bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Study of the Effects of Wall Force Upon Bubble Motion Under Static and Rolling Condition
Authors: Huang Ying , Gao Puzhen 
Published: October 2016

Much research on bubble dynamics under rolling motion has been carried out in recent years, but the effect of the wall force on bubble motion behavior is far from satisfactory and still needs to be further investigated. In order to study the interaction between bubbles and near wall under rolling condition, the paper adopts the VOF method, one momentum equation coupled with surface tension mode and the function of UDF. The bubble motion behaviors in a vertical channel under static and rolling conditions were numerically simulated.
The impact of the wall force on single bubble motions were discussed. Plus, the effect of the wall force on bubbles when they are approaching the wall was analyzed in detail. The numerical results agree well with the experimental results from other researchers. The simulating results show that as bubbles gradually approach the wall, the wall force prevents them from further coming closer to the wall and pushes them in the opposite direction. Under the rolling condition, the wall force affects the bubble motion only when the wall is located in the positive bubbles’ rolling direction.
Design Approach for Mitigation of Air Ingress in High Temperature Gas-Cooled Reactor
Authors: Sato Hiroyuki , Ohashi Hirofumi , Nakagawa Shigeaki 
Published: October 2016

One important safety design consideration for high temperature gas-cooled reactor (HTGR) is air ingress following a rupture of the reactor pressure boundary such as primary piping. The air intrusion to the reactor core held at high temperature through the break will results in significant oxidation of graphite components and fuels. Such oxidation may leads to the weakening of core support structures as well as fuel element damage and subsequent fission product release.
This paper intends to propose a practical solution to protect the reactor from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR to study the effectiveness of the proposed design concept using simplified lumped element models. The analysis considers internal decay heat generation and transient conduction from inner to outer regions at the reactor core, cooling of vessel outer surface by radiation and natural convection, and natural circulation flow in reactor. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.
Phased Array Ultrasonic Inspection Technique for Cast Austenitic Stainless Steel Parts of Nuclear Power Plants
Authors: Yamamoto Setsu , Semboshi Jun , Sugawara Azusa , Ochiai Makoto , Tsuchihashi Kentaro , Adachi Hiroyuki , Higuma Koji 
Published: October 2016

For safety operation of nuclear power plants, soundness assurance of structures has been strongly required. In order to evaluate properties of inner defects at plant structures quantitatively, non-destructive inspection using ultrasonic testing (UT) has performed an important role for plant maintenances. At nuclear power plants, there are many structures made of cast austenitic stainless steel (e.g. casings, valve gages, pipes and so on). However, UT has not achieved enough accuracy measurement at cast stainless steels due to the noise from large grains. In order to overcome the problem, we have developed comprehensively analyzable phased array ultrasonic testing (PAUT) system.
We have been noticing that dependency of echo intensity from defect is different from grain noises when PAUT conditions (for example, ultrasonic incident angles and focal depths) were continuously changed. Analyzing the tendency of echoes from comprehensive PAUT conditions, defect echoes could be distinguished from the noises. Meanwhile, in order to minimize the inspection time on-site, we have developed the algorithms and the full matrix capture (FMC) data acquisition system. In this paper, the authors confirmed the detectability of the PAUT system applying cast austenitic stainless steel (316 stainless steel) specimens which have sand-blasted surface and 3 slits which made by electric discharge machining (EDM).
Water Level Control for Steam Generator in Nuclear Power Based on Port-Controlled Hamiltonian Method
Authors: Tao Mo , Yu Rui , Li Xianling , Ke Zhiwu , Zhao Zhenxing , Xiao Qi 
Published: October 2016

This article designs the water level of steam generator control system applying PCH system modeling and control principle of Energy-Shaping.
The basic principles of port-controlled Hamiltonian (PCH) system modeling and control principle of Energy-Shaping are provided, including passivity, dissipation and so forth. Meanwhile, port-controlled Hamiltonian (PCH) system, energy-shaping principle, matching method and its corresponding realization and stability analysis are presented. The water level of steam generator is accomplished; the controlled object is mathematically modeled with principle modeling method. Then the mathematic model is translated into PCH model. the proper Hamiltonian function is finely selected. The water level equilibrium point of steam generator system is respectively determined; the PCH controller based on energy-shaping method is ultimately obtained.
The proposed modeling and control methods provide a potential idea for the other control methods applied method on steam generator system. Moreover, the proposed methods can be put into industrious process control.
Impact of Deviations Handling on Outages
Authors: Matahri Naoëlle 
Published: October 2016

