bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Study of the Processes of Gas-Steam Pressurizer Insurge Transients
Authors: Bolong Wang , Weihua Li , Haijun Jia , Jun Li , Wentao Hao 
Published: October 2020

Abstract
Small reactors have received more and more attention for their high safety, reliability, low power density, and short construction period. And the gas-steam pressurizer is widely used in small reactors due to its characteristics of simple structure, saves the heating and spray equipment, and prevents the coolant from boiling. The gas-steam pressurizer is a pressure control equipment for the reactor coolant system, and its characteristic of transient response is an important factor that affect operation stability of nuclear reactor systems. An experimental system was established to study the effect of pressure response for an insurge transient and influence factors were analyzed quantitatively. Experimental investigation shows that for the gas-steam pressurizer, the increase of coolant loading capacity (insurge) can cause system pressure rising. And the change of system pressure has much consistency with the change of liquid level and gas space temperature. The liquid phase exists temperature fluctuations and overall shows a downward trend during the insurge transient. And there exists a temperature gradient from bottom to top in the pressurizer liquid phase region during the insurge transient. The change of water vapor quantity curve is the oscillating curve during the transient and water vapor quantity is in a decreasing trend overall during the insurge transient. What’s more, the experiments also analyzed the pressure response and temperature response during the insurge transient.
Protection of Nuclear Power Plants Against Natural Hazards: Protection Principles Against Rare and Severe Natural Hazards for New Nuclear Power Plants
Authors: Coulon Vincent , Christophe Sébastien , Grammosenis Laurence , Guinard Luc , Cordier Hervé 
Published: October 2020

Abstract
The field of protection against external natural hazards (eg.: rare and severe hazards) has regularly evolved since the design of the first NPPs (Nuclear Power Plants) to take into account the experience feedback. Following the Fukushima Daiichi accident in March 2011, consideration of rare and severe natural hazards has considerably increased in the international context. Taking rare and severe natural hazards into account is a challenge for operating nuclear reactors and a major issue for the design of new nuclear reactors.
In Europe, considering lessons learnt from the Fukushima Daiichi accident, European safety authorities released new reference levels in the framework of WENRA 2013 (Western European Nuclear Regulators Association) standards for new reactors [1] to address external hazards more severe than the design basis hazards. Considering this input, the French and UK nuclear regulators have released specific guidelines (Guide No. 22 related to design of new pressurized water reactors [2] for France and ONR Safety Assessment Principles SAPs [3] for the UK) to describe how to apply those principles in their national context.
To comply with those different guidelines, EDF has developed different approaches, called Beyond Design Basis (BDB) approaches, related to rare and severe natural hazards issue in the French and UK context for nuclear new build projects.
Those two approaches are presented in the present technical paper with the following structure:
- safety objectives;
- hazards to consider;
- SSCs (Structures, Systems, and Components) required to meet safety objectives;
- study rules and assumptions;
- analysis of deterministic or probabilistic nature, thereby including the following:
○ analysis of available margins (margin between 10−4 per annum exceedance frequency of hazard site level or equivalent level of safety and the chosen Design Basis Hazard level also called ‘inherent margin’);
○ Fukushima Daiichi accident Operating Experience feedback;
○ probabilistic safety analyses.
This technical paper highlights common characteristics and differences between the two approaches considering the French and UK regulatory contexts.
Impact of Climate Change on EDF’s Nuclear Facilities: Climate Watch Approach
Authors: Guinard L. , Parey S. , Cordier H. , Grammosenis L. 
Published: October 2020

Abstract
According to the Periodic Safety Review Process, the safety level is re-assessed every ten years, considering national and international operational feedback, evolution of knowledge and best available practices. Protection against natural hazards is part of this safety level re-assessment. In the current global change context, climate change impact has to be integrated in external natural hazards estimations, such as climatic hazards or external flooding.
EDF has consequently implemented a climate watch approach. Undertaken approximately every 5 years, roughly in line with the publication of the assessment reports of the Intergovernmental Panel on Climate Change (IPCC) and with the update of safety licensing basis during Periodic Safety Reviews, this approach is intended to:
- revisit the climatic hazards which present a plausible or certain upward trend, and could lead to an increased reference hazard level,
- monitor the reach of target levels which should trigger a thorough analysis (concept of Major Climate Event) to ensure the robustness of the reference hazard level between two periodic reviews.
This climate watch approach is developed in partnership with the scientific community and is based on the following activities:
- compile and analyze datasets on hazards that are subject to changes with climate change (observed and modelled time series),
- develop knowledge of associated climatic phenomena (models, projections).
The application of this approach is presented in two steps:
- the key implications of the last climate watch exercise carried out in 2015, which identified climatic hazards whose evolution is unfavorable and is plausible or certain for the sites of EDF NPPs:
○ High air and water temperatures (for the “heat wave” hazard)
○ Sea level (for the “external flooding” hazard for coastal or estuary sites)
○ Drought or « low flow » hazard for fluvial sites;
- the results obtained for the 900 MW units, for which EDF started the 4th periodic safety review in 2019.
Such an approach, which is closely linked to periodic reviews, ensures the robustness of nuclear power plants to the climatic hazards through the consideration of the updated hazard levels.
Design of an Innovative Moisture Separator Technology for Use in Nuclear Power Plants: Numerical Approach — Part 1
Authors: Padovan Loris 
Published: October 2020

