bluuebunny/crossref_metadata_2025 huggingface dataset:
Natural Convection Heat Transfer in a Rectangular Water Pool with Internal Heating and Top and Bottom Cooling
Authors: Lee Jong K. , Lee Seung D. , Suh Kune Y.
Published: September 2008

During a severe accident, the reactor core may melt and be relocated to the lower plenum to form a hemispherical pool. If there is no effective cooling mechanism, the core debris may heat up and the molten pool run into natural convection. Natural convection heat transfer was examined in SIGMA RP (Simulant Internal Gravitated Material Apparatus Rectangular Pool). The SIGMA RP apparatus comprises a rectangular test section, heat exchanger, cartridge heaters, cooling jackets, thermocouples and a data acquisition system. The internal heater heating method was used to simulate uniform heat source which is related to the modified Rayleigh number Ra′. The test procedure started with water, the working fluid, filling in the test section. There were two boundary conditions: one dealt with both walls being cooled isothermally, while the other had to with only the upper wall being cooled isothermally. The heat exchanger was utilized to maintain the isothermal boundary condition. Four side walls were surrounded by the insulating material to minimize heat loss. Tests were carried out at 1011 < Ra′ < 1013. The SIGMA RP tests with an appropriate cartridge heater arrangement showed excellent uniform heat generation in the pool. The steady state was defined such that the temperature fluctuation stayed within ±0.2 K over a time period of 5,000 s. The conductive heat transfer was dominant below the critical Rayleigh number Ra′c, whereas the convective heat transfer picked up above Ra′c. In the top and bottom boundary cooling condition, the upward Nusselt number Nuup was greater than the downward Nusselt number Nudn. In particular, the discrepancy between Nuup and Nudn widened with Ra′. The Nuup to Nudn ratio was varied from 7.75 to 16.77 given 1.45×1012 < Ra′ < 9.59×1013. On the other hand, Nuup was increased in absence of downward heat transfer for the case of top cooling. The current rectangular pool testing will be extended to include circular and spherical pools.
Natural Convection Heat Transfer in a Rectangular Liquid Metal Pool With Bottom Heating and Top Cooling
Authors: Lee Il S. , Yu Yong H. , Son Hyoung M. , Hwang Jin S. , Suh Kune Y.
Published: September 2008

An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). A relationship is determined between the Nusselt number Nu and the Rayleigh number Ra in the liquid metal rectangular pool. Results are compared with correlations and experimental data in the literature. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with top subcooling are found to be similar to those predicted by the existing correlations or experiments. The current test results are utilized to develop natural convection heat transfer correlations applicable to low Prandtl number Pr fluids that are heated from below and cooled by the external coolant above. Results from this study are slated to be used in designing BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment at remote sites cycled with MOBIS (Modular Optimized Brayton Integral System) for electricity generation, tied with NAVIS (Naval Application Vessel Integral System) for ship propulsion, joined with THAIS (Thermochemical Hydrogen Acquisition Integral System) for hydrogen production, and coupled with DORIS (Desalination Optimized Reactor Integral System) for seawater desalination. Tests are performed with Wood’s metal (Pb-Bi-Sn-Cd) filling a rectangular pool whose lower surface is heated and upper surface cooled by forced convection of water. The test section is 20 cm long, 11.3 cm high and 15 cm wide. The simulant has a melting temperature of 78°C. The constant temperature and heat flux condition was realized for the bottom heating once the steady state had been met. The test parameters include the heated bottom surface temperature of the liquid metal pool, the input power to the bottom surface of the section, and the coolant temperature.
Development of a Probabilistic Tsunami Hazard Analysis in Japan
Authors: Sakai Toshiaki , Takeda Tomoyoshi , Soraoka Hiroshi , Yanagisawa Ken , Annaka Tadashi
Published: September 2008

