bluuebunny/crossref_metadata_2025 huggingface dataset:
Hydrogen Mixing Analyses for a VVER Containment
Authors: Kostka Pal , Techy Zsolt , Sienicki James J.
Published: March 2009

Hydrogen combustion may represent a threat to containment integrity in a VVER-440/213 plant owing to the combination of high pressure and high temperature. A study has been carried out using the GASFLOW 2.1 three-dimensional CFD code to evaluate the hydrogen distribution in the containment during a beyond design basis accident. The VVER-440/213 containment input model consists of two 3D blocks connected via one-dimensional (1D) ducts. One 3D block contains the reactor building and the accident localization tower with the suppression pools. Another 3D block models the air traps. 1D ducts represent the check valves connecting the accident localization tower with the air traps. The VVER pressure suppression system, called “bubbler condenser,” was modeled as a distributed heat sink with water thermodynamic properties. This model accounts for the energy balance. However, it is not currently possible to model dynamic phenomena associated with the water pools (e.g., vent clearing, level change). The GASFLOW 2.1 calculation gave detailed results for the spatial distribution of thermal-hydraulic parameters and gas concentrations. The range and trend of the parameters are reasonable and valuable. There are particularly interesting circulation patterns around the steam generators, in the bubbler tower and other primary system compartments. In case of the bubbler tower, concentration and temperature contour plots show an inhomogeneous distribution along the height and width, changing during the accident. Hydrogen concentrations also vary within primary system compartments displaying lower as well as higher (up to 13–20% and higher) values in some nodes. Prediction of such concentration distributions was not previously possible with lumped parameter codes. GASFLOW 2.1 calculations were compared with CONTAIN 1.2 (lumped parameter code) results. Apart from the qualitatively similar trends, there are, for the time being, quantitative differences between the results concerning, for example, pressure histories, or the total amount of steam available in the containment. The results confirm the importance of detailed modeling of the containment, as well as of the bubbler condenser and sump water pools. The study showed that modeling of hydrogen distribution in the VVER-440/213 containment was possible using the GASFLOW 2.1 code with reasonable results and remarkable physical insights.
Experimental Study on Density Wave Instability of Sodium Boiling Two-Phase Flow in an Annulus
Authors: Qiu Suizheng , Takahashi Minoru , Jia Dounan
Published: March 2009

Experiments of density wave instability in a liquid sodium boiling two-phase flow experiments in an annulus were carried out in the following parameters range: heat flux from 80kW/m2 to 976kW/m2, inlet subcooling from 25.6°C to 226.8°C, mass flow rate from 7.92kg/h to 68.9kg/h, system pressure from 2600Pa to 0.12Mpa. Not only the mechanism of the instability, critical conditions and oscillation period, but also the effects of pressure, mass flow rate and inlet subcooling on the density wave instability were explored experimentally and theoretically. From the experimental data, it was found that the lower the inlet temperature was, the higher the system pressure and the mass flow rate that could result in a more stable boiling two-phase flow were. A correlation for the density wave instability was obtained on from the dimensional analysis for the conservation equations of mass, momentum and energy.
Utilization of Plutonium and Higher Actinides in the HTGR as Possibility to Maintain Long-Term Operation on One Fuel Loading
Authors: Tsvetkova Galina V. , Peddicord Kenneth L.
Published: March 2009

Promising existing nuclear reactor concepts together with new ideas are being discussed worldwide. Many new studies are underway in order to identify prototypes that will be analyzed and developed further as systems for Generation IV. The focus is on designs demonstrating full inherent safety, competitive economics and proliferation resistance. The work discussed here is centered on a modularized small-size High Temperature Gas-cooled Reactor (HTGR) concept. This paper discusses the possibility of maintaining long-term operation on one fuel loading through utilization of plutonium and higher actinides in the small-size pebble-bed reactor (PBR). Acknowledging the well-known flexibility of the PBR design with respect to fuel composition, the principal limitations of the long-term burning of plutonium and higher actinides are considered. The technological challenges and further research are outlined. The results allow the identification of physical features of the PBR that significantly influence flexibility of the design and its applications.
Experimental Study on Thermal Striping in Mixing Tees With Hot and Cold Water
Authors: Kawamura Tsutomu , Shiina Kouji , Ohtsuka Masaya , Tanaka Isao , Hirayama Hiroshi , Tanimoto Kouichi , Fukuda Toshihiko , Sakashita Akihiro , Mizutani Jun , Minami Yasuhiko , Moriya Shoichi , Madarame Haruki
Published: March 2009

