–350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO2 aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions. In this investigation, we measured TiO2 aerosol de-agglomeration in the tube with very high flow velocities with two different TiO2 aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone. The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO2 aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter Dc = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence. It has to be noted that the particles used in this investigation were TiO2 agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.

Monodisperse, fine aerosols are needed in many applications: filter testing, experiments for testing models, and aerosol instrument calibration, among others. Usually, monodisperse fine aerosols are generated in very low concentrations, or mass flow rates, in the laboratory scale. In this work, we needed to generate aerosols with higher mass flow rate than typically available by the laboratory-scale methods, such as atomizers, nebulizers, ultrasonic generators, vibrating orifice generators, and condensation generators. Therefore, we constructed a fluidized bed aerosol generator to achieve particle mass flow rates in the range of 15–100 g/h. Monodisperse, spherical SiO2 particles of two sizes with geometrical diameters of 1.0 and 2.6 µm were used in the aerosol generator. The aerosol generator was used at both atmospheric pressure, and at high pressures up to 5 bar (abs). The particle size, mass concentration and the net average particle charge were measured after mixing the aerosol with nitrogen. The particle size distributions with both particle sizes were monodisperse, and no particle agglomerates were entrained from the fluidized bed. The behavior of the fluidized bed generator was found to be markedly different with the two particle sizes in regard to particle concentration, presumably due to different particle charging inside the generator. After determining the net average charge of the particles, an ion source Kr-85 was used to reduce the charge of the particles. This was found to be effective in neutralizing the particles.

In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver–indium–cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver–indium–cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ∼ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ∼ 1650 K, followed by a relatively steady aerosol release until core cooling by quench when the on-line measurements were stopped. Cd was released first from the control rod, followed by In, and finally, by Ag. The particle size distributions were bimodal indicating two aerosol formation mechanisms, evaporation followed by nucleation and condensation, as well as droplet and fragment generation. Generally, release is modelled as evaporation from molten regions of control rod materials. Clearly, results of this investigation give evidence of contribution by entrainment of droplets and fragmented material.

In the case of a hypothetical severe accident in a Pressurized Water Reactor (PWR), Silver-Indium-Cadmium (SIC) control rod failure occurs early during the sequence. Release of absorber melt could induce early fuel rod degradation by interaction of molten SIC alloy with fuel rod cladding, and the absorber materials could interact with the fission products, affecting significantly their speciation and transport in the primary circuit as well as their behaviour in the containment. This paper summarises the experimental and modelling progress made on this topic within SARNET over the whole project. Following a review of the status of knowledge, including the modelling in the main severe accident codes (ATHLET-CD, MAAP4, SCDAP, MELCOR, ASTEC), detailed calculations of the specific EMAIC and integral PHEBUS FPT2 experiments were performed. Accurate calculation of vapour pressure of the molten absorber elements is needed, requiring reliable values of the activity coefficients. The importance of accurate reproduction of the control rod degradation was shown, with the behaviour of absorber elements at rupture being critical as well as the thermodynamic data and speciation of the system Ag-In-Cd-Zr-H-O. The QUENCH-13 bundle experiment (FZK), conducted with a realistic integral geometry composed of 20 electrical heated rod simulators and one central SIC absorber rod, has helped to characterize the behaviour of absorber elements at the time of rod rupture, with measurements of the SIC release, using impactors (AEKI) and electrical low-pressure impactor and Berner low-pressure impactors (PSI). Coordinated pre and post-test calculations using SCDAP/RELAP5 based codes (PSI), MAAP4 (EDF), ATHLET-CD (GRS), ASTEC (IRSN) helped in defining the test and in its interpretation. Before this experiment, five tests were performed on small-scale SIC control rod samples using different designs and conditions. They helped in defining the conditions for the QUENCH-13 experiment. Five additional tests on similar small-scale samples are foreseen to help interpretation of the QUENCH-13 results. In QUENCH-13 the on-line aerosol measurements with electrical low-pressure impactors indicated control rod failure in the range 1550-1600 K; the test was terminated later at 1813 K by water reflood. Analysis of aerosols measured at sample points in the off-gas line showed significant Cd and In transport after rod failure with a smaller amount of transported Ag. Relocated SIC melt in the form of rivulets was detected in the lower part of the bundle. Investigation of SIC material properties (solidus, liquidus) by further analysis of data from QUENCH-13 is also presented. In parallel, an exhaustive review of activity coefficients of the elements in the SIC melt, including the effect of Zr was began (IRSN with the CNRS Marseille). © 2009 Elsevier Ltd. All rights reserved.

