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ASME Conference Presenter Attendance Policy and Archival Proceedings

2017;():V008T00A001. doi:10.1115/ICONE25-NS8.
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This online compilation of papers from the 2017 25th International Conference on Nuclear Engineering (ICONE25) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Computational Fluid Dynamics (CFD) and Coupled Codes

2017;():V008T09A001. doi:10.1115/ICONE25-66002.

The steam flow is simulated by FLUENT. The Lagrange-Euler method is used to simulate the droplet-laden flow in wave-type separators. Two-way coupling method is used to study the influence of the momentum transfer between droplets and carrier-phase in wave-type plate separators. A group of the trajectories of droplets with different diameters are performed in wave-type plate separator flow field. The result shows that the momentum transfer has tiny impact on the behaviors of droplets in a low velocity flow. However, the momentum transfer affects the behaviors of droplets more significantly with rising flow velocity. The one-way coupling method overestimates the diffusion of droplets. In addition, the momentum transfer affects the total pressure loss more significantly with rising volume fraction. The conclusion verifies the importance of the momentum transfer in droplet-laden flows, which could be used to simulate the behavior of droplets moving in a separator.

Topics: Momentum , Waves , Drops
Commentary by Dr. Valentin Fuster
2017;():V008T09A002. doi:10.1115/ICONE25-66025.

The passive cooling system (PCCS) for reactor containment is a security system that can be used to cool the atmosphere and reduce pressure inside of containment in case of temperature and pressure increase caused by vapor injection, which requires no external power because it works only with natural forces. However, as the driving forces from natural physical phenomena are of low amplitude, uncertainties and instabilities in the physical process can cause failure of the system. This article aims to establish a CFD simulation model for the Passive Containment Cooling System of 1000MW PWR using Code_Saturne and FLUENT software. The comparison of 4 different models based respectively on mixture model, COPAIN test, Uchida correlation and Chilton-Colburn analogy which simulate the condensing effect and the improvement of source code are based on a 3D simulation of PCCS system. To simulate the thermal-hydraulic condition in the containment after LOCA accident caused by a double-ended main pipe rupture, a high temperature vapor with the given mass flow rate are supposed to be the source of energy and mass into containment. Meanwhile the surface of three condensing island applies the wall condensation model.

The simulation results show similar transient process obtained with the 4 models, while the difference between the transient simulation and the steady-state analysis of three models is less than 3%. The large mass flow rate of water loss status inside the containment cause a high flow rate of vapor which could be uniformly mixed with air in a short time. For the self-condensing efficiency of 3 groups of PCCS system, the non-centrosymmetric injection position resulting that the condensing efficiency is slightly higher for the two heat exchanger groups nearby. During the first 2400s of simulation time, more than 75.69% of the vapor is condensed, indicating that for the occurrence of condensation at the wall mainly driven by natural convection, the effect of thermodynamic siphon could improve the flow of gas mixture inside the tubes when the velocity of mixture is not large enough, so that the vapor could smoothly enter the tube and reach the internal cooling surface then to be condensed. Besides, PCCS ensure the containment internal pressure maintained below 2 bar and the temperature maintained below 380K during 3600s.

Commentary by Dr. Valentin Fuster
2017;():V008T09A003. doi:10.1115/ICONE25-66066.

The objective was to develop a validated computational fluid dynamics (CFD) based approach for predicting hydrogen detonations and the mechanical loads. Applications of interest were scenarios relevant to hydrogen explosion risk assessment in nuclear power plant under hypothetical severe accident. Model developments were conducted within the framework of the parallel scientific computational tool GASFLOW-MPI thanks to its effectiveness, reliability and robustness in predicting all-speed flows. Validation was completed for hydrogen detonation phenomena in 3-D hemispherical hydrogen cloud. Excellent comparisons between experimental data and model predictions were observed. With the developed detonation modeling capability, the all-speed CFD code GASFLOW-MPI can be applied to model both turbulent dispersion and hydrogen detonation phenomena that occurred in the nuclear reactor containment during severe accident. Further model developments and validations will be performed for flame acceleration (FA) and deflagration to detonation transition (DDT).

Commentary by Dr. Valentin Fuster
2017;():V008T09A004. doi:10.1115/ICONE25-66167.

The current paper comprises CFD-modelling and simulation of condensation and heat transfer inside horizontal pipes. Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat transfer to the secondary side of the condenser occurs. In the case of an accident the water level in the core is decreasing, steam comes in the emergency pipes and due to the subcooled water around the pipe, this steam will condense. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the reactor core when any accident happens. The actual project is defined in order to model all these processes which happen in the emergency cooling systems. The most focus of the project is on detection of different morphologies such as annular flow, stratified flow, slug flow and plug flow. The first step is the investigation of condensation inside a horizontal tube by considering the direct contact condensation (DCC). Therefore, at the inlet of the pipe an annular flow is assumed. In this step, the Algebraic Interfacial Area Density (AIAD) model is used in order to simulate the interface. The second step is the extension of the model to consider wall condensation effect as well which is closer to the reality. In this step, the inlet is pure steam and due to the wall condensation, a liquid film occurs near the wall which leads to annular flow. The last step will be modelling of different morphologies which are occurring inside the tube during the condensation via using the Generalized Two-Phase Flow (GENTOP) model extended by heat and mass transfer. By using GENTOP the dispersed phase is able to be considered and simulated. Finally, the results of the simulations will be validated by experimental data which will be available in HZDR. In this paper the results of the first part has been presented.

Commentary by Dr. Valentin Fuster
2017;():V008T09A005. doi:10.1115/ICONE25-66211.

As one of the breeding blanket candidates for China Fusion Engineering Test Reactor (CFETR), the water-cooled ceramic blanket (WCCB) was proposed to use mono-sized beryllium pebble bed and binary mixed Li2TiO3/Be12Ti pebble bed in order to increase the packing factor and meet the tritium breeding ratio requirement. Helium (mixed with 0.1% content of H2) is used as the purge gas to sweep tritium out when it flows through the pebble beds. Purge gas flow characteristics are of great importance to tritium recovery system design and will dominate the tritium sweep capability. In this study, DEM-CFD method was used to study the flow characteristics including porosity distribution, velocity distribution and pressure drop in the pebble beds. Mono-sized pebble bed with a packing factor of 0.61 and binary mixed pebble beds with the diameter ratio of 6:1, packing factor of 0.755, and diameter ratio of 7:1, packing factor of 0.7513 were simulated. This method can be used to study the detailed flow characteristics in pebble beds and optimize the pebble bed packing parameters to obtain an appropriate pressure drop, and could be extended to study tritium sweep capability for the design of fusion blanket.

Commentary by Dr. Valentin Fuster
2017;():V008T09A006. doi:10.1115/ICONE25-66216.

In this paper, by using the CFD method, the efforts have been tried to conduct the investigation on the subcooled flow boiling in the first wall channel of Water cooled ceramic breeder (WCCB) blanket for CFETR. The detailed 3D distribution of temperature and vapor in the flowing passage have been presented under the heat flux of 0.5MW/m2 and 1MW/m2, respectively. Due to the high heat flux from plasma, the vapor distribution in the channel decreases from the plasma side to the breeder side along the radial direction. Undoubtedly, the volume fraction of vapor increases along the flowing direction because of the heating. Besides, the distribution of the channel wall along the toroidal direction presents the U-shaped tendency. As demonstrated by the results, the vapor is more likely to be generated at the corner of the square channel, and this can easily cause the Critical Heat Flux (CHF), which will destroy the structural integrity and materials melting. To avoid the enrichment of vapor in the corner, the optimization on the flowing channel has been performed by smoothing the corner. The results show that the volume fraction of vapor can be effectively decreased compared with the original square channel. Moreover, from the perspective of thermal hydraulics, the circular tube is the optimized channel which can not only avoid the concentration of vapor, but also can decrease the peak volume fraction.

Commentary by Dr. Valentin Fuster
2017;():V008T09A007. doi:10.1115/ICONE25-66246.

Hydrogen gathering in the containment may occur followed by a severe accident in a nuclear power plant. A flammable mixture can be formed when hydrogen is mixed with air. The ignition of the gas mixture could threaten the integrity of the containment. In order to provide technology base and experiment data for optimization of hydrogen safety technology of Chinese advanced pressurized water reactor CAP1400, a major project regarding hydrogen safety research of pressurized water reactor containment is underway. As an important part of the project, an experimental facility (A4Q-DH) for the study of hydrogen combustion will be built. Gas displacement method is used to filling the premixed hydrogen-air-steam mixtures into the experimental pipe. Flow behaviors of the gases in the pipe are complicated because fluid flow can be disturbed by the built-in obstacles and gas density can be changed with variation of gas composition concentration. Therefore, it is necessary to evaluate the effectiveness of the gas filling method.

In this paper, gas filling processes for the experimental pipe with different obstacles and gas composition concentrations were simulated using computational fluid dynamics software ANSYS Fluent. The results indicated that hydrogen-air-steam mixtures can be uniformly distributed in the experimental pipe within tens of seconds. The obstacles with modest blockage ratio in the pipe are conducive to shorten the required gas filling time. The hindering effect of annular obstacles is greater than the one of circular and square obstacles. The time required for air to achieve uniformly distribution increases with the increase of the inlet concentration of steam and hydrogen. However, the time required for hydrogen and steam to be evenly distributed in the pipe are relatively close regardless of the shape and blockage ratio of obstacle and the inlet gas concentration.

Topics: Pipes , Hydrogen , Steam
Commentary by Dr. Valentin Fuster
2017;():V008T09A008. doi:10.1115/ICONE25-66268.

Hydrogen may be released by Fuel and Cladding interaction with the steam at very high temperatures into the containment during a severe accident at the nuclear power plant (NPP). Locally, high hydrogen concentration may be achieved that might probably result in detonation or fast deflagration and challenges the reliability of the containment. The mixing and distribution of hydrogen is a serious safety issue for scientists to preserve the structural reliability of the containment. Although the Three Miles Island (TMI) accident in 1979 was the initiator to study the production and buildup of hydrogen in the containment. Subsequently, the hydrogen explosion in the Fukushima Dai-ichi NPP accident (2011), modeling of the gas behavior turn out to be a significant topic in the nuclear safety analysis.

Computational fluid dynamics (CFD) codes can be used to investigate the hydrogen distribution in the containment during accidental scenarios and predicts the local hydrogen concentration in different regions of the containment vessel. In such a way, the associated risk of hydrogen safety can be determined, and safety assessment and procedures can be measured. The current paper presents the results of systematic work done by using the HYDRAGON code, developed by Department of Engineering Physics, Tsinghua University, to study (I) Mesh sensitivity (II) Buoyancy-driven flows analysis, for various turbulence models i.e., a standard k-ε (SKE) model, a renormalization k-ε (RNG) model and a realizable k-ε (RLZ) model and, to demonstrate the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the NPP. The HYDRAGON code simulation results were compared to the published data performed by Jordan et al., and it has been observed that the simulation results obtained by refined mesh have given generally satisfactory results. However, the global effect of the buoyancy term in the turbulence equations on the HYDRAGON code simulated results is very small.