Thermal striping tests in mixing tees with hot and cold water were conducted for three types of flow conjunctions in order to establish an evaluation method for high-cycle thermal fatigue of piping systems. Two kinds of examinations were planned. The preliminary tests were flow visualization tests carried out using acrylic pipes to obtain flow pattern characteristics and flow temperature fluctuations. The main tests were temperature fluctuation measurement tests carried out using metal pipes to evaluate the unsteady heat transfer coefficient based on measured temperature fluctuations of fluid and pipe wall. This paper reports visualization test results. The flow patterns were visualized by injection of methylene blue and compared with flow analysis results by the k-ε turbulence model. Temperature fluctuations of fluid 3mm from the inner pipe wall were measured with C-A thermocouples. Fundamental features such as locations with a large fluctuating temperature, the fluctuating temperature amplitude and its frequency were identified.
Advanced Modeling of Multicomponent Vaporization/Condensation Phenomena for a Reactor Safety Analysis Code SIMMER-III
Authors: Morita Koji , Matsumoto Tatsuya , Akasaka Ryo , Fukuda Kenji , Suzuki Tohru , Tobita Yoshiharu , Yamano Hidemasa , Kondo Satoru
Published: March 2009

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis code SIMMER-III, which analyzes relatively short-time-scale phenomena relevant to accident sequences of CDAs. The model characterizes the V/C process associated with phase transition through heat-transfer and mass-diffusion limited models to follow the time evolution of the rector core under CDA conditions. The heat-transfer limited model describes the nonequilibrium phase-transition processes occurring at interfaces, while the mass-diffusion limited model is employed to represent effects of noncondensable gases and multicomponent mixture on V/C processes. Verification of the model and method employed in the multicomponent V/C model of SIMMER-III was performed successfully by analyzing two series of condensation experiments.
Evaluations of the Coolability Through the Inherent In-Vessel Gap Cooling in the LAVA Experiments
Authors: Kang K. H. , Park R. J. , Kim J. T. , Kim S. B. , Kim H. D.
Published: March 2009

The analysis of the LAVA (Lower-plenum Arrested Vessel Attack) experimental results focused on gap formation and in-vessel gap cooling characteristics have been performed. In the LAVA experiment, Al2O3/Fe thermite melt (or Al2O3 only) was used as a corium simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation in case there was an internal pressure load across the vessel. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were mainly determined by the possibilities of the water ingression into the gap. The possibility of heat removal through the gap in the LAVA experiments was confirmed from that the vessel cooled down with the conduction heat flux through the vessel by 70 to 470 kW/m2. Also the quantitative evaluations of the in-vessel coolability using gap cooling model based on counter current flow limits (CCFL) have been performed for the LAVA experiments in parallel. It could be inferred from the analysis for the LAVA experiments that the vessel could effectively cooldown via heat removal through the gap cooling even if 2mm thick gap should form between the interface of the melt and the vessel in the 30 kg of Al2O3 melt tests. In the case of large melt mass of 70 kg of Al2O3 melt, however, the infinite possibility of heat removal through a small size gap such as 1 to 2 mm thick couldn’t be guaranteed due to the difficulties of water ingression through the gap into the lower head vessel bottom induced by the CCFL. Synthesized the experimental results and the analytical evaluations using the CCFL model, it could be found that the coolability through gap cooling was affected mainly by the melt composition and mass and also the gap thickness.
A Study on Quenching Meshes as a Possible Controlling Tool of Hydrogen Explosion in Nuclear Power Plants
Authors: Kim Hong Jip , Hong Seong Wan , Kim Hee Dong , Yang Seung Yeon , Chung Suk Ho
Published: March 2009

Characteristics of quenching meshes for control of hydrogen combustion in severe accident in nuclear power plant are experimentally investigated. Quenching distances of hydrogen-air mixtures without steam over a range of initial pressure are measured. Those of stoichiometric hydrogen-air mixtures with steam with respect to initial pressure are also measured. In addition, parameters that should be considered to ensure the performance and validity of the installation of quenching meshes for the control of hydrogen combustion in closed compartments are obtained from a simplified phenomenological analysis and experimentally investigated.
Seismic Proving Test of Eroded Piping: Program of Eroded Piping Tests
Authors: Namita Y. , Suzuki K. , Abe H. , Ichihashi I. , Shiratori M. , Iwata K. , Kojima N. , Ishiwata M.
Published: March 2009

In 2000FY, a 3 year program of eroded piping tests was initiated with the following objectives: 1) to ascertain the seismic safety margins for eroded piping designed under the current seismic design code, 2) to clarify the elasto-plastic response and ultimate strength of eroded nuclear piping. It was intended to carry out a series of tests on eroded piping components and eroded piping systems. This paper is a report on the program of eroded piping tests.
Analysis of Unsteady Turbulent Merging Jet Flows With Temperature Difference
Authors: Yoo Geun Jong , Jeon Won Dae
Published: March 2009

