bluuebunny/crossref_metadata_2025_split huggingface dataset:
Digital Nonlinear Control for a Pressurizer in a Pressurized Water Reactor
Authors: Cappelli M. , Castillo-Toledo B. , Di Gennaro S. , Memmi F. , Sepielli M. 
Published: February 2014

Studying criteria ensuring better implementations of control laws and logics on digital programmable devices is today a crucial point in order to improve performance and safety in nuclear plants of novel design. To this aim, a mathematical model of a PWR is considered, describing the dynamics of the primary circuit. This model is simple enough to allow the determination of a control policy, but it is also sufficiently accurate to capture the nonlinear, time–varying, and switching nature of the system. A dynamic controller for the pressurizer pressure and water level is hence designed. Its main advantage is the reduction of the control response. This latter is a classical drawback, which may cause some troubles to pressure control, due to the long response of temperature sensors.
The digital implementation of the controller has been taken into account to study the important aspects of the influence of the implementation on the performance of the control law. A particularly effective and flexible environment where it is possible to analyze these aspects is Matlab/Simulink, a general–purpose tool for analysis and simulation of multi–domain dynamic systems, practically a standard in control implementations, also in industrial applications. The designed controller ensures a good performance when applied to the model used to derive them, also in the presence of unmodeled uncertainties and disturbances. Its nonlinear switching nature, reflecting the real pressurizer dynamics, ensures better transient behaviors. Since it contains a PI action, it represents an evolution and an improvement with respect to classical PID controllers, usually implemented in standard control actions. To better analyze the control performance, three steps have been considered: first, the designed nonlinear controller has been compared with a linear one. Second, it has been tested on a more detailed model, with more realistic dynamics. Finally, the digital implementation of the controller has been simulated, in order to optimize the electronic implementation on the FPGA–CPU hybrid platform.
A simulation study has been conducted with sampling times in ascending order, to further test its robustness against delays. LabVIEW development tool and technology has been chosen to create the electronic layout and to obtain preliminary results in terms of timing, resources availability and power consumption. Due to the same graphical mode of programming, an immediate link between Matlab/Scilab and LabVIEW is possible in order to optimize the use of each language where it guarantees the best performance.
The methodology here presented could be very useful to better integrate the different features of control engineering and IC design, thus giving the possibility of realizing more accurate simulations and permitting to perform a hardware in the loop (HIL) testing of a wide variety of control algorithms.
Major Improvements of French Mechanical Nuclear Components Codes: RCC-M, RCC-MRx and RSE-M
Authors: Faidy Claude 
Published: February 2014

Inside the French Association for Design, Construction and Surveillance Rules of Nuclear Power Plant Components (AFCEN) a set of 7 Codes has been developed and are periodically updated.
This paper will review the major improvements in the past 3 years of the mechanical component codes:
RCC-M: Rules for Design and Construction of Mechanical Components for PWR Nuclear Islands
RCC-MRx: Rules for Design and Construction of Mechanical Components for High Temperature, Experimental and fusion reactors
RSE-M: Rules for surveillance and ISI of Mechanical Components for PWR Nuclear Islands
After a general overview of the way of developments of these Codes, the paper will present the major improvements of these Codes since 2007, many of them have a complete new edition issued in 2010 or 2012.
The conclusion of the paper will present the major objectives and topics of development of these Codes for new international project under the pressure of Code Harmonization by Regulators (MDEP) Standards Development Organization Board (SDO) and Industry organization (CORDEL).
Water Use Management Challenges in Power Generation
Authors: Ng Kit Y. , Wan Ping K. 
Published: February 2014

Shrinking fresh water resources and increasing competing demands for water among other users have been a growing concern in many parts of the world, which compels the challenges in water use management faced by both operating and new power generation facilities. For new plants in the planning stage, the ability to demonstrate the availability of water and develop practical and achievable water use management strategies to support the long-term operation of the proposed facility in a sustainable manner has become one of the key elements in the site selection process.
The water demands of particular concerns for the power industry come primarily from cooling water needs, which traditionally constitute the majority of plant water use during the operation phase. This paper examines the water use management challenges confronted by new nuclear power plants in site selection and licensing stages as well as by existing plants in the operating stage in the United States. Using an example, it discusses the types of adaptive approaches on the selection of the heat dissipation systems and designs of cooling water systems that can be adopted by new plants to mitigate the declining water availability, and the related environmental and regulatory challenges. The use of modeling tools to estimate cooling water consumptions and availability and predict environmental impacts in the development of sustainable water management strategies are also illustrated.
Assessing the Design Margins for an Ultimate Heat Sink Sizing
Authors: Zheng Dong , Jarvis Julie M. , Vieira Allen T. 
Published: February 2014