Based on the information provided by the operators, IRSN experts select and analyze different deviations presenting a possible generic nature which could affect the safety of power plants. Some of these deviations result in non-compliance (NC) with the safety requirements.
To maintain an acceptable level of safety, an operator has to implement corrective measures for any situation of non-compliance with safety requirements. IRSN, the Technical Support Organization of the French Nuclear Authority, analyzes the different deviations to assess the impacts on the concerned NPPs safety. Based on the impact on safety, measures should be applied immediately or during the next outages, on a reactor or on several of them. The permanent corrective measures schedule is defined taking into account the “NC” safety impact. However, for some of the “NCs”, it can be difficult to define and implement swift permanent corrective measures, especially when the lack of compliance affects several similar units and requires a design change.
This paper explains the French approach of deviations treatment and specifically the relationship between the Nuclear Safety Authority, the Technical Support Organization, IRSN and the Licensee, EDF during an outage.
Use of Algebraic Heat Flux Models to Improve Heat Transfer Predictions for Supercritical Pressure Fluids
Authors: Papp Vera , Pucciarelli Andrea , Sharabi Medhat , Ambrosini Walter 
Published: October 2016

This work proposes simulations of heat transfer under supercritical pressure conditions showing improvements with respect to previous works. This is obtained by the introduction of the Algebraic Heat Flux Model (AHFM) for evaluating the turbulent heat flux in turbulence production terms, using the in-house code THEMAT and the STAR-CCM+ code. The first code makes use of the AHFM also in the energy balance equations, while for the commercial code simplifying assumptions are considered in the implementations.
Custom sets of parameters for every condition of inlet temperature and internal diameter are tuned in some cases, driven by the opinion that a single set of parameters cannot be suitable in every flow conditions, considering the complexity of the variables that concur in the heat transfer deterioration phenomenon. The AHFM model gives promising results with new sets of parameters in order to model the deterioration and the recovery phases because of its term related to the variance of temperature.
Study on Current Status and Future Developments in Nuclear-Power Industry of Ukraine
Authors: Zvorykin Alexander , Pioro Igor , Panchal Raj 
Published: October 2016

Nuclear power in Ukraine is the most important source of electricity generation. Currently, Nuclear Power Plants (NPPs) generate 45.5% of the total electricity in the country followed with coal generation – 38%, gas generation 9.6% and the rest is based on renewable sources, mainly on hydro power plants – 5.9%. Nuclear-power industry is based on 4 NPPs including the largest one in Europe – Zaporizhzhya NPP with about 6,000 MWel gross installed capacity.
These NPPs are equipped with 13 VVER-1000 and 2 VVER-440 Russian-design Pressurized Water Reactors (PWRs) with the total gross installed capacity of 13,800 MWel. Layout of these NPPs, thermodynamic diagram and thermal efficiencies are provided. Thermal efficiencies have been calculated with the IAEA Desalination Thermodynamic Optimization Programme DE-TOP and compared to the actual ones.
Two of these reactors have been built and put into operation in 70-s, ten in 80-s, one in 90-s and just two in 2004. Therefore, based on an analysis of the world power reactors in terms of their maximum years of operation (currently, the oldest reactors are 45-year old) several projections have been made for future of the nuclear-power industry in Ukraine. Unfortunately, all these projections are quite pessimistic.
There is a possibility that around 2030–2035 the vast majority of the Ukrainian reactors will be shut down, and Ukraine can be left without the basic and vital source of electricity generation. Also, current problems of Ukrainian NPPs are: 1) lower capacity factors (around 80%) compared to those in other countries (∼90%); 2) uncertainties with nuclear-fuel supply due to political situation; and 3) service and repairs of relatively old reactors.
Measurements of Fine Structures in the Lead-Bismuth Eutectic Alloy Melts by Differential Scanning Calorimetry
Authors: Ye Changqing , Li Qiulin , Wu Ping , Tang Guoyi , Liu Wei 
Published: October 2016