Commentary by Dr. Valentin Fuster
2017;():V008T09A009. doi:10.1115/ICONE25-66289.

Studies on local Fuel-Coolant Interactions (FCI) in a liquid pool are of crucial importance for the improved evaluation of severe accidents for Sodium-cooled Fast Reactors (SFR). To clarify the mechanisms underlying this interaction, several years ago a series of experiments was performed by Cheng et al. (2014) at the Japan Atomic Energy Agency (JAEA) through delivering a given quantity of water into a simulated molten fuel pool which is formed by a low-melting-point alloy. In order to acquire more evidence, numerical analysis using the SIMMER-III, an advanced fast reactor safety analysis code, was also conducted. However, through those analyses, many limitations of the experiments (esp. the uncertain initial geometry of water lump) which would impair the reliability of experimental observations and prevent the direct comparisons with calculations were realized. Focusing on those aspects, further investigations from both experimental and numerical aspects were continued at the Sun Yat-sen University in China. In this study, the effect of injection mode (namely the coolant-injection mode and fuel-injection mode) on local FCIs is studied using the FLUENT code along with its Volume of Fluid Model (VOF). It is seen that the interaction mode does have some influence on the transient behaviors, in particular the formation of steam bubbles, and time variation of temperature and pressure values. In addition, the difference on heat exchange rate due to the steam bubble formation is also confirmable. Knowledge and evidence from this study might be utilized for future development and analyses of SFR severe accident analysis codes in China.

Topics: Fuels , Coolants
Commentary by Dr. Valentin Fuster
2017;():V008T09A010. doi:10.1115/ICONE25-66292.

Liquid carryover is an important phenomenon during the small break loss of coolant accidents (LOCA). The coolant loss rate due to liquid entrained decides the decreasing height of water level and the ability that the core can be cooled. It is especially important in nuclear heating reactor as it is integral arranged without safety injection system. Therefore, the initial water volume in the reactor vessel and coolant loss rate determines whether the core could be submerged. Numerous experiments have been conducted to investigate the phenomenon of liquid carryover in the pool system. Fruitful outcomes have been proposed involving mechanism of entrained liquid produced as well as semi-empirical formula for engineering purposes. However, the semi-empirical correlations on entrainment are highly data-depended and are less applicable under different experiment settings. This paper presents a numerical method based on the VOF to study the liquid carryover in the pool system. The numerical results are analyzed with mesh independence analysis and compared with previous experiments. Several types of liquid carryover are found in the simulation. The exit effect on liquid loss rate is specifically studied including the setting of exit as well as the range that exit effect dominates.

Commentary by Dr. Valentin Fuster
2017;():V008T09A011. doi:10.1115/ICONE25-66321.

As a large scale passive pressurized water reactor nuclear power plant, CAP1400 can remove the reactor decay heat to outside containment with the air cooling in the air flow path of passive containment cooling system (PCS) during the long-term period following an accident. Flow resistance characteristic and wind neutrality characteristic are the main performances of PCS air flow path. In order to study the performance of PCS air flow path, it is necessary to carry out the PCS wind tunnel test and computational fluid dynamics (CFD) analysis to establish a suitable method for the analysis of the performance of the air flow path.

This paper comes up simulating the internal pressure and velocity distribution in the air flow path under different wind speed through CAP1400 PCS 1:100 scaled air flow path wind tunnel test to research the air flow resistance and internal flow pattern. The test shows that local uneven flow phenomenon exists in the outer annulus of the air flow path, but the wind pressure distribution of inner annulus is not affected by environment wind speed, wind direction angle, landforms and the surrounding buildings. The wind pressure is uniform at different heights on the cross section and shows the neutrality feature.

Combining with CAP1400 PCS wind tunnel test, the CFD model is built. The measured inlet wind speed, turbulent kinetic energy and turbulent dissipation rate distribution parameters are inputs and the uniform wind conditions and gradient wind conditions of simulation analysis are developed. Simulation results show that:

1) In uniform wind condition, simulation result of pressure coefficient distribution trend at each cross section is consistent with the test trend and the deviation is very small, which basically can be controlled below 5%. The simulated differential pressures between inner annulus and outer annulus at different elevation are basically identical with the test results, which increase as the elevation arises. The simulated velocity distribution is basically identical with the test. The wind velocity at the upwind and central area of the flow path outlet is larger than other area, and a large swirling region comes on the leeward side near the wall 15cm, but simulated swirling region size at leeward side is slightly smaller.

2) In gradient wind condition, the pressure coefficient distribution trends are basically identical, and the deviation between the test and CFD analysis is 5–10% approximately. Considering the stability of gradient wind condition in wind tunnel is worse than that of uniform wind conditions, and more prone to wind speed fluctuations, therefore, the deviation is slightly greater than the uniform wind condition.

According to the CFD simulation and wind tunnel test, it can be found that the simulation of air flow inside and outside annulus has a high precision though the test results are slightly affected by the instrument tubes along the two sides of test model. In general, CFD simulation and wind tunnel test results are basically identical. Therefore, CFD analysis method is well verified by PCS wind tunnel test, which can be applied to the analysis of the actual power plant.

Commentary by Dr. Valentin Fuster
2017;():V008T09A012. doi:10.1115/ICONE25-66396.

Scaling analysis based on scaling laws was performed first for a single-phase natural circulation loop. Two geometrical reduced models were built. Numerical simulations for them using a CFD code named FLUENT were carried out. The calculated heat removal capabilities of the two reduced models show good agreement with the predicted results by the RELAP5/Mod3.2. The reversely deduced heat removal capabilities of the loop prototype from the numerically calculated results agree well with the prediction by the RELAP5/Mod3.2 for full height scaling model.

Commentary by Dr. Valentin Fuster
2017;():V008T09A013. doi:10.1115/ICONE25-66450.

When fast reactor is operation, because of scouring by high pressure liquid coolant in fuel assembly, there are a lot of products in coolant channel. Using FLUENT software simulate deposition of insoluble particle in fast reactor. Using standard k-ε model predict flow field and turbulence intensity of fluid phase. Using discrete phase model track the trajectory of insoluble particle. The following are simulation results. Fuel cladding deposits lots of insoluble particle, but the concentration of insoluble particle is lower at the central of coolant; Entrance section of the insoluble particle concentration is higher than exit section; Dot deposition of insoluble particle at outlet of fuel cladding will lead to pitting phenomenon, pitting will cause deterioration of heat transfer and destroy the integrity of cladding. In view of deposition law of insoluble particle and characteristic of fuel assembly, mitigation measures of cleaning insoluble particle at fixed time and fixed position are being proposed.

Commentary by Dr. Valentin Fuster
2017;():V008T09A014. doi:10.1115/ICONE25-66497.

During hypothetical severe accidents in nuclear power plants, a large amount of hydrogen is generated rapidly as a result of Zirconium-Steam reaction and released into the containment. Hydrogen mixes with air and may come into combustion or detonation under proper conditions, which threatens the integrity of containment. Therefore, getting detailed hydrogen flow and distribution in various physical mechanisms is a key issue to resolve the hydrogen risk in containment and compartments. To study local hydrogen distribution in the containment of advanced passive PWR, an analysis model is built by 3-dimensional CFD code. Computational domain is divided by structured grid which contains over 100,000 cells. the shape and surface area of walls and obstacles of steel shell and internal structure, which have great impact on gas flow and heat transfer, are included. Hydrogen distribution in containment simulating with different turbulence models is studied, the result shows that during large amount of hydrogen release stage. In hydrogen distribution result simulating with algebraic model, hydrogen is all gathered in the dome and the peak concentration reaches 17%. When k-ε model is adopted, the peak concentration in the dome is 8%, hydrogen stratification is established in whole large space. Besides, hydrogen distribution near source also shows algebraic model cannot simulate turbulence diffusion in local compartment. It is more reasonable choosing k-ε model to study hydrogen behavior in containment. Based on adopted k-ε model, the effect of steam on hydrogen distribution is investigated. With steam injection, the hydrogen distribution is more homogeneous in upper space and average concentration is lower. In local compartment, due to diffusion enhanced by steam, the hydrogen concentration is higher in the bottom.

Commentary by Dr. Valentin Fuster
2017;():V008T09A015. doi:10.1115/ICONE25-66512.

A vortex diode is used as a highly reliable check-valve in nuclear applications, where it mainly benefits from the intrinsic properties of no moving parts and no leakage. Its basic principle is similar to the diode in an electric circuit. The typical structure of a vortex diode consists of a chamber with axial and tangential ports. When the fluid is injected through the axial port, a simple radial flow in the chamber leads to a relatively low flow resistance. On the other hand, in the reverse flow mode, a strongly swirling vortex can be set up in the chamber, resulting in a very high flow resistance.

Several experimental studies found vortex-induced vibration of a vortex diode in the reverse flow mode, where it indicated that the flow was unstable in the vortex diode. This phenomenon may affect the reliability of the vortex diode. However, the mechanism has not been investigated systematically and profoundly. In this paper, 3-D simulations are carried out to help understand the related flow characteristics in the vortex diode. Standard k-ε model was selected for forward flow, while Reynolds stress model was selected for reverse flow. We have found that the results from transient simulations are in good agreement with experimental data. The transient simulations also capture the periodic pressure fluctuation in the vortex diode. Vortex diodes with different structures and geometrical parameters are simulated at different Reynolds number conditions. It is found that the characteristics of the pressure fluctuation are determined by the structure parameters and working conditions of the vortex diode. The flow instability is mainly caused by the asymmetry of the vortex diode. The work presented in this paper will be useful to give better understanding of flows in vortex diodes and to provide some guidance for optimizing the vortex diode.

Commentary by Dr. Valentin Fuster
2017;():V008T09A016. doi:10.1115/ICONE25-66529.

The paper focuses on the development, implementation, and initial validation of the NEK-2P two-fluid two-phase model. Recent extensions of the Extended Boiling Framework (EBF) models are presented, including the implementation of a four-field two-phase topology which replaces the previous two-field two-phase approach. The paper presents results of recent NEK-2P analyses of several CHF experiments. Good agreement between the simulated wall temperatures and measured data is observed. The simulation results predict well both the dryout location and post dryout wall temperature magnitudes, illustrating the ability of the NEK-2P code and extended EBF models to simulate the CHF phenomena for a range of thermal-hydraulic conditions.

Commentary by Dr. Valentin Fuster
2017;():V008T09A017. doi:10.1115/ICONE25-66537.

Corrosion products on fuel cladding surface have a significant impact on reactor operation. These types of deposits are defined as Corrosion Residual Unidentified Deposit (CRUD) and are consist of a porous matrix of nickel and iron based oxides deposited on the fuel cladding surface. It is well known that crud deposits may cause potential Crud Induced Localized Corrosion (CILC) risk and Crud Induced Power Shift (CIPS) risk.

