Becker Technologies GmbH

Eschborn, Germany

Becker Technologies GmbH

Eschborn, Germany
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Gupta S.,Becker Technologies GmbH | Freitag M.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH
Kerntechnik | Year: 2016

The Thai (Thai = Thermal hydraulics, Hydrogen, Aerosols, Iodine) experimental programme aims to address open questions concerning the behavior of hydrogen, iodine and aerosols in the containment of water cooled reactors. Since its construction in 2000, Thai programme is being performed in the frame of various national projects (sponsored by German Federal Ministry for Economic Affairs and Energy, BMWi) and two international joint projects (under auspices of OECD/NEA). Thai experimental data have been widely used for the validation and further development of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes with 3D capabilities. Selected examples of code benchmark exercises performed based on the Thai data include; hydrogen distribution experiment (ISP-47 and OECD/NEA Thai code benchmark), hydrogen combustion behaviour (ISP-49), hydrogen mitigation by PARs (OECD/NEA Thai-2 code benchmark), iodine/surface interactions, iodine mass transfer, and iodine transport and multi-compartment behaviour (EUSARNET and EU-SARNET2), thermal-hydraulic tests (German CFD-network). In the present paper, a brief overview on the Thai experiments and their role in the containment safety assessment is discussed. © Carl Hanser Verlag, München.

Weber G.,GRS Society for plants and Reactor Safety | Bosland L.,Institute for Radiological Protection and Nuclear Safety | Funke F.,AREVA | Glowa G.,Chalk River Laboratories | Kanzleiter T.,Becker Technologies GmbH
Journal of Engineering for Gas Turbines and Power | Year: 2010

The large-scale iodine test Iod-9 of the German Thermal hydraulics, Hydrogen, Aerosols, Iodine (THAI) program was jointly interpreted by means of post-test analyses within the THAI Circle of the Severe Accident Research NETwork (SARNET)/Work Package 16. In this test, molecular iodine (I2) was injected into the vessel dome of the 60 m3 THAI vessel to observe the evolution of its distribution between water, gas, and surfaces. The main processes addressed in Iod-9 are (a) the mass transfer of I2 between the gas and the two sumps, (b) the iodine transport in the main sump when it is stratified and then mixed, and (c) the I2 adsorption onto, and desorption from, the vessel walls in the presence and absence of wall condensation. The codes applied by the THAI Circle partners were the Accident Source Term Evaluation Code (ASTEC)-IODE (IRSN, Saint Paul Lez Durance, France), Containment Code System (COCOSYS)-Advanced Iodine Model (AIM) (GRS, Garching, Germany), and Library of Iodine Reactions in Containment (LIRIC; AECL, Chalk River, ON, Canada). ASTEC-IODE and the Advanced Iodine Model (AIM) are semi-empirical iodine models integrated in the lumped-parameter codes ASTEC and COCOSYS, respectively. With both codes multicompartment iodine calculations can be performed. LIRIC is a mechanistic iodine model for single stand-alone calculations. The simulation results are compared with each other and with the experimental measurements. Special issues that were encountered during this work were studied in more details: I2 diffusion in the sump water, I 2 reaction with the steel of the vessel wall in gaseous and aqueous phases, and I2 mass transfer from the gas to the sump. Iodine transport and behavior in THAI test Iod-9 are fairly well simulated by ASTEC-IODE, COCOSYS-AIM, and LIRIC in post-test calculations. The measured iodine behavior is well understood and all measured data are found to be consistent. The very slow iodine transport within the stratified main sump was simulated with COCOSYS only, in a qualitative way. Consequently, this work highlighted the need to improve modeling of (a) the wet iodine adsorption and the washdown from the walls, (b) the I2 mass transfer between gas and sump, and (c) the I2 /steel reaction in the gaseous and aqueous phases. In any case, the analysis of the large-scale iodine test Iod-9 has been an important validation step for the codes applied. © 2010 American Society of Mechanical Engineers.

