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We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

Despite global commitments and efforts, a gender-based division of paid and unpaid work persists. To identify how psychological factors, national policies, and the broader sociocultural context contribute to this inequality, we assessed parental-leave intentions in young adults (18–30 years old) planning to have children (N = 13,942; 8,880 identified as women; 5,062 identified as men) across 37 countries that varied in parental-leave policies and societal gender equality. In all countries, women intended to take longer leave than men. National parental-leave policies and women's political representation partially explained cross-national variations in the gender gap. Gender gaps in leave intentions were paradoxically larger in countries with more gender-egalitarian parental-leave policies (i.e., longer leave available to both fathers and mothers). Interestingly, this cross-national variation in the gender gap was driven by cross-national variations in women's (rather than men's) leave intentions. Financially generous leave and gender-egalitarian policies (linked to men's higher uptake in prior research) were not associated with leave intentions in men. Rather, men's leave intentions were related to their individual gender attitudes. Leave intentions were inversely related to career ambitions. The potential for existing policies to foster gender equality in paid and unpaid work is discussed.

Poloidal asymmetries of the propagation velocity of density fluctuations perpendicular to the magnetic field measured with Doppler reflectometry have been reported in several magnetic confinement plasma devices. Careful analysis of a large variety of different low confinement mode plasma scenarios performed at the ASDEX Upgrade tokamak does not reveal such an asymmetry outside the uncertainties of the evaluation process of the measurement data. The perpendicular velocity is investigated between mid-radius and the plasma edge and follows the poloidal dependence of the E × B drift velocity regardless of the probed turbulence structure size. Compared to measurements of a charge exchange recombination spectroscopy diagnostic this points towards a significantly smaller phase velocity than the E × B drift velocity. The analysis technique is presented in a representative discharge together with a sensitivity study of the impact of density, magneto hydrodynamic equilibrium and diagnostic alignment on the interpretation of the measured Doppler shift using ray tracing and thus on poloidal asymmetries. Three more highly different plasma scenarios with poloidally symmetric velocity profiles are shown.

We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

This paper presents results from a series of integral tests performed at Framatome’s INKA test facility in Karlstein (Germany) which simulates a KERENA boiling water reac-tor (BWR). The scope of the test series was on the behaviour of and interaction be-tween the different passive systems and components under the conditions of extended loss of alternating power (ELAP). These SBO-like conditions were aggravated in three out of four tests by parallel LOCA (Loss of Coolant Accident). The scenarios of all four tests fully correspond to Design Basic Conditions (DBC). They were: main steam line break, feed water line break, reactor pressure vessel (RPV) bottom leak and station blackout (SBO, non-LOCA). In the tests, the passive systems integrated in KERENA and INKA, respectively, have fulfilled their design functions fully satisfactorily and as follows: The Passive Pressure Pulse Transmitter (PPPT) triggered the RPV depressurization without delay. The Emergency Condenser (EC) system removed decay heat along with stored energy from the RPV to the containment. The Containment Cooling Condenser (CCC) system forwarded said power to a heat sink outside of the containment. The passive containment pressure suppression system kept the containment pressure within the design range, partially displacing surplus thermal energy from the drywell to the wetwell, in particular in the early phases after occurrence of LOCA. The passive core flooding system replenished the coolant inventory of the RPV thereby ensuring water levels in the RPV which are fully sufficient for core cooling. Moreover, the systems have cooperated as anticipated by the designers, quietly and without perturbing each other. Hence the test results, which are reported and discussed more in detail within this pa-per, soundly confirm the underlying design and its passive features. Said tests were carried out as a part of the joint research project EASY (Evidence of Design Basis Accidents Mitigation solely with passive sa

We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

Systems of correlated particles appear in many fields of modern science and represent some of the most intractable computational problems in nature. The computational challenge in these systems arises when interactions become comparable to other energy scales, which makes the state of each particle depend on all other particles1. The lack of general solutions for the three-body problem and acceptable theory for strongly correlated electrons shows that our understanding of correlated systems fades when the particle number or the interaction strength increases. One of the hallmarks of interacting systems is the formation of multiparticle bound states2-9. Here we develop a high-fidelity parameterizable fSim gate and implement the periodic quantum circuit of the spin-½ XXZ model in a ring of 24 superconducting qubits. We study the propagation of these excitations and observe their bound nature for up to five photons. We devise a phase-sensitive method for constructing the few-body spectrum of the bound states and extract their pseudo-charge by introducing a synthetic flux. By introducing interactions between the ring and additional qubits, we observe an unexpected resilience of the bound states to integrability breaking. This finding goes against the idea that bound states in non-integrable systems are unstable when their energies overlap with the continuum spectrum. Our work provides experimental evidence for bound states of interacting photons and discovers their stability beyond the integrability limit.