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO 2 pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H 2 during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m). Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of ∼2300 K (compared to ∼1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.

Fine particle formation in wood combustion was studied in a laboratory scale laminar flow reactor at various flue gas chlorine and sulphur concentrations. Aerosol samples were quenched at around 850 °C using a porous tube diluter. Fine particle number concentrations, mass concentrations, size distributions and chemical compositions were measured. In addition, flue gas composition, including SO2 and HCl, was monitored. Experimental results were interpreted by thermodynamic equilibrium calculations.Addition of HCl clearly raised fine particle mass concentration (PM1.0) which was because of increased release of ash-forming material to fine particles. Especially the release of K, Na, Zn and Cd to fine particles increased. These species form chlorides which apparently increases their volatilization from the fuel. When a sufficient amount of SO2 was supplied in a chlorine rich combustion (S/Cl molar ratio from 4.7 to 7.5), most of the HCl stayed in the gas phase, release of ash-forming elements decreased and also fine particle concentrations dropped significantly. The sulphation of alkali metals is suggested to play a key role in the observed decrease in the fine particle concentration. It seems that the formation of sulphates leads to alkali metal retention in the coarse particle fraction.

Waste-to-energy applications are increasingly being used to simultaneously reduce the amount of waste and produce electricity and heat. As waste materials typically contain high concentrations of impurities that are transformed during combustion, the flue gases need to be cleaned efficiently in order to avoid harmful emissions to the environment. In this investigation, we determined experimentally fly ash particle characteristics – particle size, composition and concentration – as well as particle and trace element emissions during waste combustion in two full-scale plants using different combustion technologies but similar gas cleaning technology. The two combustion plants were a grate-fired boiler, and a circulating fluidized bed (CFB) boiler. Both boilers used selective non-catalytic reduction (SNCR) for NO x control together with novel integrated flue gas desulfurization (NID) for flue gas cleaning. The trace elements included in the investigation were As, Cd, Co, Cr, Cu, Hg, Mn, Ni, Pb, Sb, Tl, and V. Fine fly ash particles were formed from the gas phase ash-forming species by nucleation and condensation. Concentration of particles smaller than 1.0 μm by diameter (PM1.0) at grate-fired plant was 1.0–1.4 g/Nm 3 , approximately four times the PM1.0 concentration at CFB combustion, 0.25–0.31 g/Nm 3 , as determined upstream of the flue gas cleaning system NID. The average total fly ash mass concentration was higher at CFB combustion than at grate firing with 4.6 g/m 3 and 1.4 g/m 3 , respectively. Particle and trace element emissions were very low from both grate-fired and fluidized bed plants. Fabric filter particle collection efficiency was 99.99% by mass at both plants. All the measured emissions were clearly below the limit values set by European Waste Incineration Directive (WID).

Trace element behaviour in atmospheric circulating fluidized-bed combustion (CFBC) of Venezuelan bituminous coal was studied by determining particle size distributions in the CFBC flue gas. The size distributions of calcium, iron, aluminium, and 21 trace elements, Sc, V, Cr, Mn, Co, Ni, Zn, Ga, As, Se, Sr, Cd, Sb, Cs, Ba, La, Ce, Sm, Lu, Pb, and Th, in the size range 0.01–70 μm, were determined by collecting aerosols with a low-pressure impactor-cyclone sampling train from the flue gases of an 80-MW(th) CFBC boiler upstream of the electrostatic precipitator. The collected samples were analyzed gravimetrically and with instrumental neutron activation analysis (INAA), particle-induced x-ray emission analysis (PIXE), and inductively coupled plasma mass spectrometry (ICP-MS). The number size distributions of the aerosols were determined with a differential electrical mobility method in the size range 0.01–0.8 μm. In the ultrafine particle mode, i.e., D p

In order to determine penetration curve of the electrostatic precipitator (ESP) as a function of aerosol particle diameter in the range of 10–1000 nm measurement series were carried out in real scale power plant conditions. Differential mobility particle sizing (DMPS) system was used to measure the particle mobility distributions before and after ESP. MICRON -algorithm (constrained regularization) was used to invert mobility distribution to the corresponding number distributions. Penetration curve was calculated from the measured number distributions.

Aerosol formation in pulverized coal combustion have been studied experimentally at the real scale power plant. Combustion aerosol mass and number size distributions have been determined, when burning bituminous coal from Poland. Mass size distributions have been measured by low pressure impactor and number distributions by differential electrical mobility (DMA) method.