Freitag M.,Becker Technologies GmbH | Schmidt E.,Becker Technologies GmbH | Gupta S.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH
Nuclear Engineering and Design | Year: 2015

Locally enriched hydrogen as in stratification may contribute to early containment failure in the course of severe nuclear reactor accidents. During accident sequences steam might accumulate as well to stratifications which can directly influence the distribution and ignitability of hydrogen mixtures in containments. An international code benchmark including Computational Fluid Dynamics (CFD) and Lumped Parameter (LP) codes was conducted in the frame of the German THAI program. Basis for the benchmark was experiment TH24.3 which investigates the dissolution of a steam layer subject to natural convection in the steam-air atmosphere of the THAI vessel. The test provides validation data for the development of CFD and LP models to simulate the atmosphere in the containment of a nuclear reactor installation. In test TH24.3 saturated steam is injected into the upper third of the vessel forming a stratification layer which is then mixed by a superposed thermal convection. In this paper the simulation benchmark will be evaluated in addition to the general discussion about the experimental transient of test TH24.3. Concerning the steam stratification build-up and dilution of the stratification, the numerical programs showed very different results during the blind evaluation phase, but improved noticeable during open simulation phase. © 2015 Elsevier B.V.

Schmidt E.W.,Becker Technologies GmbH | Gupta S.,Becker Technologies GmbH | Freitag M.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH
International Congress on Advances in Nuclear Power Plants, ICAPP 2014 | Year: 2014

Tests combining thermally driven natural convection which dissolves steam air stratifications have been performed at the technical scale THAI test facility. This continues a series of gas distribution tests with increasing complexity intensively used for validation and development of computational fluid dynamics (CFD) and lumped parameter (LP) codes. As in an earlier test performed with a helium air stratification a naturally driven convection flow in the THAI vessel has been established by cooling the upper vessel walls while heating the lower vessel walls. At comparable wall temperature difference it has been found that the erosion of the steam air cloud is much faster than the erosion of a comparable helium air cloud. The condensation of steam at the upper plenum walls has been balanced by injecting steam at small flow rate during the erosion test phase. So the faster erosion is attributed to the lower density difference between air and steam compared to air and helium. Reducing the wall temperature difference in two additional tests reduced the erosion speed and allowed to determine erosion velocities of 1.1 mm/s and 1.7 mm/s which are still faster than 0.7 mm/s determined in the former helium stratification test.

Schmidt E.W.,Becker Technologies GmbH | Gupta S.,Becker Technologies GmbH | Freitag M.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH | Von Laufenberg B.,Becker Technologies GmbH
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Experiments have been conducted using the technical scale THAI test facility to investigate the wet resuspension of aerosols from a boiling sump. One test with a water solution of potassium iodide and cesium chloride salts and two tests with different suspensions of calcium carbonate in water (primary mass-median-diameters of 0.065 μm and 0.9 μm) have been performed. Resuspension was derived by injecting steam at the bottom of the sump. The investigated superficial velocities, ranging from 0.025 m/s to 0.13 m/s, extend previous resuspension of soluble material experiments in the THAI facility to turbulent flow regime. The aerosol released from the boiling sump into the gas space has been measured independently by using an SMPS particle counter and gas scrubbers. The droplet entrainment has been quantified, based on the concentrations of salts resuspended into the gas atmospheric flow. The entrainment derived from the gas scrubber measurements is well in accordance to previous experimental findings obtained in the THAI facility and other test facilities. A significant enrichment of insoluble aerosol concentration inside the droplets has been found. The SMPS results revealed that very small aerosols are dominantly released from a boiling sump which might remain airborne for long times. Copyright © 2015 by JSME.

Funke F.,AREVA | Langrock G.,AREVA | Kanzleiter T.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH | And 4 more authors.
Nuclear Engineering and Design | Year: 2012

The conversion of gaseous molecular iodine into iodine oxide aerosols has significant relevance in the understanding of the fission product iodine volatility in a LWR containment during severe accidents. In containment, the high radiation field caused by fission products released from the reactor core induces radiolytic oxidation into iodine oxides. To study the characteristics and the behaviour of iodine oxides in large scale, two THAI tests Iod-13 and Iod-14 were performed, simulating radiolytic oxidation of molecular iodine by reaction of iodine with ozone, with ozone injected from an ozone generator. The observed iodine oxides form submicron particles with mean volume-related diameters of about 0.35 μm and show low deposition rates in the THAI tests performed in the absence of other nuclear aerosols. Formation of iodine aerosols from gaseous precursors iodine and ozone is fast as compared to their chemical interaction. The current approach in empirical iodine containment behaviour models in severe accidents, including the radiolytic production of I 2-oxidizing agents followed by the I 2 oxidation itself, is confirmed by these THAI tests. © 2011 Elsevier B.V. All rights reserved.