In this paper, the effect of the orientation of the array elements of a uniform circular antenna array (UCA) on orbital angular momentum (OAM) communication is investigated. For the first time, UCAs are studied in which the elements are rotated in a systematic way around their center by a linearly increasing angle of rotation. It is shown here that these UCAs provide a smaller decrease in transmission over distance for some modes compared to other configurations used, but at the cost of increased mode conversion. For that reason the mode conversion and the decrease of the received mode power over distance are of particular interest. U CAs with dipoles normal and tangential to the array surface are compared. For the latter, the dipoles are rotated around their center with a rotation angle increasing linearly around the U CA. The transmission of the OAM modes is evaluated on a sphere, showing that the direction of the strongest transmission of the OAM modes varies with the rotation of the array elements. It is shown that the U CAs with radial (or azimuthal) dipoles provide the best mode transmission with little mode conversion. All investigations have been carried out using a tool based on the method of moments (MoM).

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Contains fulltext : 242602.pdf (Publisher’s version ) (Open Access)

Contains fulltext : 242602.pdf (Publisher’s version ) (Open Access)

This paper deals with the improvement and validation of numerical tools for the simula-tion of design basis accidents in nuclear power plants equipped with passive safety sys-tems. Numerical models are implemented in the framework of the 1-D thermal-hydraulic system code ATHLET developed by GRS. Experimental reference data for the validation were obtained at the INKA test facility, a model of the KERENA reactor, reproducing the passive safety systems nearly at full scale. The validation effort focuses firstly on the accuracy of the models for the single passive components, and secondly on the ability of the numerical simulation to reproduce the in-teraction of all components of the KERENA design under realistic conditions as repro-duced in the INKA test facility. Thermal-hydraulic models are presented and validated for two passive components of the KERENA reactor: the passive pressure pulse trans-mitter and the pressure-controlled flooding valve. Finally, the full model of the INKA fa-cility, including these and other passive components, is discussed and numerical results for simulations reproducing three different design basis accidents are validated by com-parison with the corresponding experimental data.

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modelling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modelling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modelling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I deals with numerical and experimental methods for modelling of condensation inside the emergency condensers and on the containment cooling condenser while part II deals with boiling and two-phase flow instabilities.

We present novel roulette schemes for rare-event sampling that are both structure-preserving and unbiased. The boundaries where Monte Carlo markers are split and deleted are placed automatically and adapted during runtime. Extending existing codes with the new schemes is possible without severe changes because the equation of motion for the markers is not altered. These schemes can also be applied in the presence of nonlinear and nonlocal coupling between markers. As an illustrative application, we have implemented this method in the ASCOT-RFOF code, used to simulate fast-ion generation by radio-frequency waves in fusion plasmas. In this application, with this method the Monte-Carlo noise level for typical fast-ion energies can be reduced at least of one order of magnitude without increasing the computational effort.

We report on the discovery and validation of TOI 813 b (TIC 55525572 b), a transiting exoplanet identified by citizen scientists in data from NASA’s Transiting Exoplanet Survey Satellite (TESS) and the first planet discovered by the Planet Hunters TESS project. The host star is a bright (V = 10.3 mag) subgiant (R⋆ = 1.94 R⊙, M⋆ = 1.32 M⊙). It was observed almost continuously by TESS during its first year of operations, during which time four individual transit events were detected. The candidate passed all the standard light curve-based vetting checks, and ground-based follow-up spectroscopy and speckle imaging enabled us to place an upper limit of 2 MJup (99% confidence) on the mass of the companion, and to statistically validate its planetary nature.