Suitable turbulence model is required in the course of establishing a proper analysis methodology for thermal stripping phenomena. For this purpose, three different turbulence models of k-ε model, modified k-ε model, and full Reynolds stress model and VLES are applied to analyze unsteady turbulent flows with temperature variation. Four test cases are selected for verification. These are vertical jet flows with water and sodium, parallel jet flow with sodium, and merging pipe flow through T-junction with sodium. The geometries of test cases well represent common places where thermal stripping might be occurred. The turbulence model computation shows overall jet flow characteristics well and good comparison of mean temperature distribution. Temperature variance (θ′2) is rather over-predicted, but location of high temperature variance is matched well with that of the large amplitude of temperature variation of experimental results. Meanwhile, mixing of hot and cold jet flow is found to be not that active.
A Model for Shortening the Length of Activities in Construction and Overhaul Planning: UPROB
Authors: Majdandzˇic´ Niko , Sebastijanovic´ Slavko , Maticˇevic´ Gordana , Sebastijanovic´ Nebojsˇa
Published: March 2009

This paper presents a mathematical model of the UPROB (planning system for construction and overhaul) system that was developed for tank assembly lines, construction of thermal energy structures, and for performing an overhaul in complex plants. Planning technology includes utilizing linear diagrams with a direct usage of input data from a plant’s database. A model has been developed to determine the critical path and also, to define steps for the most economical shortening of the entire plan. Several plan variations are developed (according to specified goals) and the management determines the optimal variation. After selecting a plan, it is possible to control and create work orders for individual tasks in certain activities. Task completion percentage, activity cost, and the total cost of the plan are also provided.
Applicability of Modified Drift Flux Model for Bubbly Flow in 2-D/3-D Rectangular Box With Various Kinds of Obstacles
Authors: Matsumoto Tatsuya , Uchibori Akihiro , Akasaka Ryo , Seki Toshinori , Kaminishi Shyuji , Morita Koji , Fukuda Kenji
Published: March 2009

In order to develop analytical tools for the analyses of multi dimensional two-phase flow in channels with obstacles, the modified drift flux model has been applied. Numerical simulations of multi dimensional gas-liquid two-phase flow in a channel, with some kinds of obstacles inserted to simulate a simple sub-channel in the fuel bundle, were carried out. Analytical results were compared with experiments, to show the validity of the modified drift flux model. Experiments were carried out with using an apparatus of 2-D/3-D rectangular box with a perforated plate or a horizontal plate with slit hole or a vertical rod inserted. Nitrogen gas-water adiabatic two phase flow was circulated in the box. The apparatus was made of acrylic resin plates and be able to make the flow inside visualized. Two-phase flow pattern were recorded with a high-speed video camera and the mass flow rate of nitrogen gas was measured with a digital gas-mass flow meter. Comparisons between the experimental results and the numerical ones showed good agreements, thus it was verified the model would be applied for predicting flows in more complex geometry with obstacles.
Analysis and Evaluation of the Liquefaction on Layered Soil
Authors: Lee Sang Hoon , Yoo Kwang Hoon
Published: March 2009

Liquefaction potential on the specific site of nuclear power plant is analyzed and reviewed. The layered site for this study consists of silt and sand. Based on the limited available soil data, maximum shear strength at critical locations using Seed & Idriss method and computer program SHAKE is calculated, and liquefaction potential is reviewed. As seismic input motion used for the assessment of liquefaction, the artificial time history compatible with the US NRC Regulatory Guide 1.60 is used. Assessment results of the liquefaction are validated by analyzing to the other typical soil foundations which can show the effects on the foundation depth and soil data.
Method of Calculation of Pipelines and Pressure Vessels Fracture Probability
Authors: Tkachev V. V. , Zheltukhin K. K.
Published: March 2009

Method of calculation of pipelines and pressure vessels fracture density probability is presented. The method takes into account operation experience of these structures in various conditions and their aging. The loading parameters, flaws and also strength and other mechanical properties of metal are used as random variables in this method.
Structure of Turbulent Flow in Subchannel of Rod Bundle Downstream of Spacer Grid With Hybrid Flow Mixing Device
Authors: Oh Dong Seok , In Wang Kee , Chun Tae Hyun
Published: March 2009

An experiment was performed in a wind tunnel to investigate the flow structure in a rod bundle with a hybrid vane grid. The hybrid vane is a flow-mixing device, which consists of two pairs of primary and secondary vanes in a cell. The test section is a rectangular channel (300 mm × 300 mm × 2400 mm) including 3 × 3 rod (75 mm diameter) array with a spacer grid. The pitch to diameter ratio of the rod array is 1.33. The flow structures downstream the grid are measured at Reynolds number of 1.2 × 105 for 35-degree deflecting angle of the hybrid flow-mixing vane. The data are obtained for the distributions of the time mean axial velocity, lateral velocity, and turbulent intensities in 3 component directions over a center subchannel along axial locations and compared with the previous results of split vane grid that has two vanes in a cell. The results show that the mixing efficiency of the hybrid vane grid could be similar with that of the split vane grid because swirl factor of the hybrid vane grid is higher than that of split vane grid and the magnitude of axial turbulent intensity, turbulent diffusion coefficient, and cross flow factor is similar to each other in spite of differences of the vane numbers and shape in a cell between hybrid and split vane grids.
Fuel Assembly Nonlinear Dynamic Model
Authors: Shah S. J. , Brenneman B.
Published: March 2009