The implementation of the Generation IV nuclear reactors and a spallation target of the accelerated driven system (ADS) concerning to the use of liquid lead-bismuth eutectic (LBE) alloy. The liquid LBE alloy should be fully characterized and especially its physical properties should be completely known to make sure the nuclear safety. Differential scanning calorimetry (DSC) experiments were employed on LBE alloy at a temperature range of room temperature (RT) to 500°C to detect the structure phase transition and obtain the thermal effect of LBE alloy.
The results of DSC curves showed that there existed two remarkable thermal signal events at the melting points zones of Pb and Bi during the melting process tested with a scan rate of 2°C/min. Even though the scan rate was increased to 5°C/min, the DSC signal still exhibited the changes of curve slop at the element melting point zone. Just this interesting DSC thermal signal change phenomenon around 300°C contributed to the relationship with the explanation on the severe embrittlement of T91 steel induced by liquid LBE alloy. The results suggest that there existed effective critical size and form of chemical clusters in liquid LBE alloy worked actively on the embrittlement of T91 Steel.
Analyses of Severe Accidents in GFR Demonstrator ALLEGRO With MELCOR 2.1
Authors: Vácha Petr , Bělovský Ladislav 
Published: October 2016

The helium-cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected for further development in the frame of the Generation IV International Forum (GIF). Since no gas cooled fast reactor has ever been built, a small demonstration reactor is necessary on the road towards the full-scale GFR reactor. A concept of this demonstrator is called ALLEGRO.
The French Commissariat à l’énergie atomique et aux énergies alternatives (CEA) developed between 2001–2009 a pre-conceptual design of both the full-scale GFR called GFR2400 and the small demonstration unit called ALLEGRO (75 MWt). Since 2013 ALLEGRO has been under development by several partners from Czech Republic, France, Hungary, Poland and Slovakia.
No severe accident study of ALLEGRO using a dedicated computer code has been published so far. This paper is the first attempt to perform computer simulations of the ALLEGRO CEA 2009 concept, using MELCOR version 2.1.
A model of the ALLEGRO CEA 2009 concept has been developed with the aim to perform safety analyses; to confirm that MELCOR can be used for such a study, to investigate what scenarios lead to a severe accident and to study in detail the progression of the severe accident during the in-vessel phase.
Several pressurized and depressurized protected scenarios were investigated; four of them are presented in this paper. It was observed that even long-lasting station blackout (SBO) without further failures of the passive safety systems does not lead to a severe accident as long as there is enough water in the decay heat removal (DHR) system. Loss of coolant (LOCA) transients with DHR system in the forced-convection mode can lead to peak cladding temperatures causing limited core damage in the early phase of the accidents, but without further development into core meltdown. On the other hand, LOCA combined with SBO leads to excessive core melting in orders of minutes, which represents a weak point of ALLEGRO 2009 concept. Recommendations were formulated for the further development of the ALLEGRO concept.
The Early Depressurization Method for the Vapor Generation at the Top of Reactor Vessel
Authors: Tsuruoka Hokuto , Miyata Shinya , Maruyama Manabu , Kanamori Daisuke 
Published: October 2016

Mitsubishi Heavy Industries (MHI) is developing improved reactor cooling process in accident conditions, by using steam generator (SG) on accident in order to enhance safety of operating plants.
In case of station black out (SBO), leak from the reactor coolant pump (RCP) seal should be terminated by depressurizing the primary system. In this process, operators use the main steam relief valves (MSRVs) to reduce primary coolant temperature and pressure. However, coolant temperature in vessel head and upper plenum tends to remain higher than reactor temperature, and the coolant in these regions eventually starts boiling because of depressurization. This phenomenon inhibits smooth decrease in primary coolant temperature and pressure by forming large void region in the primary system. MHI is investigating improved process to achieve smooth decrease of primary temperature and pressure. In the process, condensation of the void generated should be appropriately achieved.
In this study, experimental tests on void condensation in the vessel head were performed using Large Scale Test Facility (LSTF). In addition, applicability of M-RELAP5, a plant transient computer code used in safety analysis for Japanese PWRs, to actual plant behavior modeling was confirmed, in terms of void generation/condensation in vessel head in accidental condition.
Confirmation of Seismic Integrity of HTTR Against 2011 Great East Japan Earthquake
Authors: Ono Masato , Iigaki Kazuhiko , Shimazaki Yosuke , Shimizu Atsushi , Inoi Hiroyuki , Tochio Daisuke , Hamamoto Shimpei , Nishihara Tetsuo , Takada Shoji , Sawa Kazuhiro , Kondo Toshinari , Kojima Keidai 
Published: October 2016

Japan Atomic Energy Agency (JAEA) has carried out research and development to establish the technical basis of High Temperature Gas-cooled Reactor (HTGR) by using High Temperature engineering Test Reactor (HTTR). On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR.
The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the “inspection of equipment and instrument”. The other is the “seismic response analysis” for the building structure, equipment and instrument using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria.
As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.
Vented Explosion Phenomena: Hydrogen Combustion Benchmark on Confined Vessel Experiments
Authors: Manzini Giovanni , Kljenak Ivo , Povilaitis Mantas 
Published: October 2016

