bluuebunny/crossref_metadata_2025 huggingface dataset:
Generic CFD for Vortex Induced Acoustic Resonance in Deep Cavities
Authors: Lillberg Eric , Nilsson Per
Published: July 2009

Extended power uprates (EPUs) of nuclear power plants result in changes in steam velocity and temperature. These changes increase the risk of flow induced vibrations due to shear layer instabilities and vortex shedding amplified by acoustic resonance. In the present study the possibility to predict vortex induced acoustic resonance is investigated using compressible Navier-Stokes equations (NSE) for viscous fluid flow with focus on generality. For this purpose the NSE are solved using Large Eddy Simulation (LES) technique and the open source solver OpenFOAM which is available under GPL. Validation is made using experiments as well as a other simulation using a pseudo compressible formulation. In the present study several LES models are used including the more recent Implicit LES (ILES) approach.
Study on Mechanism of Condensation Heat Transfer of Water Jet in Steam Injector
Authors: Shimizu Takahiro , Takahashi Yuhki , Koizumi Yasuo , Ohtake Hiroyasu , Miyashita Toru , Mori Michitsugu
Published: July 2009

It has been proposed to introduce a steam injector into boiling water reactors as a feed water heat exchanger and a safety injection pump. In the present paper, the heat transfer characteristics in the steam injector were examined. The nozzle size of the water jet in the steam injector that was used in the present experiments was 5 mm. The length of the mixing section of the water jet and the steam flow in the injector was 53 mm. Subcooled water was supplied to the water nozzle. Saturated steam at approximately 0.1 MPa was also supplied to the mixing section of the steam injector. Water jet velocities tested in the present experiments were in the range of 9.7 ∼ 21 m/s. The velocities corresponded to the Reynolds number of 5.9×104 ∼ 1.5×105. Radial and axial temperature distributions of the water jet in the steam injector were measured. Velocity distributions of the water jet were also measured. From the measured temperature and the velocity distributions, heat exchange rates from the steam flow around the water jet to the water jet were derived. The obtained results indicated that the heat exchange rates were greatly larger than those of usual turbulent flow in a pipe. A flow state of the water jet was also visually examined. The results of the visual observation revealed that the interface between the water jet and the steam flow was very wavy. It was supposed that the wavy motion on the water jet surface created the effective-large-internal circulation flow in the water jet, which resulted in the tremendously effective heat transport into the center portion of the water jet. From the pictures of the water jet surface recorded by a high speed video camera, characteristics of waves on the surface; the wave height, the wave velocity, the wave length and the wave frequency, were obtained. The heat transfer of the water jet in the steam injector was correlated with the wave characteristic properties. The heat transfer of the water jet was also analyzed by using the commercial CFD code of STAR-CD. When the wavy interface was introduced into the STAR-CD code analysis, the radial heat transport was drastically improved. This analytical result also supported that the tremendously more effective radial heat transport than that of the usual turbulent flow was caused by the wavy motion of the water jet surface.
Tension Metastable Fluid Nuclear Particle Detector: Qualification and Comparisons
Authors: Lapinskas J. , Smagacz P. , Xu Y. , Taleyarkhan R. P.
Published: July 2009

Fluids in states of tension metastability offer unique capabilities for detection of nuclear emissions from fission and other nuclear phenomena. This includes the ability to completely avoid photonic interference when detecting neutrons in addition to being able to detect neutrons over eight orders of magnitude with spectroscopic capabilities, and the ability to provide directionality information, all from the same instrument; altogether, representing an unsurpassed capability for next-generation application of multi-disciplinary technology for diverse fields of application. In this paper we present the underlying principles of detection using tension (negative, i.e., sub-zero) pressure fluid states at room temperature and present results of qualification of performance in terms of intrinsic efficiency of detection of neutrons from fission and fusion sources. It is found that, unlike present day systems where intrinsic efficiencies are limited to about 20% for fast neutrons, the tension metastabile fluid detector (TMFD) systems offer intrinsic efficiencies of over 90% with the ability to readily scale-up in size for vastly improved effective detection.
Comparision of Wind and Nuclear Power Economics
Authors: Rogers Vern C. , Sandquist Gary M. , Kunze Jay
Published: July 2009

The Renewable Energy Community is actively promoting Renewable Portfolio Standards (RPS) for adoption by states as legal mandates for compliance by electrical power utilities. These standards, if implemented, would require utilities to acquire “renewable energy” from available sources or purchases often without full consideration of ratepayer cost impact. An economic analysis of the costs for production of equivalent electrical power base load capacity is developed and compared for both wind and nuclear general electrical power. Various cost components, including utility impact for providing base load supplies, are examined and compared. This analysis allows the anticipated costs to ratepayers for projected wind and nuclear generated electricity to be compared.
Option Study on a Steam Pressure Control Logic for SMART
Authors: Kang Han-Ok , Park Cheon-Tae
Published: July 2009

