bluuebunny/crossref_metadata_2025 huggingface dataset:
An Experimental Study on Air-Water Two-Phase Flow in Pebble Beds
Authors: Bai Bofeng , Liu Maolong , Zhang Xiaojie
Published: February 2010

An experimental study was conducted on the air-water two-phase flow patterns and pressure drop in the bed of rectangular cross section containing spheres of regular distribution. Three kinds of glass spheres with different diameters (3mm, 6mm and 8mm) were used for the establishment of the test section. By means of visual observations of the two-phase flow through the test section, it was discovered that five different flow patterns occur in the experimental parameter ranges, namely bubbly flow, bubbly-slug flow, slug flow, slug-annular flow and annular flow. A correlation for bubble and slug diameter in packed beds was proposed, which is an extension of the Tung/Dhir model, Jamialahmadi’s model and Schmidt’s model. Three correlations were proposed to calculate the void friction of flow regime transition in bubble flow, slug flow and annular flow based on the bubble model in the pore region. The experimental result shows that the modified Tung and Dhir’s model of flow pattern transition is a better agreement with the experimental data compared with Tung and Dhir’s model.
Dynamics Integrated Valve Analysis for Gas Turbine System
Authors: Kang Seong G. , Suh Kune Y.
Published: February 2010

A methodology and theory are developed to precisely determine the gas mass flow rate for the design of a gas turbine system manufactured for the operating conditions of nuclear and fossil power plants. From the heat exchanger or boiler to the first assembly of turbine blades, the gas passes by a stop valve, control valves and first nozzle, each device of which is connected with piping. The corresponding gas flow rate can be computed when the thermal and hydraulic conditions are defined at the stop valve, control valve and pipes. Such thermophysical properties as pressure, temperature, enthalpy and velocity specified at the inlet of each device are changed at the outlet of that device due to its structural characteristics. DIVA (Dynamics Integrated Valve Analysis) is written to predict the gas mass flow rate with reference to the Widows’ Creek type control valve.
The Overview of Natural Circulation Characteristics of a Marine Reactor Under Ocean Condition
Authors: Zhong Jia , Yang Xingtuan , Jiang Shengyao
Published: February 2010

Marine reactor would be influenced by ocean conditions such as rolling, inclination and heaving which made system coolant fluctuate and changed the ability of natural circulation of the system. The paper summarizes the study of effects of ocean conditions upon flow characteristics of natural circulation, including the researches of Japan, South Korea and China. The result shows that the ocean conditions have some effects upon natural circulation, and the structure of the loop would affect the experiment result. Generally, Inclination would reduce the ability of natural circulation, the larger the inclination angle, the more decrease the ability of natural circulation. Rolling would cause core flow rate oscillation and the amplitude of oscillation increases as rolling angle increase and rolling period decrease. Heat transfer in the core is enhanced by the rolling motion and the enhancement is thought to be caused by the internal flow due to rolling motion. The flow oscillation caused by rolling motion would overlap the density wave oscillation which made the system more instable. Heaving would also cause core flow rate oscillation and the amplitude of oscillation increases as heaving acceleration increase and heaving period increase. It is also found that the improved RETRAN procedure can be used to simulate the thermal-hydraulic characteristics of marine reactor under ocean conditions. The parameters of the current experiment have some differences with the actual condition of the marine reactor, so some further research should be done.
Multidimensional Analysis of UPTF Upper Plenum Injection Test With MARS-KS Code
Authors: Lee Kyung Won , Cho Yong Jin , Ahn Seung Hoon
Published: February 2010

This study aims to assess the multi-dimensional fluid model of MARS-KS (KINS standard version of MARS) code. The UPTF upper plenum injection test (UPI Test, Test No. 20) is modeled with the MARS in order to assess the code capability for predicting the multi-dimensional behavior of UPI water in the upper plenum during a reflood phase of large break loss-of-coolant accident. The coolant piping is simulated by one-dimensional components, while the reactor vessel is modeled by the cylindrical multi-dimensional component with five rings in radial direction, eight sectors in azimuthal direction, and fourteen axial levels. The results of MARS calculation are compared with the experimental data of UPTF test, particularly the ECC downflow area in the core flow area, the water carryover rate to hot legs, and the water level distribution in upper plenum. The calculation results indicate that the size of the downflow area is about 16.7% of the core flow area, comparable to the result of 10% in UPTF test. The MARS overestimates the amount of water carryover to the hot legs. As seen in the UPTF UPI test, the collapsed water level formed in the upper plenum is uniform at the entire core flow area, except at the down flow region where more water accumulates. These results demonstrate that MARS can provide a reasonable prediction of multi-dimensional behavior of UPI water in the upper plenum.
Experiences With Implementation, Application and Improvement of an Integrated Management System
Authors: Berg H. P. , Beckmerhagen I. A.
Published: February 2010