The paper presents a Computational Fluid Dynamic (CFD) method of predicting the crud effect on the thermal hydraulic performance. The effect of the crud roughness is mainly considered in the simulation, the flow near the wall of the crud is solved by modifying wall function in the prism layer.

The simulation object is a span of typical 17×17 rod bundle with a mid grid in PWR, all the structures including grid straps, springs, dimples, mixing vanes and welding spots are included. Thicknesses of grid and fuel cladding are considered in order to precisely simulate the fluid-solid conjugate heat transfer. The crud is set to be covered on the full span downstream of the grid. The simulation is based on the CILC risk pre-analysis and the computed information in the mostly likely crud deposit position is used as boundary condition.

Based on the simulation results, the crud effects on the flow characteristics including vortex structures, circulation, turbulent intensity and second flow intensity and the heat transfer characteristics including rod temperature, fluid temperature and heat transfer coefficient are discussed in detail.

Commentary by Dr. Valentin Fuster
2017;():V008T09A018. doi:10.1115/ICONE25-66545.

A new type of dry storage system is designed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI), which can efficiently remove the decay heat of the hexagonal spent fuel assemblies such as VVER fuel assemblies. The dry storage system includes a Ventilated Concrete Cask (VCC) and a Multi-assembly sealed basket (MSB). Decay heat is removed by natural circulation with helium and air, heat conduction and thermal radiation heat transfer.

Thermal performance of the dry storage system has been investigated by two different numerically methods, i.e., the Computational Fluid Dynamics (CFD) method and the lumped parameter method. The CFD method is utilized based on the commercial software STAR-CCM+, and fuel assemblies are modeled as a porous medium characterized by effective conductivity and the permeability and inertial resistance factor, while other geometry including the lids, base plates, inner and outer shell are modeled explicitly with necessary simplifications. The lumped parameter method is utilized based on the system code GOTHIC, the geometry and the fuel assemblies are divided and represented by 44 volumes. The flow of the air and helium are modeled by flow path which connects the related volumes, and the heat transfer between fluid and solid structures are modeled by thermal conductor models. Heat transfer by convection, conduction and thermal radiation is modeled in both of the two methods.

The maximum temperature of spent fuel assembly can be obtained by both of the two methods, which can be a design basis for investigations attempting to improve the performance of the dry storage system. It is found that the simulation results calculated by the lumped parameter method are more conservative than those calculated by the CFD method. Both methods indicate that after the storage of 7.5 years, the dry storage system is able to remove the decay heat from the hexagonal spent fuel assemblies, keeping maximum cladding temperature below the design limit. Besides, detailed flow characteristic are obtained by CFD simulation. Furthermore, effects of MSB normal operating pressure and the ambient temperature are studied.

Commentary by Dr. Valentin Fuster
2017;():V008T09A019. doi:10.1115/ICONE25-66546.

The helical-coil once-through steam generator (OTSG) is usually used in the nuclear power plant when the compactness of equipment was taken into consideration. The investigation of flow parameters in the primary side is valuable for the optimization of the OTSG. The purpose of this research is to obtain a further understanding of fluid behaviors in the primary side of the OTSG to achieve a more rational design. Using ANSYS ICEM and ANSYS FLUENT, a three-dimensional (3D) computational fluid dynamics (CFD) model was created and analyzed. Through a series of cases, the velocity profiles and pressure drop through the primary side of the helical-coil OTSG have been calculated, and the influences of different structure designs on the coolant flow parameters have also been tested. Ultimately some pertinent suggestions for improvements were proposed, and insight is obtained into the importance of various modeling considerations in such a model with a complicated structure and large-scale grids.

Commentary by Dr. Valentin Fuster
2017;():V008T09A020. doi:10.1115/ICONE25-66547.

A closed vessel of 1.9 m length and 0.9 m in internal diameter is being designed and will be constructed to experimentally investigate combustion and propagation behavior of the hydrogen-air mixture. Before the experiment is performed, the benchmark and pre-analyses on the hydrogen combustion and propagation behavior in this vessel are carried out by using FLUENT CFD code. The benchmark results indicated that the peak overpressure verse time agrees well with the experimental results of published data. Furthermore, the effect factors such as the hydrogen concentrations, the ignition position, the initial temperature and the pressure were simulated. The results show that the ignition position, initial temperature and pressure have significantly influence on hydrogen behavior. In addition, the overpressure increases with the equivalent ratio and reaches the peak in the vicinity of stoichiometric ratio. However, the overpressure peak is considerably lower than the design pressure of the vessel.

Commentary by Dr. Valentin Fuster
2017;():V008T09A021. doi:10.1115/ICONE25-66582.

During the design process of Reactor Coolant Pump (RCP) test circuit, a flow conditioner is considered to be necessary in view of the fact that the RCP test circuit has a large flow volume and a relatively limited space, and it must meet the need of stabilizing the flow as quickly as possible. This article bases on the demands of the function of the flow conditioner in reactor coolant pump test circuit, utilizes the CFD analysis software to carry out 3-dimensioning modeling of several flow conditioners, and simulates the characteristic of the fluid in circuit under different flow conditioners. Through the analysis of the pressure distribution in the downstream of the flow conditioner, different stabilizing effects are obtained, and the final structure of the flow conditioner is determined considering the real demand of the test circuit.

Commentary by Dr. Valentin Fuster
2017;():V008T09A022. doi:10.1115/ICONE25-66656.

The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). REA is usually an asymmetric transient where neutron kinetics and the thermal hydraulics are strongly coupled (through doppler feedback). With the aim of increasing simulation accuracy by eliminating conservatism at code interfaces, the coupled code SMART-FLICA which links 3D-neutron kinetics code SMART and thermal-hydraulic code FLICA III-F is used to perform REA transient. The analysis will characterize the PWR real plant phenomenology more accurately in the RIA category of accidents.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2017;():V008T09A023. doi:10.1115/ICONE25-66660.

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and local parameter distribution with different bubble diameter models in circular tube under static and rolling condition. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be obtained. The calculation results of local void fraction are compared with experiment data to verify the accuracy of the numerical calculation for subcooled boiling flow under rolling condition. The Zeitoun bubble mean diameter model which the most recommended correlation for bubble diameter under low pressure and several fixed bubble diameters are applied to simulate the same condition in low pressure. These results are compared, include the distribution of void fraction, velocity distribution and radial flow induced by rolling motion.

A good agreement with the experimental data has been achieved when Zeition bubble mean diameter and 2 mm fixed bubble diameter are used to describe vapor diameter in static condition. The local void fraction fluctuation has the same period with the rolling motion, and the fluctuation amplitude increases with the increase of rolling amplitude and rolling frequency. The difference shown in rolling condition between calculation results and experimental data demonstrates that better agreement with the experimental data has been achieved in the near-wall region about local void fraction which has bigger fluctuation amplitude. Higher void fraction has gotten using Zeition bubble mean diameter model to describe bubble diameter in subcooled boiling flow, tiny difference has showed in temperature, velocity and radial velocity in different bubble diameter model. Accurate vapor diameter model or method to describe vapor diameter coupled with suitable interphase force model is needed in rolling condition under low pressure to fit the calculation of subcooled boiling better under rolling condition.

Commentary by Dr. Valentin Fuster
2017;():V008T09A024. doi:10.1115/ICONE25-66666.

The multi-inlet condenser has two steam inlets, the main steam inlet of main steam turbine and the steam inlet of auxiliary steam turbine. In this paper, it establishes a two-dimensional model for the multi-inlet condenser. The porous medium concept is used in the simulations. A porosity factor is incorporated into the governing equations to account for the flow volume reduction due to the tube bundles, baffles and other internal obstacles. The mathematical model is based on the fundamental governing conservation equations of mass and momentum, and the air mass fraction conservation equation. Then, the equations are solved by the SIMPLEC algorithm. Research results indicates that : 1) The physical fields, including pressure, velocity, thermal resistances and condensation rate, are obtained by numerical simulation of the multi-inlet condenser. 2) The performance of multi-inlet condenser is better than the performance of single-inlet condenser from the point of view of pressure drop. 3) There is an optimum angle of auxiliary steam inlet to make the minimum pressure drop. The optimum angle of the multi-inlet condenser used in this study is 50.65°.

Commentary by Dr. Valentin Fuster
2017;():V008T09A025. doi:10.1115/ICONE25-66725.

Two kinds of three-dimensional model are built to simulate the gas entrainment process through a small break in the horizontal coolant pipe at the bottom of the stratified flow. The results were compared with the two-dimensional simulation results and the experimental data. In terms of the two-phase distribution, the simulation results agree well with the experimental data and show much superiority compared with the two-dimensional model. The results verify the reliability of model building, condition setting and calculating method qualitatively and quantitatively. In general, after gas entrainment, the average velocity over cross section increases obviously, but the mass flow rate decreases contrarily. This is because that void fraction meanwhile reduces the fluid density. In addition, it is found that the larger the void fraction of vapor is, the higher the average discharge velocity of the fracture cross-section fluid is. Besides, with the larger internal and external pressure difference, the gas volume fraction and the flow velocity in the break increase, resulting in the mass flow rate increasing along with them. However, since the critical height increases as well, the total loss amount of liquid in the stable effluent stage decreases, and the time before entrainment becomes shorter.

Commentary by Dr. Valentin Fuster
2017;():V008T09A026. doi:10.1115/ICONE25-66729.

During a LOCA accident, the debris caused by the action of high energy fluid discharged from the break may transport to the containment sump, then may be entrained into the core by the ECCS water. The debris may cause the blockage of fuel assembly. The air may also enter the reactor with water. Numerical simulations are performed to analyze the air-water flow and particle-water flow through the blocked fuel assembly. The pressure drop in fuel assembly will impact the long-term core cooling capability. The effects of different parameters on the pressure drop over the fuel assembly are analyzed. Pressure drop increases as blockage percentage, mass flow rate or inlet velocity increases for both two-phase flow. The decrease of air volume fraction and the increase of particle volume fraction all cause the increase of pressure drop. Pressure drop increases slightly as bubble diameter increases, and the tiny effect of particle diameter on pressure drop was found as particle diameter varying at an increment of 10um.

Commentary by Dr. Valentin Fuster
2017;():V008T09A027. doi:10.1115/ICONE25-66782.

