bluuebunny/crossref_metadata_2025_split huggingface dataset:
Influence of Two-Phase Flow Structure on Operating Characteristics of the Supersonic Steam Injector
Authors: Sato Kyosuke , Abe Yutaka , Kaneko Akiko , Mori Michitsugu 
Published: November 2014

Steam injector (SI) is a passive jet pump with a converging-diverging structure. A SI operates without an electrical power source by direct contact condensation of a supersonic steam flow and a subcooled water jet in a mixing section. Furthermore, there are advantages that a SI has high heat-transfer performance and discharges water at high pressure. Therefore a SI is expected to apply to the safety system that is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant. However it is not cleared about the operating range of a SI which is taken account of the discharged flow structure.
The objective of the present study is to reveal the influence of two-phase flow behavior on operating limits of a SI. The test section of the SI is made by transparent material to observe flow structure in it. The pressure distributions along the flow direction and the discharge pressure were measured by changing the inlet steam pressure and a load on exit of the SI. At the same time, the discharged flow at the diffuser was observed with a high speed camera.
From the observation results, it was confirmed that a boundary which the flow structure changed in the discharged flow. The area from the throat to this boundary is seemed a two-phase region that steam which has not been condensed completely in the mixing nozzle remains. It was found that this boundary moved to upstream as the load on the exit increased and significant pressure rise occurred at the position the boundary reached. Additionally white propagations toward downstream were observed. This propagation is seemed a pressure wave propagation. The velocity of the propagation was estimated by image processing. Assuming a pressure wave propagates at sonic speed, void fraction at the discharged flow was estimated by existing homogenized model.
From the above, the influence of two-phase flow in discharged flow on the operating limits of the SI is discussed.
The Historical Development of French Nuclear Regulatory Organizations and the Major Factors in Their Changes
Authors: Kimura Kenji , Kimura Hiroshi 
Published: November 2014

At the beginning of the French nuclear history, CEA controlled all the nuclear development including its safety. In 1950s, this situation was changed by the participation of many industrial companies, which means industrialization of nuclear technology. This change became clearer when they adopted PWR in 1970. And so they needed and established a system to make safety regulation by putting the multiple actors’ opinions together.
After the accident of Chernobyl, antinuclear public opinion has increased. And in 1997, The Greens obtained the post of the Minister of Ecology. These facts required more independent and transparent regulatory system, so in 1998, a report which proposed to establish a new regulatory organization was published. On the basis of this report, they founded ASN in 2006.
From this French history, we can say that as the relationship between nuclear technology and the society changed, the regulatory body also changed to meet the necessity.
SCDAP Model Improvement With QUENCH-06 Analysis
Authors: Madokoro Hiroshi , Okamoto Koji , Ishiwatari Yuki 
Published: November 2014

One of the important severe accident management measures in a Light Water Reactor (LWR) is water injection to the reactor core. Reflooding of the uncovered reactor core is essential to prevent total core degradation. The series of QUENCH tests have been conducted to acquire knowledge of the reflooding. A number of analyses on QUENCH tests have also been done by different computer codes for code validation and improvements.
In this study, the modeling of the QUENCH-06 experiment was performed with RELAP/SCDAPSIM computer code. The input deck was modified and the SCDAP model was improved in order to represent the experimental facility more precisely. The uncertainties regarding the electrical resistance distribution in the heater rod system and the thermal properties of the shroud insulator were assessed, respectively.
The main purpose of this study is to identify the models to be improved. The rather good agreement between the calculation results and the measurement was acquired than the past studies [1]. In addition, more accurate modeling of the electrical resistance and the thermal properties of shroud insulator was done and their importance was indicated. The temperature profile and oxide thickness still showed similar tendencies with the original case. Further improvements are required mainly in the heat transfer model and the oxidation model in the SCDAP code.
Uses of Supercritical Fluids and Their Characteristics Within Deteriorated Heat Transfer Region
Authors: Gupta Sahil , Saltanov Eugene , Pioro Igor 
Published: November 2014