An integrated management system encompasses all management and assessment activities. The organizational structure of the Federal Office for Radiation Protection (BfS) quality management system is able to merge existing and future requirements. In this context the BfS management system encompasses quality management and radiation protection requirements as well as safety management, mining requirements and those of testing and calibration laboratories. Its structure allows the integration of function-specific management systems as well as specific requirements. The quality management system has taken into account the existing structure of documents to improve acceptability for the extended approach. Integrating management systems should enhance managerial and operational effectiveness. Whether or not this goal can or has been achieved is still a lingering question. A well-proven method to assess the effectiveness of any management system is auditing. Audits are designed to determine adequate compliance with applicable standards and guidelines. Moreover, they can and should be instrumental in identifying problem areas and potential improvements, along with corrective and preventive actions.
An Application of Main Steam Flow-Based Power Monitoring to OPR1000 Plants
Authors: Jung Byung Ryul , Song In Ho , Kim Jung Hoon , Lee Byung Jin , Baik Se Jin , Kim Joon Sung , Lee Jong Sun , Choe Teuk Ki
Published: February 2010

It is extremely important in managing nuclear power plants to accurately confirm the thermal power of reactor for controlling and monitoring the plant power output. For pressurized water reactors (PWRs), feedwater flowrates have been typically used for mass and energy balance calculations to confirm the reactor thermal power. The feedwater flowrates are calculated from pressure drop measurements across venturis in plant feedwater systems. There have been so far many reports on the fouling of venturi, which has lead to erroneous high pressure drop measurements. This gives high mass flowrates, leading to high calculated reactor thermal power. The high calculated thermal power causes plant operators to reduce the power output to comply with the licensed operating power level. This causes losses in plant electrical output, typically about 2 ∼ 3% of the plant rating. One method to overcome the overestimation of reactor power due to high calculated feedwater flowrates due to venturi fouling is using steam flow measurements in reactor thermal power calculations. Plant operating experiences have shown that steam flows are a closer indicator of plant power. In addition they are known to show little fouling problem. A new power calculation methodology based on steam flow measurements is proposed herein for mass and energy balance calculations of reactor thermal power. A mathematical mass flowrate equation is developed for measured steam flows in a similar way to feedwater flows. The steam flow correction factors are determined from the operating data gained at the beginning of operation cycle. This is a way to calibrating steam flow measurements to feedwater flow indications to allow for the calorimetric heat balance to be based on steam flow rather than feedwater flow. Routine calibration testing procedure is developed to determine discharge coefficients for each operation cycle. The results of application to the Optimized Power Reactor 1000 (OPR1000) plants show good performance in plant power monitoring.
Agents and Processes Design for TRU Elements Back Extraction in TRPO Process
Authors: Wei Meng , Liu Xuegang , Chen Jing
Published: February 2010

To reduce the long-term risk of the high-level liquid waste (HLLW) and the waste disposal cost, transuranium (TRU) elements should be removed from HLLW. A so-called TRPO process has been developed by Chinese scientists to partition HLLW. In this process, the extractant, trialkyl phosphine oxide (TRPO), is able to extract TRU elements into organic phase completely, which makes the treatment and disposal of raffinate HLLW much easier. However, the treatment of extracted TRU elements in organic phase, in return, becomes new troublesome issue. Generally, there are three promising ways to treat the extracted TRU elements: (1)transmutation; (2)conditioning; (3)recycling U+Pu in Purex-TRPO Integrated Process. In any of the three ways, the back extraction agents and processes play significant roles. In this paper, the investigations on back extraction agents for TRU elements, such as TTHA, DTPA, AHA, HEDPA, DOGA, and carbonates are introduced. The corresponding back extraction processes and experimental results are reviewed.
Safe-End Replacement on Pressurizer Upper Nozzles
Authors: Pitoiset Xavier , Dellopoulos Georges , Weverbergh Jacques , Che´rasse Jean-Marc , Flaherty John , Evans Paul
Published: February 2010