The Ultimate Heat Sink (UHS) is a large body of water supply that can be used to cool vital nuclear power plant systems during normal operation and for accident conditions. Due to more stringent environmental and water permit requirements, many new nuclear design proposals have selected the relatively smaller sized mechanical-draft cooling tower with a basin for their UHS. UHS sizing analysis is a critical licensing task for some new generation nuclear power plants Combined Operating License Applications (COLA).
In this paper, a potential UHS is sized for a representative new generation nuclear power plant considering worst case design inputs and modeling assumptions. Over 30 years of historical site meteorological data are processed using an automated technique to identify limiting conditions based on resulting worst UHS design parameters, such as the maximum basin evaporative water loss and the highest basin temperature. The impacts of the cooling tower entrance recirculation effect to these design parameters are also investigated. This paper models the transient plant heat loads in detail for various design basis accident conditions. The large-break LOCA heat load is determined to be bounding for the basin evaporative water loss, while a small-break LOCA heat load may result in the highest basin water temperature. This paper also illustrates that the bounding basin water temperature can result when the peak wet bulb temperature is coincident with the peak UHS heat load. The results of this paper are of interest for new generation nuclear power plants as the paper determines impacts of limiting conditions in assessing the design margins for UHS sizing.
The Experimental Approach to Observe Atomistic Behaviors of Metallic Surfaces by a Linked Facility Between an Ion Accelerator and a STM
Authors: Hirabayashi Junichi , Sato Motohiro , Murakami Kenta , Okita Taira 
Published: February 2014

This study focuses on an experimental approach to quantifying irradiation-induced defects in structural materials used in nuclear power plants at an atomic scale. As a preliminary step, a procedure for observing metallic surfaces using a facility comprising an ion accelerator and a scanning tunneling microscope was developed.
Using this experimental setup, atomic-resolution images of Au(111) surfaces were successfully obtained. A herringbone structure, which is characteristic of reconstructed Au(111) surfaces, was clearly observed, with individual atoms being distinguishable.
Numerical Analysis of Natural Circulation Phenomena of Supercritical Fluids
Authors: Samuel Jeffrey , Harvel Glenn , Pioro Igor 
Published: February 2014

The feasibility of operating with natural circulation as the normal mode of core cooling has been successfully demonstrated for a few small sized nuclear reactors. Natural circulation is being considered for cooling the core of a nuclear reactor under normal operating conditions in several advanced reactor concepts being developed today.
Although studies have been conducted in natural circulation for many decades, using natural circulation as the primary cooling mechanism for nuclear reactors or as a passive safety system requires a comprehensive understanding of local and integral system phenomena, validated benchmark data, accurate predictive tools, and reliability analysis methods. As full-scale experiments of supercritical water are expensive, scaling laws can be applied to develop test matrices using modelling fluids to reproduce similar conditions in a scaled-down experimental thermalhydraulic loop.
The main aim of this work is to understand the natural circulation phenomena by analyzing water and modelling fluids such as Carbon dioxide (CO2) and Freon 134a (R-134a). The use of the modelling fluids at subcritical, pseudocritical and supercritical pressures is discussed along with fluid-to-fluid scaling techniques. The results from a one-dimensional numerical model developed using MATLAB to calculate the steady-state mass flow rate and heat transport characteristics of an experimental natural circulation test loop are presented and analyzed.
Fluorescence Emission Behavior of Eu(III) Sorbed on Calcium Silicate Hydrates Formed With No Dried Process
Authors: Niibori Yuichi , Narita Masayuki , Kirishima Akira , Chida Taiji , Mimura Hitoshi 
Published: February 2014