Taking into account the expected increase in global energy demands and increasing climate change issues there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a big part in being part of the energy mix since it offers a relatively clean, safe and reliable source of energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as flexibility in design to adapt to different countries and situational needs. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF, 2002). The main design goals for the Generation IV nuclear system concept were to improve economic gains, enhance safety, extend sustainability, and strengthen proliferation resistance.
To achieve high thermal efficiencies of up to 45–50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed. An important step towards development of SCF applications in novel Generation IV Nuclear Power Plant (NPP) designs and in other industries; is to understand the thermal hydraulic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions.
Heat transfer under SC turbulent conditions is generally very complex and is extremely sensitive to the test geometry and operational flow parameters. Detailed sets of experiments have been conducted around the world in tubes and other geometries with SCFs to study the basics of heat transfer. By variation of operational parameters and test geometries, fundamental heat transfer data sets are collected that will help in our understanding of SC heat transfer phenomena.
Applications for SCFs are not limited to power industry; recent advancements have indicated the use of SCFs in a much wider range of applications due to its unique and attractive heat transfer characteristics. In this paper, proposed uses of SCFs are presented and basic concept of DHT within SCFs is analyzed.
Safety Evaluation of KAERI Pyroprocess Facilities
Authors: You G. S. , Choung W. M. , Ku J. H. , Moon S. I. , Kim H. D. 
Published: November 2014

KAERI (Korea Atomic Energy Research Institute) has been studing the pyroprocess since 1997. For a demonstration of the pyroprocess, KAERI developed two facilities, the ACPF (Advanced spent fuel Conditioning Process Facility) and PRIDE (PyRoprocess-Integrated inactive DEmonstration facility), by 2012. From 2013, KAERI has been studying a pre-conceptual design of the ESPF (Engineering-Scale Pyroprocess Facility).
In this paper, the safety evaluation methods and results for these three facilities are described.
Development of a PWR-W and an AP1000® Containment Building 3D Model With a CFD Code for Best-Estimate Thermal-Hydraulic Analysis
Authors: Jiménez Gonzalo , Bocanegra Rafael , Fernández Kevin , Queral César , Montero-Mayorga Javier 
Published: November 2014

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. However, for the detailed study of the thermal-hydraulic behavior in every room and compartment of the containment building, it could be more convenient to model the containment with a more detailed 3D representation of the geometry of the whole building.
The main objective of this project is to obtain a standard PWR Westinghouse as well as an AP1000® containment model for a CFD code to analyze the thermal-hydraulic detailed behavior during a design basis accident. In this paper the development and testing of both containment models is presented.
The CFD Analysis of Hydrogen Behavior in HTR-PM During the Water-Ingress Accident Caused by Double-Ended Guillotine Break of Two Steam Generator Heating Tubes
Authors: Zhang Ruidong , Sun Ximing , Dong Yujie 
Published: November 2014

Water ingress into the primary circuit of HTR-PM (High Temperature Gas-cooled Reactor Pebble-bed Modular) is one of the beyond-design-basis accidents, which will lead to chemical corrosion of graphite and metal material in the reactor, resulting in the generation of combustible gases such as hydrogen and carbonic oxide. Besides, the rise of pressure in the reactor containment caused by the incoming of abundant of gas will threaten to the safety of radioactive material releasing. Therefore, it is essential to analyze the gas behavior in the containment of HTR-PM under the condition of water ingress.
Based on the preliminary design of 250MW HTR-PM, the paper analyzes the transport and distribution of hydrogen, helium and other kinds of gas in the containment under the condition of water ingress accident resulted from a double-ended guillotine break of two steam generator heating tubes which is a typical beyond design basis accident of HTR-PM with the multi-dimensional computational fluid dynamics CFD code GASFLOW. In addition, the density and distribution of hydrogen will be investigated in relation to possible hydrogen explosion.
Seismic Test on Double-Layer Model of HTR-PM Graphite Structure
Authors: Lai Shigang , Sun Libin , Shi Li , Zhang Zhengming 
Published: November 2014

The HTR-PM (High Temperature Reactor Pebble bed Module) reactor core consists of assemblies that include graphite bricks, carbon bricks and keys/dowels. The double-layer structure, which is composed of two layers of bricks connected by keys and dowels, is the basic load-bearing unit of the graphite internals. A series of seismic experiments have been carried out to study the dynamic characteristic of this multi-body structure. As a part of them, the present study aimed to evaluate the nonlinear dynamic response of a full-size graphite/carbon unit and investigate the integrity of this assembly.
Dynamic excitations were applied at the bottom of the double-layer structure. The experimental random-motion excitation ranged from 0.2g to 0.75g. The sine-sweep excitation frequency band ranged from 5–100Hz, sweep rate was 1oct/min, with acceleration magnitudes of 0.3g and 0.5g. The sine beat wave excitation frequencies were 5Hz, 10Hz, 20Hz and 40Hz with acceleration of 0.5g. Response data of acceleration and displacement at certain points on bricks were measured. PSD curves and displacement time-histories were acquired. Results indicated that the fundamental frequency of the assembly is 42∼49Hz and the damping rate is 4.9%. The dynamic response of the assembly exhibited as an integrated component only when the excitation is lower than 0.35g.
Validation of ATHLET Code by LOCA-Induced Pressure Wave Propagation Tests
Authors: Trosztel István , Tóth Iván , Ézsöl György 
Published: November 2014