Following the occurrence of cracking in various nozzle Dissimilar Metal (DM) butt welds in different PWR plants around the world (VC Summer, Ringhals, Tsugura and more recently Wolf Creek), an integrated approach to develop a material degradation management program was adopted for the Belgian PWR Units. Electrabel, together with Tractebel Engineering and Vinc¸otte have qualified enhanced nondestructive examination procedures and extended in-service inspections. Over the last decade, a very effective monitoring of the potential cracking in the pressurizer nozzles to safe-ends DM welds has been conducted. Electrabel and Tractebel Engineering are now initiating a mitigation process. Within this frame, Westinghouse Electric Company has developed a replacement solution of the pressurizer (PRZ) upper nozzle safe-ends of the 4″ and 6″ lines with implementations at Doel Units 3 & 4 and Tihange Units 2 & 3. This paper outlines the development of the replacement solution. It provides an overview of the replacement process itself. It outlines the comprehensive qualification program developed to achieve approval for implementation.
Degradation of Aged Plants by Corrosion: Radiation-Induced Corrosion Cells Inducing “Long-Cell” Action
Authors: Saji Genn
Published: February 2010

In the previous papers, the author has established various ‘long cell’ corrosion configurations that should exist in nuclear power plants. With these corrosion mechanisms in place, the plant can be characterized as an assembly of gigantic short-circuited electrical batteries, inducing electrochemical corrosion at localized anodic sites. If these corrosion cells are involved at nuclear power plants, macroscopic electrochemical potential differences must be demonstrated between anodic sites where dissolution of metal (i.e. corrosion) is taking place and cathodic sites where deposition (also called sedimentation) of corrosion products are often observed. Among these, the radiation-induced corrosion cell is an important mechanism of corrosion issues among nuclear power plants, since it plays a major role in the corrosion problems found in primary water, including PWSCC and AOA in PWRs and IGSCC in the BWRs. There is numerous experimental evidence indicating a potential difference induced by radiation, however, the exact mechanism of such phenomena has not been investigated from the ‘long cell action’ corrosion hypothesis point of view. The author investigated the basic mechanism by combining radiation chemistry, electrochemistry and corrosion science to confirm the existence of radiation-induced ‘long-cell’ action (macro) corrosion cell. By performing a competition kinetic study, which is a simplified approach to determine which of several competing reactions will predominate, the hydrated electrons, e−aq, reacting mainly with stable molecules, are found responsible for inducing a large portion of the potential difference both in the PWR and BWR water chemistry environment. The hydrated electrons react with a cathodic half-cell included in the stable solutes thereby inducing redox reactions in the mixed cell configuration with both reducing and oxidizing actions. This method reproduces the reported experimentally observed ECP variation to a certain extent (observed in the INCA Test Loop in Sweden and NRI-Rez BWR-2 Loop in Czech Republic) which was measured by widely changing the solute concentrations, such as dissolved hydrogen and oxygen. The author believes the results support the assumed major reactions acting in the redox process.
Proposed Experiment for Validation of CFD Methods for Advanced SFR Design: Upper Plenum Thermal Striping and Stratification
Authors: Pointer W. David , Lomperski Stephen , Fischer Paul , Obabko Aleksandr
Published: February 2010

In response to the goals outlined by the U.S. Department of Energy’s Advanced Fuel Cycle Initiative, an effort is underway to develop an integrated multi-physics, multi-resolution thermal-hydraulic simulation tool package for the evaluation of nuclear power plant design and safety. As part of this effort, initial guidance has been proposed for the development of experiments to supply validation data sets for the CFD-based thermo-fluid simulation capability. To demonstrate that the proposed data requirements can be achieved using current generation measurement methods and to refine correlation and data comparison methods suitable for very large data sets, an initial experiment focused on turbulent mixing in the upper plenum of an advanced sodium fast reactor has been proposed. Prior validation efforts to support the use of one-dimensional lumped parameter models in the analysis of reactor safety performance relied primarily on data from carefully scaled integral system experiments to validate and tune correlations used to represent the physics associated with a particular transient in a particular reactor design. Unlike the correlation-based lumped parameter codes, computational fluid dynamics simulations reduce the reliance on experimentally derived correlations to the prediction of local turbulence effects rather the prediction of integral quantities like pressure drop and heat transfer coefficients. As a consequence, simpler separate effects experiments, which capture the turbulence effects but not necessarily the integral effects within a specific component of a system, can be utilized as the primary validation basis for the CFD codes. However, while the need for large carefully scaled integral experiments is reduced, the high spatial and temporal resolution of these codes requires that experimental data be collected at fine spatial and temporal resolutions. An initial series of simulations has been completed to support the development of the proposed experimental facility using air as a surrogate for the sodium coolant. Design options considered in RANS simulations using the commercial CFD code Star-CCM+ include mixing facility dimensions, the number of inlet jets to be included and outlet position. The use of RANS simulations is supported by an initial benchmarking comparison with predictions from the spectral element large eddy simulation code Nek5000 for the nominal experimental geometry.
Numerical Simulation and Experimental Study on Resonant Acoustic Chambers: For Novel, High-Efficiency Nuclear Particle Detectors
Authors: Wang Jing , Archambault Brian , Xu Yiban , Taleyarkhan Rusi P.
Published: February 2010