A computational fluid dynamics based simulation is performed to optimize the design of the flow distribution device in the lower plenum of the intermediate heat exchanger (IHX) of a pool-type sodium-cooled fast reactor (SFR) in this work. As a typical shell and tube heat exchanger, hot primary sodium flows in the IHX from the top and flows over the tube bundles, called shell-side. The secondary sodium (tube-side) runs through heat transfer tubes and its inlet plenum is specified at the bottom. The flow distribution device is arranged in the lower plenum of IHX, to change the flow distribution of the secondary sodium before into the heat transfer tubes. The CFD tool used in the work is ANSYS Fluent code. Two separated flow distribution devices have been simulated and compared. First, the orifice plates, three flow distribution orifice plates with different positions in the cylinder of lower plenum are respectively set as the model 1, 2 and 3. Secondly, the conical disk model, which is arranged at the bottom of the lower plenum, is established as model 4. And changing the size of the conical disk, the model 5 is established to predict the influence of the size of the conical disk on flow distribution. The results show that all of these models have similar velocity distributions at the outlet of lower plenum, which can be divided into three separate regions, where the flow velocity is higher at the inner and outer, and the velocity in the middle is lowest. When the orifice plate is set at the higher position, the overall velocity distribution is more uniform at the outlet. And the larger conical disk could make a more uniform velocity distribution as well.

Commentary by Dr. Valentin Fuster
2017;():V008T09A028. doi:10.1115/ICONE25-66840.

For AP1000 reactor, passive containment cooling system is a vital way to release heat to the environment, so an accurate prediction of distribution of temperature or density in large layered space plays an important role on the reactor optimization design and safety analysis. This paper investigates in comparing the results of different kinds of models in FLUENT with the experiment results in order to find out a more effective model and more suitable mesh number to simulate the mixing and stratification phenomenon. When LOCA or MSLB occurs in containment, the radius, position, and angle of the break can affect the containment mixing and thermal stratification. So this paper also studies the influence of height of the break and angle of the break on stratification with Fluent, and makes comparative analysis with the experiment results.

Commentary by Dr. Valentin Fuster
2017;():V008T09A029. doi:10.1115/ICONE25-66857.

The motion of CRDM (Control Rod Drive Mechanism) is affected by the electro-magnetic force and the fluid resistance, also the movement process affected the control system. In this paper, we considered all the coupled factor and try to decouple the problem to several simple problems. We supposed 3 decoupling methods, and try to verification each assumption, and finally we found that an assumption is valid — the electromagnetic force is mainly affected by the displacement of CRDM in the attract process of kinetic jaw armature and kinetic jaw magnetic pole. We studied the fluid resistance force by CFD simulation, and exported the force to the electromagnetic model which is built by Infolytica/Magnet, the control system model is built by Matlab/Simulink coupling interface.

In order to simulate the continuous movement process of CRDM, we studied the boundary data at each time period, and linked each period by in-house code. Then we tested the coupling simulation, listed the duration time of electric current and the movement duration time of mechanism. Compare to the test data, the maximum error is 17%. On the whole, this solution coupled CFD, electro-magnetic and control system, the simulation result can help engineers to optimize the control algorithm and shape design in a certain extent.

Commentary by Dr. Valentin Fuster
2017;():V008T09A030. doi:10.1115/ICONE25-66911.

Supercritical water-cooled reactor (SCWR) is one of the Gen-IV reactors, which shows higher economy and safety. Reactor core of a SCWR employs a tight lattice in order to efficiently remove heat from nuclear fuels. In the narrow sub-channels of a tight lattice reactor core, a spacer has been used as a turbulence generator and a space-keeper between the fuel rods. Meanwhile, the spacer has significant influences on the heat transfer of the fuel rods. In order to investigate the effects of simple spacers on heat transfer to upwardly flowing supercritical fluid in a vertical annular channel, an experiment is underway at the supercritical model fluid thermal hydraulics test facility (SMOTH) with supercritical R134A. The equivalent diameter of the annular channel is 6 mm. The outer metal circular tube of the annular channel test section is electrically-heated. The blockage ratio of the simple spacers ranges from 0.2 to 0.4. Based on the geometry parameters of the test section, preliminary numerical investigations were carried out for the effects of simple spacer on the local heat transfer performance of supercritical R134A using commercial CFD code FLUENT. Heat transfer characteristics in the spacer downstream were analyzed with respect to the variations of heat flux, mass flux, pressure, blockage ratio and local enthalpy. And the reason for the different heat transfer enhancement under different conditions is given preliminarily. Finally, existing empirical correlations were selected to be compared with the results of CFD numerical simulation. The applicability of conventional subcritical heat transfer enhancement correlations for spacer grids to supercritical fluid was discussed.

Commentary by Dr. Valentin Fuster
2017;():V008T09A031. doi:10.1115/ICONE25-66955.

The fatigue damage and lift force caused by vortex induced vibration occur very often in the core of the Pressurized Water Reactor (PWR) [1] It is extremely complex to illustrate the mechanism of vibration which induced by Cross-flow. With the spacer grids and wings, the flow direction which in axial direction at the inlet will change and create swirls, so there are many flow directions in the nuclear fuel component. Assumed the tube endure cross-flow only in this article to simplify the fluid model. Most researchers in this field often ignore the displacement of structure induced by the cross flow because the value is so small that not enough to change the fluid region. In truth conditions, the motion of the cylinder caused the wake oscillation and strengthen the vortex shedding, in turn, the vortex shedding will aggravate the vibration amplitude. According that, one way FSI (Fluid Solid Interaction) can’t capture the influence from the cylinder vibration. In this article, Two-way FSI method was executed to get the vibration in time history in order to get the random vibration induced by the cross flow more close to the actual project. Using Finite Volume Method to discrete the fluid control equation and finite element method to discrete structure control equation combined with moving mesh technology. An interface between the fluid region and the structure region was created to transfer the fluid force and the structure displacement. Coupling CFD code and CSD (Computational Solid Dynamics) code to solve the differential equation and obtain the displacement of the cylinder in time history. A Fast Fourier Transfer (FFT) has been done to get the vibration frequency. An Analysis of the vortex shedding frequency and vibration frequency to find the correlation between the vortex shedding and the vibration frequency has been done. A modal analysis for the cylinder without water has been done to get the natural frequency. Results shows the cylinder has different response to the vortex shedding at different position of the cylinder in the same condition. There are more works need to be done aim to get the vibration mechanism in tandem tube and parallel tube to get clearly mechanism of vortex induced vibration in nuclear fuel assembly. The research of the vortex induced vibration in this article is a key to get on the follow research in more tubes array in different methods.

Commentary by Dr. Valentin Fuster
2017;():V008T09A032. doi:10.1115/ICONE25-67020.

The CAP1400 reactor internal is going to use a new component termed the “Even Flow Distributor (EFD)”, instead of the existing flow skirt (FS) design, to help distribute the incoming flow more evenly to the fuel assemblies. To verify the effect of the EFD, a scale model of the reactor and internals was built and hydraulic tests of both the EFD and the FS configurations were conducted. In addition, numerical simulations of the flow fields, using CFD, of both designs were also carried out.

From the scale model test results, the overall flow distribution of EFD is better than that of the FS. The core inlet flow distribution taken from the CFD results is slightly better than that from the hydraulic test. The differences between CFD result and test results are less than 3 percent for the most of fuel assemblies, and about 5 percent for a few assemblies. Based on this study, it is concluded that the EFD is a very effective means of controlling core inlet flow distribution in a CAP1400 reactor.

Topics: Flow (Dynamics)
Commentary by Dr. Valentin Fuster
2017;():V008T09A033. doi:10.1115/ICONE25-67026.

Pressurizer surge lines are essential pipeline structure in NPPs, and the thermal stratification in surge line is recognized as one of the possible cause of thermal fatigue. In this paper, a Computational Fluid Dynamic (CFD) method has been adopted to simulate temperature fluctuations on the process of temperature rising in a pressurizer surge line under rolling motion of single degree of freedom. This work focuses on a fundamental description of differences of thermal stratification between the surge line rolling around the coordinate X-axis condition and that in a static state. The Large-eddy simulation (LES) model is employed to capture the details of temperature change in surge line. Temperature distributions near the inner wall of a surge line pipe with or without swinging were monitored and compared. The temperature differences between the top and bottom of the pipe sections are employed to represent the maximum temperature differences at all the monitored sections. As the surge line swinging, the pattern of temperature distribution and the length of thermal stratification development are different from that in a static. Fluid temperature fluctuation in surge line occur periodically during the fluid temperature rising when the surge line is rotated with the X-axis, and the temperature difference between top and bottom of the surge line is reduced in the same motion mode compared with the static state.

Commentary by Dr. Valentin Fuster
2017;():V008T09A034. doi:10.1115/ICONE25-67027.

During accidental situations, large quantities of hydrogen are generated and released due to metal-water reaction. Eventual stratification formed in the top region of containment and locally high concentration threatens the integrity of nuclear power plants. The stratification of hydrogen may be eroded by several measures, such as air jet, spray, ventilation. So far, several activities have been carried out on the hydrogen stratification’s break-up study. The well-known OECD/SETH-2 project used PANDA and MISTRA test facilities to study this phenomenon with variable test conditions. These test cases are also wildly used to validate the computational fluid dynamics (CFD) codes.

The large-scale containment test facility – COntainment Thermal-hydraulics and Hydrogen Distribution (COTHYD) was designed and constructed at CGN to study the containment thermal hydraulics behavior and hydrogen risk in PWR during severe accidents. A series of tests including helium and air stratification, helium and air stratification eroded by air jet, as well as steam dispersion and condensation are planned to be performed on this test facility. The objective is to set up high quality database for code validation and physical phenomena research.

During the test cases preparation, the relevant tests carried out since 2000 in MISTRA, TOSQAN, THAI, PANDA were collected to give the reference for the cases design in COTHYD. For the tests of stratification’s break-up on COTHYD, helium is discharged from top and it accumulates at the top region of containment. On the preparation stage, Code_Saturne was used to define the test scenario and predict helium distribution. Both dead volume and open volume (with ventilation) are modeled to investigate helium stratification formation and helium-rich layer concentration evolution. These results will be used as test matrix configurations. Code_Saturne, EDF in-house open-source software, has been used in the simulation of hydrogen dispersion. In 2015, B. Hou presented their work on the simulation of the break-up phenomenon of helium stratification by air (ST1_7 and ST1_10 test cases of OECD/SETH-2 project) (Hou et al., 2015) [1]. Simulation results demonstrated that Code_Saturne can well predict and simulate this phenomenon. This software is, as a consequence, used as the case design and validation tools in the pre/post experiment steps.

Commentary by Dr. Valentin Fuster
2017;():V008T09A035. doi:10.1115/ICONE25-67038.

The CAP1400 reactor coolant main pipe plays a very important role in the plant safety operation, as it transports coolant among the Reactor Vessel, the Steam Generator and the Reactor Coolant Pumps, simultaneously connects the main equipments. The flow resistance of CAP1400 coolant main pipe is calculated by classic flow resistance calculation formula at present, but a big tolerance would be caused by the big size, complex structure and multi system/ instruments interfaces of the main pipe. Otherwise, CAP1400 Reactor Coolant System employs elbow flow instrument to measure coolant flowrate, the big size piping nozzle like ADS 4th and surge line might influence the stability of flow field, which affects the accuracy of the elbow flow instrument. A CFD software STAR-CCM+ is used to model CAP1400 main pipe hot leg in this paper and analyze the detailed flow field to obtain more precise calculation results, and verify if the elbow flow instrument arrangement is reasonable.