Design features of SMART such as a large coolant inventory with a relatively low flow rate and the existence of a once-through steam generator require new steam control logic capable of coping with a prompt load change without inducing severe operational parameter fluctuations. A new MMS SMART model was developed to study the load-following capability and the system parameter manageability of three candidate control logics: the reactor leading, the turbine leading, and the feedwater leading logics. The MMS SMART model was composed of several interacting MMS modules with numerical data, each of which represented a component of the SMART plant and control logic. The Reactor Coolant System, and the Steam and Power Conversion System with their control logics were modeled using default modules such as a pipe, a pump, and a tank. The candidate control logics had been implemented in the model and their dynamic characteristics for the case of a 100%-50%-100% load-following operation with a 25%/min rate were examined. With the reactor-leading control logic implemented, the turbine power was changed with a considerable time delay, which was mainly due to coolant temperature signal retardation to the feedwater controller. The steam pressure variation was very limited for the reactor-leading control logic. With the turbine-leading control logic, the turbine power was manipulated well to match the reference value, whereas relatively large fluctuations of the steam pressure and the coolant temperature occurred. The steam pressure swung with a comparatively large amplitude and the peak value of the fluctuation was not reduced even with larger gain values of the PI controller. This steam pressure swing was considerably decreased with the feedwater leading control logic, while the reactor power and the coolant temperatures had similar trends to those of the turbine leading control logic.
Digital I&C System in the US-APWR
Authors: Shirasawa Hiroshi , Ishimoto Shunsuke , Mashio Kenji
Published: July 2009

The digital safety Instrumentation and Control (I&C) system has been developed and approved in Japan. Our digital I&C system has been applied to many non-safety systems and approved for full I&C system including a safety system. The fully digitalized I&C system will soon be in operation in a newly constructed plant in Japan. Based on this proven technology, the digital I&C system will be applied to the I&C system in the US-APWR. This digital I&C system is slightly improved and modified to conform to the U.S. codes and standards. The I&C system of the US-APWR has some specific features. The system consisting of four systems capable of diverse echelons can ensure defense-in-depth and diversity which also achieves countermeasure against software common cause failures. In addition, the I&C system has a four-train configuration that applies to the safety human-system interface systems, safety protection system and safety plant component controls.
Specificity of Flow Structure in a Dual Elbow and Its Visualization by PIV Measurement
Authors: Yoshida Kazuhiro , Kazuhisa Yuki , Hashizume Hidetoshi , Toda Saburo
Published: July 2009

A large number of pipe failures caused by wall thinning have been reported in nuclear power plants, some of which occur in a dual elbow or the vicinity of it. These pipe failures could be influenced by complex flow induced in the elbow. This study, therefore, aims at predicting the whole flow structure in the dual elbow as the first step by taking a secondary flow after the elbow by PIV measurement. A test section consists of two elbows that are 2-dimentionally connected with/without a straight pipe. They are made of acrylic resin. The diameter of the elbow is 56mm and the curvature radius ratio is 1.5. Reynolds number in this experiment is 4×104. It is confirmed that the flow structure in the dual elbow has specificity depending on the inlet flow condition to the elbow and that the secondary flow itself has swirling motion in a streamwise direction. The dual elbow seems to generate more complex and unstable flow field even when the flow field at the inlet of the elbow slightly changes from a fully developed flow. However, there is a strong possibility that putting a straight pipe between the two elbows makes it ease the occurrence of the complex flow field.
Effect of Hydrated Salt Additives on Film Boiling Behavior at Vapor Film Collapse
Authors: Arai Takahiro , Furuya Masahiro
Published: July 2009

A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a high-speed digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature, however, approaches a constant value when the slat concentration is close to its saturation concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts.
Validation of a Computer Code for Analysis of Elasto-Plastic Crack Propagation
Authors: Malik Afzaal M. , Khawar Jawad
Published: July 2009

Finite Element Method (FEM) is routinely used for a better understanding of the behavior of structures subjected to various loads during their service life. A class of such problems is related to defects present in materials in form of cracks, which propagate, leading eventually to fracture and structural failure. Such problems are of crucial importance to many applications such as nuclear, aerospace, and aeronautical industries. In this research paper, an FEM code is developed and is used to study crack propagation in ductile materials under specified load conditions. This code performs crack propagation analysis based on J-Integral method. This method of crack propagation analysis can be used to study crack propagation in brittle as well as ductile materials. The results from the analysis are compared with experimental and analytical data available in the literature.
Numerical Simulation of Upward Two-Phase Flow in Narrow Channel Between Two Flat Plates With Bilateral Heating
Authors: Li Lei , Zhang Zhijian , Liu Jiange
Published: July 2009