Acoustic-structural-electromagnetic coupled models have been successfully set up for Resonant Acoustic Chambers (RACs) which have important applications in radiation detection, sonoluminescence and sonofusion; the goal being to simulate transient acoustically driven metastable states and structural responses so that the designs of RACs can be optimized for advanced applications. The simulation predictions have been benchmarked with experimental data in two designs of RACs, Open Chamber System (OCS) and Closed Chamber System (CCS). A framework was developed for benchmarking and validating the predicted resonant frequency and oscillatory pressure mapping profiles with and without scattering centers. Experiments were conducted with and without external neutron-induced cavitation bubble clusters. Comparison of measurements versus experimental data demonstrated the applicability of the modeling-cum-simulation framework. Studies have provided insights into the significant and complex influences of fluid-structure-electromagnetic coupling and on the influence of scattering center inclusions on the system’s acoustic responses. The framework appears reasonable for design of advanced, high-powered RACs; however, significant technical challenges remain for capturing the overall system performance upon evolution and transport of transient bubble clusters.
Behavior of a Sequence of Geometric Transformations for a Truncated Ellipsoid Geometry in Transport Theory
Authors: Panta Pazos Rube´n
Published: February 2010

The neutron transport equation has been studied from different approaches, in order to solve different situations. The number of methods and computational techniques has increased recently. In this work we present the behavior of a sequence of geometric transformations evolving different transport problems in order to obtain solve a transport problem in a truncated ellipsoid geometry and subject to known boundary conditions. This scheme was depicted in 8, but now is solved for the different steps. First, it is considered a rectangle domain that consists of three regions, source, void and shield regions 5. Horseshoe domain: for that it is used the complex function: f:D→C,definedasf(z)=12ez+1ezwhereD=z∈C−0.5≤Re(z)≤0.5,−12π≤Im(z)≤12π(0.1) The geometry obtained is such that the source is at the focus of an ellipse, and the target coincides with the other focus. The boundary conditions are reflective in the left boundary and vacuum in the right boundary. Indeed, if the eccentricity is a number between 0,95 and 0,99, the distance between the source and the target ranges from 20 to 100 length units. The rotation around the symmetry axis of the horseshoe domain generates a truncated ellipsoid, such that a focus coincides with the source. In this work it is analyzed the flux in each step, giving numerical results obtained in a computer algebraic system. Applications: in nuclear medicine and others.
Effect of Nanofluid on the Film Boiling Behavior at Vapor Film Collapse
Authors: Arai Takahiro , Furuya Masahiro
Published: February 2010

A high-temperature stainless-steel sphere was immersed into Al2O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto a surface of the sphere. A direct contact between the immersed sphere and Al2O3 nanofluids is quantified by the acquired electric conductivity. The Al2O3 nanofluid concentration is varied from 0.024 to 1.3 vol%. A film boiling heat transfer rate of Al2O3 nanofluid is almost the same or slightly lower than that of water. A quenching temperature rises slightly with increased the Al2O3 nanofluid concentrations. In both water and Al2O3 nanofluid, the direct contact signals between the sphere and coolant were not detected before vapor film collapse.
Release of Materials During the Decommissioning of the Belgonucleaire Dessel MOX Fuel Fabrication Plant
Authors: Cuchet Jean-Marie , Vandergheynst Alain , Libon Henri
Published: February 2010

BELGONUCLEAIRE has been operating the Dessel plant from the mid-80’s at industrial scale. In this period, over 35 metric tons of plutonium (HM) has been processed into almost 100 reloads of MOX fuel for commercial west-european light water reactors. In late 2005, the decision was made to stop the production because of the shortage of the MOX fuel market remaining accessible to BELGONUCLEAIRE. As a significant part of the decommissioning project of this Dessel plant, about 170 medium-sized glove boxes and about 1.300 metric tons of structure and equipment outside the glove boxes are planned for dismantling. The dismantling works are expected to start in the second quarter of 2009. On account of stringent internal rules of alpha-containment during over 25 years of operation, there is no significant contamination of the plant, outside the glove boxes; that assumption has been confirmed by radiological surveys performed by independent bodies in 2001 and 2008. Therefore most of the materials outside the glove boxes that were not a priori destined for radioactive waste will be released without restriction on the basis of the applicable legal regulations in Belgium (ARBIS), along with the buildings and the plant site. In this paper, after having reviewed the different regulations in Belgium, the authors introduce the different options considered for release of materials, and the main decision criteria (process, safety aspects, radiological, etc) for the different expected types of materials (inert materials, metals, plastics, electrical cabinets and cables and electronics) are analysed. Besides the regulatory aspects, the technological and economical aspects are considered (as an example, comprehensive metal smelting is implemented, as a favourite solution because it provides with decontamination, homogeneization and volume characterization).
Recent Design Technologies of Moisture Separator Reheater
Authors: Manabe Jun , Kasahara Jiro , Kojima Toshiki , Fujita Issaku
Published: February 2010