Commentary by Dr. Valentin Fuster
2017;():V008T09A036. doi:10.1115/ICONE25-67042.

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

Commentary by Dr. Valentin Fuster
2017;():V008T09A037. doi:10.1115/ICONE25-67057.

The great spray area of a Super Large-Scale Natural Draft Cooling Tower (SLNDWCT) makes it difficult to achieve an uniform wind field, and non-uniform water spray distributions are adopted in engineering. In this paper, to improve the cooling performance, optimized non-uniform water spray distributions are designed by utilizing network hydraulic calculations and numerical simulations. In the network calculations, the node-formula is applied to figure out the water pressure and flow rate of each spray nozzle, providing more precise data in simulations for the heat and mass transfer. Simulations for operating in summer, Spring/Autumn and winter seasons, which are different in water spray density, have been presented. In the operating in summer, the spray zone is divided into two regions (inner and outer regions), and by adjusting the water spray density and areas of the two regions, an improved water distribution is achieved.

Commentary by Dr. Valentin Fuster
2017;():V008T09A038. doi:10.1115/ICONE25-67061.

Neutron noise analysis has been done over the decades to predict fuel assembly vibrations and to evaluate safety related issues. Neutron noise occurs due to several reasons: the vibration of the fuel rods, flow obstacles such as rod bending and crud deposition, the moderator temperature and time dependent changes caused by varying flow distributions within a fuel assembly, etc. In order to have a better insight of the neutron noise, a fluid mechanics, structural and neutronics coupled code is developed to perform detailed multiphysics simulations at the level of the fuel rods inside a fuel assembly. In this paper the coupling routine of both steady state and transient calculation is described and the outcome is discussed under several scenarios to understand the influence of rod vibration, moderator temperature and flow distribution on the neutronic field. This paper presents the methodology to couple the multiphysics Computational Fluid Dynamics (CFD) code ANSYS-CFX 16.0 with the 3D neutron diffusion code PARCS v3.0. The model for a 16×16 Pressurized Water Reactor (PWR) fuel assembly is set up for ANSYS-CFX. A sensitivity analysis is carried out to obtain the optimal mesh parameters which results in a good accuracy, as well as a small need for computation capability. Transient cases are studied on a quarter fuel assembly applying oscillating moderator inlet boundary conditions in which the inlet moderator temperature and the inlet moderator velocity are varying over time. In order to simulate the vibration of the fuel rod, the fuel rod part is implemented as immersed solid in ANSYS-CFX. Different vibration modes are applied to both cases: individual single rods of the fuel assembly, and all rods of the fuel assembly. The results of each case are shown in this paper giving a better understanding of how axial power distribution develops with varying flow conditions and vibrating fuel rods.

Commentary by Dr. Valentin Fuster
2017;():V008T09A039. doi:10.1115/ICONE25-67150.

Deflagration to detonation transition (DDT) is a quite challenging subject in computational fluid dynamics both from a standpoint of the phenomenon nature understanding and from extremely demanding computational efforts. In the article the hybrid DDT combustion model is introduced as an efficient method to simulate the DDT problems. As verification, two DDT experiments made in experimental facility MINI RUT are used.

Commentary by Dr. Valentin Fuster
2017;():V008T09A040. doi:10.1115/ICONE25-67179.

Reactor coolant pump (RCP) is one of the most critical devices in third generation of pressurized water reactor nuclear power plant. EMD shield pump and KSB wet winding pump are two representative kinds of RCPs without complex shaft seal system. Due to cancellation of shaft seal system, the entire rotors (including the flywheel) are immersed in the coolant. The losses in RCPs take one third of the total power including rotation loss caused by rotor in the water, electromagnetic loss in the shielding sleeve,the heat transferred through high temperature coolant, and heat generated by bearing.Because of the losses listed above, bearing and winding are heated,and the losses make temperature rise. in order to ensure that the motor is working properly at low temperatures, the company EMD and KSB design the RCP internal cooling circulation which brings the heat out to ensure the normal operation of the RCPs.

The RCP internal cooling circulation includes inlet flow area, auxiliary impeller, thrust bearing, the lower flywheel, motor can, upper radial bearing, upper flywheel, outlet flow area, and external heat exchanger,etc. Flow characteristics in every flow path determine the flow distribution and heat transfer, and the flow distribution determines whether the cooling performance of RCP internal cooling circulation meets the requirements. In order to control operating temperature of motor and bearing, and to optimize heat transfer, adjusting the size of flow area and changing the flow characteristics arecritical. flow field and temperature field in RCP internal cooling circulation need overall analysis.

Flow distribution can be obtained theoretically through the calculation of an overall three-dimensional model.But on the one hand, the calculation time is long due to a complex three-dimensional model with a large quantity of grids, on the other hand, it is easier to casue errors in local processing and the errors are difficult to find or correct. For rapid analysis and optimization of flow and heat transfer in RCP internal cooling circulation, ensure the motor winding and bearing operate at an appropriate temperature, the local characteristics of RCP internal cooling circulation are studied, one-dimensionalanalysis method of RCP internal cooling circulation is developed. This one-dimensional analysis method can be used to predict the flow distribution of each part of RCP internal cooling circulation according to change of the channel geometry parameters, key dimensions, boundary conditions and rotor speed. The geometric parameters are optimized by analyzing the flow distribution, and the purpose of design guidance are achieved.

Commentary by Dr. Valentin Fuster
2017;():V008T09A041. doi:10.1115/ICONE25-67203.

The plate-type fuel element is widely used in ship-based nuclear reactors. The typical coolant channel in such reactor is a narrow rectangular one with a relatively large aspect ratio. The thermal-hydraulic characteristics in narrow rectangular channels may be different from that in conventional pipe. It could be expected that the geometry of narrow channels has some impact on the bubble behaviors in subcooled flow boiling.

In this paper, the bubble behaviors were simulated by using CFD software FLUENT in narrow rectangular channel with different widths under both stagnation conditions and flow conditions. The main concern was placed on the parameters of bubble departure time, departure diameter and departure velocity.

In the model setup process, the VOF model and PISO algorithm were selected. The results were compared against the experimental data and showed the models good capability of capturing the interface between gas and liquid. Furthermore, for stagnation cases, the width of the channel played an important role for bubble departure diameter. In flow cases, the width also had a great impact on departure diameter and departure velocity as well. The departure time and departure velocity showed strong relevance with each other under different contact angle conditions.

Commentary by Dr. Valentin Fuster
2017;():V008T09A042. doi:10.1115/ICONE25-67222.

Sub-channel thermal-hydraulics program named CORTH and assembly lattice calculation program named KYLIN2 have been developed in Nuclear Power Institute of China (NPIC). For the sake of promoting the computing efficiency of these programs and achieving the better description on fined parameters of reactor, the programs’ structure and details are interpreted. Then the characteristics of linear systems of these programs are analyzed. Based on the Generalized Minimal Residual (GMRES) method, different parallel schemes and implementations are considered. The experimental results show that calculation efficiencies of them are improved greatly compared with the serial situation.

Topics: Linear systems
Commentary by Dr. Valentin Fuster
2017;():V008T09A043. doi:10.1115/ICONE25-67256.

Dynamic characteristics of wire-wrapped fuel rod was studied when the structure was coupled with fluid. In this paper, the wet modal analysis of wire-wrapped fuel rod in the axial flow was investigated by using the finite element method, which relied on fluid-structure interaction calculation. This numerical simulation is compared to experiments in the literature and the results agree well. The Lead-Bismuth Eutectic (LBE) was used as the working fluids. The CFD Simulations of a wire-wrapped fuel rod were performed with different boundary conditions. By comparing the modal characteristics of the fuel rod in the LBE and in vacuum, the results showed that the added mass of the fluid has important effect on vibration frequency of the wire-wrapped fuel rod. The different three frequencies and modal shape can be obtained under the different inlet velocity. According to the simulation result, it could be found that the maximal equivalent elastic strain occurred in the constrained end.

Commentary by Dr. Valentin Fuster
2017;():V008T09A044. doi:10.1115/ICONE25-67277.

Study of gaseous explosions and their effects on structures is helpful in designing offshore platforms. Specifically, reliable methods for the prediction of overpressures in offshore explosions are highly useful and are extensively researched. The selection and/or development of means of prevention, control and mitigation of explosions often depends on the comprehensive analysis of their probability of incidence and damage potential. This involves a number of factors, such as explosive gas leak size, location, composition, wind direction, and characteristics of probable ignition. This paper presents a 3D transient CFD based analysis tool for such purposes and the results of some simulations done using it. The first set of simulations is a validation exercise, which involves hydrogen leakage and explosion, and the computational results are compared with the experimental data. The second set of calculations involved simulation of a hydrogen gas leakage scenario on an offshore platform, followed by explosion studies for different scenarios to find the effect of various guidelines for the initial conditions in the reactive cloud. These results show that, the maximum explosion pressure occurs when stoichiometric initial mixture conditions are applied in the dispersed flammable region. The worst case explosion scenario thus observed has maximum over pressures and maximum blast wall displacement of about 18 to 20 times higher than the base case explosion.

Commentary by Dr. Valentin Fuster
2017;():V008T09A045. doi:10.1115/ICONE25-67278.

This work describes the activity performed at the University of Pisa concerning the application of an in-house developed coupling methodology between a modified version of RELAP5/Mod.3.3 and the ANSYS Fluent commercial CFD code to a pool system. Mono-dimensional codes, like RELAP5, are commonly used for thermal-hydraulic analysis of entire complex systems. Nevertheless, their one-dimensional feature represents a limit in the analysis of such problems where significant 3D phenomena are involved. On the other hand, CFD codes standalone are usually employed to simulate relatively small domains. The use of System Thermal-Hydraulic + CFD coupled calculations can overcome these issues, allowing the simulation of a complete system, but with a part of the domain reproduced with the CFD code.

In this work, the coupled calculation technique was used to simulate a PLOHS + LOF transient in the HLM experimental facility CIRCE (CIRCulation Experiment), located at the ENEA Brasimone research centre. The paper initially calls up the coupling procedure adopted, consisting in a “two-way” coupling. MATLAB software, used as external interface, manages the exchange of data between the system and the CFD code. The numerical method adopted for the coupling is the implicit scheme. Then, the main features of the CIRCE facility are briefly described, so are the two computational domains employed in this study. In particular, the CFD code was used to model the CIRCE pool (8 m high) and the Decay Heat Removal (DHR) heat exchanger. Due to the long duration of the transient simulated, a 2D axial-symmetric domain was chosen in order to reduce the computational time. The test section, placed inside the pool and consisting in a heat source and a heat sink, and the secondary side of the heat exchanger, were modeled with RELAP5. The use of the coupling tool allowed to set realistic boundary conditions in the calculation, more representative of the experimental ones. The main numerical results obtained from the PLOHS + LOF coupled calculation were compared with experimental data. Calculated LBE mass flow rates in the test section and in the DHR showed good agreement with experimental data. Some discrepancies with respect to the experimental trends were noticed for LBE temperatures; these should be related to some simplifications introduced in the model. Nevertheless, obtained outcomes represent a preliminary guideline for the improvement of the modeling for future works.

