bluuebunny/crossref_metadata_2025_split huggingface dataset:
AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology
Authors: Wang G. , Sapienza P. , Fetterman R. J. , Young M. Y. , Secker J. R. , Nelson S. W. , Oelrich R. L. , Sung Y. 
Published: October 2013

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization.
In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.
Study of Aerosol Scavenging by Rain in Case of a Radioactive Contamination of the Atmosphere
Authors: Quérel Arnaud , Lemaitre Pascal , Monier Marie , Porcheron Emmanuel , Flossmann Andrea 
Published: October 2013

The analysis of the radioactive aerosol scavenging by rain after the Chernobyl accident highlights some differences between the modelling studies and the environmental measurements. Part of this gap is due to the uncertainties on the scavenging efficiency of aerosol particles by raindrops, in particular for drops with a diameter larger than one millimeter. The IRSN (Institut de Radioprotection et de Sûreté Nucléaire) has decided to launch an experimental study to measure with a better accuracy the scavenging efficiency of large raindrops.
The scavenging efficiency of aerosol has been determined by measuring precisely the mass of aerosol particles collected by a single drop after its path through an atmosphere loaded with particles.
The collection efficiencies for drop diameters of 2 mm and 2.6 mm (previously unknown for atmospheric aerosols) are measured. The impact of these new data on modeling of the washout of the atmosphere by the rain is noticed.
Improving Air-Cooled Condenser Performance and Availability Through Innovative Cleaning, Inspection and In-Situ Sleeving
Authors: Fischer Gary , Ripley Craig 
Published: October 2013

Whether a power plant chooses an air-cooled condenser (ACC) because of siting issues, or because of changing environmental laws, the utilization of this dry cooling option has seen a dramatic level of growth over the past decade around the world. Like the steam surface condenser, the ACC is prone to tube fouling, albeit to the external finned tube surfaces instead of internal surfaces. This tube fouling can dramatically impact the performance of the ACC causing plant heat rate to suffer, increasing the consumption of fuel, increasing CO2 emissions and reducing megawatt output. In addition to fouling, ACC units are exposed to dramatic swings in ambient temperature. Units exposed to multiple freeze-thaw cycles each year often develop numerous points of inleakage into the vacuum boundary, also causing unit efficiency to suffer. Unlike the steam surface condenser, the ACC can be elevated at 60–75 feet in the air which poses considerable challenges to locating and repairing the sources of leaks.
This paper will examine the performance improvements realized by the Yellowstone Power Plant, located in Billings, Montana on their ACC unit after utilizing an innovative cleaning technology as well as the use of a new finned tube “sleeve” to repair leaking tubes.
Coupled Code SOCRAT-BN Development for Safety Analysis of Sodium-Cooled Fast Reactors
Authors: Usov Eduard V. , Kudashov Ivan G. , Zhigach Sergey A. , Butov Anton A. , Pribaturin Nikolay A. , Lezhnin Sergey I. , Chalyy Ruslan V. , Yakush Sergey E. , Vinogradova Uliya 
Published: October 2013

SOCRAT-BN is a software package to simulate design and severe accidents of sodium-cooled fast reactors. The package consists of modules for calculating damage to the reactor’s core, thermohydraulic processes and neutron physics.
The thermohydraulic module has been developed to calculate one- and two-phase flows in channels with different geometry and bundles. The module is based on a two-fluid model for equal pressures of phases.
In this paper we present an explanation of the deciding constitutive models for equations used in the system. Validation of the module was performed on the experimental data for one- and two-fluid flows in complex geometry channels and on calculation of running a first loop of the reactor BN-600 in nominal mode.
Dynamic Simulation Research on Primary Coolant System of Sodium Cooled Fast Reactor
Authors: Yu Yang , Guo Yun 
Published: October 2013

The Chinese Experimental Fast Reactor (CEFR) is a 65MWt/20MWe sodium cooled fast reactor. It is a pool-type reactor where the reactor and other internals such as pumps and intermediate heat exchangers (IHX) are immersed in a sodium pool. In this paper a one-dimensional dynamic code was developed to model the primary sodium circuit which included the reactor core, IHX, pumps, hot and cold pool etc. Moreover, the model of the property of sodium flow and heat transfer correlations was collected and compiled. This paper also discusses the mathematical models of various components of the primary sodium circuit, the numerical techniques to solve the models, the thermal-hydraulic studies of some design basis events such as the loss of primary pump or secondary pump accident etc, the comparison of the results of the code with that of the safety analysis report.
Studies were conducted simulating both full and low power operating conditions. The dynamic code has been validated, and the results show that it has a benign response to some typical accidents. Finally, the model and code derived in this paper could be used in the safety analysis of pool-type sodium cooled fast reactor, and adopted in the development of CEFR simulation platform.
Research on Low Cycle Fatigue in 26m2 Low Pressure Inner-Casing of Nuclear Power Plant
Authors: Hu Yifeng , Chen Gang , Kang Ming 
Published: October 2013