We report on the discovery and validation of TOI 813b (TIC 55525572 b), a transiting exoplanet identified by citizen scientists in data from NASA's Transiting Exoplanet Survey Satellite (TESS) and the first planet discovered by the Planet Hunters TESS project. The host star is a bright (V = 10.3 mag) subgiant ($R_\star=1.94\,R_\odot$, $M_\star=1.32\,M_\odot$). It was observed almost continuously by TESS during its first year of operations, during which time four individual transit events were detected. The candidate passed all the standard light curve-based vetting checks, and ground-based follow-up spectroscopy and speckle imaging enabled us to place an upper limit of $2 M_{Jup}$ (99 % confidence) on the mass of the companion, and to statistically validate its planetary nature. Detailed modelling of the transits yields a period of $83.8911_{ - 0.0031 } ^ { + 0.0027 }$ days, a planet radius of $6.71 \pm 0.38$ $R_{\oplus}$, and a semi major axis of $0.423_{ - 0.037 } ^ { + 0.031 }$ AU. The planet's orbital period combined with the evolved nature of the host star places this object in a relatively under-explored region of parameter space. We estimate that TOI-813b induces a reflex motion in its host star with a semi-amplitude of $\sim6$ ms$^{-1}$, making this system a promising target to measure the mass of a relatively long-period transiting planet., Comment: Accepted for publication in MNRAS (16 pages, 10 figures, 3 tables)

The Project PANAS (acronym for the German “PAssive Nachzerfallswärme-AbfuhrSysteme” = passive residual heat removal systems) deals with the assessment of the heat transfer mechanisms and the operational stability of passive residual heat removal systems applied in GENERATION III and III + nuclear reactors. The project is funded by the German Federal Ministry of Education and Research and executed by a consortium consisting of the German Partners: Technische Universität Dresden (TUD), Technische Hochschule Deggendorf (THD), Helmholtz Zentrum- Dresden- Rossendorf (HZDR), Framatome GmbH and PreussenElektra GmbH. Based on experimental data recorded during the course of the project at the high-pressure test facility COSMEA at HZDR and the low-pressure test facility GENEVA at TUD state of the art heat transfer models are examined. The test facilities are equipped with innovative two-phase-flow instrumentations and heat transfer probes. At the COSMEA test facility the angular resolved radial heat transfer measurement is combined with a 2D image of the liquid/vapor phase distribution inside a heat exchanger pipe. Since drastic deviations between experimentally obtained and calculated heat transfer have been detected depending on the operational conditions, available heat transfer models are modified or new models are proposed. This model adaptation and development process is supported by 3D CFD modeling of the flow inside the tubes. In addition, the operational stability of low-pressure natural circulation systems has been assessed applying RAM-ROM methodology. The paper gives an overview of the first project phase recently being completed and outlines the objectives of the subsequent project phase, which is planned to be completed until end of 2020.

The Project PANAS (acronym for the German “PAssive Nachzerfallswärme-AbfuhrSysteme” = passive residual heat removal systems) deals with the assessment of the heat transfer mechanisms and the operational stability of passive residual heat removal systems applied in GENERATION III and III + nuclear reactors. The project is funded by the German Federal Ministry of Education and Research and executed by a consortium consisting of the German Partners: Technische Universität Dresden (TUD), Technische Hochschule Deggendorf (THD), Helmholtz Zentrum- Dresden- Rossendorf (HZDR), Framatome GmbH and PreussenElektra GmbH. Based on experimental data recorded during the course of the project at the high-pressure test facility COSMEA at HZDR and the low-pressure test facility GENEVA at TUD state of the art heat transfer models are examined. The test facilities are equipped with innovative two-phase-flow instrumentations and heat transfer probes. At the COSMEA test facility the angular resolved radial heat transfer measurement is combined with a 2D image of the liquid/vapor phase distribution inside a heat exchanger pipe. Since drastic deviations between experimentally obtained and calculated heat transfer have been detected depending on the operational conditions, available heat transfer models are modified or new models are proposed. This model adaptation and development process is supported by 3D CFD modeling of the flow inside the tubes. In addition, the operational stability of low-pressure natural circulation systems has been assessed applying RAM-ROM methodology. The paper gives an overview of the first project phase recently being completed and outlines the objectives of the subsequent project phase, which is planned to be completed until end of 2020.