Commentary by Dr. Valentin Fuster
2017;():V008T09A046. doi:10.1115/ICONE25-67304.

Accurate prediction of condensation plays an important role in the development of high efficiency turbo-machines working on condensable fluid. Therefore it demands modeling of poly-disperse characteristic of number distribution function while modeling condensation. Two such kind of models are considered in this work and they are namely, quadrature method of moments (QMOM) and multi-fluid method (MFM) models. The vital difference between these two models lies in the method of discretisation of the droplet size distribution. Further, their numerical aspects like ease of implementation in general purpose computational fluid dynamics solvers, accuracy and associated computational cost are discussed. In order to obtain accurate thermodynamic properties, the real gas formulations defined in IAPWS-IF97 are used. These algorithms are applied to the compressible Navier-Stokes solver of Fluidyn MP and tests are carried on Laval nozzle and compared with the experimental measurements.

Commentary by Dr. Valentin Fuster
2017;():V008T09A047. doi:10.1115/ICONE25-67320.

The thorium molten salt experimental reactor with solid fuels (TMSR-SF1) was one of the conceptual designs developed by Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP, CAS). The fuel pebbles at the reactor core and the pool type structure of the vessel increase the complexity of thermal-hydraulic (T-H) analysis of the reactor. In order to analysis the T-H feature of TMSR-SF1 in case of the postulated Loss Of Forced Cooling (LOFC) accident, and investigate whether its external air cooling system with nature convection is capable of removing the residual heat, the Computational Fluid Dynamics (CFD) method was used to model the reactor and simulate the transient. In this research, an integrated pseudo 2-D thermal-hydraulic model of the core was developed and a simulation and analysis of the LOFC accident has been conducted. The preliminary calculation results using CFD method show that the external air cooling system has the capability of removing the residual heat. The calculation results also indicate that the peak temperatures of the fuel pebbles, key components and structures of TMSR-SF1 remain under the safety limits and the temperature of the molten salt remains below boiling point.

Commentary by Dr. Valentin Fuster
2017;():V008T09A048. doi:10.1115/ICONE25-67345.

During a postulated main steam line break (MSLB) event of a Pressurized Water Reactor (PWR) initiated at the Hot Zero Power (HZP) condition, increased heat removal from the broken steam generator (SG) on the secondary side that significantly reduces the coolant temperature on the primary side, and cold primary coolant enters the reactor vessel through the affected loop resulting in asymmetric temperature and mass flux distributions into the reactor core. A plant safety analysis under the MSLB condition needs to account for the thermal and mass flux asymmetry effects on the reactor core response due to the colder water flowing from the affected SG and the reactor coolant system (RCS) to reactor vessel.

High resolution computational fluid dynamics (CFD) methodology with ANSYS CFX (Version 16.1) software was applied to analyze the flow behaviors and thermal-hydraulic phenomena and to study the thermal mixing and asymmetry effects in the downcomer and lower-plenum of a typical Westinghouse design four-loop PWR under the MSLB conditions. Two scenarios were considered for the CFD simulation distinct by reactor coolant pump status:

(1) Low-flow case: without offsite power where the reactor core is cooled through natural circulation

(2) High-flow case: with offsite power available and the reactor coolant pumps in operation

The CFX CFD modeling and simulation were based on the reactor vessel boundary conditions from a system code transient simulation at the limiting time steps with respect to thermal margin of the fuel design. The geometric model included the vessel downcomer and the lower internals up to the reactor core inlet below the fuel assemblies. The results of CFD simulation show the different flow patterns and temperature distributions at the reactor core inlet for the low-flow case and for the high-flow case. Thermal asymmetric effect exists in both cases, but in the low-flow case, cold flow enters into core inlets at the opposite side of faulted loop located, and in the high-flow case cold flow enters into core inlets at the same side of faulted loop located. A mass flux asymmetric effect exists in both cases, but for the low-flow case, the core inlet mass flow distribution is more uniform than that for the high-flow case. The reactor core inlet distributions under the MSLB condition were further evaluated through comparisons with the results from the STAR-CCM+ (Version 10.04.01) CFD modeling and simulation. The evaluation showed that the simulation results are in good agreement with the STAR-CCM+ predictions and consistent with the phenomenon observed in an experiment published in open literature and site engineer judgment based on the available detected data.

Commentary by Dr. Valentin Fuster
2017;():V008T09A049. doi:10.1115/ICONE25-67375.

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of nuclear containment building. Therefore, getting detailed hydrogen flow and distribution is a key issue to resolve the hydrogen risk in containment and compartments. In this study, local hydrogen behavior has been investigated in a multi-subcompartments using Computational Fluid Dynamics method, concerning the local concentration in the multi-subcompartments. The multi-subcompartments containment is represented by four vessels interconnected by pipe. Hydrogen inlet pipe is located in source vessel. The analysis model is built by 3-dimensional Computational Fluid Dynamics code ANSYS-CFX in Cartesian coordinates. Computational domain is discretized in about 47559 cells, 87484 cells and 126388 cells, respectively. Taking a full consideration of the computational time and accurate, the medium mesh scheme is adopted to simulate the hydrogen transport.

With standard k-ε turbulence model, the effects of the connecting pipe parameters and direction on hydrogen distribution in the multi-compartments are investigated. Local hydrogen behavior showed remarkably different in simulations with the change of the pipe parameter. Increasing the connection pipe diameter and decreasing the connection pipe length are helpful for helium flow between compartments. Compare with the vertical connection pipe, horizontal connection pipe is not conducive to the diffusion of hydrogen from source compartment to non-source compartment.

Commentary by Dr. Valentin Fuster
2017;():V008T09A050. doi:10.1115/ICONE25-67721.

In this paper,we studied the characteristic of four kernel functions of Moving Particle Semi-implicit method (MPS).In order to find the dependence of pressure oscillation on kernel functions, the dam break problem was selected for as a benchmark.Results showed that proper value of a kernel function was required in the compact support region to smooth the pressure distribution. The value of the kernel function should approach zero when the particle distance was close to the edge of compact support region to avoid large replusive force.When the distance between particles was close to zero,the value of kernel function should be large enough to avoid particle clustering. At the same time,the kernel function should be simple to stabilize the simulation process and decrease the computational time.

Commentary by Dr. Valentin Fuster
2017;():V008T09A051. doi:10.1115/ICONE25-67746.

The type-II 200MW nuclear heating reactor (NHR200-II) designed by Tsinghua University is a novel pressurized water reactor, whose main specifications are as follows, the heat power is 200 MW, the design pressure of major loop is 10 MPa, the inlet and outlet temperature of coolant are 230 °C and 278 °C. In this paper, a quarter three-dimensional model of the lower plenum of the reactor pressure vessel is set up for analysis. The flow velocity distribution and the pressure distribution on the core supporting structure are calculated by method of three-dimensional numerical simulation. The results show that the lower part of the core produce symmetric vortex due to the existence of support structure. The production of the symmetric vortex, to some extent, increases the instability of the flow. On the other hand, the existence of the vortex is good for uniformity of flow distribution in the outlet holes. The flow rate in the flow channel of support structure is lower at the center and larger in the margin of core inlet. The results show that the maximum of the velocity in the flow channel is 5% higher than the minimum one.

Commentary by Dr. Valentin Fuster
2017;():V008T09A052. doi:10.1115/ICONE25-67766.

Oil coolers are widely-used in nuclear power plants which cool oil flowed through the bearing of steam turbines to a suitable temperature in order for the safe operation of steam turbines. Owing to the high viscosity, the flow state of oil is generally laminar flow or transition flow, which easily leads to poor heat transfer capability and thus a large volume of the oil coolers. The insertion of twisted tapes in circular tube is a passive method widely-used for enhancing heat transfer of laminar or transition flow. However, little research focuses on the heat transfer of highly-viscous fluids inside tube using the twisted tapes. The article will present the numerical simulations of the swirling flow induced by a new coaxial cross twisted tape inserts in a heat transfer tube with lubricating oil. The effects of the clearance ratios and twist ratios on oil side heat transfer coefficient, friction factor and performance evaluation criterion will be numerically investigated using CFD computer software STAR-CCM+. The clearance ratios are 0.077, 0.154 and 0.231. The twist ratios are 2.0 and 4.0. The boundary condition of simulation is constant wall temperature for the Reynolds number ranging from 200 to 1300. The results indicate that the new coaxial cross twisted tapes are efficacious in enhancing the heat transfer of the lubricating oil inside tube. When the clear ratio is 0.077, the effect of heat transfer enhancement of the coaxial cross twisted tapes is better than that of traditional twisted tapes. Furthermore, the highest performance evaluation criterion is up to 2.3.

Commentary by Dr. Valentin Fuster
2017;():V008T09A053. doi:10.1115/ICONE25-67767.

The flow rate distribution at the entrance of the core plays a key role for reactor design since it has important implications for the performance, and efficient safety of a nuclear reactor. When the coolant passes from the downcomer to the core, it changes direction due to the inertia force and the curvature of the bottom vessel head. The internal components inside lower plenum work to homogenize the flow distribution. Their purpose is to prevent the formation of instabilities and the creation of vortices due to the flow reversal. In the frame of EDF’s new reactor design there is a desire to identify an optimal flow diffuser. The future intention is to study five different types of flow diffuser including EPR, VVER, Konvoï, APR+ and Westinghouse to look at the pros and cons of each design. The authors underline that the geometries of each Reactor Pressure Vessel (RPV) and associated diffuser device are quite different therefore a generic form needs to be used to make an equivalent comparison. The goal of the present work is to find the optimal mesh refinement and associated numerical parameters for the simulation of the lower plenum flow. This work is a preliminary step for a future study to compare existing diffuser concepts. Thus in the future work only the section containing the flow diffuser structure will be changed.

The PIRT methodology is applied to better define the physical phenomena and key parameters that will influence the flow distribution at the entrance of the core. In order to better understand the fluid distribution and the function of the diffuser component, 3D computational fluid dynamics (CFD) simulations are launched to improve our knowledge on the flow pattern inside the lower plenum.

Both the geometry and mesh are generated by Salomé1. Simulations are carried out using Code_Saturne2, an EDF in-house open-source CFD code. The generic test case is a 1/5 scale EDF “BORA” 4 loop mock-up with a flow rate of 0.1 m3/s injected into each cold leg. The unsteady flow algorithm with standard k-epsilon turbulence model has been used with a full explicit meshing except for the reactor core where a porous approach is adopted. The physical time for each calculation case is 5s for a converged simulation. Mesh sensitivity tests have been carried out ranging from 8 million cells to 28 million cells. A mesh of 22 million cells is found to provide the most appropriate balance between simulation quality and feasibility. Due to the size of the simulations, high performance computers are necessary to provide timely results. The results indicate that CFD can provide extra capacity to engineers for reactor design to evaluate the pros and cons of different existing diffuser concepts.