This paper introduces the development of the current model Moisture Separator Reheater (MSR) for nuclear power plant (NPP) turbines, commercially placed in service in the period 1984–1997, focusing on the mist separation performance of the MSR along with drainage from heat exchanger tubes. A method of predicting the mist separation performance was devised first based on the observation of mist separation behaviors under an air-water test, then developed for the application to predict under the steam conditions, followed by the verification in comparison with the actual results of a steam condition test. The instability of tube drainage associated with both sub-cooling and temperature oscillation, which may adversely affect the seal welding of tubes to tube sheet owing to thermal fatigue, was measured on an existing unit both to clarify the behaviors and to develop a method to suppress them. Both methods were applied to current model MSR and the effectiveness of the methods was demonstrated. A new concept MSR for 1,700 MW class APWR units is put in perspective based on the technologies, alongside a multidisciplinary optimum design evaluating the heat exchanger tube bundle.
The Application of Noise Analysis to Find the Miss-Connected LPRM Cables Before Plant Startup
Authors: Kubota Osamu , Maki Akira , Uegata Tomomichi
Published: February 2010

The miss-connected LPRM (Local Power Range Monitor) cables were occasionally found at the BWRs (Boiling Water Reactors) in Japan. These miss-connections were confirmed by visual inspection after the plant shutdown. In 2007, we proposed a method to find the miss-connection by applying noise analysis to LPRM signal data during plant operation. Actually, in 2008, the miss-connection was found by the method during plant operation at a BWR in Japan. It showed the effectiveness of the method. However, it should be more convenient and helpful, if the miss-connection could be found before the plant startup. A new method is proposed here to find the miss-connections before plant startup, that is, before the primary containment vessel is closed after plant startup preparation.
Development of Repair Technologies for J-Groove Welds of Drain and Instrument Nozzle in Steam Generator
Authors: Park Kwang Soo , Kong Chang Sig , Lee Seon Ho , Kim Woo Sung
Published: February 2010

SG drain & instrument nozzles and their welds fabricated with alloy 600 and alloy 82/182 is susceptible to Primary Water Stress Corrosion Cracking (PWSCC). In Korea, the cracks due to PWSCC were discovered in the drain nozzle of Yongkwang units 3 & 4. Doosan has developed a system for steam generator to repair damaged drain nozzle & welds and to prevent further damage on the instrument nozzle & welds. The repair system consists of machining, welding equipment and installation tool for this equipment. The machining equipment is used to remove the nozzle and J-groove welds. The process is called mechanical machining and the main equipment is installed on steam generator’s outer wall. The welding equipment is designed for the machined J-groove welds and overlay. The auto welding equipment consists of welding head, controller, monitoring tool and Gas Tungsten Arc Welding (GTAW) power supply. Doosan has developed remote welding process using the monitoring tool. The installation tool consists of automatic installment tool for instrument nozzle and manual installment tool for drain nozzle. Doosan successfully completed a mockup test and field application for Yongkwang unit 3.
The Study for Installation of the Autocatalytic Flammability Control System in New ABWR Plants
Authors: Suzuki Haruki , Takii Taichi , Oda Shingo
Published: February 2010

As a measure to prevent the generation of combustible hydrogen/oxygen gases expected in an ABWR accident, a flammability control system (FCS) needs to be installed. The autocatalytic FCS, which has been introduced mainly in Europe and the US, is composed of multiple passive autocatalytic recombiners (PARs) and is capable of recombining hydrogen and oxygen at room temperature (low activation energy) by catalysis. Unlike conventional forced-circulation and heating FCS, the autocatalytic FCS is a passive system that does not need operator startup; it also has a simple structure, without components requiring power supplies, such as blowers and heaters. Thus, it is superior in terms of reliability and operability. This system is considerably smaller than a conventional FCS, and can be installed in portions within the pressure containment vessel (PCV); conventional FCSs are installed inside the reactor building (outside the PCV). So autocatalytic FCS makes it possible to reduce the size of the building, and construction costs as well. Though multiple tests have been performed in Europe and the US on the autocatalytic FCS for product development and to check performance[1],[2],[3],[4], before introducing this FCS to a new ABWR plants, some additional tests have been performed with consideration for Japanese BWR accident conditions like PCV spray, influence of low oxygen condition and reaction inhibition in the collaborative research by Japanese electric power companies. These include examining the influences of low-oxygen conditions and reaction inhibitors. This paper presents the details of additional studies made after considering results of the additional tests performed in Japan to install the autocatalytic FCS instead of the thermo-reactive FCS.
Experimental Study on Thermal-Hydraulic Performance of Single-Phase Water Flow in Narrow Rectangular Channel
Authors: Wang Mei , Wen Yan , Qiu Suizheng , Su Guanghui , Ni Weifeng
Published: February 2010