The implementation of passive safety systems in nuclear reactors provide the opportunity to enhance the nuclear safety. On the other hand, an accurate and reliable prediction of the heat removal behavior is not ensured because the operating conditions of certain types of passive systems like containment cooling systems differ from the validity ranges of the established heat transfer correlations. Therefore, a generic and detailed investigation is still necessary for passive systems. Against this background, the test facility GENEVA was erected at Technische Universität Dresden in 2012. Since the commissioning, generic experiments concerning the system and stability behavior of this facility, which emulates a low pressure and low flow (LPLF) natural circulation system, were provided. Nevertheless, the investigation of the heat transfer behavior remained an open issue. On this account, the instrumentation in the heat transfer region inside GENEVA was improved to gather the necessary temperature and void fraction profiles. The performed experiments provide a generic and wide database concerning boiling in a LPLF natural circulation systems. Within this paper, the development of the wall and bulk fluid temperature as well as the axial and center line void fraction profile in a slightly inclined tube for different heat flow rates are discussed. Furthermore, flow patterns could be identified on behalf of the void fraction measurements. To conclude the experimental analysis, the development of the heat transfer coefficient was estimated. These experimental data provide the basis for a simulation with the lumped-parameter thermal-hydraulic code ATHLET and serve as validation reference. However, the comparisons between the experimental and computational results show insufficient agreements. Mainly, the simulation misses the saturation point of the experiments, which leads to great differences of the void fraction values. Moreover, inaccuracies appear as well with the heat tr

The implementation of passive safety systems in nuclear reactors provide the opportunity to enhance the nuclear safety. On the other hand, an accurate and reliable prediction of the heat removal behavior is not ensured because the operating conditions of certain types of passive systems like containment cooling systems differ from the validity ranges of the established heat transfer correlations. Therefore, a generic and detailed investigation is still necessary for passive systems. Against this background, the test facility GENEVA was erected at Technische Universität Dresden in 2012. Since the commissioning, generic experiments concerning the system and stability behavior of this facility, which emulates a low pressure and low flow (LPLF) natural circulation system, were provided. Nevertheless, the investigation of the heat transfer behavior remained an open issue. On this account, the instrumentation in the heat transfer region inside GENEVA was improved to gather the necessary temperature and void fraction profiles. The performed experiments provide a generic and wide database concerning boiling in a LPLF natural circulation systems. Within this paper, the development of the wall and bulk fluid temperature as well as the axial and center line void fraction profile in a slightly inclined tube for different heat flow rates are discussed. Furthermore, flow patterns could be identified on behalf of the void fraction measurements. To conclude the experimental analysis, the development of the heat transfer coefficient was estimated. These experimental data provide the basis for a simulation with the lumped-parameter thermal-hydraulic code ATHLET and serve as validation reference. However, the comparisons between the experimental and computational results show insufficient agreements. Mainly, the simulation misses the saturation point of the experiments, which leads to great differences of the void fraction values. Moreover, inaccuracies appear as well with the heat tr

The Project PANAS (acronym for the German “PAssive Nachzerfallswärme-AbfuhrSysteme” = passive residual heat removal systems) deals with the assessment of the heat transfer mechanisms and the operational stability of passive residual heat removal systems applied in GENERATION III and III + nuclear reactors. The project is funded by the German Federal Ministry of Education and Research and executed by a consortium consisting of the German Partners: Technische Universität Dresden (TUD), Technische Hochschule Deggendorf (THD), Helmholtz Zentrum- Dresden- Rossendorf (HZDR), Framatome GmbH and PreussenElektra GmbH. Based on experimental data recorded during the course of the project at the high-pressure test facility COSMEA at HZDR and the low-pressure test facility GENEVA at TUD state of the art heat transfer models are examined. The test facilities are equipped with innovative two-phase-flow instrumentations and heat transfer probes. At the COSMEA test facility the angular resolved radial heat transfer measurement is combined with a 2D image of the liquid/vapor phase distribution inside a heat exchanger pipe. Since drastic deviations between experimentally obtained and calculated heat transfer have been detected depending on the operational conditions, available heat transfer models are modified or new models are proposed. This model adaptation and development process is supported by 3D CFD modeling of the flow inside the tubes. In addition, the operational stability of low-pressure natural circulation systems has been assessed applying RAM-ROM methodology. The paper gives an overview of the first project phase recently being completed and outlines the objectives of the subsequent project phase, which is planned to be completed until end of 2020.