High reliability and long life endurance are very important factors in the operation of large steam turbines for nuclear power plant (NPP). The low cycle fatigue (LCF) is necessary take into account problem in the design of turbine component. Thus the calculation method and assessment criterion of LCF are important for the design and development of the low pressure (LP) inner-casing.
The main activity of this study was focus on the research of the low cycle fatigue in 26m2 LP inner-casing of steam turbines for NPP, which is manufactured by Shanghai Electric Power Generation Equipment Co., Ltd: STP. And detailed calculation method and assessment criterion combined with FEA analysis, fatigue curve and linear cumulative damage theory are developed to judge the life of the LP inner-casing, The approach is applied in the process of the design work of large steam turbine inner-casing.
Visualization on Inert Gas Jets Impinging to a Glass Tube Submerged in Liquid Sodium
Authors: Kudoh Hideyuki , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki , Kurihara Akikazu 
Published: October 2013

For the sodium-water reaction accident, it is important to grasp the structure of gas jets submerged in liquid sodium and associated droplet sizes. In this study, we successfully obtained visualized images of inert gas jets injected into liquid sodium. Formation processes of liquid sodium droplets entrained into the gas jets and drop-size distributions are discussed.
Visualization was conducted on upward argon gas jets impinging to a glass tube in stationary liquid sodium pool. The gas jet behavior was observed by using a bore scope inserted inside of the glass tube. The gas jet velocities at the nozzle exit were 13 < u0 < 125 m/s.
For u0 > 50 m/s, liquid droplets began to be formed. The formation processes of liquid droplets were classified into (1) spray type; a liquid film formed on the nozzle is blown and atomized by the gas jets and (2) entrainment type; attributed to the instability of a gas cavity-liquid phase interface. But the difference in the drop-size distribution of the both formation processes was not significant. Mean droplet diameter of liquid sodium was confirmed to be agreed well with an empirical equation suggested from a model experiment using water.
Estimation of Fragmentation on Jet Breakup in Coolant
Authors: Kuroda Taihei , Abe Yutaka , Kaneko Akiko , Yuzuru Iwasawa , Nariai Hideki , Sakaba Hiroshi , Koyama Kazuya , Matsuo Eiji 
Published: October 2013

Fast Breeder Reactor (FBR) is designed with safety in mind. However, there is billion to one possibility that a hypothetical Core Disruptive Accident (CDA) occurs. When CDA occurs, the Post Accident Heat Removal (PAHR) must be achieved. In the PAHR, the molten material is required to be fragmented and solidified in sodium coolant. In order to estimate whether the molten material jet is completely solidified in sodium coolant or not, it is significant to estimate jet breakup length. Although, the jet breakup length is influenced with fragmentation behavior, the correlation between them is not clear yet. Therefore, it is strongly required to clarify the mechanism of the fragmentation behavior on the jet surface. The objective of the present study is to estimate fragmentation on jet breakup in coolant experimentally. Tap water and Fluorinert™ (FC-3283) are used as simulated coolant and molten material, respectively. Flourinert is transparent and colorless liquid and its density is higher than water, therefore we can observe internal flow structure of Fluorinert. Fluorinert injected into water, and the jet breakup behavior and the fragmentation behavior of the jet are observed by using high speed video camera.
In order to estimate fragmentation on liquid jet, we identified the position of the interface with back lighting technique and also, we conducted velocity measurement with Particle Image Velocimetry (PIV) technique simultaneously. It is observed that interfacial waves of the jet are generated. Waves are pulled with surrounding liquid and grown up. Finally, a fragment is separated as a droplet from front edge of the wave. Also, the vorticity is evaluated from the velocity data in order to investigate influence of the flow field in detail. From the result of calculating vorticity, the high value was estimated when jet was fragmented. It is suggested that fragmentation behavior correlates with the surrounding flow field. And the energy ratio contributing to fragmentation is calculated from velocity field. The energy ratio is important to investigate the amount of the fragmentation on liquid jet. Fragmentation on jet breakup in coolant is estimated.
Development of a Core Management Tool for MYRRHA: Work in Progress
Authors: Jaluvka David , Van den Eynde Gert , Vandewalle Stefan 
Published: October 2013