The convection heat transfer characteristic in narrow channel is superior. Therefore, narrow channels are suitable for being manufactured into compact heat transfer components with high heat transfer performance. This new technology of heat transfer enhancement with no source is widely applied to various fields such as microelectronic heat sink, cryogenic industry, chemical industry, aeronautic & astronautic industry and nuclear engineering. In integrated reactor, the coolant channels between the plate fuels are narrow channels. As to the investigation of two-phase flow and heat transfer in narrow channels, reports are focused on experimental research with low pressure, at home and abroad, but reports on numerical simulation research are relatively small. Based on separated flow, a theoretical two-fluid model predicting for upward flow in a rectangular narrow channel with bilateral heating has been developed in this paper. The theoretical model is based on fundamental conservation principles: the mass, momentum and energy conservation equations of liquid films and the momentum conservation equation of vapor core. The model assumed that liquid film covers the surface of the channel while the vapor with entrained droplets flows in the central core. And there exists the mass transfer between the liquid droplets in the vapor core and the liquid films. Through numerically solving these equations, liquid film thickness, radial velocity and temperature distribution in liquid films and heat transfer coefficient are obtained. The calculated results shows that the width of the narrow channel, heat pattern, heat flux have great influences on heat transfer coefficient and the thickness of liquid film. The heat transfer coefficient will increase with the decrease of the channel width. That is, the heat transfer may be enhanced with small rectangular narrow channel. But when the channel width is less than a certain value, the model may not be proper anymore. So the further research should be necessary. As the applications of the present model, the critical heat flux and critical quality are calculated.
Assurance of Seismic Safety by Defense-in-Depth and Safety Goals: Lessons Learned From the Chuˆetsu-Oki Earthquake (2007)
Authors: Saji G.
Published: July 2009

Although the basic safety concerning control of reactivity, residual heat removal and confinement, was assured in the Kashiwazaki–Kariwa Nuclear Power Plants at the time of the Chuˆetsu-Oki Earthquake (2007), the accident caused great public concern as to the seismic safety of NPPs. The earthquake resulted in severe economic impacts, far exceeding the actually negligible environmental effect. The public is calling for a reassessment of the seismic safety of NPPs, as they are unable to understand the basic safety approaches of the Japanese seismic design practice. The earthquake significantly exceeded the design basis ground motion in all units. Due to this the seismic consequences, especially those malfunctions and damages in the lower seismic class items, are not surprising. The following three topics are highlighted from the lessons learned and the author’s reappraisal of the current seismic safety approach, namely: (1) Prevention of seismic consequences of the lower seismic class items, (2) Measures to ensure the seismic safety by including defense-in-depth, and (3) Reduction of seismic safety risks as low as reasonably achievable referring to ‘safety goals for seismic events.’ The author believes that ‘reduction of seismic risks as low as practical (S-ALAP)’ should be a new guiding principle for lower safety class items, in line with the new concept (gensai) developed in light of the Great Hanshin-Awaji Earthquake (1995), acknowledging difficulties of coping with earthquakes just by conservative design. For a reasonable reduction of seismic risks, it is necessary to answer to question of ‘how safe is safe enough.’ The author developed a safety goal for seismic consequences by integrating the International Nuclear Event Scale (INES) and Farmer’s probabilistic siting criteria. It is an extension of the author’s quantitative safety goals for non-seismic events already published in a series of previous papers (including RESS Vol. 80-2, pp. 143–161, 163–172; PSA’05-139985, 139989, 139990: ICONE14-89351).
Dynamics and Responses of a Single Flexible Block on a Vibrating Floor
Authors: Kim Dong-Ok , Choi Woo-Seok , Park Keun-Bae , Lee Won-Jae
Published: July 2009

Graphite blocks constituting the reactor core of a typical prismatic type HTGR form a group of graphite block columns. The seismic analysis of a HTGR core structure composed of stacked prismatic graphite blocks has been an important design issue. In a tall column of stacked graphite blocks, the flexibility of blocks can not be ignored in the dynamic analysis. This paper proposes a dynamic model of flexible block and presents response analysis results of a single flexible block on a vibrating floor, and compared them to the classical ones of with a rigid assumption.
Dynamic Stability of Pipes With Leakage
Authors: Torres Martin R.
Published: July 2009