It is meaningful for tsunami assessment to evaluate phenomena beyond the design basis as well as seismic design. Because once we set the design basis tsunami height, we still have possibilities tsunami height may exceeds the determined design tsunami height due to uncertainties regarding the tsunami phenomena. Probabilistic tsunami risk assessment consists of estimating for tsunami hazard and fragility of structures and executing system analysis. In this report, we apply a method for probabilistic tsunami hazard analysis (PTHA). We introduce a logic tree approach to estimate tsunami hazard curves (relationships between tsunami height and probability of excess) and present an example for Japan. Examples of tsunami hazard curves are illustrated, and uncertainty in the tsunami hazard is displayed by 5-, 16-, 50-, 84- and 95-percentile and mean hazard curves. The result of PTHA will be used for quantitative assessment of the tsunami risk for important facilities located on coastal area. Tsunami hazard curves are the reasonable input data for structures and system analysis. However the evaluation method for estimating fragility of structures and the procedure of system analysis is now being developed.
Simulation of Steam Explosion with a General Purpose CFD Code
Authors: Leskovar Matjazˇ , Koncˇar Bosˇtjan
Published: September 2008

An ex-vessel steam explosion may occur when during a severe reactor accident the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles at later times, during the expansion of the highly pressurized water vapor, that may endanger surrounding structures. In contrast to specialized steam explosion CFD codes, where the steam explosion is modeled on micro-scale using fundamental averaged multiphase flow conservation equations, in the presented approach the steam explosion is modeled in a simplified manner as an expanding high-pressure premixture of dispersed molten fuel, liquid water and vapor. Applying the developed steam explosion model, a comprehensive analysis of the ex-vessel steam explosion in a typical PWR reactor cavity was done using the CFD code CFX-10. At four selected locations, which are of importance for the assessment of the vulnerability of cavity structures, the pressure histories were recorded and the corresponding pressure impulses calculated. The pressure impulses determine the destructive potential of the steam explosion and represent the input for the structural mechanical analysis of the cavity structures. The simulation results show that the pressure impulses depend mainly on the steam explosion energy conversion ratio, whereas the influence of the premixture vapor volume fraction, which is a parameter in our model and determines the maximum steam explosion pressure, is not significant.
Isolation and Damping of Shocks, Vibrations, Impact Load and Seismic Movements at Buildings, Equipment and Pipe Networks by SERB-SITON Method
Authors: Panait A. , Serban V.
Published: September 2008

The paper presents SERB–SITON method to control, limit and damp the shocks, vibration, impact load and seismic movements with applications in buildings, equipment and pipe networks (herein called: “components”). The elimination or reduction of shocks, vibration, impact load and seismic movements is a difficult problem, still improperly handled theoretically and practically because many times the phenomena are random in character and the behavior of components is non-linear with variations of the properties in time, variations that lead to the increase or decrease of the energy & impuls transfer from the dynamic excitation to the components. Moreover, the existing supports and dampers applied today, are not efficient enough in the reduction of the dynamic movement for all the frequency ranges met with in the technical application field. The stiffness and damping of classic supports do not allow a good isolation of components against shocks and vibrations so to eliminate their propagation to the environment and neither do they provide a satisfactory protection of the components sensitive to shocks and vibrations and seismic movements coming from the environment. In order to reduce the effects of shocks, vibrations impact and seismic movements on the components, this paper presents the results obtained by SITON on the concept, design, construction, experimental testing and application of new types of supports, devices and thin lattice structure, called “SERB”, capable to overtake large static loads, to allow displacements from impact, thermal expansions or yielding of supports and which, in any work position, can elastically overtake large dynamic loads or impact loads which they damp. The new supports and devices and thin lattice structure allow their adjustment without the occurrence of overstressing in the components due to their non–linear geometric behavior, and the contact pressure among the elements is limited to pre-set values to avoid blocking phenomena that generates great stresses induced by thermal expansion for example. Due to their characteristics of adjustment to the actual position and level of stress, SERB supports, devices and thin lattice structure show minimal effects on the components stress condition whenever the installation and computation errors. Herein below it is a presentation of the actual results obtained by SITON in the isolation of heavy equipment and pipe networks and others in process of application for buildings. Due to the very good results obtained in the isolation against shocks, vibrations and seismic movements at components in the conventional industry, there is the proposal to implement SERB-SITON method to the increase of the safety level at new or existing Nuclear Power Plants or to protect nuclear building against missiles and airplane crush impact.
Degradation Assessment in Complex Systems
Authors: Matuzas Vaidas , Augutis Juozas , Uspuras Eugenijus
Published: September 2008