Currently, the advanced nuclear irradiation facility MYRRHA is under development at the Belgian Nuclear Research Centre SCK•CEN in Mol, Belgium. MYRRHA is intended as a multi-purpose machine for material and fuel irradiations, radioisotopes production, silicon doping, transmutation research, education, etc. A special effort is being made within the project to develop an in-core fuel management tool for MYRRHA.
Various requirements have been imposed on the core management tool; e.g., the tool should allow a user to specify a variety of optimization objectives and constraints. Based on that, a modular design of the tool was developed that makes it independent of the particular metaheuristic method used for the optimization as well as of the reactor analysis code used for an evaluation of candidate solutions during the search.
In the paper, the current version of the core management tool for MYRRHA is presented together with some results on an illustrative in-core fuel management optimization problem.
AREVA’s Advanced Thermal-Hydraulic Codes and Methods for Levels III and IV Crud Risk Assessment
Authors: Jones John H. , Anghelescu Mihnea S. , Bradbury Michael S. , Martin Mathieu G. , Galimov Azat Y. , Hatman Anca , Harne Richard L. , Keheley Thomas H. , Wende Uta , Bigot Jérôme 
Published: October 2013

From a crud deposition perspective the achievement of zero fuel failures requires the integration of core neutronics, core thermal-hydraulics, and plant chemistry disciplines. The various level of detail required in the assessment of crud risk is based on guidelines and a checklist developed and published by EPRI. In the guidelines EPRI defined two levels of assessment where core neutronics and thermal-hydraulics are coupled with plant chemistry. These two levels are called Levels III and IV in the EPRI guidelines.
AREVA developed a process using the standard licensing tools and a number of specialized application tools and interfaces that allow Level III and IV calculations. The Level III calculations are performed using a typical reload licensing subchannel node scale (scale of several centimeters) whereas the Level IV calculations are performed on fractions of a millimeter scale.
This paper provides an overview of the neutronics and thermal-hydraulic process used to perform Levels III and IV assessments and provides some results for a typical B&W-designed 177-fuel assembly operating plant. This same process could be applied to other plant types. The details of the plant chemistry process are not covered in this paper, as they are covered in other publications.
High-Speed Steam Turbine Systems for Small-Scale Power Generation Applications
Authors: Petrov Miroslav P. , Fridh Jens , Göransson Ake , Fransson Torsten H. 
Published: October 2013

Energy utilization from low-grade fuels of either fossil or renewable origin, or from medium-temperature heat sources such as solar, industrial waste heat, or small nuclear reactors, for small-scale power generation via steam cycles, can be reasonably enhanced by a simple technology shift. This study evaluates the technical feasibility of a compact power generation package comprising a steam turbine directly coupled to a high-speed alternator delivering around 8–12 MW of electrical power. Commercial or research-phase high-speed electrical generators at MW-scale are reviewed, and a basic thermodynamic design and flow-path analysis of a steam turbine able to drive such a generator is attempted.
High-speed direct drives are winning new grounds due to their abilities to be speed-controlled and to avoid the gearbox otherwise typical for small system drivetrains. These two features may offer a reasonable advantage to conventional drives in terms of higher reliability and better economy.
High-speed alternators with related power electronics are nowadays becoming increasingly available for the MW-size market. A generic 8 to 12 MW synchronous alternator running respectively at 15,000 to 10,000 rpm, have been used as a reference for evaluating the fundamental design of a directly coupled steam turbine prime mover.
The moderate steam parameter concept suits well for converting mid-temperature thermal energy into electrical power with the help of low-tech steam cycles, allowing for distributed electricity production at reasonable costs and efficiency. Steam superheat temperatures below 350°C (660°F) at pressures of maximum 20 bar would keep the steam volumetric flow sufficiently high in order to restrain the turbine losses typical for small-scale turbines, while helping also with simpler certification and safety procedures and using primarily established technology and standard components. The proposed steam turbines designs and their characteristics thereof have been evaluated by computer simulations using the in-house code ProSteam and its sub-procedures AXIAL and VaxCalc, by courtesy of Siemens Industrial Turbomachinery and its steam turbine division located in Finspong, Sweden.
The first results from this study show that high-speed steam turbines of the proposed size and type are possible to design and manufacture based on conventional components, and can be expected to deliver a very satisfactory performance at variable power output.
Experiments and Simulations on the Steady-State and Transient Behavior of Gas/Liquid Interfaces in Impulse Pipes for Hydrostatic Level Measurements
Authors: Schulz Stephan , Hampel Rainer 
Published: October 2013