Dynamic stability is of concern in submerged piping configurations, which experience leakage flow such as nuclear reactor internals and heat exchangers. Based on actual industry problems it was noticed that leakage flow in piping slip joints could incur damaging vibration amplitudes. Slip joints are often designed into piping systems to accommodate static loads or thermal expansion. A simple geometry consisting of two cantilever pipes joined at their free ends with a slip joint was studied experimentally. In the experimental set up, fluid was pumped through the pipes, which were submerged in a tank. The experiment showed that an instability due to leakage does indeed exist for this configuration.
Allowable Load Derivation of Single Angle Members for Small Bore Pipe Supports Welded on 4-Bolt CEA Plate in APR1400
Authors: Lee Joon-Ho , Choo Hee-Cheon , Bae Jae-Hwan
Published: July 2009

Single angle members have been rarely used as supporting structures in nuclear power plant because they are open sections which have significantly reduced capacities when considered in comparison to closed sections, and have weakness for twisting load such as local torsion caused by loading eccentricity of geometric center and shear center. However, in APR1400 (Advanced Power Reactor 1400 MW-class in Korea), the extended application of single angle members for supporting structures of small bore piping systems is considered to enhance the constructability and economics of plant. Furthermore, although it is general guideline for support design in APR1400 that supporting structures for equipment should be directly welded to embedded plates or steel structures in buildings as far as possible, in the case of small bore piping system, for the low level priority of construction in site, supporting steel structures for small bore piping could not be evitable to be welded onto the CEA (Concrete Expansion Anchor) plate. Per the ACI 318, ACI 349 and ACI 355.2, most CEA plate designs and anchor bolt load determinations are now based on finite element models that many applications have been individually made for CEA plates. If single angle members are attached onto these plates, integrated finite element models should be developed and analyzed in detail accroding to NRC IEB 79-02. Such a detailed analysis may appear to be excessive to small bore pipe supports which have diverse design materials and frequently subjected to field changes requiring rapid revision. Consideration should be given to reviewing current practices and reducing the level of effort being used for the integrated analysis of support and CEA plate by developing consensus standard regarding reasonable support and CEA plate designs. In this paper, allowable piping loads for single angle members such as L2×2×1/4, L2-1/2×2-1/2×1/4, L3×3×3/8, and L4×4×1/2 welded on the 4-bolt CEA assembly are derived and reviewed for general use for small bore pipe support design, and L2-1/2×2-1/2×1/4 and L3×3×3/8 welded onto 4-bolt (3/8″Φ sleeve type) CEA plate (1/2″×9″×9″) are recommended as standard small bore pipe supports with post-installed anchor system in APR1400.
Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (III)
Authors: Tamura Masataka , Makino Yoshinobu , Hino Takehisa , Kawano Shohei , Kouno Wataru , Matsunaga Keiji
Published: July 2009

Recently, stress corrosion cracking (SCC) has been observed at aged components of nuclear power plants under water environment and high exposure of radiation. Toshiba has been developing both an underwater laser welding directly onto surface of the aged components as maintenance and repair techniques. This paper reports underwater laser cladding and seal welding using INCONEL 52/52M.
Design and Test Plan of the Supercritical CO2 Compressor Test Loop
Authors: Ishizuka Takao , Muto Yasushi , Aritomi Masanori
Published: July 2009

Supercritical carbon dioxide (CO2) gas turbine systems can generate power at a high cycle thermal efficiency, even at modest temperatures of 500–550°C. That high thermal efficiency is attributed to a markedly reduced compressor work in the vicinity of critical point. In addition, the reaction between sodium (Na) and CO2 is milder than that between H2O and Na. Consequently, a more reliable and economically advantageous power generation system can be created by coupling with a Na-cooled fast breeder reactor. In a supercritical CO2 turbine system, a partial cooling cycle is employed to compensate a difference in heat capacity for the high-temperature — low-pressure side and low-temperature — high-pressure side of the recuperators to achieve high cycle thermal efficiency. In our previous work, a conceptual design of the system was produced for conditions of reactor thermal power of 600 MW, turbine inlet condition of 20 MPa/527°C, recuperators 1 and 2 effectiveness of 98%/95%, Intermediate Heat Exchanger (IHX) pressure loss of 8.65%, a turbine adiabatic efficiency of 93%, and a compressor adiabatic efficiency of 88%. Results revealed that high cycle thermal efficiency of 43% can be achieved. In this cycle, three different compressors, i.e., a low-pressure compressor, a high-pressure compressor, and a bypass compressor are included. In the compressor regime, the values of properties such as specific heat and density vary sharply and nonlinearly, dependent upon the pressure and temperature. Therefore, the influences of such property changes on compressor design should be clarified. To obtain experimental data for the compressor performance in the field near the critical point, a supercritical CO2 compressor test project was started at the Tokyo Institute of Technology on June 2007 with funding from MEXT, Japan. In this project, a small centrifugal CO2 compressor will be fabricated and tested. During fiscal year (FY) 2007, test loop components will be fabricated. During FY 2008, the test compressor will be fabricated and installed into the test loop. In FY 2009, tests will be conducted. This paper introduces the concept of a test loop and component designs for the cooler, heater, and control valves. A computer simulation program of static operation was developed based on detailed designs of components and a preliminary design of the compressor. The test operation regime is drawn for the test parameters.
Study on Turbulent Behavior of Water Jet in Supersonic Steam Injector
Authors: Fukuichi Akira , Abe Yutaka , Fujiwara Akiko , Kawamoto Yujiro , Iwaki Chikako , Narabayashi Tadashi , Mori Michitsugu , Ohmori Shuichi
Published: July 2009