Components condition could change dynamically because of various dependencies existing in complex systems since usually not only physical connections exist in the system. Various processes (as degradation or wear out, reliability decrease, failures) affecting one component have influence to condition of other components as well. Current paper is devoted to the analysis and development of mathematical models to the reliability assessment of network systems. Degradation as the main process in the network systems is taken into account. The main attention in present work was paid to the mathematical description of the degradation spreading mechanism in network systems. Developed recursive mathematical model allows assess degradation in network systems during various time moments. Developed methods can be applied for assessment of various risk estimates of hazardous systems such as systems at nuclear power plants.
Thermal-Hydraulic Experimental Studies on Safety and Efficiency of NPP with VVER
Authors: Blinkov Vladimir N. , Melikhov Oleg I. , Elkin Ilya V. , Melikhov Vladimir I. , Parfenov Yuriy , Boltenko Eduard A. , Lipatov Igor A.
Published: September 2008

New design of nuclear power plant with pressurized water reactor entitled “NPP-2006” is under progress in Russia. It is evolutionary type of design based on existing NPP with VVER-1000 reactors. Intensive thermal-hydraulic experimental activities support design development. They are aimed at increasing of NPP safety and efficiency. Integral test facility PSB-VVER, modeling main components of primary system of VVERs, is the main tool for experimental studies. Volumetric-power scale of the test facility is 1:300, elevations are 1:1. A core model is an assembly of 168 fuel rod imitators. About 20 experiments at the PSB-VVER have been performed to generate experimental data base for validation of best estimate thermal-hydraulic system codes for application to analyses of VVERs transients and accidents. An important direction of PSB-VVER experimental activities is connected with the development and verification of accident management procedures. 15 experiments have been conducted to support development of such procedures. For new plant design (hydroaccummulators, operating under low pressure) is under experimental studies at the PSB-VVER. The experimental program also includes the approach, based on investigations of local thermal-hydraulic parameters in the fuel rod imitators assembly at the large-scale test facility. This approach is supposed to decrease a level of design conservatism and to enhance NPP efficiency.
Perspectives of Increasing Efficiency and Productivity of Electromagnetic Induction Pumps for Mercury Basing on Permanent Magnets
Authors: Bucenieks I.
Published: September 2008

In the next generation neutron sources the HLM (heavy liquid metals) such as lead, lead based eutectic alloys and mercury will be used both as spallation target material and simultaneously as the cooling liquid. In this aspect the design of safe and effective pumps for HLM recirculation at high pressure heads and big flow rates becomes important. For this purpose electromagnetic inductions pumps having no problems of hydraulic seals being in contact with liquid metal (electromagnetic forces in the liquid metal are induced by magnetic system located outside of the channel of pump) are more perspective from the point of view of their safety for operation at high temperature and radiation conditions in comparison with mechanical pumps. At the Institute of Physics of University of Latvia (IPUL) the design concept of electromagnetic induction pumps basing on the principle of rotating permanent magnets (PMP) have been developed. Such design concept of electromagnetic induction pumps has many advantages in comparison with traditionally used electromagnetic induction pumps basing on 3-phase linear flat or cylindrical inductors. The estimations of parameters of powerful pumps (such as overall dimensions of the active magnetic system, power of motor needed for pump drive, the efficiency of pump) for mercury for the developed by pump pressure heads in the range up to 10.0 bar and provided flow rates in the range up to 20 litres per second are demonstrated.
Optimization of the Mode of the Uranium-233 Accumulation for Application in Thorium Self-Sufficient Fuel Cycle of CANDU Power Reactor
Authors: Bergelson Boris , Gerasimov Alexander , Tikhomirov Georgy
Published: September 2008