We report on the discovery and validation of TOI 813b (TIC 55525572 b), a transiting exoplanet identified by citizen scientists in data from NASA's Transiting Exoplanet Survey Satellite (TESS) and the first planet discovered by the Planet Hunters TESS project. The host star is a bright (V = 10.3 mag) subgiant ($R_\star=1.94\,R_\odot$, $M_\star=1.32\,M_\odot$). It was observed almost continuously by TESS during its first year of operations, during which time four individual transit events were detected. The candidate passed all the standard light curve-based vetting checks, and ground-based follow-up spectroscopy and speckle imaging enabled us to place an upper limit of $2 M_{Jup}$ (99 % confidence) on the mass of the companion, and to statistically validate its planetary nature. Detailed modelling of the transits yields a period of $83.8911_{ - 0.0031 } ^ { + 0.0027 }$ days, a planet radius of $6.71 \pm 0.38$ $R_{\oplus}$, and a semi major axis of $0.423_{ - 0.037 } ^ { + 0.031 }$ AU. The planet's orbital period combined with the evolved nature of the host star places this object in a relatively under-explored region of parameter space. We estimate that TOI-813b induces a reflex motion in its host star with a semi-amplitude of $\sim6$ ms$^{-1}$, making this system a promising target to measure the mass of a relatively long-period transiting planet., Comment: Accepted for publication in MNRAS (16 pages, 10 figures, 3 tables)

To guarantee the nuclear safety which means keeping the radioactivity inside the fuel rods it is necessary to remove the decay heat in all circumstances of normal and abnormal operation situations. Decay heat removal systems of the present reactor fleet are based on active components like pumps, motor driven valves, electrically I&C etc. They depend on the supply of additional energy which could fail how it was imposingly demonstrated in Fukushima. Virtually all new reactor designs of generation 3+ are characterized by the implementation of various passive safety components. Passive Residual Heat Removal (PRHR) systems use heat transfer induced density differences to provide sufficient driving forces to establish a system mass flow in natural circulation loops of various configurations. Thus in order to design and model the system performance the determination of the heat transfer resistances is a crucial part since it influences the quality of calculation results on two sides: the heat transferred to the coolant as well as the mass flow of the coolant. Today’s state-ofthe-art PRHR systems use mostly the phase transition between the liquid and the vapor phase of the coolant to maximize the system mass flow and thus the performance. Precondition for the adoption of PRHR systems in nuclear reactors is the verification of the functional capability in all operation modes of the power plant. Therefore a comprehensive experimental work at different mockup scales combined with theoretical investigations (e. g. CFD and system codes) has to be undertaken.

The KERENA reactor with 1,250 MW electrical power is an evolutionary boiling-water reactor (BWR) concept jointly developed by AREVA GmbH and PreussenElektra GmbH. It is a Generation III+ reactor with innovative passive safety systems such as emergency and containment cooling condenser, core flooding system and pressure pulse transmitter (PPPT) to complement the safety concept of a BWR. One design goal of the KERENA reactor concept is, that in case of an accident the core can be cooled for at least 72 hours by passive safety systems only. The INKA test facility at AREVA in Karlstein was built to investigate the heat removal capabilities and the interaction of the passive safety systems and components of the KERENA concept during different accidental scenarios. This test facility represents the KERENA main components like RPV, flooding and pressure suppression pool, drywell and shielding/storage pool, emergency condenser and containment cooling condenser at a sophisticated geometrical and power scaling. In summer 2017 at the INKA test facility a feed water line break, a leak at the lower head of the RPV and a station blackout were experimentally simulated to investigate the integral plant behaviour and the designated safety functions of each single passive component. An existing ATHLET input deck of the INKA test facility, which was already validated against the INKA experiment of a main steam line break, was extended by a PPPT model and the break lines for the loss of coolant experiments. Pre- and post-test calculations for the “leak at the lower head of the RPV” experiment were conducted to assess and validate the input deck. The experiment has shown that the passive safety systems are capable to remove the decay heat and the core flooding system was also triggered in this accident sequence. Comparing the ATHLET simulations with the experimental data, some deviations were found, which are currently being investigated and treated by ATHLET input data adjustments.