Gupta S.,Becker Technologies GmbH
Nuclear Engineering and Technology | Year: 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and 60 m3 volume) are discussed in the light of the Fukushima accident. © 2015, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.

Gupta S.,Becker Technologies GmbH | Kanzleiter T.,Becker Technologies GmbH | Poss G.,Becker Technologies GmbH
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015 | Year: 2015

The Fukushima Daiichi nuclear accident of 11 March 2011 indicated that in the event of a severe accident in Light Water Reactors (LWR), large amounts of hydrogen may be generated by core degradation and released into the containment. Even if several hydrogen mitigation measures such as, containment inerting, igniter, and Passive Autocatalytic Recombiners (PAR) are available, hydrogen deflagration at relatively low H2 concentrations, e.g. below 10 or 12 % in the presence of an operating PAR or even by a random ignition source might not be completely avoided. Depending on locally available hydrogen concentration, turbulence and structural configurations within the containment, ignition initiated by a PAR can cause deflagration or under certain conditions even local detonations. The resulting pressure and temperature levels may pose a threat to the integrity of the containment building. Therefore, PAR inlet conditions at which ignition may occur are important to be determined experimentally in order to assess the possible combustion mode upstream of a PAR by using validated safety analysis tools for the reactor case. In the present paper, PAR-inlet conditions leading to an ignition and the resulting hydrogen deflagration behaviour in a closed vessel atmosphere are discussed. Tests reported in this paper have been conducted in the frame of OECD/NEA Thai project (2007 - 2009). Performance of three different commercially available PAR designs (based on plate- And pellet-type catalysts) have been investigated in the Thai test facility (H = 9.2 m, D = 3.2 m, V = 60 m3) under accidental conditions. Results indicate that a PAR exposed to hydrogen concentration higher than about 5.5 vol % can act as ignition source for the hydrogen-air-steam mixture present in the PAR environment and initiate a hydrogen deflagration. PAR induced ignition is directly correlated with the catalyst surface temperature which in turn depends on the H2 concentration present at the PAR inlet. Following an ignition, flame propagation starts always at the upper end of the PAR in upward direction. The course and strength of a hydrogen deflagration initiated by PAR ignition depends on the available gas composition nearby the PAR outlet as well as gas distribution in the surrounding vessel atmosphere. © Copyright (2015) by American Nuclear Society All rights reserved.

Fischer K.,Becker Technologies GmbH | Freitag M.,Korea Atomic Energy Research Institute | Kang H.S.,Korea Atomic Energy Research Institute
Nuclear Engineering and Design | Year: 2014

Mass transfer of molecular iodine (I2) at the water pool-gas interface can be modeled by means of a water surface film renewal model superimposed to the established two-film theory, where the water-side I2 mass transfer coefficient kw is related to the I2 molecular diffusivity in water D and the air-water contact time a according to The present paper describes a mechanistic approach to determine the contact time from the water flow distribution. The method makes use of a numerical simulation of the poolwater flow, and a numerical evaluation of the contact time distribution at the pool surface. Owing to the numerical treatment it can be applied to pool geometries of any kind, which makes it applicable for nuclear reactor safety studies in general kw = √D/πa The present paper describes a mechanistic approach to determine the contact time from the water flow distribution. The method makes use of a numerical simulation of the poolwater flow, and a numerical evaluation of the contact time distribution at the pool surface. Owing to the numerical treatment it can be applied to pool geometries of any kind, which makes it applicable for nuclear reactor safety studies in general. © 2014 Elsevier B.V.

Becker Technologies GmbH | Date: 2012-08-29

Battery monitoring devices that may be attached to a battery to monitor the performance of the battery and operating software for use therewith, sold as a unit.

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