Abstract
The Moisture Separator Re-heater (MSR) is a key component of Nuclear Power Plants (NPP) both in terms of performance and avoiding erosion and erosion/corrosion damage. Wet steam is usually dried in a MSR by inertial separation using separator elements. Depending on the design of a MSR, the technology of separator elements contributes significantly to its size and performance, hence is seen as a subject for in-depth investigation, improvement and innovation.
Computational Fluid Dynamics (CFD) has been used to understand the working principles of moisture separating devices, in particular the OpenFOAM platform has been utilized for this scope. Eulerian/Lagrangian models, wall-droplet interaction and water film formation models have been adopted to determine the physical phenomena occurring during the moisture separation process. Additional sub-models have been implemented to make a more robust solver and to solve in a comprehensive way all the possible physical processes: in particular a two-layers turbulence model and a film breakup model have been implemented. An out-of-the-box thinking approach was adopted to devise a new proposed chevron vane. Aerodynamic principles were used to design an innovative concept of separator panel, which can entrap the moisture droplets and water rivulets through a subsequent formation of recirculating steam representing artificial slots (hidden pockets) within the separator channel. The control of the steam separation on precise regions of the separator panel wall, helps the drainage of the water film without the utilization of physical obstacles (pockets or drainage channels). To validate the results achieved from the numerical simulation and to characterize separation performance of a new kind of separator technology, a bespoke test rig has been designed, built and put into operation at typical MSR operating conditions [1]. Throttling calorimeter methodology has been adopted to measure, with very good accuracy, the residual moisture content after the separator.
The design developed has shown excellent separation performance. Particularly, this solution will allow improved MSR performance and significantly reduced MSR size. This represents an innovative technology which is a major advance on current technology available within the industry. The novel design features have been patented by General Electric. The first operation of a MSR with this technology is eagerly awaited.
Development of Advanced Fire Zone Model Applicable to Fire PRA for Nuclear Power Plant
Authors: Ji Junghoon , Shirai Koji , Tasaka Koji , Udagawa Toshiko 
Published: October 2020

Abstract
In implementing the fire PRA for nuclear power plants, a highly predictive fire model is required for more realistic fire scenarios and fire risk assessment. The fire simulation zone model BRI2002 developed in Japan has been continuously improved to allow analysis considering the characteristics of a compartment fire.
In this study, a heat feedback phenomenon was introduced in BRI2002, in which combustion of a fire source can be accelerated by radiant heat transfer inside the compartment during a compartment fire. Not only the thermal radiation from the flame and smoke layers, but also radiation from the hot ceiling surface and the ceiling jet flame were considered when the flame impinges with the ceiling.
In addition, in the zone model, the existing model for predicting the oxygen concentration in a compartment was improved so that the oxygen concentration could be predicted considering the vertical location of a fire source (height from the floor).
The prediction results were verified by full-scale compartment fire test results. As a result of the calculation in which the fire source is installed at 2 m above the floor, the prediction results for the burning rate and zone temperature were well consistent with the test results.
Analytical Study on Removal Mechanisms of Cesium Aerosol From a Noble Gas Bubble Rising Through Liquid Sodium Pool
Authors: Miyahara Shinya , Kawaguchi Munemichi , Seino Hiroshi 
Published: October 2020