Commentary by Dr. Valentin Fuster
2017;():V008T09A054. doi:10.1115/ICONE25-67870.

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan Atomic Energy Agency, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor during the core disruptive accident. Thermal-hydraulics evaluations in FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, thermal-hydraulics in FAIDUS are investigated by using a subchannel analysis code ASFRE, which is applicable to a wire-wrapped fuel pin bundle with a distributed resistance model and a simplified turbulence mixing model. At first, the distributed resistance model was validated by comparison of pressure drop coefficients with experimental data obtained in water experiments with simulated FAs under the condition of wide-range Reynolds number. And then, the turbulence mixing model was validated by comparison of temperature distribution in the pin bundle with experimental data obtained in sodium experiments with simulated FAs. After the applicability of ASFRE to FAs was confirmed through these validations, thermal-hydraulic analyses of a FA with 271 fuel pins without the inner duct and a FAIDUS with 255 fuel pins were conducted. The obtained results indicate that no significant asymmetric temperature distribution occurs in a FAIDUS as a FA without an inner duct. In addition, the temperature distribution of FAIDUS with 255 fuel pins under the low flow rate condition tended to be the same as that of a FA with 271 fuel pins due to the local flow acceleration and the flow redistribution caused by the buoyancy force.

Commentary by Dr. Valentin Fuster
2017;():V008T09A055. doi:10.1115/ICONE25-67876.

Thermal striping on the core instrumentation plate (CIP) at the bottom of the upper internal structure (UIS) of an advanced loop-type sodium-cooled fast reactor in Japan (Advanced-SFR) has been numerically investigated. At the top of the core below the CIP, the sodium at high temperature flows out from the fuel subassemblies (FSs) and the sodium at low temperature flows out from the primary control rod (PCR) and backup control rod (BCR) channels, and also the radial blanket fuel subassemblies (RBFSs) at the outer side of the core. In order to predict the thermal striping on the CIP caused by mixing fluids at different temperatures from the FSs, the PCR and the BCR channels, and the RBFSs, a numerical estimation method using a spatial connection methodology between the upper plenum analysis and the local area analysis for the target area has been developed. By using the connection methodology, the numerical simulation considering the influence of the transversal flow in the UIS and the external flow around the UIS in the upper plenum can be performed to improve the accuracy of the estimation results. In this paper, the outline of the spatial connection methodology including data transfer technique from the upper plenum analysis to the local area analysis was described. As a validation process, numerical simulation of the water experiment using the test apparatus named TAFUT which was 1/3-scaled 1/6 partial model of the upper plenum of the Advanced-SFR was performed to confirm applicability of the spatial connection methodology to a practical thermal striping problem. The numerical result of temperature distribution was compared with the measured result in TAFUT experiment. Additionally, mesh sensitivity of the local area analysis model to the numerical results was indicated by using a small and a large area models in order to suggest an appropriate local area analysis model.

Commentary by Dr. Valentin Fuster

Nuclear Education, Public Acceptance and Related Issues

2017;():V008T12A001. doi:10.1115/ICONE25-66020.

The EAGLE project was a Euratom FP7 which helped to identify and disseminate good practices in information and communication processes related to ionizing radiation.

For this purpose, the consortium reviewed national and international data, tools and methods as well as institutional work in order to identify education, information and communication needs.

Generally in high school the first concepts on radioactivity and ionizing radiation (IR) are introduced mainly in the subjects of physics or physical chemistry.

There are a number of concepts in relation with IR and nuclear topics, and different ways to teach them: theoretical, mathematical, historical or practical. The question also rose, to what extend the various topics related to ionizing radiation (health, environment, history) are dealt with.

As already mentioned, all these questions let to the idea to compare the content dealing with radioactivity and nuclear topics in different physics school books and more specifically schoolbooks for high school students (in the age 17 to 18). The method was as follows:

- For the review the different partners of EAGLE have sent the schoolbooks used for the target group, or scanned documents.

- Spanish schoolbooks and English schoolbooks were purchased to extend the review to other EU countries.

- IRSN works in partnership with a high school based in the French town Vichy.

- Each book was analyzed in detail to list with precision the content. A matrix helped to compare them.

The paper presents the comparison of the contents of these books and their analysis. Some recommendations coming from the Eagle project will be discussed.

Commentary by Dr. Valentin Fuster
2017;():V008T12A002. doi:10.1115/ICONE25-66074.

Nuclear emergency is the important component part of the defense in depth philosophy of nuclear safety, which is the last barrier to protect public in a nuclear accident. According to Chinese codes and standards which are Regulations on the Nuclear Accident in Nuclear Power Plant and Emergency Preparedness and Emergency Response of the Research Reactor (HAD002/06), the research reactor operation units should carry out a comprehensive emergency drill at least once every two years, in order to check the effect of the emergency training about the on-site emergency organization and verify the usefulness and feasibility of the emergency plan. Therefore, the Institute of Nuclear and New Energy Technology (INET), Tsinghua University, had drew up the comprehensive emergency plan, conducted the emergency practice in December 2015, and tested the emergency preparedness of multiple research nuclear installations including the 10MW high temperature gas-cooled reactor (HTR-10), 5MW test heating reactor, shielding-experimental reactor, etc. The purpose, accident scenarios, process, and key content of the emergency drill were descripted in details, the design of the guiding worksheets which were adopted in the emergency drill of INET for the first time were distinctly introduced, and the effect of these guiding worksheets was explicitly discussed. The whole process of the drill in INET in 2015 was reviewed and summarized, the experience feedback of which can supply useful reference information and recommendations for the development of the emergency drill of research reactors.

Commentary by Dr. Valentin Fuster
2017;():V008T12A003. doi:10.1115/ICONE25-66297.

In France, since 2006, the legal framework has been reinforced in order to have a better compliance with the safety features of nuclear installations but also a relevant communication with the public about the nuclear risks and the nuclear decisions. Starting with the Nuclear Safety Transparency law of 2007 defining transparency in the nuclear field as “the set of provisions adopted to ensure the public’s right to reliable and accessible information on nuclear safety”, afterwards reinforced by the Law on Energy Transition and Renewable Energy in 2015. This law reinforced the transparency provisions, requiring not only transparent one way information but also public participation.

Even before the first transparency law, in France, communication consisted in a classic one way information process based on reporting incidents or events which occurred on Nuclear Power Plants or installations. Next it was decided to publish all technical inspection notifications on nuclear installations on the regulatory website. The French Technical Support Organization IRSN, and the Nuclear safety Authority ASN, promote these reports and publications through press conferences and nowadays also through twitter.

IRSN, way before the transparency requirements, made it one of their priorities to develop different methods and tools for the improvement of communication between Experts and Public promoting visibility and trustworthiness. Thanks to the new legal framework, the development of new tools to inform and engage citizens is accelerated. The traditional tools available are annual reports, newsletters, websites, magazines, Press data center, press conferences, etc...today completed with new tools such as YouTube, twitter, Facebook.

In addition, IRSN developed ways and tools promoting direct contact with the public, such as “Open House Days” allowing the public to discover work on site and to dialogue with Experts. In line with the “face to face” formula, IRSN implemented an Information and Education Strategy for the Public to enhance their Radiation Protection and Nuclear Safety Culture.

The objective of this article is to explain further each method developed and the support used to enhance the Public’s Radiation Protection and Nuclear Safety Culture.

Topics: Safety , Education
Commentary by Dr. Valentin Fuster
2017;():V008T12A004. doi:10.1115/ICONE25-66386.

At present, the most influencing factor on nuclear power’s development in China the public’s attitude and acceptance. This paper studies the public perception of nuclear power risk in China, and provides several feasible methods to improve the quality and effectiveness of public perception level. Therefore, the public acceptance of nuclear power can be ameliorated greatly, which will help the development of nuclear power industry.

For decades, the environment pollution has become one of the most serious and urgent problems in China. Since nuclear power has been proved to be a type of low-carbon and environment-friendly energy, striving to develop nuclear power is a good solution to China’s environment issues. However, by the end of 2015, China’s nuclear power’s electricity production share was only 3.03%, which was far below the average level of developed countries. This situation might be partly due to technical and economic reasons, but the essential cause of the restricted development of nuclear power in China is the public perception of nuclear power risk is far from objective and comprehensive, which leads to the public acceptance of nuclear power is not as high as expected.

This paper states that public perception process of nuclear power risk is a dynamic, complex and closed system, which consists of the risk, the transformation of the risk and related information (both truth and rumors), the public perception process and public’s acceptance of the risk, and the public’s actions after receiving the information. The public’s actions often react on the risk perception. Nuclear power risk, unfortunately, is usually magnified, and it makes people become more frightened and oppose nuclear power more seriously.

In order to solve those issues, in this paper, the public risk perception’s characteristics and external influencing factors are studied, a model describes the public perception process of nuclear power risk is developed and analyzed, and the causes of current acceptance level of nuclear power (which is relatively low) are explained. In addition, based on this model, methods to conduct effective risk communication and public education on nuclear power are provided, the future of nuclear power in China can be much better.

Topics: China , Nuclear power , Risk
Commentary by Dr. Valentin Fuster
2017;():V008T12A005. doi:10.1115/ICONE25-66613.

The harmonious development of economic, energy and environment is an important premise to realize the objective of China’s modernization. Currently, different parts of China have different main energy source, while nuclear power development has many opportunities and challenges. This paper considers the current trends of energy needs in China, and discusses the different influencing factors of energy needs throughout China.

In addition, this paper will focus on the potential nuclear power development in China, which mainly focuses on policy, technology, nuclear security and social attitude. Then it will focus on the application in Shenzhen (Daya Bay). Based on previous analysis, technical/engineering feasibility and site feasibility are considered in this part. Finally, a number of recommendations for nuclear development management in China will be given. These recommendations will help the public to have a basic understanding of nuclear power management, and to improve the social attitude of China’s nuclear energy development.

In all, this paper puts forward the management methodology of nuclear power industry, which has positive significance for the field of nuclear power education. Meanwhile, the paper will play a positive role on popularize the knowledge of nuclear power to the public.

Topics: China , Nuclear power
Commentary by Dr. Valentin Fuster
2017;():V008T12A006. doi:10.1115/ICONE25-66789.

With the development of public awareness on environmental protection, especially after the Fukushima nuclear accident, the opposition to nuclear power due to NIMBY (not in my back yard) effect begins to hinder the rapid development of Chinese nuclear industry. For example, in recent years several large-scale mass incidents with appealing to stop the siting and construction of nuclear facilities in China have put related projects (including nuclear power plant and nuclear fuel cycle facility) into termination, resulting in certain financial loss and unnecessary social unstabilization, thus causing more and more concern from administrative authority, research institution and nuclear industry.