Fuel assembly finite-element dynamic models are developed to perform the core seismic analysis. The fuel assembly is modeld by a single vertical beam, which represents the cross-sectional inertias of the fuel rods, guide thimbles and instrument thimble, and a series of rotational springs. The rotational springs are located at the intermediate spacer grid locations. Benchmarking to fuel assembly natural frequencies determined by testing is accomplished by adjusting the moment of inertia and the grid rotational stiffness to find their effective values. Most often these models are linear and are appropriate for the small amplitude stiffness representation of the fuel assembly. Large deflection problems are approximated by choosing a fuel assebly stiffness value appropriate to the average deflection range. Some loss of accuracy will naturally result from this approach. This paper presents a nonlinear model to approximate the hysterisis and free vibration response for large amplitudes fuel assembly motion. The force required to impose the initial displacement (pluck) and the free vibration responses are used to compare the nonlinear model’s behavior with the test data. This model correctly predicts fuel assembly deflection shapes as a function of axial position for various lateral loads for several fuel assembly designs. Displacement hysterisis is primarily due to fuel assembly to grid slippage, which is a strong function of grid preload. “Tight” and “relaxed” prototypes were tested to account for grid preload effect. The model correctly analyzes the grid preload effect. A nonlinear fuel assembly model provides better matching of grid impact loads determined by fuel assembly lateral impact testing and also prvides better matching of all of the initial conditions (initial deflection, initial force, initial energy and impact velocity). In this test, the fuel assembly impacts a test wall to determine grid internal stiffness. This value is used in the core model for seismic analyses.
Reliability Methods for Passive Safety Functions
Authors: Marque`s M. , Pignatel J. F. , D’Auria F. , Burgazzi L. , Mu¨ller C. , Cojazzi G. , La Lumia V.
Published: March 2009

This paper presents the study performed within the framework of a European Project called Reliability Methods for Passive Safety Functions (RMPS). Its objective is to propose a specific methodology to assess the passive system thermal-hydraulic (T-H) reliability. The methodology is tested on an example of industrial passive system: the Isolation Condenser System (ICS). The T-H calculations are performed using the RELAP5, ATHLET and CATHARE computer codes. In this paper, the present state of the methodology and its application to the example are described. Owing to the recent start of the project, only results concerning the first objective of the project (identification and quantification of the sources of uncertainties and determination of the important variables) are presented. A sensitivity analysis was carried out on 69 computation results performed with the RELAP5 code. This first analysis highlights the significant parameters influencing the performances of the ICS and shows a non-monotonous behavior of the system.
Mechanistic Prediction of Dryout Heat Flux in Annular Two-Phase Flow
Authors: Okawa Tomio , Kim Saeyun , Kataoka Isao , Naitoh Masanori
Published: March 2009

Incorporating the recently developed correlations for the rates of droplet deposition and entrainment with the film flow model, critical heat flux due to liquid film dryout in steam-water annular flow was predicted. In the present calculations, the predicted critical heat flux is affected not only by the deposition and entrainment rates but also by the entrainment fraction at the transition to annular flow (Transition quality was estimated by Wallis’s correlation). For this reason, the entrainment fraction at the transition was correlated in terms of dimensionless transition length to annular flow and density ratio. A simple correlation for the occurrence of dryout was also proposed. It was demonstrated by numerical simulations that the present method is to predict the available 1,340 data of dryout heat flux with the mean square relative error of 0.068.
Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients
Authors: Ratnayake Ruwan K. , Ergun S. , Hochreiter L. E. , Baratta A. J.
Published: March 2009

In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena.
Passive Safety of the STAR-LM HLMC Natural Convection Reactor
Authors: Sienicki James J. , Petkov Plamen V.
Published: March 2009