Results of calculation studies of the first stage of self-sufficient thorium cycle for CANDU reactor are presented in the paper. The first stage is preliminary accumulation of 233U in the CANDU reactor itself. Parameters of active core and scheme of fuel reloading were accepted the same as those for CANDU reactor. It was assumed for calculations, that enriched 235U or plutonium was used as additional fissile material to provide neutrons for 233U production. Parameters of 10 different variants of the elementary cell of active core were calculated for the lattice pitch, geometry of fuel channels, and fuel assembly of the CANDU reactor. The results presented in the paper allow to determine the time of accumulation of the required amount of 233U and corresponding number of targets going into processing for 233U extraction. Optimum ratio of the accumulation time to number of processed targets can be determined using the cost of electric power produced by the reactor and cost of targets along with their processing.
Corrosion Phenomena of Eurofeer Steel in Pb-17Li Stationary Flow at Magnetic Field
Authors: Platacis E. , Bucenieks I. , Muktupavel F. , Shishko A.
Published: September 2008

Search of new energy sources draws the increasing attention to use for this purpose of reactors. In the Europe some years the program EUROATOM uniting scientific of the many countries for the decision of constructive problems at designing of fusion reactors operates. One of the main things in this program is the problem of liquid metals breeder blanket behaviour. Structural material of blanket should meet high requirements because of extreme operating conditions. Therefore the knowledge of the effect of metals flow velocity, temperatures and also a neutron irradiation and a magnetic field on the corrosion processes are necessary. At the moment the eutectic lead -lithium (Pb-17Li) is considered as the most suitable tritium breeder material [1–3]. In turn as a structural material have been tested both many austenitic and ferritic-martensitic steels [2–4]. As the optimum variant is considered steel EUROFER 97, which corrosion rate in liquid Pb-17Li eutectic is the least [3,4]. However, these results have been received without taking into account influence of a strong magnetic field. At the same time, this influence should be essential, as because of change of hydrodynamics of a liquid metal flow, and because of interaction of a magnetic field with a ferromagnetic steel. It has been shown in [5,6] that the magnetic field leads to increase of corrosion rate for austenitic (316L) and martensitic (1,4914) steels. Experimental data for EUROFER 97, and also a theoretical substantiation of the phenomenon are absent, that creates essential difficulties for forecasting working capacity of blanket construction. The aim of presented work were the theoretical and experimental investigations of magnetic field influence on the corrosion of EUROFER 97 steel exposed to flowing Pb-17 Li in specific designed loop.
Application of Entry-Time Processes to Asset Management in Nuclear Power Plants
Authors: Nelson Paul , Wang Shuwen , Kee Ernie J.
Published: September 2008

The entry-time approach to dynamic reliability is based upon computational solution of the Chapman-Kolmogorov (generalized state-transition) equations underlying a certain class of marked point processes. Previous work has verified a particular finite-difference approach to computational solution of these equations. The objective of this work is to illustrate the potential application of the entry-time approach to risk-informed asset management (RIAM) decisions regarding maintenance or replacement of major systems within a plant. Results are presented in the form of plots, with replacement/maintenance period as a parameter, of expected annual revenue, along with annual variance and annual skewness as indicators of associated risks. Present results are for a hypothetical system, to illustrate the capability of the approach, but some considerations related to potential application of this approach to nuclear power plants are discussed.
Structural Integrity Demonstration of High Burnup Fuel Rod
Authors: Liu Tong , Song Danrong , Zhao Hua , Wang Kaiming
Published: September 2008

This paper introduces the current status and the work has been done for structural integrity demonstration of high burnup fuel in China. The FRAMATOME ANP COCCINEL, FLAMBE and MISTIGRI design codes are used in the structural integrity demonstration of fuel rod. This paper presents the results of structural integrity demonstration of AFA3G fuel rod in GNPS. All the design criterion are met. Based on CAFA program, the new fuel models are planning to be developed in the near future.
Thermal Analysis of Spent Fuel Transportation Cask
Authors: Ali Seik Mansoor , Goyal P. , Verma Vishnu , Ghosh A. K. , Kushwaha H. S.
Published: September 2008