The KERENA reactor with 1,250 MW electrical power is an evolutionary boiling-water reactor (BWR) concept jointly developed by AREVA GmbH and PreussenElektra GmbH. It is a Generation III+ reactor with innovative passive safety systems such as emergency and containment cooling condenser, core flooding system and pressure pulse transmitter (PPPT) to complement the safety concept of a BWR. One design goal of the KERENA reactor concept is, that in case of an accident the core can be cooled for at least 72 hours by passive safety systems only. The INKA test facility at AREVA in Karlstein was built to investigate the heat removal capabilities and the interaction of the passive safety systems and components of the KERENA concept during different accidental scenarios. This test facility represents the KERENA main components like RPV, flooding and pressure suppression pool, drywell and shielding/storage pool, emergency condenser and containment cooling condenser at a sophisticated geometrical and power scaling. In summer 2017 at the INKA test facility a feed water line break, a leak at the lower head of the RPV and a station blackout were experimentally simulated to investigate the integral plant behaviour and the designated safety functions of each single passive component. An existing ATHLET input deck of the INKA test facility, which was already validated against the INKA experiment of a main steam line break, was extended by a PPPT model and the break lines for the loss of coolant experiments. Pre- and post-test calculations for the “leak at the lower head of the RPV” experiment were conducted to assess and validate the input deck. The experiment has shown that the passive safety systems are capable to remove the decay heat and the core flooding system was also triggered in this accident sequence. Comparing the ATHLET simulations with the experimental data, some deviations were found, which are currently being investigated and treated by ATHLET input data adjustments.

To guarantee the nuclear safety which means keeping the radioactivity inside the fuel rods it is necessary to remove the decay heat in all circumstances of normal and abnormal operation situations. Decay heat removal systems of the present reactor fleet are based on active components like pumps, motor driven valves, electrically I&C etc. They depend on the supply of additional energy which could fail how it was imposingly demonstrated in Fukushima. Virtually all new reactor designs of generation 3+ are characterized by the implementation of various passive safety components. Passive Residual Heat Removal (PRHR) systems use heat transfer induced density differences to provide sufficient driving forces to establish a system mass flow in natural circulation loops of various configurations. Thus in order to design and model the system performance the determination of the heat transfer resistances is a crucial part since it influences the quality of calculation results on two sides: the heat transferred to the coolant as well as the mass flow of the coolant. Today’s state-ofthe-art PRHR systems use mostly the phase transition between the liquid and the vapor phase of the coolant to maximize the system mass flow and thus the performance. Precondition for the adoption of PRHR systems in nuclear reactors is the verification of the functional capability in all operation modes of the power plant. Therefore a comprehensive experimental work at different mockup scales combined with theoretical investigations (e. g. CFD and system codes) has to be undertaken.

To guarantee the nuclear safety which means keeping the radioactivity inside the fuel rods it is necessary to remove the decay heat in all circumstances of normal and abnormal operation situations. Decay heat removal systems of the present reactor fleet are based on active components like pumps, motor driven valves, electrically I&C etc. They depend on the supply of additional energy which could fail how it was imposingly demonstrated in Fukushima. Virtually all new reactor designs of generation 3+ are characterized by the implementation of various passive safety components. Passive Residual Heat Removal (PRHR) systems use heat transfer induced density differences to provide sufficient driving forces to establish a system mass flow in natural circulation loops of various configurations. Thus in order to design and model the system performance the determination of the heat transfer resistances is a crucial part since it influences the quality of calculation results on two sides: the heat transferred to the coolant as well as the mass flow of the coolant. Today’s state-ofthe-art PRHR systems use mostly the phase transition between the liquid and the vapor phase of the coolant to maximize the system mass flow and thus the performance. Precondition for the adoption of PRHR systems in nuclear reactors is the verification of the functional capability in all operation modes of the power plant. Therefore a comprehensive experimental work at different mockup scales combined with theoretical investigations (e. g. CFD and system codes) has to be undertaken.

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.

Beside the nuclear reactor and its primary circuit the spent fuel pool is yet another safety-critical part in a nuclear power plant which has gained increasing focus after the Fukushima accident. Loss of coolant or enduring loss of cooling conditions would ultimately result in loss of cladding integrity at elevated temperatures with excessive release of fission products and hydrogen. To predict the available response time and to assess the efficacy of mitigating measures computer simulations can be employed. Their validity, however, needs to be proven by dedicated experiments at lower scale but relevant thermal hydraulic conditions. For that purpose, the test facility ALADIN was designed, which enables conduction of experiments on a single BWR fuel element mock-up under loss of coolant and loss of cooling accident conditions. In this paper we introduce the facility and its instrumentation, with a focus on temperature sensors and a new thermal anemometry grid sensor for spatially resolved flow velocity measurement of the superheated steam in the subchannels together with the affiliated calibration procedure.