The STAR-LM 300 to 400 MWt class modular, factory fabricated, fully transportable, proliferation resistant, autonomous, reactor system achieves passive safety by taking advantage of the intrinsic benefits of inert lead-bismuth eutectic heavy liquid metal coolant, 100+% natural circulation heat transport, a fast neutron spectrum core utilizing high thermal conductivity transuranic nitride fuel, redundant passive air cooling of the outside of the guard/containment vessel driven by natural circulation, and seismic isolation where required by site conditions. Postulated loss-of-heat sink without scram, overcooling without scram, and unprotected transient overpower accidents are analyzed for the 300 MWt STAR-LM design using a coupled thermal hydraulics-neutron kinetics plant dynamics analysis computer code. In all cases, STAR-LM is calculated to exhibit passive safety with peak cladding and coolant temperatures remaining within the existing database for lead-bismuth eutectic coolant and ferritic steel core materials.
Tracking Interfaces in Vertical Two-Phase Flows
Authors: Aktas Birol
Published: March 2009

The presence of stratified liquid-gas interfaces in vertical flows poses difficulties to most classes of solution methods for two-phase flows of practical interest in the field of reactor safety and thermal-hydraulics. These difficulties can plague the reactor simulations unless handled with proper care. To illustrate these difficulties, the US NRC Consolidated Thermal-hydraulics Code (TRAC-M) was exercised with selected numerical bench-mark problems. These numerical benchmarks demonstrate that the use of an average void fraction for computational volumes simulating vertical flows is inadequate when these volumes consist of stratified liquid-gas interfaces. In these computational volumes, there are really two regions separated by the liquid-gas interface and each region has a distinct flow topology. An accurate description of these divided computational volumes require that separate void fractions be assigned to each region. This strategy requires that the liquid-gas interfaces be tracked in order to determine their location, the volumes of regions separated by the interface, and the void fractions in these regions. The idea of tracking stratified liquid-gas interfaces is not new. There are examples of tracking methods that were developed for reactor safety codes and applied to reactor simulations in the past with some limited success. The users of these safety codes were warned against potential flow oscillations, conflicting water levels, and pressure disturbances which could be caused by the tracking methods themselves. An example of these methods is the level tracking method of TRAC-M. A review of this method is given here to explore the reasons behind its failures. The review shows that modifications to the field equations are mostly responsible for these failures. Following the review, a systematic approach to incorporate interface tracking methods is outlined. This approach is applicable to most classes of solution methods. For demonstration, the approach to incorporate the tracking method into the field equations of TRAC-M is described in steps. The success of this approach is demonstrated by exercising TRAC-M with the same benchmark problems that were previously used to illustrate the difficulties the code suffered from in the presence of interfaces. Besides improvements to the accuracy of the code predictions, one of the benchmark problems, which simulates a strong condensation at the liquid-gas interface, shows that the code’s runtime is improved significantly where the alternative methods like water packing fails.
Numerical Simulation of a Compartment Fire in a Nuclear Power Plant Containment Building
Authors: Floyd Jason
Published: March 2009

The current trend towards the increased use of risk assessment in the regulation of nuclear power plants will inevitably result in changes in the analysis of fire protection systems and the methods of analysis. Before fire protection can be regulated on a risk basis, a consensus must be reached on a number of issues. One key issue is what types of computational tools will be allowable for analyzing fire events, and what types of scenarios those tools will be approved for use. Reaching this consensus will require an understanding of the types of computational tools available and their inherent advantages and disadvantages. To aid with this understanding, three different methods of fire simulation are applied to an oil pool fire test in the HDR (Heiss Dampf Reaktor) containment test facility [1]. These methods are a hand calculation, the zone model code CFAST (Consolidated Model of Fire Growth and Smoke Transport) [2], and the computational fluid dynamics code FDS (Fire Dynamics Simulator) [3]. Each is applied to a steady-state portion of the test using, to the extent possible, the same set of input parameters. The results of the computation are compared to the test data. The comparisons show that each method is potentially suitable for use depending on the information required from the simulation. Each method will potentially have a role to play in risk based regulation depending on the scenario.
Performance and Unavailability Analysis of RPS/ESFAS in Korea Standard Nuclear Power Plant (KSNPP)
Authors: Jang Seung-Cheol , Min Kyung-Ran , Han Sang-Hoon
Published: March 2009

The safety-related performance analysis of plant protection system (PPS) comprising reactor protection system/engineered safety features actuation system (RPS/ESFAS) was performed from the operating experience of the Korean standard nuclear power plant (KSNPP). The PPS operational data was collected from the trouble reports (TR) to record details of test and maintenance activities at sites. The total operating experience of 8.63 commercial reactor years at four units during the period 1995 through 2000 was studied. The system unavailability analysis was also performed through the detailed fault tree models, using plant specific data based on observed operational experience. Estimated were the unavailabilities on 11 automatic trip parameters for the RPS and 6 signals for the ESFAS. Results of the data analysis and system unavailability were close to ones published for other CE-supplied plants, though this study included a lot of failures occurred in the beginning periods of commercial operation without percolation. This study was performed to provide technical basis for risk-informed applications like technical specification improvement.
Study of Interfacial Structures: Bubbly Flow in 1.27 cm Diameter Pipe
Authors: Paranjape S. , Kim S. , Ishii M. , Uhle J.
Published: March 2009