Spent fuel transportation casks are required to meet among others, the regulatory thermal test conditions in order to demonstrate their ability to withstand specified accidental fire conditions during transport. This paper describes the transient thermal analysis performed with the above intention for a transportation cask that has not undergone drop test and is not damaged. The original dimensions are used for computations. The analysis was carried out using a heat conduction code employing the Finite Element Method (FEM). At the outset, a benchmark exercise was carried out with the present code to ascertain its capability to carry out calculations involving phase change. The benchmarking was done against the solution given by commercial CFD code STAR CD. After assessing the suitability of the FEM code, it was employed for thermal analysis of the transportation cask. The computation covers normal transport condition, half an hour fire test at 800°C and also the post fire cool down period. The objective of the analysis was to determine the maximum outer surface temperature for normal transport conditions and to assess the extent of melting of lead during fire and post fire period.
Dependence of Vapor Void Fraction on Fundamental Bubble Parameters in Subcooled Flow Boiling
Authors: Kubota Hayato , Ishida Tatsuhiro , Okawa Tomio , Kataoka Isao , Mori Michitsugu
Published: September 2008

A visual study of water subcooled flow boiling was conducted to clarify the mechanism of triggering the net vapor generation (NVG). The test section was a transparent sapphire grass tube of 20 mm in inside diameter; a high-speed camera was used to capture the behavior of vapor bubbles. In the present experiments, the vapor void fraction in the heated tube was expressed as the function of the following bubble parameters: nucleation site density, frequency of bubble release, bubble lifetime, and bubble size. Among these four bubble parameters, the bubble size had a particularly strong influence on the vapor void fraction: the void fraction was approximately proportional to the forth power of mean bubble diameter. Consequently, mean bubble diameter should be large enough for the vapor void fraction to increase rapidly with the wall heat flux. In low flowrate experiments, bubbles generated at nucleation sites were relatively large at the onset of nucleate boiling (ONB). The heat flux at ONB hence appeared the reasonable approximation of that at NVG. Whereas, in high flowrate experiments, bubbles were small at ONB and much higher heat flux was necessary to obtain large bubbles. Thus, the heat flux required to trigger NVG was much higher than that at ONB. It was concluded in the present experimental conditions that accurate evaluation of mean bubble diameter was of significant importance in predicting the onset of net vapor generation.
Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code
Authors: Wang Shisheng , Rineiski Andrei , Maschek Werner , Ignatiev Victor
Published: September 2008

Recent developments extending the capabilities of the SIMMER-III [1, 2] code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmutater. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed.
Metal Surface Decontamination by the PFC Solution
Authors: Won Hui-Jun , Kim Gye-Nam , Choi Wang-Kyu , Jung Chong-Hun , Oh Won-Zin
Published: September 2008

PFC (perfluorocarbon) spray decontamination equipment was fabricated and its decontamination behavior was investigated. Europium oxide powder was mixed with the isotope solution which contains Co-60 and Cs-137. The different shape of metal specimens artificially contaminated with europium oxide powder was used as the surrogate contaminants. Before and after the application of the PFC spray decontamination method, the radioactivity of the metal specimens was measured by MCA. The decontamination factors were in the range from 9.6 to 62.4. The spent PFC solution was recycled by distillation. Before and after distillation, the turbidity of PFC solution was also measured. From the test results, it was found that more than 98% of the PFC solution could be recycled by a distillation.
Ligament Rupture Pressure of Fretted SG Tubes of PWRs
Authors: Hwang Seong Sik , Jung Man Kyo , Kim Hong Pyo , Kim Joung Soo
Published: September 2008