Abstract
In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released from the failed pin as an aerosol such as cesium iodide and/or cesium oxide together with a fission product noble gas such as xenon and krypton. As the result, the xenon and krypton released with cesium aerosol into the sodium coolant as bubbles have an influence on the removal of cesium aerosol by the sodium pool in a period of bubble rising to the pool surface. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion from a noble gas bubble rising through liquid sodium pool was analyzed by constructing a computer program which deals with the expansion and the deformation of the bubble together with the aerosol absorption. In the analysis, initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration in the bubble were changed as parameter, and the sensitivities of these parameters on decontamination factor (DF) of cesium aerosol were investigated. From the results, it was concluded that the initial bubble diameter was most sensitive parameter to the DF of cesium aerosol in the rising bubble due to the inertial deposition. It was found that the sodium pool depth, the aerosol particle diameter and density have also important effect on the DF of cesium aerosol, but the sodium temperature has a marginal effect on the DF.
To meet these results, the experiments for the investigation of cesium aerosol absorption behavior from rising noble gas bubble through sodium pool are under planning to validate the results for the sensitivities of above-mentioned parameters on the DF of cesium aerosol in the analysis.
An Experimental Study of Two-Phase Flow in a Tight Lattice Using Wire-Mesh Sensor
Authors: Zhang Hengwei , Xiao Yao , Gu Hanyang 
Published: October 2020

Abstract
Tight lattice bundle can improve the conversion ratio and the heat transfer coefficient between the fuel bundle and the coolant, which is widely used in the innovative reactor fuel bundle design. The P/D ratio of a tight lattice bundle is usually less than 1.1, which is smaller than that of a conventional rod bundle. In the small-break loss-of-coolant accident (LOCA), the steam-water two-phase flow will occur in the reactor. The investigation of gas-liquid two-phase flow in the tight lattice is very important to the reactor safety analysis. A dual sub-channels tight lattice was designed in this study. The original reference of the channel is the annular fuel bundle, with the fuel diameter of 15.52mm, pitch of 16.51mm, P/D = 1.06. The original reference of working condition is the stream-water two-phase flow under the pressure of 15.5MPa. The experimental condition is the air-water two-phase flow at the normal temperature and pressure. According to the ratio of a critical bubble diameter in the reactor (steam-water) to that in atmospheric conditions (air-water), the channel is zoomed in 2.7 times. The diameter of the rod in the dual sub-channels tight lattice is 42mm and the pitch is 44.6mm. The total length of the dual sub-channels tight lattice is 3m. A self-developed 16 × 32 Wire-mesh sensor (WMS) was used to measure the void fraction distribution of air-water two-phase flow in the dual sub-channels tight lattice channel. The spatial resolution of the WMS is 2.79mm and the temporal resolution is 5000fps. The WMS was installed at a distance of 2.5m from the channel inlet and 0.5m from the outlet, which can avoid the influence of outlet on bubbles. The experimental range of flow condition is 0.921–1.84m/s for the superficial liquid velocity and 0.0884–1.07m/s for the superficial gas velocity. The instantaneous and time-averaged void fraction distributions in the channel was measured. With the increase of superficial gas velocity, the distribution of void fraction distribution changed from the wall peak to the core peak. The characteristics of bubbles in the sub-channel were also discussed in this study.
Numerical Investigation on Turbulence Models in Bubbly Flow Using Euler-Lagrange Approach
Authors: Zhou Yujia , Du Jingyu , Zhao Chenru , Bo Hanliang 
Published: October 2020

Abstract
To accurately simulate the bubble motions in the turbulent flow liquids, two way coupled Euler–Lagrange method is adopted in this work. The continuous phase is solved with the Navier–Stokes equations based on the Euler grid, while the individual bubble is tracked by using the Lagrange frame of reference. Two–way coupling is realized by transferring the interaction forces to the momentum equation of the continuous phase at each fluid flow timestep and in turn influencing the bubble motion. The interaction forces including the drag force, lift force, wall lubrication force and virtual mass force are carefully selected. Turbulent models are of significance to capture the fluid flow conditions. Comparing the different cases treated by the k-ε, PANS and LES models with the experimental data, the results of the PANS with RWM and BIT, and LES can dynamically display the oscillation of the bubble plume and the calculated time averaged axial liquid velocity reasonably agrees with the measured data. In the case of bubble injective flow conditions in large containers, the bubble induced turbulent effect can be almost neglected. However, the RWM describes the fluctuating velocity should be considered applied with the PANS model. In addition, the saturated vapor–liquid system under high pressure of 6.9 MPa is numerically simulated by implementing the PANS with R WM and BIT.
Aerosol Decontamination Behavior in Two Phase Flow During Pool Scrubbing
Authors: Yoshida Kohei , Fujiwara Kota , Nakamura Yuki , Kaneko Akiko , Abe Yutaka 
Published: October 2020