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.

The KERENA reactor with 1,250 MW electrical power is an evolutionary boiling-water reactor (BWR) concept jointly developed by AREVA GmbH and PreussenElektra GmbH. It is a Generation III+ reactor with innovative passive safety systems such as emergency and containment cooling condenser, core flooding system and pressure pulse transmitter (PPPT) to complement the safety concept of a BWR. One design goal of the KERENA reactor concept is, that in case of an accident the core can be cooled for at least 72 hours by passive safety systems only. The INKA test facility at AREVA in Karlstein was built to investigate the heat removal capabilities and the interaction of the passive safety systems and components of the KERENA concept during different accidental scenarios. This test facility represents the KERENA main components like RPV, flooding and pressure suppression pool, drywell and shielding/storage pool, emergency condenser and containment cooling condenser at a sophisticated geometrical and power scaling. In summer 2017 at the INKA test facility a feed water line break, a leak at the lower head of the RPV and a station blackout were experimentally simulated to investigate the integral plant behaviour and the designated safety functions of each single passive component. An existing ATHLET input deck of the INKA test facility, which was already validated against the INKA experiment of a main steam line break, was extended by a PPPT model and the break lines for the loss of coolant experiments. Pre- and post-test calculations for the “leak at the lower head of the RPV” experiment were conducted to assess and validate the input deck. The experiment has shown that the passive safety systems are capable to remove the decay heat and the core flooding system was also triggered in this accident sequence. Comparing the ATHLET simulations with the experimental data, some deviations were found, which are currently being investigated and treated by ATHLET input data adjustments.

To guarantee the nuclear safety which means keeping the radioactivity inside the fuel rods it is necessary to remove the decay heat in all circumstances of normal and abnormal operation situations. Decay heat removal systems of the present reactor fleet are based on active components like pumps, motor driven valves, electrically I&C etc. They depend on the supply of additional energy which could fail how it was imposingly demonstrated in Fukushima. Virtually all new reactor designs of generation 3+ are characterized by the implementation of various passive safety components. Passive Residual Heat Removal (PRHR) systems use heat transfer induced density differences to provide sufficient driving forces to establish a system mass flow in natural circulation loops of various configurations. Thus in order to design and model the system performance the determination of the heat transfer resistances is a crucial part since it influences the quality of calculation results on two sides: the heat transferred to the coolant as well as the mass flow of the coolant. Today’s state-ofthe-art PRHR systems use mostly the phase transition between the liquid and the vapor phase of the coolant to maximize the system mass flow and thus the performance. Precondition for the adoption of PRHR systems in nuclear reactors is the verification of the functional capability in all operation modes of the power plant. Therefore a comprehensive experimental work at different mockup scales combined with theoretical investigations (e. g. CFD and system codes) has to be undertaken.

The KERENA reactor with 1,250 MW electrical power is an evolutionary boiling-water reactor (BWR) concept jointly developed by AREVA GmbH and PreussenElektra GmbH. It is a Generation III+ reactor with innovative passive safety systems such as emergency and containment cooling condenser, core flooding system and pressure pulse transmitter (PPPT) to complement the safety concept of a BWR. One design goal of the KERENA reactor concept is, that in case of an accident the core can be cooled for at least 72 hours by passive safety systems only. The INKA test facility at AREVA in Karlstein was built to investigate the heat removal capabilities and the interaction of the passive safety systems and components of the KERENA concept during different accidental scenarios. This test facility represents the KERENA main components like RPV, flooding and pressure suppression pool, drywell and shielding/storage pool, emergency condenser and containment cooling condenser at a sophisticated geometrical and power scaling. In summer 2017 at the INKA test facility a feed water line break, a leak at the lower head of the RPV and a station blackout were experimentally simulated to investigate the integral plant behaviour and the designated safety functions of each single passive component. An existing ATHLET input deck of the INKA test facility, which was already validated against the INKA experiment of a main steam line break, was extended by a PPPT model and the break lines for the loss of coolant experiments. Pre- and post-test calculations for the “leak at the lower head of the RPV” experiment were conducted to assess and validate the input deck. The experiment has shown that the passive safety systems are capable to remove the decay heat and the core flooding system was also triggered in this accident sequence. Comparing the ATHLET simulations with the experimental data, some deviations were found, which are currently being investigated and treated by ATHLET input data adjustments.

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.