The objective of the present research is to study the flow regime map, the detailed interfacial structures, and the bubble transport in an adiabatic air-water two-phase flow mixture, flowing upward through a vertical round pipe having 1.27 cm. inner diameter. The flow regime map is obtained by processing the characteristic signals acquired from an impedance void meter, using a self-organized neural network. The local two-phase flow parameters are measured by the state-of-the-art four-sensor conductivity probe at three axial locations in the pipe. The measured local parameters include void fraction (α), interfacial area concentration (ai), bubble frequency (fb), bubble velocity (Ub) and bubble Sauter mean diameter (Dsm). The radial profiles of these parameters and their development along the axial direction reveals the structure of the two phase mixture and the bubble interaction mechanisms.
Risk-Informed Assessment Methodology Development and Application
Authors: Chi Sung Goo , Park Seok Jeong , Choi Chul Jin , Ritterbusch S. E. , Jacob M. C.
Published: March 2009

Westinghouse Electric Company (WEC) has been working with Korea Power Engineering Company (KOPEC) on a US Department of Energy (DOE) sponsored Nuclear Energy Research Initiative (NERI) project through a collaborative agreement established for the domestic NERI program. The project deals with Risk-Informed Assessment (RIA) of regulatory and design requirements of future nuclear power plants. An objective of the RIA project is to develop a risk-informed design process, which focuses on identifying and incorporating advanced features into future nuclear power plants (NPPs) that would meet risk goals in a cost-effective manner. The RIA design methodology is proposed to accomplish this objective. This paper discusses the development of this methodology and demonstrates its application in the design of plant systems for future NPPs. Advanced conceptual plant systems consisting of an advanced Emergency Core Cooling System (ECCS) and Emergency Feedwater System (EFWS) for a NPP were developed and the risk-informed design process was exercised to demonstrate the viability and feasibility of the RIA design methodology. Best estimate Loss-of-Coolant Accident (LOCA) analyses were performed to validate the PSA success criteria for the NPP. The results of the analyses show that the PSA success criteria can be met using the advanced conceptual systems and that the RIA design methodology is a viable and appropriate means of designing key features of risk-significant NPP systems.
Use of Influence Diagrams and Fuzzy Theory to Develop Assessment Method of Organizational Influences on Component Maintenance
Authors: Kim Yoonik , Ahn Kwang-Won , Chung Chang-Hyun , Kim Kil Yoo , Yang Joon-Eon
Published: March 2009

Organization can make influences on all the systems. Especially in case of nuclear power plants in which safety is established to be one of the most important operating goals, there have been a lot of research efforts for the hardware advancement. However in recent years, it has been widely recognized that organizational factors in nuclear power plants have an important influence on the safety attitudes and the safe behavior of individuals. Until now, any means to include assessments of organizational structure in probabilistic risk assessments have not been universally accepted. The objective of this work is to develop a method to assess organizational influences on component maintenance. Influence diagrams are introduced in this method as a decision making tool and fuzzy theory is used to reflect the vagueness in considering relevance of human activities in maintenance tasks. Introducing fuzzy theory to assess the organizational factors is deemed to a somewhat new trial, which makes it possible to convert linguistic vague descriptions into mathematical ones. Fuzzy linguistic descriptions offer an alternative and often complementary language to conventional, i.e., analytic approaches to modeling systems. Among the existing methodologies to assess organizational factors, the concept of the ω-factor model is utilized and the mechanism that organizational factors have influences on component maintenance is evaluated through composing influence diagrams. These influences go to failure rates and eventually affect component unavailability. Further study will make it possible that the influences of organizational factors on human error probabilities are incorporated into human reliability analysis and furthermore probabilistic safety assessment.
RBEC-M Lead-Bismuth Cooled Fast Reactor: Optimization of Conceptual Decisions
Authors: Mikityuk K. , Vasiliev A. , Fomichenko P. , Schepetina T. , Subbotin S. , Alekseev P.
Published: March 2009

A concept of the RBEC lead-bismuth fast reactor is a synthesis, on one hand, experience in development and operation of fast sodium power reactors and reactors with Pb-Bi coolant, and, on the other hand, of large R&D activities on development of the core concept for modified fast sodium reactor. The paper presents improved project RBEC-M, characterized by a number of innovative decisions, which allow to improve safety and cost parameters compared to the basic RBEC project. These innovative decisions include application of nitride fuel based on 15N, two-circuit scheme without main reactor pumps, gas lift system in the primary circuit, passive reactor auxiliary cooling system, etc.
Numerical Prediction of Cooling Capability in Hemispherical Gap Flow Passage for In-Vessel Core Retention
Authors: Uchibori Akihiro , Fukuda Kenji , Morita Koji , Matsumoto Tatsuya
Published: March 2009