The purpose of this study is to discover the differences of pressure drop and heat transfer of single-phase water flow between conventional channels and narrow rectangular channels. Furthermore, the differences between the level and the vertical channel have been studied. The gap of the test channel is 1.8mm. Compared with conventional channels, the narrow rectangular channel showed differences in both flow and heat transfer characteristics. The critical Reynolds number of transition from laminar flow to turbulent flow is 900∼1300, which is smaller compared with conventional channels. The friction factor is larger than that of the conventional channels and the correlation of friction factor with Reynolds number was given by experimental results. From the relation graph of Nusselt number and Reynolds number, the demarcation of the laminar flow region and turbulence flow region is obvious. In laminar region, Nusselt number almost remained constant and approximately consistent with numerical simulation results. While in turbulent region, Nusselt number increased significantly with increasing Reynolds number. A new Nusselt number correlation was obtained based on Dittus-Boelter equation, and the coefficients were less about 13% than that of Dittus-Boelter equation.
Design of a Decay Heat Removal System for EFIT Plant
Authors: Leoncini Barbara , Alemberti Alessandro , Mansani Luigi , Pini Prato Alessandro
Published: February 2010

The Integrated Project EUROTRANS, funded by the European Commission as part of the 6th European framework program, provides the advanced design of a multi purpose research oriented Accelerator Driven System (ADS), called XT-ADS (eXperimenTal-ADS), and the preliminary design of an industrial scale ADS, called EFIT (European Facility for Industrial Transmutation) [1]. EFIT [2] [3] is fuelled with Minor Actinides (U-free fuel) and its main design options overcome the limits of previous ADS projects, particularly as concerns the economic aspects, maintaining meanwhile the high safety level, the high reliability and low investment risks. EFIT is endowed with three different systems to remove decay heat, this paper deals with the Decay Heat Removal System 2 (DHR2). DHR2 mainly consists of an isolation condenser (IC) connected to the secondary side of the Steam Generators (SGs). DHR2 is a passive safety system: system operation is based on condensation of the steam coming from the SG unit and gravity head injection of the condensate back to the SG inlet. The system has been designed to satisfy the following two requirements: • to evacuate the decay power maintaining acceptable values of the vessel and fuel temperature; • to avoid lead solidification also in the configuration of maximum efficiency.
Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress
Authors: Oh Chang H. , Kim Eung S. , Schultz Richard , Petti David , Kang Hyung S.
Published: February 2010

A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (∼160 sec) would be significantly earlier than the previous predictions (∼150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.
Management of Construction at Latest ABWR “Shimane 3”
Authors: Hirata Norihiko , Fushiki Katsumi , Nishimoto Masahide , Murayama Kouichi
Published: February 2010

The Chugoku Electric Power Co., Inc has been constructing Shimane nuclear power station unit No.3 as our top priority, which is the fifth ABWR (Advanced Boiling Water Reactor) and one of the largest nuclear power plants (Electrical output: 1,373 MWe) in Japan. This will enable us to attain “3Es”: (Environment, Energy Security, Economy) simultaneously. After issue of the first construction license, the excavation of main buildings started in 2006, and we have been progressing smoothly toward its commercial operation in December 2011. In this project, we coordinate all subcontractors in civil, building and mechanical fields and manage among each company in order to control construction work progress. “Modularization” is applied for the construction as one of the latest construction technologies. Plant components such as equipment, piping and structures are modularized in a shop or fabrication areas in a construction site and are installed by a large crawler crane at the site. This method contributes to improve construction safety and quality of work, as well as to shorten the construction schedule.
Westinghouse AP1000 Power Production Reliability Classification Methodology
Authors: Evans Matthew C. , Schulz Terry L.
Published: February 2010