Calcium silicate hydrate (CSH) is a main component of cement-based material required for constructing the geological repository. As in many countries, since the repository in Japan is constructed below water table, we must consider the interaction of radionuclide with cement materials altered around the repository after the backfill. Using fluorescence emission spectra, so far, the authors have investigated the interaction of Eu(III) (as a chemical analog of Am(III)) with CSH gels formed with no dried process, considering a condition saturated with groundwater. However, in such fluorescence emission behaviors, a deexcitation process of OH vibrators of light water and a quenching effect caused by Eu-Eu energy transfer between Eu atoms incorporated in the CSH gel must be considered.
This study examined the fluorescence emission behavior of Eu(III) sorbed on CSH gels formed with no dried process, by using La(III) (non-fluorescent ions) as a diluent of Eu(III). Furthermore, the CSH samples were synthesized with CaO, SiO2, and heavy water (D2O) as a solvent in order to avoid the obvious deexcitation process of OH vibrators of light water. This study prepared CSH samples with the Ca/Si ratio set to 1.6, 1.0, and 0.5. A 1 mM solution of a given combination of Eu(III) and La(III) (Eu(III) content: 100%, 67%, 50% or 33%) was added into CSH gel sample. The contact time-period of the CSH gel with the Eu(III)/La(III) solution was set to 60 days.
In the results, the peak around 618 nm was split into two peaks of 613 nm and 622 nm in the cases of Ca/Si = 1.0 and 1.6. Then, the peak of 613 nm decreased with increment of Eu(III)/La(III) ratio. This means that the relative intensity of 613 nm is useful to quantify the amount of Eu(III) incorporated in CSH gel. Besides, the intensity peak of 584 nm decayed with increment of Eu/La ratio, suggesting a quenching effect due to Eu-Eu energy transfer. However, the decay behavior of the fluorescence emission did not depend on the Eu/La concentration ratio. That is, such a quenching effect is neglectable. Additionally, the low Ca/Si ratio samples underwent slow attenuation of fluorescence and showed profiles similar to those of high Ca/Si ratio samples. Therefore, low Ca/Si ratio samples also include the reaction forming a complex on the surface of CSH gel with Eu(III). In other words, even if Ca/Si ratio is lower than 1.0, CSH gels would retard the migration of radionuclides released from the repository.
Numerical Simulation of Accident Scenario in HTGR (Pebble Bed Reactor) Using COMSOL® Code
Authors: Peter Geoffrey J. 
Published: February 2014

High Temperature Gas Cooled Reactor (HTGR) development and operation is expanding in the United Kingdom, Russia, USA (Generation IV Reactors), and France (Pebble Bed Modular Reactor, PBMR). A prototype pebble bed reactor producing 10 MW thermal, High Temperature Reactor (HTR-10) is in operation in China. However, the general public remains skeptical of the safety and the perceived dangers of possible accidents.
Of particular concern are blockages caused by local variations in flow and heat transfer that lead to hot spots within the bed. This paper models the accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors using the commercially available computer code COMSOL. Numerical modeling of flow and heat transfer in a packed bed produces an Elliptical Non-Linear Partial Differential equation that requires custom made computer codes. Previously published results obtained from the use of a custom-made verified computer code are limited to one accident scenario and involve considerable modification to study different accident scenarios. Thus the use of a commercially available computer code that can simulate many different accident scenarios is of considerable advantage. Further, this paper compares numerical solutions obtained from custom-made computer code with COMSOL simulation and discusses the advantages and limitations of both codes.
Preliminary Thermal-Hydraulic Analyses for Designing an Experimental Model of a Molten Salt Reactor Concept
Authors: Yamaji Bogdán , Aszódi Attila 
Published: February 2014

Based on the MSFR (Molten Salt Fast Reactor) reactor concept presented within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project preliminary three-dimensional thermal-hydraulic analyses and the discussion of scaled experimental modelling will be presented.
The MSFR concept is a single region, homogeneous liquid fuelled fast reactor. The reactor concept uses fluoride-based molten salts with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up into sixteen identical modules distributed around the core.
Purpose of the three-dimensional computational fluid dynamics (CFD) calculations is to study the possibility of experimental investigation of the fluid flow in the core of the proposed MSFR concept using a scaled model and Particle Image Velocimetry (PIV) flow measurement technique.
First the main properties of the proposed MSFR concept are introduced, and the information on other experimental thermal-hydraulic modelling of different reactors, including MSRE (Molten Salt Reactor Experiment) are summarised.
With a scaled plexiglas MSFR model it would be possible to carry out flow field measurements under laboratory conditions using PIV method. Possible way of scaling are presented and a series of preliminary CFD calculations are discussed. Possibilities and limitations of such scaling and segmenting of a model and the use of water as substitute fluid for the experimental mock-up will be discussed.
Objectives of such a measurement series would be validation, benchmarking of CFD calculations and codes, application of CFD modelling experience in the detailed thermal-hydraulic design of the MSFR concept, possible measurements for the study of specific problems or phenomena, for example refinement of inlet geometry, effects of internal structures, coolant mixing.
Neutronic Analysis and Fuel Cycle Simulation of the MIT Reactor Using MCODE-FM and Experimental Validation
Authors: Sun Kaichao , Ames Michael , Newton Thomas , Hu Lin-wen 
Published: February 2014