Propagation of pressure waves inside the reactor vessel after a large break LOCA is an issue since it affects pressure drop across core internals and, as a result, induces stresses in different components, like core barrel, core structures and even fuel. For reactor safety analysis pressure wave propagation is traditionally performed by systems codes. However, strong dispersion among the calculated results calls for test results to validate the calculations.
The pressure wave propagation following a larger LOCA is being systematically addressed by experiments in the PMK-2 integral-type test facility. In order to capture the high speed propagation of pressure waves special pressure transducers (capable to resolve the pressure variation with a frequency of 4 kHz) have been installed. The first four tests were conducted with rupture disks for opening the break, but a special quick opening valve will be installed for future tests, allowing the adjustment of the opening time between 12 and 50 ms.
The paper presents results of validation of the ATHLET code by the test results. The low-frequency oscillation of the measured system pressure was shown to be caused by flow rate coming from the pressuriser that compensates mass lost via the break: the frequency of the oscillation was slightly under-predicted. The propagation of the first rarefaction wave from the top of the downcomer to the upper plenum is very well calculated by ATHLET: in spite of the first order discretisation no numerical diffusion can be observed. The calculated pressure differences between two different locations in the system are of primary interest, since they define the loads on primary system internals. ATHLET somewhat overestimates the amplitude of the pressure difference pulses, while it fairly well describes the frequency of oscillations. First analyses indicate an effect of the calculated break flow rate. ATHLET calculates a slower attenuation of the pressure oscillations as compared to test results. This can be the consequence of rigid walls assumed in the analysis. The tendency of increasing first pressure peak with increasing system pressure is well predicted by the code.
In summary, it can be stated that ATHLET calculations produce slightly conservative results based on comparison with measured data.
A CFD Analysis of Flow Blockage Phenomena in ALFRED LFR DEMO Fuel Assembly
Authors: Di Piazza I. , Tarantino M. , Magugliani F. , Alemberti A. 
Published: November 2014

A CFD study has been carried out on fluid flow and heat transfer in the HLM-cooled Fuel Pin Bundle of the ALFRED LFR DEMO.
In the context of GEN-IV Heavy Liquid Metal-cooled reactors safety studies, the flow blockage in a Fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR Fuel Assembly. The present paper is a first step towards a detailed analysis of such phenomena, and a CFD model and approach is presented to have a detailed thermo-fluid dynamic picture in the case of blockage. The closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED has been modeled and computed. At this stage, the details of the spacer grids have not been included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, have not been included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900°C, are reached with blockage larger than 30% in terms of area fraction.
Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages (<10%). The blockage area has been placed at the beginning of the active region, so that both over-mentioned phenomena can fully take place. The mass flow rate at the different degree of blockage has been imposed from preliminary system code simulations.
Transient analyses with fully resolved SST-ω turbulence model have been carried out and results indicate that a blockage of ∼15% (in terms of blocked area) leads to a maximum clad temperature around 800 °C, and this condition is reached in a characteristic time of 3–4 s without overshoot. Local clad temperatures around 1000 °C can be reached for blockages of 30% or more.
CFD simulations indicate that Blockages >15% could be detected by putting some thermocouples in the plenum region of the FA.
Performance Evaluation of SCS for AHTR and Time Assessment of Operation Procedures
Authors: Ogura Takahito , Nakata Junya , Mori Mititsugu , Sakashita Hiroto , Miwa Shuichiro 
Published: November 2014

The Advanced High-Temperature Reactor (AHTR) is a new nuclear power reactor concept being investigated in some countries including the United States. The coolant is a liquid salt with a melting point of 460°C and a boiling point of 1430°C. The AHTR uses Silo Cooling System (SCS) as the decay heat removal system in a Beyond-Design-Basis Accident (BDBA). SCS has two accident mitigations. The first component is low-cost, and thick steel rings which conduct heating up the silo wall for BDBA. The second component is an annular ring of an inexpensive, solidified BDBA salt, which is heated from the bottom and melts when the temperature of the salt increases above the melting point, then flows into the silo, and floods the whole silo to its top level. SCS could make AHTR free from catastrophic accidents, where core melting or vessel failure never takes place since the BDBA salt near the top of silo passively absorbs decay heat. On the other hand, AHTR decreases its heat removal ability to avoid freezing of the salt and blocking the flow of the liquid when the temperatures are low. We performed the numerical calculation of AHTR heat removal system and evaluated whether it has the ability to remove decay heat with the robustness for a long-time cooling operation after BDBA. Furthermore, we need to build up and optimize the operation plan of SCS in AHTR, taking its thermal characteristics of this system into account. It is essential to avoid severe accidents which we can suppose as the possible catastrophic scenario.
In this paper, we calculated temperature distributions using the thermal-hydraulics code developed for AHTR, and assessed the performance in a long term cooling period under BDBA conditions. Finally, we investigated the temperature distributions of the whole plant, predicting the accident scenario without air-cooled passive decay heat-removal system. We obtained important conclusion about SCS of the AHTR that its heat removal ability was enough to avoid catastrophic accidents under Loss of Heat Sink (LOHS) conditions.
Experimental Study of the Cadmium Effects on Iodine Transport in the Primary Circuit During Severe Nuclear Accident
Authors: Gouëllo M. , Kalilainen J. , Rantanen P. , Kärkelä T. , Auvinen A. 
Published: November 2014