For Boiling Water Reactors (BWR) and steam generators, the water level is a safety-relevant process variable. The most commonly applied measuring method is based on the calculation of the liquid level from geodetic pressure differences to a reference column of defined height and density. However, transition processes occurring under operational and accident conditions may lead to dynamic changes in the reference level and therefore to fluctuations in the differential pressure signal.
This paper presents experiments and numerical simulations on the steady-state and transient behavior of gas/liquid phase boundaries in “Zero Chamber Level Vessels” (ZCLV). In these slightly inclined miniature tubes, the constant reference level is provided by surface tension forces and the capillary effect, respectively.
To investigate the basic topology of gas/liquid interfaces under simplified conditions (environmental parameters, no heat transfer), a test facility with optical access was developed. The construction allows for variations of the inner tube diameter, inclination angle and liquid mass flow rate, respectively. By this means, experiments on phase boundaries were carried out for ethanol/air and water/air. The results provide information about the impact of geometry parameters and their interactions on the interface topology. In addition, the dynamic draining of excess liquid mass at the free end of the tube and at artificial weld seams, which is supposed to be the reason for temperature fluctuations observed in ZCLV during power operation of BWR, was experimentally analyzed.
The measurements represent the basis for an experimental validation and optimizations of the numerical flow code ANSYS CFX 12.0. In the next step, water/vapor phase boundaries at 286 °C and 70 bar will be investigated by applying x-ray radiography to a scale model. The results will be discussed in context with the hydrostatic level measurement in BWR.
Pressure Drop and Flow Pulsation in Tight-Lattice Rod Bundle
Authors: Lee Chi Young , Shin Chang Hwan , Park Ju Yong , Oh Dong Seok , Chun Tae Hyun , In Wang Kee 
Published: October 2013

In order to ensure the compactness and high-power density of a nuclear power reactor, the research on tight-lattice fuel bundle, with a narrow gap distance between fuels, has been highlighted. Recently, KAERI (Korea Atomic Energy Research Institute) has been developing dual-cooled annular fuel to increase a significant amount of the reactor power in OPR1000 (Optimized Power Reactor), a PWR (Pressurized Water Reactor) optimized in the Republic of Korea. The dual-cooled annular fuel is configured to allow a coolant flow through the inner channel as well as the outer channel. To introduce the dual-cooled annular fuel to OPR1000 is aiming at increasing the reactor power by 20% and reducing the fuel-pellet temperature by 30%, as compared to the cylindrical solid fuel, without a change in reactor components. In such a case, due to larger outer diameter of a dual-cooled annular fuel, the dual-cooled annular fuel assembly exhibits a smaller P/D (Pitch-to-Diameter ratio) than the conventional cylindrical solid fuel assembly. In other words, the dual-cooled annular fuel array becomes the tight-lattice fuel bundle configuration, and such a change in P/D can significantly affect the thermal-hydraulic characteristics in nuclear reactor core.
In this paper, the pressure drop and flow pulsation in tight-lattice rod bundle were investigated. As the test sections, the tight-lattice rod bundle of P/D = 1.08 was prepared with the regular rod bundle of P/D = 1.35. The friction factors in P/D = 1.08 appeared smaller than those in P/D = 1.35. For P/D = 1.08, the twist-vane spacer grid became the larger pressure loss coefficients than the plain spacer grid. In P/D = 1.08, the flow pulsation, quasi-periodic oscillating flow motion, was visualized successfully by PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques. The peak frequency and power spectral density of flow pulsation increased with increasing the Reynolds number. Our belief is that this work can contribute to a design of nuclear reactor with tight-lattice fuel bundle for compactness and power-uprate and a further understanding of the coolant mixing phenomena in a nuclear fuel assembly.
A Probabilistic Assessment of Flow-Accelerated Corrosion Rate in Pipe Bends With Unknown Initial Thickness
Authors: Pandey M. D. , Lu D. , Riznic Jovica 
Published: October 2013

In CANDU reactors, the feeder pipe system supplies the primary coolant from the reactor core to steam generator. The flow-accelerated corrosion (FAC) of the extrados of the first bend of the outlet feeder is a life limiting factor. Therefore, accurate prediction of FAC rate is needed to determine the scope of inspection and replacement of feeder pipes during an outage. To estimate the FAC rate, the initial wall thickness of the extrados of the bend is required. However, initial wall thickness is not known precisely, because the bending of pipe during installation causes the thinning of extrados section. Another difficult is the sizing error or noise in the wall thickness measurements introduced by the inspection probes.
The objective of this paper is to develop a sound probabilistic method for the FAC rate estimation considering uncertainties in the initial wall thickness of bend and the sizing error (or noise). A Bayesian model is developed and applied to predict the end of life of the outlet bend on a risk-informed basis. A practical case study is presented to illustrate the use of the proposed method.
Break Size Effects on CET Response in an Upper Head SBLOCA Transient
Authors: Gallardo S. , Querol A. , Verdú G. 
Published: October 2013

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced.
This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS).
The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.