A numerical method for thermal hydraulic phenomena in a hemispherical narrow gap flow passage was developed to evaluate a cooling capability with gap formation between the molten core and the reactor pressure vessel.The gap cooling mechanism was modeled as gas-liquid two-phase flow in the narrow gap with two-dimensional spherical coordinate system. The analytical model is based on a modified drift flux model for multi-dimensional two-phase flow analysis. Numerical results showed that liquid phase intrusion into the gap in the counter direction of gas phase upward flow kept down a rise of void fraction as gap cooling phenomena. Under the high heat flux condition,expansion of the high void fraction region due to the counter-current flow limitation was reproduced as a dryout phenomenon. Characteristics of gap cooling limitation predicted by the numerical analyses were verified by comparison with various experimental data and correlations of critical heat flux.
Motor-Operated Valve Importance Categorization Study Using Level 1 Internal PSA
Authors: Lee Yong Suk , Chung Chang Hyun , Sung Key Yong
Published: March 2009

Recently GL 89-10 static and dynamic Motor-Operated Valve (MOV) tests are being performed in Korean nuclear power plants to ensure MOV operability in design basis condition. And GL 96-05 MOV periodic verification testing will be followed after GL 89-10 testing. MOV risk importance categorization will be incorporated in Korean MOV periodic testing program for effective MOV test. In this study, MOVs were preliminarily categorized based on risk importance measures FV and RAW using level 1 internal event PSA. Configuration effect and importance uncertainty effect to risk importance of MOVs were also investigated.
On-Line Monitoring of Instrument Channel Performance in Nuclear Power Plant Using PEANO
Authors: Fantoni Paolo F. , Hoffmann Mario , Shankar Ramesh , Davis Eddie L.
Published: March 2009

On-Line monitoring evaluates instrument channel performance by assessing its consistency with other plant indications. Industry and EPRI [1] experience at several plants has shown this overall approach to be very effective in identifying instrument channels that are exhibiting degrading or inconsistent performance characteristics. On-Line monitoring of instrument channels provides information about the condition of the monitored channels through accurate, more frequent monitoring of each channel’s performance over time. This type of performance monitoring is a methodology that offers an alternate approach to traditional time-directed calibration. On-line monitoring of these channels can provide an assessment of instrument performance and provide a basis for determining when adjustments are necessary. Elimination or reduction of unnecessary field calibrations can reduce associated labor costs, reduce personnel radiation exposure and reduce the potential for miss-calibration. PEANO [2–9] is a system for on-line calibration monitoring developed in the years 1995–2000 at the Institutt for energiteknikk (IFE), Norway, which makes use of Artificial Intelligence techniques for its purpose. The system has been tested successfully in Europe in off-line tests with EDF (France), Tecnatom (Spain) and ENEA (Italy). PEANO is currently installed and used for on-line monitoring at the HBWR reactor in Halden. This paper describes the results of performance tests on PEANO with real data from a US PWR plant, in the framework of a co-operation among IFE, EPRI and Edan Engineering, to evaluate the potentials of PEANO for future installations in US nuclear plants.
Passive Safety Features in Sodium Cooled Super-Safe, Small and Simple Reactor
Authors: Ueda N. , Kinoshita I. , Nishi Y. , Minato A. , Matsumiya H. , Nishiguchi Y.
Published: March 2009

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.
Natural Circulation of Lead-Bismuth in a One-Dimensional Loop: Experiments and Code Predictions
Authors: Agostini P. , Bertacci G. , Gherardi G. , Bianchi F. , Meloni P. , Nicolini D. , Ambrosini W. , Forgione F. , Fruttuoso G. , Oriolo F.
Published: March 2009

The paper summarizes the results obtained by an experimental and computational study jointly performed by ENEA and University of Pisa. The study is aimed at assessing the capabilities of an available thermal-hydraulic system code in simulating natural circulation in a loop in which the working fluid is the eutectic lead-bismuth alloy as in the Italian proposal for Accelerator Driven System (ADS) reactor concepts. Experiments were performed in the CHEOPE facility installed at the ENEA Brasimone Research Centre and pre- and post-test calculations were run using a version of the RELAP5/Mod.3.2, purposely modified to account for Pb-Bi liquid alloy properties and behavior. The main results obtained by the experimental tests and by the code analyses are presented in the paper providing material to discuss the present predictive capabilities of transient and steady-state behavior in liquid Pb-Bi systems.
Influence of Water Sub-Cooling on Discharge Characteristics in a Water Storage Tank
Authors: Chen Bingde , Zheng Haifeng , Li Jie
Published: March 2009