One of the most interesting devices for Next-generation nuclear reactor systems aimed at simplified system, improvement of safety and credibility is a supersonic steam injector. Supersonic steam injector is a passive jet pump without rotating machine and high efficient heat exchanger because of the direct contact condensation between the supersonic steam and the subcooled water jet. It is considered that flow behavior in the supersonic steam injector is governed by the complicated turbulent flow with large shear stress induced by velocity difference between the steam and water and direct contact condensation. There are various studies about formulation of operating characteristic of the steam injector and jet structure in it [1][5]. According to these investigations, heat and momentum transfer of the steam injector are supposed to be related to the turbulent eddy diffusion. However, the studies about the turbulent flow under the large shear stress with direct contact condensation are not enough. Especially mechanisms of the momentum and the heat transfer are not clarified in detail. Objective of the present study is to investigate the turbulent behaviors of the water jet and the interface between the water and the supersonic steam that play an important role in the heat and the momentum transfer. Radial velocity distribution was measured by a special pitot tube. Fluctuation of the total pressure which doesn’t depnd on the density of the fluid is measured by the pitot measurement. It was found that the streamwise velocity increased as it approaches downstream of the injector. Fluctuation of the total pressure is large at the mixing region of the water and the steam. Visual measurement of the turbulent water jet was conducted by high speed camera in order to identify the position of unstable interface and its behavior. It was confirmed that wave propagated on the interface. And its velocity is obtained.
Environmental Fatigue Evaluation of a Korean Nuclear Power Plant
Authors: Oh Chang-Kyun , Kim Hyun-Su , Youm Hag-Ki , Jin Tae-Eun , Kim Young-Jin
Published: July 2009

In accordance with the recommendation of USNRC and the U.S. license renewal experiences, the effects of reactor coolant environment on the fatigue life have to be considered for the continued operation of operating nuclear power plants as well as for the design of new plants. Although various evaluation methodologies have been suggested to date, a wide range of comparison of the existing methodologies has not been performed. The purpose of this paper is to evaluate the environmental effects on the fatigue life of a reactor pressure vessel and ASME Class 1 piping by the various methodologies and to investigate the effects of the pressure and moment stress histories on the environmental fatigue evaluation. The evaluation results show that the environmental fatigue evaluation results based on the design cumulative usage factors for the reactor pressure vessel and ASME Class 1 piping satisfy the requirement of the ASME code except for charging nozzle. However, when using operating cumulative usage factor, the environmental fatigue evaluation result for the charging nozzle satisfies the ASME code allowable. And the effects of the pressure and moment stress histories on the environmental fatigue evaluation are considered to be small when using the modified rate approach.
Steam Generator Blockage: A Thermal-Hydraulic Approach Based on CATHARE 2
Authors: Vergnault Alexis , Dorel Robin , Goux Fre´de´ric
Published: July 2009

In some French steam generators (SG) with broached tube support plates (TSP), the TSP flow paths have become blocked by deposits. Due to these blocked broached TSP, flow paths decrease the flow area and result in increased void fraction in the tube bundle region and pressure drop across the TSP. As a consequence, it also leads to a decrease of the recirculation ratio which can be observed through the SG level measurement. The major impacts of this phenomenon can be classified in three main categories: the mechanical impact on the TSP due to an increase of the pressure drop, the impact on the safety demonstration itself due to the less secondary side inventory that have to be considered for the thermal-hydraulic studies and finally the impact on the normal transients mainly due to the level oscillation encountered in the steam generator. This paper which focuses on the last item presents the Insitute for Radiological Protection and Nuclear Safety (IRSN) approach to calculate with CATHARE 2 [1] code the response of steam generators experiencing the blockage: both are considered transients with only a “stand alone” steam generator and transients performed with our SImulator for Post-Accident so called “SIPA” [2 to 4] which enables to simulate normal and accident scenarios. Sensitivity calculations on the blockage ratio, on the blockage distribution and on the power level are made to be able to assess this phenomenon. Some conclusions are arrived at the power levels that should be adopted to deal with blockage in order to restore the water level stability.
Development of General-Purpose Software to Analyze the Static Thermal Characteristic of Nuclear Power Generation System
Authors: Nakao Yoshinobu , Koda Eiichi , Takahashi Toru
Published: July 2009