The AP1000 is a pressurized water reactor with passive safety features and extensive plant simplifications that provide significant and measurable improvements in safety, construction, reliability, operation, maintenance and cost. The design of the AP1000 has incorporated a standardization approach, which results in a plant design that can be constructed in multiple geographical regions with varying regulatory standards and expectations. The AP1000 uses proven technology, which builds on over 30 years of operating PWR experience. In order to provide additional assurance that the AP1000 will meet its aggressive reliability and availability goals, Westinghouse has developed a Power Production Reliability Methodology. This methodology is used to identify structures, systems, and components (SSCs) that are important to reliable plant operation and applies appropriate procurement actions to these SSCs. Internationally, nuclear facilities have backfit equipment reliability processes in order to efficiently provide high levels of safe and reliable plant operation. These processes are characterized by the classification of SSCs based on the impact of component failure on plant operation. This classification is used to prioritize and simplify plant maintenance programs, optimize spare parts inventories, and improve component and system performance monitoring. The AP1000 Power Production Reliability Methodology has been developed to utilize compatible classification process to support the design and operation of domestic and international AP1000 plants. The implementation of this process has enabled Westinghouse to identify reliability-important SSCs during the design phase and to establish procurement requirements for these SSCs to be supplied for the AP1000. Plans are being developed to expand the use of the AP1000 Power Production Reliability Methodology to include requirements for storage, transport, construction, and startup testing. The ultimate goal of this process will be the establishment of a site equipment reliability program in accordance with the plant operator’s processes to support effective commercial operation. The execution of the AP1000 Power Production Reliability Methodology will provide an additional level of confidence that the AP1000 will meet the design reliability and availability goals for commercial operation.
Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor
Authors: Wang Peng , Qiu Suizheng
Published: February 2010

The new concept Molten Salt Reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutrons precursors are drifted by the fuel flow, which leads the spread of delayed neutrons distribution to non-core parts of the primary circuit, and it also can result in a reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the Molten Salt Reactor is quite different from the conventional nuclear reactors using solid fissile materials, and no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particle considering the flow effect of the fuel salt in the Molten Salt Reactor, while the thermal-hydraulic model uses the fundamental conservation laws: the mass, momentum and energy conservation equations. Then the neutronic and thermal-hydraulic calculations were coupled and the influences of inflow temperature and flow velocity on the reactor physical properties were obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff. While the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron flux and delayed neutron precursors slightly, and makes significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady state, but affects the delayed neutron precursors’ distribution significantly.
Measurement of Surface Heat Flux and Surface Temperature in Nucleate Pool Boiling Using Micro-Thermocouples
Authors: Liu Wei , Takase Kazuyuki
Published: February 2010

In this paper, a measurement system for surface temperature and surface heat flux was developed to study heat transfer mechanism in boiling process. The system was consisted by two parts: (1) inner block temperatures were measured using micro-thermocouples set at two layers inside heating block; (2) with using the measured temperatures, inverse heat transfer analysis was performed to get surface heat flux and surface temperature. For the inner block temperature measurement, special T-type micro thermocouples with a common positive pole were developed. Totally 20 thermocouples were set at two layers at the depths 3.1μm and 4.905mm beneath the boiling surface, in a radius of 5mm. The developed system was used to research the change of surface heat flux and surface temperature in a boiling process. Experiments were performed to pool boiling at atmospheric pressure. The experiments showed the developed special T-type micro thermocouples could trace temperature change in boiling process successfully. With comparison to images from a high-speed camera, temperature change tendencies in boiling process were tried to understand. Then one dimensional inverse heat conduction problem was solved to get surface heat flux and surface temperature. Increase in surface heat flux with the generation of big bubble was derived successfully.
COSI Test Case Simulation With CFX Code
Authors: Glantz Tony
Published: February 2010

During a Small Break Loss-of-Coolant Accident (SB-LOCA) in a Pressurized Water Reactor (PWR), cold water (9°C) is injected into the cold leg and mixes with hot primary fluid (285°C) which flow down to the Reactor Pressure Vessel (RPV) downcomer. Such a scenario may result in a Pressurized Thermal Shock (PTS) on the structural components, first of all the walls. In order to access to the structural behaviour of the RPV during a PTS, reliable predictions of thermal loads on the RPV walls and fluid temperature distribution in the downcomer are needed. This second parameter depends on temperature mixing in the cold leg between cold water coming from Emergency Core Cooling (ECC) and hot primary fluid. The two-phase flow issues are Direct Contact Condensation (DCC), stratified flow, jet impacting a free surface or a cursed wall. A Computational Fluid Dynamic (CFD) simulation is needed in order to have an accurate prediction of these phenomena and then to calculate the distribution of fluid temperature in the cold leg and the downcomer. The CFD-code CFX seems to be appropriate to deals with such issues but it is not validate to simulate physical phenomena which occur in the cold leg during a PTS. The paper presents CFX assessment on COSI experiment. COSI facility is a scaled model of a 900 MWe PWR under LOCA thermal hydraulic conditions and has been developed to study condensation which can occur in the cold leg during a Pressurized Thermal Shock. The paper focuses on one test corresponding to a situation with a high water level in the cold leg. The COSI test has been simulated with CFX code where a specific model coming from literature has been used for heat transfer between liquid and interface description. The use of this model is necessary in order to obtain a good agreement between COSI experimental and simulated results. In fact, with this model, the code is able to reproduce correctly all the main features, mean error between experimental and simulated results being about 10%.
Study on Spacer Grid Span Pressure Loss Under High Reynolds Number Flow Condition
Authors: Ikeda Kazuo , Yamaguchi Yasunao , Shimizu Juntaro , Okamoto Kaoru
Published: February 2010