The influences of water sub-cooling on water discharge behaviors from the water storage tank were studied experimentally with a test rig, which consists of a water vessel, a steam supplier, valves and piping. From this study, it is found that the gravitational discharging process is effected strongly by the steam condensation, taking place between steam and cold water in the vessel during discharge. Increased sub-cooling of the water, enhances the steam condensation, and promotes a deeper penetration of steam into water. The two modes of condensation identified in this study are: steam-supply-limit-mode and steam-condensation-limit-mode. The first mode occurs at the primary discharge period if water sub-cooling is large than a certain value, for example, 50°C sub-cooling in this study. Conclusions were derived based on this study for the impact of subcooling on discharge characteristics with respect to flow, pressure, and other parameters. Some suggestions to avoid or reduce undesired phenomena are also presented.
Stability Analysis on Single-Phase Natural Circulation in Argonne Lead Loop Facility
Authors: Wu Qiao , Sienicki James J.
Published: March 2009

One-dimensional linear stability analysis was performed for single-phase lead-bismuth eutectic natural circulation. The Nyquist criterion and a root search method were employed to find the linear stability boundary of both forward and backward circulations. It was found that the natural circulations could be linearly unstable in a high Reynolds number region. Increasing loop friction makes a forward circulation more stable, but destabilizes the corresponding backward circulation under the same heating/cooling conditions. The characteristic wavelength of an unstable disturbance is roughly equal to the entire loop length.
Design Strategies for Lead-Alloy-Cooled Reactors for Actinide Burning and Low-Cost Electricity Production
Authors: Hejzlar Pavel , Todreas Neil E. , Driscoll Michael J. , MacDonald Philip E. , Buongiorno Jacopo , Weaver Kevan D.
Published: March 2009

A multi-year project at the INEEL and MIT is investigating the potential of lead or lead-bismuth (lead-alloy) cooled fast critical reactors for producing low-cost electricity as well as for burning actinides from LWR spent fuel. While these two goals are the primary thrust in the development of a conceptual design, the proliferation resistance of the fuel and the plant safety are also important constraints incorporated into the design process. Thus, this concept addresses all Generation IV reactor goals, which involve favorable economics, enhanced safety, and sustainability. This paper outlines the objectives of the project, the challenges shaping the design strategy, and the approaches adopted to achieve the design goals. The most promising path forward is also identified. The four key factors that influence the direction of the design and also require compromise are the actinide destruction rate, safety, economy, and proliferation resistance. Achieving a maximum actinide destruction rate per MWth requires fertile-free fuels. However, the achievement of safe reactivity coefficients in such cores is difficult. If the total elimination of actinides from LWR spent fuel is pursued, multiple reprocessing with high recovery efficiency is necessary. This will probably significantly increase the fuel cycle costs, thus negatively affecting the economics. On the other hand, in-situ breeding and burning of plutonium in cores initially loaded with U235 can be cost effective. However, such a system does not achieve any reduction in the actinide inventory, and the discharge fuel contains relatively pure Pu239, which poses a potential proliferation threat. To reconcile these competing goals, a number of approaches have been investigated to achieve a balanced design that is cost competitive with other alternatives for electricity generation, attains excellent safety, helps in the reduction of transuranics from the spent LWR fuel, and has discharged fuel that is at least as proliferation resistant as spent LWR fuel from a once-through cycle. The preliminary design of the reactor concept that has the best potential to achieve these characteristics is identified and briefly described. This concept incorporates a supercritical carbon dioxide power conversion cycle that achieves thermal efficiencies up to 45% at a core outlet temperature of 550°C. However, conventional steam cycles are also an option.
Estimation of a Decommissioning Program Considering the Reuse of Demolition Materials
Authors: Sakata Eiichi , Ozaki Sachio , Hironaga Michihiko , Ogane Daisuke , Usui Tatsuo , Kono Yutaka
Published: March 2009

As for a decommissioning job in Japan decontamination and dismantling including safety store are executed during the prior period of the total project. The reprocessing and disposal of dismantling wastes as well recycling is to be practiced during the succeeding period. An expert system proposed in this paper has a faculty of furnishing motivations for decommissioning planners by correlative estimations of the project between its prior part and the succeeding recycle part of total process. This paper presents both a summarized configuration and an algorithm of the proposed model and indicates some contents of essential data bases (D/B) to be prepared as well some additional data to embody an assumed scenario for recycling. The proposed model provides the useful outputs concerning commercial cost, required procedures or licenses, regional encouragement, obstacles to be surmounted and so on. These outputs are available to explain the outline of the project both to inside and to outside of a plant corporate entity by combining each other. Simulated cases for concrete structures in non-controlled areas bring some information about both feasibility and comparisons of assumed recycling scenarios taking account of quality requirements in relevant technical standards, ordinances and the ability of reprocessing facilities.