To strengthen public acceptance on nuclear power, related enterprises such as CGN and CNNC have made great efforts in information disclosure to eliminate mysterious feelings towards nuclear power and expect to build new impression as clean energy. Domestic institutions and universities carry out plenty of work on methods to help public correctly perceive nuclear risk and present strategies for effective public communication. Administrative authority also issued detailed guidance on public communication required to be fulfilled during plant’s siting phase, which provided explicit provisions on the responsibility and job content of different entities.

Here we will take one public communication practice of one nuclear power project located in south Zhejiang region as an example. In this scenario, we face more difficulty than other projects, such as doubt from local government, complexity of public types, and large amount of stakeholders. In this paper, we will make summary on endeavors to improve public acceptance, such as large amount of NPP visits, comprehensive scientific popularization, direct communication with stakeholders and integration development between local society and nuclear industry. And we will discuss the feasibility of innovative practice, combining several similar tasks needed in different subjects, such as environmental impact assessment and social stabilization assessment, to fulfill at once. To achieve this goal, we design specific questionnaire and use it to survey the opinion of more than 800 people in the fairly large region across different provinces, covering 30km radius area of site, which gains satisfactory results.

By comparing outcomes of opinion surveys carried out before and after the practice, we will put forward to the considerable effect of public communication in improving public acceptance to nuclear power, and analysis the pros and cons of this example. Moreover, we also expect the good experience in practice can be promoted to overall processes of nuclear power plant, including siting, construction, commission and life extension, helping nuclear power gain more public acceptance.

Topics: Nuclear power
Commentary by Dr. Valentin Fuster
2017;():V008T12A007. doi:10.1115/ICONE25-66963.

In case of a severe accident, all the staff members of a nuclear power plant (NPP), members of the Emergency Response Organization (ERO) on site as well as operators in the main control room (MCR) are required to take necessary actions to mitigate the consequential effects of the accident. Therefore, Nuclear Engineering Ltd (NEL) has been implementing education and exercise for severe accident (SA) management both at nuclear power plants and Nuclear Power Division of Kansai Electric Power Company (KANSAI) since FY 2014. For the education of commanders who take a lead at the ERO in case of an accident, table top exercise is provided by using simulators developed by NEL, including functions to respond to a SA involving core melt. Continuous implementation of this education and exercise program is expected to enhance KANSAI’s severe accident management ability and their voluntary safety improvement activities in the future.

Commentary by Dr. Valentin Fuster
2017;():V008T12A008. doi:10.1115/ICONE25-67082.

At present, there are hundreds of nuclear power plants in operation around the world. Anti-nuclear movements continue in many places, although the nuclear power plants have good operating records. It has some factors, and the first factor that the public knows little about nuclear industry, results in regarding the nuclear power plant mysterious. This condition relates to destructive scene by nuclear weapon with nuclear industry, deeming it unacceptable to take this risk. Secondly, construction of nuclear power plant and off site emergency may occupy large land. The public hopes to be rewarded more to offset the risk by their imagination. Last, it relates to the political environment of one country. Every country has its own situation, so the strategies of developing nuclear power plant are widely different. The public is not familiar with other nuclear engineering projects except nuclear power plants, and hence the boycott happens more frequently. Sino-French cooperation on nuclear fuel cycle project is the first large-scale commercial spent fuel reprocessing plant, which is the biggest cooperative project between China and France until now. AREVA is responsible for technology, and CNNC is responsible for building. Spent fuel reprocessing is the most important part of nuclear fuel cycle back end, which separates uranium and plutonium from spent fuel, and manufactures MOX fuel with recycled resources for using in nuclear reactor again. This will make the best use of the uranium resources. After that process, the fission products needed to be disposed reduce significantly. And it is good for environmental protection. The public protest happened in one of the candidate sites, when CNNC carried out the preliminary work of site selection. For meeting the enormous energy demands, the fossil energy may be exhausted in the future due to the greenhouse gases emission. Chinese government speeds up the development of new energy. Nuclear energy is the only technology with no emission of greenhouse gases and will be rapidly developed. Along with the nuclear power units continuing to increase, they become the critical factors in restricting the sustainable development of nuclear energy. That is efficient utilization of uranium resources, spent fuel intermediate storage, reprocessing, and geologic disposal of high level radioactive waste. To this project, it not only has a great current demand, but also closely relates to transition of energy structure. The public has different views in the project progressing, which results in wide concern and discussion. The article took this event for example, and analyzed the reason from all directions. Besides, the author put forward own views for the public acceptance events about nuclear engineering projects except nuclear power plant.

Commentary by Dr. Valentin Fuster
2017;():V008T12A009. doi:10.1115/ICONE25-67330.

Companies involved in the nuclear energy domain, like component and platform manufacturers, system integrators and utilities, have well established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees, or indoctrinations of temporary staff. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on IT, the daily office work, test bay or construction site work environment, and some data protection and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrators and utilities.

Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new Nuclear Power Plant (NPP). The engineers working on safety, physical protection and cybersecurity, must be aware of these interrelations to jointly elaborate a robust I&C architecture (defense-in-depth, design basis events, functional categorization and systems classification) and a resilient security architecture (security by design, security grading, zone model or infrastructure domain, security conduits, forensic readiness, Security Information and Event Management).

This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines, will be discussed.

Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the inter-disciplinary cooperation in parallel to the allocation of resources, e.g. on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.

Commentary by Dr. Valentin Fuster
2017;():V008T12A010. doi:10.1115/ICONE25-67689.

Started with the concept and significance of Public Participation (PP), and based on the current situation of internet voluntary action of nuclear related affair in China, this paper indicated that in China, PP in Environmental Impact Assessment (EIA) of Nuclear Power Project (NPP) was inadequate and cannot play the appropriate role in decision-making of NPP. By surveying the situation of PP of nuclear power industry in the developed countries, which focused on the incepts in the legal basis in PP, the participants, the participation stages and the methods, some issues and flaws in PP of China were indicated. Meanwhile, some suggestions were proposed as a reference of PP in EIA of NPP in China and in favor of the development of nuclear power safety.

Topics: China , Nuclear power
Commentary by Dr. Valentin Fuster
2017;():V008T12A011. doi:10.1115/ICONE25-67825.

To ensure that adequate protective actions are in place for the public, a salient lesson learned from Fukushima is the necessity to improve the effectiveness of the off-site response, namely the effective implementation of protective actions in a nuclear emergency.

Among recent research on nuclear emergency, little attention has been paid to public participation, where the disconnect between the public and authorities, and its negative effect on emergency response exist. This study conducted an analytic discussion on the effectiveness of off-site nuclear emergency, from the standpoint of public participation.

The two key factors contributing to effective emergency responses in a nuclear emergency were identified to be the feasibility of emergency plans and the adequacy of emergency preparedness (EP), to which the passive role the public has been playing does no good. First, nuclear emergency plans are developed unilaterally by emergency managers and authorities, without the public involved. This government-centric planning process usually fails to meet the actual needs of the residents should a nuclear accident occur, consequently impairing the feasibility of emergency plans. As regards EP, emergency management’s efforts have long been dedicated to maintain the response capabilities of emergency response personnel, while overlooking the EP of the public. In this case, the public will not be well-prepared for an emergency.

Corresponding to the deficiencies stated above, possible solutions to improve the overall effectiveness of off-site emergency response were proposed, from the perspective of increasing public participation. First, to make emergency plans feasible and comprehensive, 1) the public can be incorporated in planning process to consider their needs in emergency plans, 2) emergency plans should be periodically assessed and updated accordingly, based on the up-to-date socio-demographic information. Second, to ensure the effective implementation of EP, 1) the public should be educated more on the knowledge of radiation protection and emergency response, in a participatory rather than informational way, 2) More-realistic nuclear exercises, such as evacuation drills of the population-at-risk, could be cautiously carried out, to test whether the public are well-prepared under emergency conditions. Finally, a precondition of broad public participation is that the public have interest in nuclear emergency. To this end, information communication technologies, should be widely utilized in nuclear emergency to generate public interest, by facilitating two-way communication and displaying the emergency-related information in an easy-to-read way.

This study indicates that nuclear emergency should not be a process dominated by emergency managers alone, since the public are not only the protected but also the true first responders in nuclear accidents. Wider public participation should be incorporated into the whole process of emergency management, from planning to preparedness, to maximize the effectiveness of the off-site response to a nuclear emergency.

Topics: Emergencies
Commentary by Dr. Valentin Fuster
2017;():V008T12A012. doi:10.1115/ICONE25-67828.

Public participation systems in environmental impact assessment started late in our country. Relevant laws, regulations, and work protocols need to be further improved. In this study, extensive research was conducted on the public participation systems in the environmental impact assessment of foreign nuclear power plants. Analyze the current status of our public participation systems were drawn from legal aspect and the aspect of implementation. Together with case analysis, main problems of public participation systems in environmental impact assessment of China’s nuclear power plant were summarized from this study: (1) delayed information disclosure; (2) the scope of public participation need to be widened; (3) interactive platforms are required for convenient and efficient public participations instead of a single participation approach; (4) timely response to the platforms and more supervision over the participation systems are desired. Solutions to each problem are proposed to help develop relative regulations and the implementation of these regulations.

Commentary by Dr. Valentin Fuster
2017;():V008T12A013. doi:10.1115/ICONE25-67942.

Nuclear safety law and regulation related to the field of public acceptance and nuclear education of nuclear power (NP) is in blank, which seriously restricts the sustainable development of nuclear energy. In order to make up for nuclear safety legal system in the field of public acceptance of nulcear power plant (NPP) and education of NP and related fields, this paper provides data and gives some advice. The investigation used a random sample survey method to made a field survey of 301 people living near inland NPP. In this survey, 55.48% of people are concerned about NPPs development in their daily life, 73.09% of people support the development of NPPs, 63.79% of people are against NPPs built in the vicinity of their own, 30.90% of people think that NPPs is safe, 76.08% of people worry about the harmful impact on the surrounding residents, 81.06% of people know and learn the knowledge of NP by television and internet, 19.93% of people know the related knowledge of NP by government propaganda. In conclusion, most of people support and pay attention to the development of inland NPP, but they do not have a deep understanding of nuclear safety so they do not fully accept the inland NPP. It is suggested that we should increase information publicity of public network, promote popular science propaganda about NPP by government, improve nuclear safety laws and regulations, and strengthen the study on the psychological mechanism of public NP. If we can do all of this, it will help to improve the development of inland NPP Using the internet and television popularizes the NPP Invite officials from the site of the new NPP station and the relevant people in the surrounding regions to visit the NPP which has been built and participate in nuclear safety related knowledge training. Speeding up the process of “nuclear safety” law and perfecting the laws and regulations for protecting the safety and rights of public.

Commentary by Dr. Valentin Fuster

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