We have developed a general-purpose software by which static thermal characteristics of a power generation system can be analyzed easily. This software has following notable features. - It has new algorithm that can solve non-linear simultaneous equations used in analyzing static thermal characteristics of various power generation systems, namely the heat and mass balance and the efficiency. - It has flexibility over setting calculation conditions. - It can be executed on a personal computer easily and quickly. We have ensured that it can produce the heat and mass balance diagrams of a nuclear power plant’s main steam system and then it can calculate power output and efficiency of the system. Furthermore, we have evaluated various measures to recover heat from steam generator blowdown water and proved that this software is a useful operation aid for planning effective changes in support of power stretch.
Laser Joining of Ceramics: A Contribution to High Temperature Range Application of Ceramic Components
Authors: Lippmann Wolfgang , Herrmann Marion , Hille Carmen , Hurtado Antonio
Published: July 2009

Non-oxide ceramics, such as silicon carbide (SiC) and silicon nitride (Si3N4), have excellent properties that make the materials interesting for application also in the nuclear sector. Due to their exceptional resistance to high temperatures, aggressive and abrasive media as well as nuclear radiation, the materials seem to be particularly suitable for developments in such fields as high-temperature reactors ((V)HTR) and peripheral systems (e.g. for hydrogen production). To simplify and thus to enable the technical application of these high-tech ceramics, the Dresden University of Technology has developed a laser beam joining process. This opens up many possibilities, e.g., to encase HTR fuel elements (as well as spheres and composites) in SiC, to encapsulate highly radioactive waste in SiC or to build a highly efficient heat transformer using high-temperature energy from VHT reactors. The progress made in laser beam technology in the last few years is a major element that has contributed to the developments achieved to date. Research has been focused mainly on the following three areas: (1) optimization of the laser parameters in combination with the advancement of oxide brazing fillers, (2) transfer of the basic technology to other high-tech ceramics like oxide ceramics, and (3) application of the laser process to develop electrically conductive joints. The possibility to laser join also Al2O3 and ZrO2 ceramics has created the opportunity to produce full ceramic sensors for (V)HTR specific applications at low cost. This requires adaptation of laser technology to the special properties of oxide ceramics. These are markedly less resistant to thermally induced stress than non-oxide ceramics, placing high requirements on laser process control. Another peculiarity is the property of oxide ceramics to be partly transparent to the laser wavelengths emitted by diode lasers (808 nm and 940 nm), with the result that the ceramic material is not heated primarily at the surface but inside its volume. This produces joint seams inside ceramic components even without any excessive thermal stress. The R&D work has made it possible to produce novel sensors for the high-temperature range that are also highly resistant to aggressive media. It is considered a further advantage that this joining technology has no special requirements regarding the process atmosphere such as vacuum or inert gas, which ensures that the process lends itself well to automation.
Modeling Approach and Predictions of Boron Mixing in Core Bypass of ESBWR
Authors: Boyd Christopher F.
Published: July 2009

Computational fluid dynamics (CFD) is used to predict the mixing and transport of a Boric acid solution in the core bypass region of the General Electric (GEH) Economic Simplified Boiling Water Reactor (ESBWR) during an anticipated transient without scram (ATWS) scenario. Transient boundary conditions are synthesized from a set of system code predictions. Full-scale CFD predictions are completed using a representative geometry. Symmetry is used to reduce the domain to 1/8th of the core bypass. Forty-five million finite volume cells are used to define the computational domain inside the core barrel which is made up of the small passages between the fuel channels and an outer peripheral open space. The size of the model posed special challenges in the FLUENT CFD model development and execution. The solution required a large number of iterations with reduced solver under-relaxation factors to ensure stable convergence. Steady-state predictions were completed to quantify the sensitivity of the results to variations in selected boundary conditions. These results show good transport of the Boron into the inner regions of the core bypass with no significant sensitivity of the result to changes in the distribution of flow from the top surface. The results also showed no significant impact to changing the specific gravity of the incoming jets from 1.05 to 1.0. Only a significant change in the leakage rate from the bypass and/or a change in the mass flow rate of injected Boron solution were able to impact the distribution and quantity of the Boron in the core. These results are in line with expectations. The predictions indicate efficient penetration of the Boron into the interior regions of the core bypass during an ATWS.
Advanced Test Reactor: A National Scientific User Facility
Authors: Stanley Clifford J. , Marshall Frances M.
Published: July 2009

