[en] Steam generator tube rupture (SGTR) events have a low probability of occurrence, but if they occur, they might lead to the escape of radio nuclides from the primary circuit to the secondary side, and subsequent emissions to the environment bypassing the containment. In western pressurized water reactors (PWR), SGTR is a design basis accident (DBA) and the plants are designed to cope with it. However, a particular safety challenge arises from an SGTR in combination with other failures resulting in a core melt, and radioactive fission product release from the core. The source term into the environment is determined by the release of fission products from the core and their retention in their release path through the steam generator. Based on the need for aerosol retention data during an SGTR, EU-SGTR project was carried out in 2000-2002. It was followed by international collaboration projects ARTIST (Aerosol Trapping In a Steam Generator, 2003-2007) and ARTIST II (2008-2011), coordinated by the Paul Scherrer Institut (PSI). In these projects, aerosol retention in the steam generator was investigated both experimentally and analytically. Integral experiments for aerosol retention in the ARTIST facility, which is a scaled-down model of the FRAMATOME 33/19 type steam generator, showed significant retention of aerosols in the steam generator during SGTR. In this paper, we will briefly summarize some of the main conclusions of the ARTIST projects. The main emphasis of this work is the extension of the research by estimating the offsite consequences (doses) that could be expected from an SGTR, and the reduction in these consequences due to the acquired understanding of the behavior of the fission product transport and retention in the secondary side of a PWR steam generator from the ARTIST projects. MELCOR calculations have been performed for such an accident and the results for releases of a Spontaneous Steam Generator Tube Rupture (S-SGTR) accident are used for the main part of the work. Assessment of consequences is performed using the USNRC code MACCS2. (authors)

[en] The Paul Scherrer Institut is proposing a large scale, two-pronged experimental program to study a variety of PWR steam generator phenomena which are of high safety relevance. The first project part called 'Reflux', addresses reflux condensation available for the cooling of the reactor during design basis and severe accidents. The second project part, called 'Mixing and recirculation' addresses mixing in the steam generator (hot leg, inlet plenum and gas recirculation in the steam generator tubes) during PWR severe accidents. Inlet plenum mixing is a critical parameter determining the degree of thermal challenge to steam generator tubes during postulated severe accidents, and consequently, the potential for induced tube rupture. In this work, we briefly describe the whole project, but the main emphasis is on the 'Mixing and recirculation' part of the project. We present the design of the experimental facility, and analytical work conducted to investigate the steam generator mixing and recirculation using CFD tools. In addition, the future experimental work will be described. (authors)

[en] Filtered Containment Venting System (FCVS) is a safety system equipped in nuclear power plants to release steam and non-condensable gases from the containment during a severe accident scenario, in order to avoid over-pressurizing the containment, while retaining most radioactive fission products inside the plant. This paper presents the RELAP5/Mod3.3-p3 and TRACE/V5.0-p2 simulations of the experiments performed in the VEFITA (Venting Filter Assessment) test facility at Paul Scherrer Institut (PSI). This experimental program is designed to investigate thermal-hydraulic phenomena in the FCVS such as internal recirculation, water inventory, and level swell. The water level swell during the gas injection is particularly interesting since it would affect the design and operation of the venting filtration system. Test section is a vessel with a 0.58 m diameter and 12 m total height. The lower wet scrubber section is filled with water to a certain level above the gas injection nozzle. Non-condensable gas at various mass flow rates are fed through the injection nozzle into the test section. RELAP5 and TRACE code models are developed for the test facility and several series of experiments are simulated. The system code simulation results, as well as a previous physical model based on homogeneous-heterogeneous regimes transition in bubble columns, are compared with the experimental data. (authors)

[en] Highlights: • Source term analyses in a PWR of mitigated thermally-induced SGTR scenario performed. • Experimental ARTIST program results on aerosol scrubbing efficiency used in analyses. • Results demonstrate enhanced aerosol retention in a flooded steam generator. • High aerosol retention cannot be predicted by current theoretical scrubbing models. - Abstract: 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.

[en] Molten Salt Reactors (MSRs) are rather category of reactors than a single concept. In majority of the concepts the fuel cycle performance profits from the consideration that selected gaseous, volatile, non-soluble fission products could be removed during the operation from the liquid fuel. Many concepts also foresee integration of fuel cleaning, or actually processing, unit in the same complex. Accordingly, the distribution or radiotoxicity in the MSR system may strongly differ from classical reactors with solid fuel. At the same time, the presence of driving forces can be eliminated.

[en] Highlights: • Coupling of a sever accident code with a thermodynamic modelling package. • Improved description of the Mo and Ba species release from the nuclear fuel. • The developed code reproduced fission products release during the VERDON experiment. • The species partitioning and release behaviour depends on the redox conditions. 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.

[en] The containment filtered venting system (CFVS) filters the atmosphere of the containment building and discharges a part of it to the outside environment to prevent containment overpressure during severe accidents. The Korean CFVS has a tank that filters fission products from the containment atmosphere by pool scrubbing, which is the primary decontamination process; however, prediction of its performance has been done based on researches conducted under mild conditions than those of severe accidents. Bubble behavior in a pool is a key parameter of pool scrubbing. Therefore, the bubble behavior in the pool was analyzed under various injection flow rates observed at the venturi nozzles used in the Korean CFVS using a wire-mesh sensor. Based on the experimental results, void fraction model was modified using the existing correlation, and a new bubble size prediction model was developed. The modified void fraction model agreed well with the obtained experimental data. However, the newly developed bubble size prediction model showed different results to those established in previous studies because the venturi nozzle diameter considered in this study was larger than those in previous studies. Therefore, this is the first model that reflects actual design of a venturi scrubbing nozzle

[en] 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.

[en] This MELCOR 2.2 severe accident was used to study two accident scenarios in the Chinese high temperature reactor HTR PM. A model of the reactor core, thermal insulator and metal structures, reactor pressure vessel and the residual heat removal system was created using the information from the open source literature. A steady state calculation was performed and used as an initial condition for the simulation before the accident sequence commenced. In pressurized loss of forced cooling accident the coolant helium flow was decreased to zero and the reactor was shut down. A natural convective flow was established in the pebble bed core due to the high pressure in the primary system and the peak fuel temperature during the accident was 1438 K. In de pressurized loss of forced cooling accident the pressure of the primary system is lost causing the coolant flow to stop and reactor to shut down. The heat was removed from the core mainly by thermal radiation and heat conduction leading to the peak fuel temperature of 1730 K. The results were compared to previously published simulation data obtained using a THERMIX code. The peak fuel temperature results didn’t differ significantly between the both simulations, with the main differences probably caused by the uncertainties in the input parameter used in this study. In the future work, MELCOR code is planned to be used in the simulation of fission product release and transport in the pebble bed high temperature reactors during accident normal operation and in accident conditions. (author)

[en] Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: 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.