Abstract
In some scenarios in severe accidents (SAs) of nuclear power plants (NPPs), non-condensable gas containing vapor and small particles of fission products (FPs) will be released from the reactor vessel, pass through the wet well of a suppression chamber. In these conditions, water inside the wet well will trap FPs before released to the atmosphere. This trapping effect is called pool scrubbing. Since pool scrubbing is assumed to have a good decontamination effect, researches about the pool scrubbing are important from the viewpoints of NPPs safety. Therefore, it is necessary to understand the two-phase flow behavior in pool scrubbing.
The mechanism of the FPs removal behavior in pool scrubbing is not clearly understood due to the complicated hydrodynamic phenomena. In addition, the evaluation of the validity of physical models used in pre-existing SA analysis codes are not enough. Therefore, a reliable model to evaluate particle decontamination efficiency by pool scrubbing is highly regarded.
Currently, decontamination factor (DF) of pool scrubbing is calculated by analysis code such as MELCOR. In the MELCOR code, some assumptions and empirical formula are used, e.g. bubble diameter and rising velocity. However, the correlation between DF and flow structure characteristics are not understood, and dominant factor of decontamination behavior in two phase flow is required.
Therefore, the aim of this study is to elucidate the dominant factor of decontamination by measuring DF and flow structure of bubbly plume. As to understand two phase flow characteristics during pool scrubbing, we focus on bubble diameter and swarm rise velocity. In Air-particle mixture flow injected into the experimental facility gas velocity, void fraction and bubble diameter distribution were measured by a wire mesh sensor (WMS) in each height. In addition, DF by pool scrubbing was derived from aerosol measurements by an aerosol spectrometer. The measured warm rise velocity were compared with equation of single bubble terminal velocity. The swarm rise velocity is higher than single bubble velocity. Furthermore, Large bubble are identified to be still remained at swarm region defined in MELCOR. In addition, dominant factor of decontamination behavior is defined as particle diameter, neither aerosol solubility nor particle density.
HTGR Power-Level Control Only by Regulating Helium Flowrate
Authors: Dong Zhe 
Published: October 2020

Abstract
Currently, most of the power-level control methods of nuclear reactors are given based upon the scheme of inducing reactivity directly via control rods. However, the reactivity can also be injected indirectly via the negative temperature feedback effect. Motivated by this, a cascaded power-level control of high temperature gas-cooled reactor (HTGR) is proposed by only regulating primary helium flowrate, which can be able to provide globally asymptotical closed-loop stability. This new HTGR power-level control is composed by a helium temperature controller and a neutron flux controller. The former controller, which is located in the inner loop, regulates the primary helium flowrate according to the setpoint of helium temperature. While, the latter one in the outer loop revises the setpoint of helium temperature so as to regulate neutron flux. Numerical simulation results verify the theoretical result while showing the satisfactory performance as well as the influence from the controller parameters.
The Dependence of Pool Scrubbing Decontamination Factor on Particle Number Density: Modeling Based on Bubble Mass and Energy Balance
Authors: Sibamoto Yasuteru , Sun Haomin , Hirose Yoshiyasu , Kukita Yutaka 
Published: October 2020

Abstract
The dependence of pool scrubbing performance on particle number density is studied through numerical simulation of experimental results. The DF values obtained from the authors’ experiments (Sun et al., Sci. Technol. Nucl. Inst., Article ID 1743982, 2019) indicate a sharp decrease with an increase in the inlet particle number density beyond 1011/m3. The mechanisms underlying such dependence is yet to be studied.
In this paper, a simple model is developed to study the factors affecting the experimentally observed dependence of DF. The test results suggest that the condensational growth of particles plays an essential role in the inertial capture. The vapor condensation on the particles has an effect to deplete the vapor supersaturation in the bubble by both lowering the vapor concentration and raising the temperature. This effect will become important at high particle number densities.
The bubble mass and energy balance is calculated to derive the particle growth and the inertial DF as a function of the bubble rise distance through the pool water. The balance is assumed to be quasi-steady, and the vapor concentration and the temperature to be uniform in the bubble. It is shown that the model reproduces the tendency observed in the experimental DF. The model predicts that the degree of supersaturation is affected when particle concentration exceeds 1011/m3, curbing the condensational growth of particles, and thereby retarding the inertial capture.
Selection and Validation of Fast and Synchronous Interface to the Controller of a Space Nuclear Reactor Power System
Authors: Hahn Andrew , Schriener Timothy , El-Genk Mohamed S. 
Published: October 2020

Abstract
Enabling transmission of state variables and controller command signals between physics-based model and physical or emulated PLC requires developing fast and synchronous communication interface. This research investigates four methods of interface communication: (a) reading/writing to text files, (b) serial communication, (c) TCP communication, and (d) shared memory. The shared memory interfacing method provides the best performance, reliability, and fidelity. It is validated by linking a representative PLC controller to a Simulink dynamic simulation model of a fully integrated space reactor power system with multiple closed Brayton cycle loops for energy conversion (DynMo-CBC). Transient performance results of the PLC controlled dynamic simulation model are identical to those obtained using the Simulink model with internal control.