This presentation and associated paper provides an overview of the research and irradiation capabilities of the Advanced Test Reactor (ATR) located at the U.S. Department of Energy Idaho National Laboratory (INL). The ATR which has been designated by DOE as a National Scientific User Facility (NSUF) is operated by Battelle Energy Alliance, LLC. This paper will describe the ATR and discuss the research opportunities for university (faculty and students) and industry researchers to use this unique facility for nuclear fuels and materials experiments in support of advanced reactor development and life extension issues for currently operating nuclear reactors. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration (Fig. 1) of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and chemistry can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ∼1.0 x1015 n/cm2-sec with a maximum fast flux of ∼5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.
Experimental Investigation on the Heat Transfer Characteristics of a Minichannel in Transition Flow
Authors: Masson Viviana , Silin Nicola´s , Azcona Alejandra , Delmastro Dari´o , Garci´a Juan Carlos , Mateos Daniel
Published: July 2009

In MTR reactors the fuel elements are usually of flat plate-type and the cooling water flows through gaps of 2 to 4 mm. Therefore, correlations developed for parallel flat plates are usually used for their study. Nevertheless, in the hydrodynamic regime of transition, the flow is not stable and, under high heat fluxes, there are no reliable correlations that can be applied in this geometry. The aim of the present work is to determine experimentally the heat transfer coefficient for water in this geometry in the rank of 3000< Re <10000 and for relatively high heat fluxes (between 24 W/cm2 and 32 W/cm2). The experimental setup consists of a 200 liters water tank, pressurized to 1.7 atm, which discharges its content through the vertical test section. The flow is downwards and it is controlled by a valve located downstream. The flow rate is measured with a orifice plate connected to a DPcell and the pressure with a JUMO pressure digital sensor. The test section is a rectangular channel constructed with two 62 cm long, 6 cm wide and 6mm thickness aluminum plates, separated with Teflon strips which set the gap for the flow passage. Gaps or plate separations of 2.7 mm and 3.8 mm were studied. The aluminum plates were electrically heated from the rear using a Bruker BMN 70/700 direct current power supply. The tests were made at 18 kW and 24 kW. The wall temperatures were measured with K-type thermocouples placed in different axial positions. The temperature measurements, the DPcell signal and the pressure signal were acquired with a digital card connected to a PC. Two different inlet flow temperatures were considered 10°C and 38°C. With this arrangement, measurements of wall temperatures evolutions were obtained and the local coefficient of convection h(z) was calculated. Comparisons with turbulent correlations for flat plates indicate that the measured temperatures of wall are greater than expected. It was also observed that there are zones where subcooled boiling is reached. There are some particular aspects that could affect the reliability of correlations for flows between flat plates. The first phenomenon is the change of viscosity close to the wall due to the very high heat fluxes. There may also appear buoyancy effects, though we believe they have a minor importance. And finally the long thermal and hydrodynamic development distances, which may delay the appearance of turbulence in the channel.
Prediction of Boiling Water Reactor Assembly Void Distributions Using a Two-Phase Computational Fluid Dynamics Model
Authors: Pointer W. David , Tentner Adrian , Sofu Tanju , Lo Simon , Splawski Andrew
Published: July 2009

This paper presents recent results obtained as part of the on-going integral validation of an advanced Eulerian-Eulerian two-phase (E2P) computational fluid dynamics based boiling model that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel assembly. The code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. Simulations of a prototypic BWR fuel assembly experiment have been completed as an initial assessment of the applicability of the E2P model to realistic BWR geometries and conditions. Initial validation has focused on comparison with measured sub-channel averaged data to enable the benchmarking of the accuracy of the E2P against the current predictive capabilities of the sub-channel methods. The paper will discuss the effects of modeling assumptions, assumed coefficient values and the computational mesh structure used to describe the fuel assembly geometry on the accuracy of the sub-channel averaged void fraction.
Transient Forced Convection Heat Transfer Due to Exponentially Increasing Heat Input for Helium Gas Flowing on a Narrow Plate
Authors: Shibahara Makoto , Liu Qiusheng , Fukuda Katsuya
Published: July 2009