A fretting/wear degradation at the tube support in the U-bend region of a steam generator (SG) of a pressurized water reactor (PWR) has been reported. Simulated fretted flaws were machined on SG tubes of 195 mm in length. A pressure test was carried out with the tubes at room temperature by using a high pressure test facility which consisted of a water pressurizing pump, a test specimen section and a control unit. Water leak rates just after a ligament rupture or a burst were measured. Tubes degraded by up to 70% of the tube wall (TW) showed a high safety margin in terms of the burst pressure during normal operating conditions. Tubes degraded by up to 50% of the TW did not show a burst. Burst pressure depended on the defect depths rather than on the wrap angles. The tube with a wrap angle of 0° showed a fish mouth fracture, whereas the tube with a 45° wrap angle showed a three way fracture.
Transient Thermo-Hydraulic Analysis of the Windowless Target System for the Lead Bismuth Eutectic Cooled Accelerator Driven System
Authors: Bianchi Fosco , Ferri Roberta , Moreau Vincent
Published: September 2008

The target system, whose function is to supply an external neutron source to the ADS sub-critical core to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. A windowless option was chosen as reference configuration for the target system of the LBE-cooled ADS within the European PDS-XADS project in order to reduce the material damage and to increase its life. This document deals with the thermo-hydraulic results of the calculations performed with STAR-CD and RELAP5 codes for studying the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes needed for this analysis. The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and the beam trips lasting more than one second.
Single- and Two-Phase Diversion Cross-Flows Between Triangle Tight Lattice Rod Bundle Subchannels: Data on Flow Resistance and Interfacial Friction Coefficients for the Cross-Flow
Authors: Higuchi Tatsuya , Kawahara Akimaro , Sadatomi Michio , Kudo Hiroyuki
Published: September 2008

Single- and two-phase diversion cross-flows arising from the pressure difference between tight lattice subchannels are our concern in this study. In order to obtain a correlation of the diversion cross-flow, we conducted adiabatic experiments using a vertical multiple-channel with two subchannels simplifying the triangle tight lattice rod bundle for air-water flows at room temperature and atmospheric pressure. In the experiments, data were obtained on the axial variations in the pressure difference between the subchannels, the ratio of flow rate in one subchannel to the whole channel, the void fraction in each subchannel for slug-churn and annular flows in two-phase flow case. These data were analyzed by use of a lateral momentum equation based on a two-fluid model to determine both the cross-flow resistance coefficient between liquid phase and channel wall and the gas-liquid interfacial friction coefficient. The resulting coefficients have been correlated in a way similar to that developed for square lattice subchannel case by Kano et al. (2002); the cross-flow resistance coefficient data can be well correlated with a ratio of the lateral velocity due to the cross-flow to the axial one irrespective of single- and two-phase flows; the interfacial friction coefficient data were well correlated with a Reynolds number, which is based on the relative velocity between gas and liquid cross-flows as the characteristic velocity.
RELAP5-3D Code Includes ATHENA Features and Models
Authors: Riemke Richard A. , Davis Cliff B. , Schultz Richard R.
Published: September 2008

Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, sf6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper.
Experience and Results of Calculations in Case of Vertical Cable Tray Fires
Authors: Ro¨wekamp M. , Klein-Heßling W. , Riese O. , Berg H. P.
Published: September 2008

An “International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications” (ICFMP) has been started in 1999 to improve the reliability and the applicability of different fire models with respect to modeling fires for application to nuclear plants. Within the frame of this project several benchmark exercises have been conducted from different organizations to represent various fire scenarios. For one of these benchmark exercises, four full scale cable tray fire experiments have been performed. The goal of these experiments was to gain more insights on the consequences and effects of cable fires, in particular with respect to the functional failures of electric equipment, as well as for validation of state-of-the-art fire simulation codes. The test series considered preheating effects and includes time dependent electrical functional failure measurements. In this paper, an overview of the experimental as well as the calculated results of two different types of codes is given.
Two-Phase Frictional Pressure Loss in Horizontal Bubbly Flow with 90-Degree Bend
Authors: Kim Seungjin , Park Jung Han , Kojasoy Gunol , Kelly Joseph
Published: September 2008