A neutronic analysis of the Massachusetts Institute of Technology Research Reactor (MITR) is performed using state-of-the-art computational tools: the continuous-energy Monte Carlo code MCNP5 and the point-depletion code ORIGEN2.2. These codes are externally coupled by the in-house code package, MCODE (MCNP-ORIGEN Coupled Depletion Program), more recently, it being extended to MCODE-FM (Fuel Management). The latter features automated input file generation, data manipulation, and post-processing of the output data for the fuel cycle analysis, so that it is used to simulate the fuel management of the MITR. MCODE-FM also has an optional criticality search algorithm to simulate control blade movement.
The code validation is carried out by comparing the calculated results to experimental data. Two sets of the comparisons are made in the present paper: 1) the Xe-135 reactivity effect during the reactor start-up and shutdown and 2) the thermal and fast neutron flux in an irradiation capsule in the reactor core. Good agreements have been found.
The validated MCODE-FM is therefore useful for neutronic analysis and the fuel cycle simulation of the MITR. The time dependent variation of the key parameters, viz. the control blades’ axial position (maintaining criticality) and the fissile inventory in the fuel, is presented.
Safety Evaluation of Nuclear Plant Fire Using CFD Analysis
Authors: Matsuda Akihiro , Ishibashi Takashi , Tsuchino Susumu 
Published: February 2014

In this paper, investigations of the safety of electrical cables in nuclear plant fire were described by CFD analysis and by flammability tests of power and control cables. For CFD analysis, thermal configurations of a nuclear power plant room in a fire accident was calculated by the Fire Dynamics Simulator (FDS). For flammability tests of cables, the cone-calorimeter test machine was modified to use the cable tray type specimen. The modified cone-calorimeter test machine is able to test the flammability of cables and insulation resistance from room temperature to 800°C.
In the flammability test, electric power and control cables of the polyethylene sheathed cables which had been applied to the nuclear plants were adopt.
The cable specimens of which length were 600mm, were fixed on a cable tray specimen. The temperature of the cable specimens was raised to from 200 to 570°C using the cone-type electrical heater. During the tests, electrical resistance and temperature was measured to evaluate safety of insulation resistance of cables.
As results, the destruction of the insulation resistance of a polyethylene sheathed control cable was observed in the examinations, temperatures of which were 400°C and 500°C.
In the numerical simulation using Fire Dynamics Simulator (FDS), an electrical switchgear room in nuclear plants was supposed as an evaluation model. As the fire source, fire in motor control cabinet in the switchgear room on nearby cable tray was applied to CFD analysis. From the results of numerical simulations, the heat configuration and temperature in the switchgear room was evaluated and radiation heat from fire to the cable trays was calculated.
Considering these results given by flammability tests of electrical cables and CFD analysis, the safety of a switchgear room in nuclear plants was shown.
Development of Experimental Setup of High Temperature Helium Loop and Preparation for In-Pile Operation
Authors: Miletić Marija , Fukač Rostislav , Pioro Igor , Dragunov Alexey 
Published: February 2014

Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuels supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow’s needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts which will utilize thermal neutron spectrum is a Very High Temperature Reactor (VHTR).
This reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behaviour within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Centre Rez, Ltd.
The purpose of the High Temperature Helium Loop (HTHL) is to simulate technical and chemical conditions of VHTR’s coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and Helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.
Ultrasonic Examination of Noble Metal Precipitate Detection in Glass Melter
Authors: Koikegami Hajime , Yasunaga Shigeyoshi , Kakihara Toshiaki , Yanagihara Arisa , Kawai Hiroki , Hatanaka Hiroaki , Maruyama Tadashi 
Published: February 2014

This paper describes the results of ultrasonic examination of the build-up of noble metal precipitates in the glass melter using a wave guide rod. The glass melter is used for the vitrification process of high-level radioactive liquid waste (HLLW).
A high density isotropic graphite T-6 (Ibiden Co., Ltd.) and a nickel-based corrosion and heat resistant alloy 690 were selected as candidate materials of wave guide rod. Ultrasonic propagation experiment with the frequency 0.5, 1.0 and 2.25 MHz was performed using the wave guide rods. The T-6 graphite wave guide rod in the simulant fluid (glycerin) showed a good capability of ultrasonic propagation and signal detection. However, the propagation of ultrasonic wave into molten glass was not possible due to the poor wetting of molten glass with the graphite. On the other hand, it was possible to propagate the ultrasonic wave into molten glass with the wave guide rod made of the nickel-based alloy 690. However, the intensity of reflecting echo signal was very small due to the large ultrasonic attenuation in the nickel-based alloy. The simulation study on the ultrasonic wave propagation using the three dimensional finite element method code ComWave showed good agreement with the experimental results.