Steady and transient forced convection transient heat transfer due to exponentially increasing heat input to a heater is important as a database for safety assessment of the transient heat transfer process not only in a high temperature gas cooled reactor (HTGR) due to an accident in excess reactivity but also in high heat flux gas cooling devices such as a gas turbine and a rocket engine. In this research, forced convection transient heat transfer for helium gas at various periods of exponential increase of heat input (Q0exp(t/τ)) to a horizontal narrow plate was numerically solved based on a turbulent flow model. The platinum plate with a length of 50 mm was used as test heater. The velocities ranged from 4 to 10 m/s, the gas temperatures ranged from 313 to 353 K, and the periods of heat generation rate, τ, ranged from 46 ms to 8.6 s. The values of numerical solutions for surface temperature and heat flux were compared and discussed with authors’ experimental values. It was obtained that the surface temperature difference and heat flux increase exponentially as the heat generation rate increases with the exponential function. Then the temperature within the boundary layer also increases with the increase of the surface temperature. It is understood that the gradient of the temperature distribution near the wall of the plate is higher at a higher surface temperature difference. The values of numerical solutions for surface temperature and heat flux at the velocity of 6 m/s agree well with the experimental data, though they show some differences at other velocities. And also, heat transfer coefficients at the velocity of 6 m/s agree well with the experimental data, though they show some differences at other velocities. They agree within 15% at various periods and velocities.
Link Between Operational Experience Data and Pre-Accidental Data
Authors: Matahri Naoe¨lle
Published: July 2009

RECUPERARE method has been developed for operating feedback analysis and has been built on the French Human Reliability Analysis (HRA) principles. It is used to study the causes of human errors or technical failures occurred in French PWRs and the recovery process of events. Based on an event classification (6 categories) model according to the nature of the link between failure and recovery, the identified and recorded data are: • the causes of the defects (technical, human, organizational) and the context in which they appear; • the factors of the recovery performance (depending on technical and organizational aspects); • a chronological analysis, designed to collect delays between failures and their detection/recovery for each event. About 4500 events reported in French PWRs (1997–2006) have been reviewed through this model. Initially, the weight of factors and the most important factors, which influenced the detection and recovery delay, are defined. For this purpose, the regression Partial Least Square (PLS) is used. Then, to link RECUPERARE results with pre-accidentals data, conditional probabilities of events linked between them by a cause and effect relationship are calculated. For this, the Bayesian method with a Bayesian network is built with the PLS obtained results and applied. This constitutes a first approach to take into account the human and organizational factors in HRA highlighted by operating feedback.
Study for the Space Craft for Interplanetary Cruise: (3) Examination of the Thermal Efficiency
Authors: Yoshida Yutaka , Honma Yuji , Narabayashi Tadashi , Shimazu Yoichiro
Published: July 2009

In order to explore the deep space, such as Mars, Jupiter, Saturn, etc in the future, a spacecraft that will be driven by nuclear power should be developed [1]. At present, satellites or space probes have been using mainly electric source of chemical battery, fuel battery, solar battery, and RI battery. However, considering highly developed and extensive space exploration in the future, it is obvious that larger electric power is required over the long term space travel more than several years. Additionally, the solar battery used in space will be fundamentally impossible to use in planetary exploration father away form Mars because sunlight is attenuated. Therefore, lager electric power source must be installed in the space craft. In this study, we consider about co-generation system for heat and electricity using nuclear power. We think that the nuclear power is appropriate for using in deep space because of a long time operation without refueling and possibility in downsizing due to higher power density. We selected the fast reactor system of about 18 MWth compared with other type of reactors, such as PWR and high temperature gas reactor [2]. With regard to a power generation system, we examined about efficiency of Stirling engine compared with a gas-turbine engine. Theoretical efficiency of Stirling engine [3] is much higher than that of gas-turbine engine. Therefore, we selected Stirling engine and we have started the model test of a Stirling engine. Total power generation at International Space Station (ISS) that has been built since 1998 is about 110kWe. We estimated that about 5times as much electricity as that of ISS is enough to explore or develop the space. In that case, 2.5MWe will be generated by the system, number of crews will be about 10 and 2MW will be used to electric propulsion.
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: 1 — Introduction of the Project and Development of Structural Mechanics Module
Authors: Koshizuka Seiichi , Liu Jie , Shirakawa Noriyuki , Uehara Yasushi , Naitoh Masanori , Yamamoto Yuichi
Published: July 2009

Sponsored by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan, a computer code, named COMPASS, has been developed employing the Moving Particle Semi-implicit (MPS) method for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In this and the following five papers [1–5], the outcomes of the project in FY2006 are presented. In FY2006, development of the basic functions of the COMPASS was completed and fundamental verification calculations were carried out. In this paper, the outline of the project and a part of COMPASS, the structural mechanics part, are described, including formulation of MPS method and the conceptual design of COMPASS, and the results of the basic verification calculations for structural mechanics module.