The two-phase pressure drop due to the minor loss in horizontal bubbly two-phase flow is studied. In particular, geometric effects of a 90-degree elbow is of interest in the present study. Experiments are performed in air-water two-phase flow near atmospheric pressure condition in round glass tube with inner diameter of 50.3mm. Along the test section, 90-degee elbow is installed at L/D = 206.6 from the two-phase mixture inlet. Experiments are performed in 15 different flow conditions and the local static pressures are measured at five axial locations. Characteristic pressure drop due to the elbow is clearly demonstrated in the profiles of local pressure data along the axial direction. It is also found that the elbow effect propagates and is more significant further downstream than immediate downstream of the elbow. The overall two-phase frictional pressure loss between L/D = 0 and 329 can be predicted well with the Lockhart-Martinelli correlation with parameter C = 25, which is higher than the generally accepted value of C = 20. A correlation for the two-phase pressure loss, including the minor loss due to the 90-degree elbow is developed by employing the approach analogous to that of Lockhart-Martinelli’s. The newly developed correlation suggests that the modified parameter, C = 65 fits best with the experimental data. In addition, the two-phase minor loss factor for the 90-degree elbow is found to be k = 0.58, 50% higher than that recommended for single-phase flow.
Measurement and Analysis of Void Fraction in a Multiple-Channel Simplifying Triangle Tight Lattice Rod Bundle
Authors: Sadatomi Michio , Kawahara Akimaro , Kudo Hiroyuki , Shirai Hiroshi
Published: September 2008

In order to know the effects of reduced surface tension on void fraction, adiabatic experiments were conducted for both air-water and air-water with surfactant systems at room temperature and pressure. Void fraction data were obtained for bubbly, slug, churn and annular flows in a vertical channel with two subchannels simplifying a triangle tight lattice rod bundle. The void fraction was found to be lower in air-water system than air-water with surfactant one. In addition, the void fractions for both systems were found to be lower than those calculated by various correlations in literatures for circular pipe flow. In order to study the cause of the above data trend, for annular flows as a first step, the void fraction has been calculated by a subchannel analysis using wall and interfacial friction correlations in literatures as constitutive equations, and by assuming the liquid film to be uniform over the wall perimeter. The best agreement between the calculation and the experiment has been obtained when NASCA correlation for wall friction force and modified RELAP5/MOD2 correlation incorporating reduced surface tension effects for interfacial friction force were used.
Development of 1000 MWe Advanced Boiling Water Reactor
Authors: Hisajima Kazuo , Uchida Ken , Matsumoto Keiji , Kondo Koichi , Yokoyama Shigeki , Miyagawa Takuya
Published: September 2008

1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29m of 1350 MWe ABWR to 26.5m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage.
Development and Application of Laser Peening System for PWR Power Plants
Authors: Yoda Masaki , Chida Itaru , Okada Satoshi , Ochiai Makoto , Sano Yuji , Mukai Naruhiko , Komotori Gaku , Saeki Ryoichi , Takagi Toshimitsu , Sugihara Masanori , Yoriki Hirokata
Published: September 2008

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it’s too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described.
Development of the ISI Device for Fast Breeder Reactor MONJU Reactor Vessel
Authors: Tagawa Akihiro , Ueda Masashi , Yamashita Takuya
Published: September 2008

In-service inspection (ISI) is carried out to confirm the integrity of the main components of the Fast Breeder Reactor (FBR) “MONJU”. The weld-joints are examined by using an inspection device which has a glass fiber scope for visual examination and a horizontally polarized shear (SH) wave electromagnetic acoustic transducer (EMAT) for volumetric testing. The ambient temperature during the inspection is 200°C and the irradiation field is 10 Sv/hr. A new inspection device has been developed in order to improve the visual test performance, volumetric test performance and controllability of the inspection device reflecting the experience of the original test. In this paper, detail of the new inspection device and the test results of sensors such as the CCD camera, EMAT and bead sensor are reported. The paper also reports on the CCD camera cooling system and other components.
Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant
Authors: Kawabata Osamu , Kajimoto Mitsuhiro
Published: September 2008