Pressure loss coefficient of spacer grid is used as a key parameter for PWR core thermal hydraulic design. It has been obtained by single-phase hydraulic testing for many years. However, it is necessary to develop design tool for precise estimation of pressure loss of spacer grids as well as hydraulic tests to meet the needs of the worldwide nuclear fuel market. Recently, Computational Fluid Dynamics (CFD) analysis has been applied for estimation of flow field in a fuel rod bundle. In this study, the numerical simulation in a range of bare rod Reynolds numbers of the reactor flow condition is performed to examine the applicability of the CFD model for estimating spacer grid span pressure loss. For verification of the numerical estimation, the span pressure loss of 5×5 rod bundle with spacer grid is measured in Nuclear Development Corporation (NDC) hydraulic test facility up to bare rod Reynolds number as high as 500,000. The simulation shows good agreement with experimental data in the range of Reynolds numbers. The CFD model is also utilized to investigate the pressure loss as a function of distance from last passed spacer grid and to discuss the turbulent flow characteristics in the rod bundle with spacer grid under high Reynolds number flow condition.
High-Cycle Thermal Fatigue in Mixing Tees: New Large-Eddy Simulations Validated Against New Data Obtained by PIV in the Vattenfall Experiment
Authors: Odemark Ylva , Green Torbjo¨rn M. , Angele Kristian , Westin Johan , Alavyoon Farid , Lundstro¨m Staffan
Published: February 2010

New data was obtained for a previously studied T-junction experimental setup [1] for a range of flow ratios between hot and cold flows in order to validate new Large Eddy Simulations (LES). The instantaneous velocity field downstream of the T-junction was measured with two-component Particle Image Velocimetry (PIV) in several horizontal and vertical planes at the centre line downstream of the T-junction. The generated PIV database enables a thorough validation of CFD turbulence statistics. The turbulence statistics are shown to be well predicted despite the fact that the mesh in the LES is rather coarse. By usage of time resolved PIV the temporal evolution of the predominant low frequent large-scale structures, responsible for much of the mixing and the high amplitude temperature fluctuations on the walls, were captured. Those structures are, however, weaker in LES than in PIV, being in line with the fact that the wake region behind the penetrating vertical hot jet is underpredicted in LES. Tests regarding the influence of the LES-results to the shape of the inlet boundary conditions (developed or flat symmetric mean-velocity profiles) were carried out and the sensitivity in the results was shown to be small. Furthermore, the results show good agreement with the experimental data independent of the flow ratio between the hot and the cold flows.
Development of Code, PNFENT, Based on Using Finite Elements for Neutron Transport
Authors: Zolfaghari Ahmad , Minuchehr Hamid , Abbasi Mohammadreza
Published: February 2010

A variational treatment of the finite element method for neutron transport is used based on a version of the even parity Boltzman equation for the general case of anisotropic scattering and sources. The theory of maximum principles is based on the Cauchy-Schwartz inequality and the properties of a leakage operator G and a removal operator C. For system with extraneous sources a maximum principle is used in boundary free form to ease finite element computations. The global error of an approximate variational solution is shown. The energy dependence of the angular flux is treated by the multi-group method. In this paper the spatial dependence of the angular flux is given in a finite element representation. The directional dependence of angular flux is represented preferably by a spherical harmonic expansion. The above method has been developed and implemented in the finite element program PNFENT. A homogenous slab of a pure absorber along edge-cell and a two dimensional problems are solved with an accuracy as good as the best problem techniques.
Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel
Authors: Allison Leyland J. , Grande Lisa , Mikhael Sally , Rodriguez Prado Adrianexy , Villamere Bryan , Pioro Igor
Published: February 2010

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.