In case of a severe accident in a light water reactor, iodine is one of the most important fission products in the context of reactor safety because of its significant total fuel inventory, high specific activity and radiotoxicity. Consequently, understanding its behavior under severe accident transient conditions is a major point in the optimization of the accident management and mitigation.
An experimental study has been launched at VTT investigating the behavior of iodine on primary circuit surfaces during a severe nuclear accident. The paper presents results obtained from the heating of metallic cadmium and caesium iodide in a crucible at temperature below 400°C under three different atmospheres. Aerosols and gaseous species released from the reaction crucible were sampled at 150°C on filters and liquid scrubbers and analyzed with HR-ICP-MS.
At first, experiments were conducted with one precursor (cadmium or caesium iodide). It was observed that cadmium is predominantly present as aerosol in all atmospheres but represents less than 1% of mass what has been recovered from the facility leaching. To the contrary, caesium iodide experiment showed that sampled iodine exists mainly as gaseous species.
By comparison to one-precursor studies, a change in the behavior of element was noticed when cadmium was added to caesium iodide. Different observations let to suggest a reaction between caesium iodide and cadmium in the crucible. The reaction would lead to the formation of compound made of cadmium and iodine in condensed phase below 150°C. More gaseous iodine was found from the sampling line with Ar/H2O/H2 atmosphere than in Ar/H2O atmosphere, suggesting that the cadmium is more reactive toward iodine in hydrogen-free atmosphere.
Post-Fukushima Assessment of the AP1000® Plant
Authors: Schulz Terry L. , Kindred Thomas A. , Friedman Bryan N. , Glavin John E. , Malinowski Adam D. , Mathewson Phil A. 
Published: November 2014

The AP1000 plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety with reduced plant costs. The AP1000 passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at the Fukushima Daiichi nuclear power station without compromising core and containment integrity. Without AC power, using passive safety technology, the AP1000 plant provides cooling for the core, containment and spent fuel pool for more than 3 days without the need for operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. With the re-supply of fuel oil the coping time may be extended for an indefinite time. Connections for a few, small, easily transportable components provide a diverse backup means of extending passive system operation after the first 3 days. As a result, the AP1000 design provides very robust protection of public safety and the utility investment.
Following the accident at the Fukushima Dai-ichi nuclear power station in Japan, several initiatives were launched worldwide to assess the lessons learned. These include, but are not limited to, the European Nuclear Safety Regulators Group (ENSREG) stress tests, the Office for Nuclear Regulation (ONR) Final Report, the International Atomic Energy Agency (IAEA) Expert Mission Report, and the U.S. NRC Near-Term Task Force Recommendations. The AP1000 design has been assessed against these initiatives and lessons learned.
The purpose of this paper is to describe:
• How the accident at the Fukushima Dai-ichi nuclear power station was evaluated and translated into conclusions and recommendations for nuclear power plants worldwide
• How the AP1000 plant was evaluated in light of the recommendations resulting from the various post-Fukushima assessments
• The key conclusions resulting from the post-Fukushima evaluation of the AP1000 design
ÚJV Řež, a.s. Research Programme Supporting Development of Deep Geological Repository in the Czech Republic
Authors: Havlová Václava 
Published: November 2014

ÚJV Řež, a.s. as a company with a long term experience in radioactive waste management (RWM) has been running a comprehensive research programme, supporting development of deep geological repository (DGR) in the Czech Republic.
Recently ÚJV Řež, a.s. research has focused on the different aspects of safety functions that DGR barriers should provide. Moreover, the research has also recently paid strong attention to real conditions that can be present in DGR (anaerobic reducing conditions, increased T due to heat generation by radioactive waste, contact of different materials within repository, real scale of the rock massive etc.). Both types of experiments, laboratory and in-situ experiments in underground laboratories, were included in the research programme. The presentation gives a brief overview of experimental trends, being conducted for materials and conditions, concerned in Czech repository concept.