In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHEONICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the start of steam release.
World’s First PWR Reactor Internals Replacement
Authors: Uchiyama Junichi
Published: September 2008

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of the upper and lower reactor internals of the 566 MW Ikata Unit No.1 of Shikoku Electric Power Co., Inc. The event marks the world’s first all-in-one-piece extraction and replacement work of its kind in a pressurized water reactor. The author wishes to introduce an outline and features of this milestone.
Study on Jet Breakup Behavior at Core Disruptive Accident for Fast Breeder Reactor
Authors: Matsuo Eiji , Abe Yutaka , Nariai Hideki , Chitose Keiko , Koyama Kazuya , Itoh Kazuhiro
Published: September 2008

It is important to estimate the cooling possibility of the molten jet in coolant during a core disruptive accident (CDA) of a fast breeder reactor (FBR). In the present study, the molten jet of U-alloy 78 simulating the core material is injected into the water simulating the coolant. The visual data of the molten jet breakup behavior is observed by using the high-speed video camera. The front velocity of the molten jet is estimated by using the image processing technique from the visual data. It shows that the front velocity of the molten jet can be divided into three time regions. In the first region, the front velocity of the molten jet increases. In the second region, the front velocity of the molten jet suddenly decreases. In the third region, the front velocity of the molten jet keeps at low and steady. In first region, the column diameter of the molten jet decreases with the passage of time. At the location between first region and second region, the column of the molten jet breaks up and disappears. In the present study, the jet breakup length is defined as the distance from the water surface to the location where the jet column disappears. The results show that the jet breakup length depends on the injection nozzle diameter, but does not depend on the jet penetration velocity. This tendency agrees with the prediction by Epstein’s equation. After the experiment, the solidified fragments are collected and the mass median diameter is measured. The mass median diameter is compared with the existing theories. Furthermore, a model to estimate the cooling possibility during a CDA of a FBR is constructed, reflecting the above-mentioned results.
Pressure Drop Correlations of Single-Phase and Two-Phase Flow in Rolling Tubes
Authors: Cao Xia-Xin , Yan Chang-Qi , Gao Pu-Zhen , Sun Zhong-Ning
Published: September 2008

A series of experimental studies of frictional pressure drop for single phase and two-phase bubble flow in smooth rolling tubes were carried out. The tube inside diameters were 15mm, 25mm and 34.5mm respectively, the rolling angles of tubes could be set as 10° and 20°, and the rolling periods could be set as 5s, 10s and 15s. Combining with the analysis of single-phase water motion, it was found that the traditional correlations for calculating single-phase frictional coefficient were not suitable for the rolling condition. Based on the experimental data, a new correlation for calculating single-phase frictional coefficient under rolling condition was presented, and the calculations not only agreed well with the experimental data, but also could display the periodically dynamic characteristics of frictional coefficients. Applying the new correlation to homogeneous flow model, two-phase frictional pressure drop of bubble flow in rolling tubes could be calculated, the results showed that the relative error between calculation and experimental data was less than ± 25%.
Analysis of Potential Hydrogen Risk in the PWR Containment
Authors: Jian Deng , Cao Xuewu
Published: September 2008

Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential inconsecutive burnings. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity.
One-Loop Operation of Primary Heat Transport System in MONJU During Heat Transport System Modifications
Authors: Goto T. , Tsushima H. , Sakurai N. , Jo T.
Published: September 2008

MONJU is a prototype fast breeder reactor (FBR). Modification work commenced in March 2005. Since June 2004, MONJU has changed to one-loop operation of the primary heat transport system (PHTS) with all of the secondary heat transport systems (SHTS) drained of sodium. The purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the decay heat by an acceptable margin but the capacity of preheaters to keep the sodium within the primary vessel at about 200°C must be maintained. With regard to the heat loss and the decay heat, the estimated heat loss in the primary system was in the range of 90–170kW in one-loop operation, and the calculated decay heat was 21.2kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the decay heat. As for preheaters, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the PHTS was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the PHTS and the modification period was shortened.