Confined vented explosion is a very complex topic as many parameters affect the phenomena, mainly because the flame front develops from an ignition source and travels through a medium which may involve complex boundary conditions and obstructions of various geometries.
Therefore, in the plant safety assessing step, it is important to provide correct estimates of the flame spreading rates as well as overpressures which may result from various explosion initiation scenarios. This will help designers for plant layout optimisation with the aim to minimize the risk associated with those events.
Although hydrogen explosion in unvented compartments was often simulated in the past, there were not many opportunities, so far, to simulate explosion in a vented room. With this purpose, a benchmark exercise was organized, based on simple hydrogen combustion experiments, performed in a vented compartment (Chamber for View of Explosion – CVE) at the Scalbatraio laboratory of University of Pisa (Italy). In that activity, many tests were performed by varying the initial hydrogen concentration and the obstacles inside the compartment.
The numerical codes used in the benchmark were lumped-parameter (LP) ones (ECART, ASTEC), which remain, for the time being, the customary tools for simulating hydrogen combustion accidents in current NPPs, because of their fast-running calculation capabilities also for large-scale scenarios.
Study on Power Transmission Plan of the Nuclear Power Base in Shandong Peninsula
Authors: Yin Chengzhu , Wang Miao 
Published: October 2016

The nuclear power site resource is very rich in Jiaodong Peninsula of Shandong Province. It is suitable for construction of the large nuclear power base. The transmission scope and direction of Jiaodong Peninsula nuclear power base is analyzed, and optional transmission plans of Plant 1, Plant 2, Plant 3 and Plant 4 are proposed. The transmission plans are recommended based on technical and economic comparison, which provide good references for construction of the large-scale nuclear power base and power grid development planning.
Jiaodong Peninsula nuclear power base is planned to be built in the year of 2016–2030, planning capacity of which is 30500MW. The site of nuclear power base is 100∼400km away from the power load center. The nuclear power will use AC transmission and mainly meet the demand of local power load. The early 15500MW gensets will be accessed to the power grid at 500kV, as the following 15000MW gensets will be accessed at 1000kV UHV (Ultra-high voltage) grid. As the accessing of many large-capacity gensets will produce huge impact to the short-circuit current, sectionalized double-bus configuration is recommended in the 500kV main electrical wiring to reduce the short-circuit current of 500kV bus of nuclear power plant. Double bus section cross wire connection is presented to make sure that every two generators on each bus will be connected to different substations on two transmission lines which are set up on different poles and in different paths, to improve the reliability of the power plant. Through analysis and provement, the construction of large nuclear power base must be based on large and stronge power grid, especially the UHV (Ultra-high voltage) AC grid, to meet the demand of huge nuclear power transmission, and to improve the ability of power exchange and ensure the safety of regional power supply.
Also, as the nuclear power plant should better be in base-load operation, the construction of large-scale nuclear power base, would make the system load-control demands increase, which leads to more prominent problems. In order to avoid adding additional depth of peaking power operation and reducing the overall economic operation of power system, power grid should have the necessary means to load-control. Namely the construction of peaking units, such as pumped storage units or gas-fired units at about 5000MW.
By analyzing and demonstration, large-scale nuclear power base must rely on large-scale power grid, particularly the support of UHV power grid in order to meet the demond of large-scale power transmission and electricity exchange, and also to ensure regional security of electricity supply.
Air Ingress Experiments Within High Temperature Graphite Flow Channels
Authors: Gould Dan , Bindra Hitesh , Schlaikjer Eric , Liu Hanwen 
Published: October 2016

The graphite-fuel matrix in the core region of HTGRs can come into direct contact with air at high temperature during air ingress accident. Nuclear grade graphite has been shown to undergo oxidation when subjected to oxidizing flow at high temperatures. There is, however, no agreement on the relative importance of the numerous factors that can contribute to the overall rate of oxidation. Examples of some of these potential factors include graphite temperature, fluid temperature, fluid composition, graphite composition, and graphite surface conditions. Separate effects experiments have, in general, not been able to fully capture these complex interactions.
In this work, a new experimental apparatus was designed to conduct mixed-effect experiments to understand the complicated interactions that would influence the oxidation rate and heat removal rate of graphite exposed to a high temperature air ingress. Utilizing thermographic methods, experiments detailing the local temperature response of a representative graphite flow channel were conducted at high temperatures (1173 K) in both oxidizing and inert gaseous environments.