bluuebunny/crossref_metadata_2025 huggingface dataset:
Evolution of Microstructure, Microtexture and Texture in Dilute Zirconium Based Structural Components of Pressurised Heavy Water Reactors
Authors: Srivastava D. , Mani Krishna K. V. , Neogy S. , Dey G. K. , Samajdar I. , Banerjee S.
Published: February 2010

Microstructure and texture of Zr alloys plays a significant role in deciding the material properties and its performance in nuclear reactor. In order to understand the development of microstructure, it is important to know the phase transformations mechanism in these alloys. Dilute Zr-Nb alloys exhibit a range of diffusional, diffusion less and hybrid phase transformations which includes martensitic omega and hydride transformation. With these points in view detail studies on the physical metallurgy aspects of the binary, ternary and quaternary Zirconium base alloys has been carried out in this study. Electron Diffraction and X-ray diffraction techniques have been used to determine the phases, defects and their crystallographic features. Hydriding is most important corrosion problem in the Zirconium cladding alloys. The mechanism of hydride phase transformation and their crystallographic and microstructural aspects has been presented in detail in some Zirconium alloys. Zirconium alloys being crystallographically asymmetric it shows deformation and annealing texture. Result of the studies on bulk and micro texture in some zirconium alloys and some micro-textural aspects on hydride formation is presented in this paper.
Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels in Japan
Authors: Yamashita Norimichi , Iwasaki Masanobu , Dozaki Koji , Soneda Naoki
Published: February 2010

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material ageing issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation based on the appropriate recognition of the neutron irradiation embrittlement is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complementally use of surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old BWR RPV steels with high impurity content. Some discrepancies of PWR RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201–2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December, 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long-term surveillance program”, which is to be applied for the operation beyond 40 years. The use of the reloading, re-irradiation and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201–2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.
Managing CANDU Plant Ageing Using a Risk Informed Engineering Approach
Authors: Dinnie Keith S.
Published: February 2010

A common challenge to operators of plants nearing the end of design life or undergoing life extension is to maintain safe and economic operation where multiple components are degrading simultaneously due to ageing effects. Typically, the management of ageing is carried out on a component-by-component basis but the real challenge is to ensure that the collective impacts of degradation are controlled such that the risk posed by continued operation of the plant remains acceptably small. The strategy being proposed to the Canadian industry is to use a risk-informed approach that derives failure frequency targets for individual components in a manner that ensures that the total risk remains within established limits. These frequency limits can then be embodied in fitness for service guidance to manage component reliability. The approach is to use the component importance measures in the plant PSA to derive the failure frequency that would result in a risk contribution of 1E−06 or less per reactor-year to the severe core damage frequency. Given that the safety goal limit for existing plants is 1E−04 per reactor-year, this would allow a number of components to be managed in this way without a significant increase in severe core damage frequency relative to the limit, where a cumulative increase of 1E−05 per reactor-year or more would be considered “significant”. A limit is placed on the derived “allowable” failure frequency for any individual component by deterministic considerations, in that the frequency is not permitted to exceed the maximum for the event class for which it was licensed in Canada. The frequency is also reviewed for economic and operability implications to ensure such risks are not unreasonably high. This approach helps to achieve a balanced allocation of inspection and maintenance resources as well as maintaining an adequate safety margin. The paper summarizes some of the challenges facing the current CANDU fleet, and provides examples of how the proposed approach could be applied to selected components. It should be noted that the approach is under consideration by the Canadian industry but is not committed at the present time.
Fragmentation of a Single Molten Metal Droplet Penetrating Sodium Pool: Comparisons of Thermal and Hydrodynamic Fragmentation
Authors: Zhang Zhi-gang , Sugiyama Ken-ichiro
Published: February 2010

In order to characterize the relationship between hydrodynamic and thermal fragmentation of molten metallic fuels with the interaction of the coolant of sodium in the wide range of thermal and hydrodynamic conditions, the present paper focuses on a series of fragmentation experiments of a single molten copper droplets (1∼5 g) with wide ambient Weber number (102 ∼ 614) and superheat condition (15 ∼ 574 °C), which penetrated a sodium pool. The intensive fragmentations of single molten copper droplets were clearly observed even at the instantaneous contact interface temperatures (Ti) below the melting point besides higher Ti. The size distributions of molten droplet with high Weber number somewhat incline to be less than those with low Weber number in the same thermal condition. When approximately Wea>200, the effect of hydrodynamic fragmentation becomes predominant over that of thermal fragmentation in the low superheat condition. In the high Weber number or high superheating condition separately, the size distributions of the single metal droplet and jet reported by Schins’ group with different mass of thousand times can keep very similar. Moreover, the values of Dm/D0 must be quite small. The relatively larger size distributions of the single metal droplet and the jet reported by Nishimura et al. will only arise in the both low Weber number and low superheating condition.
Flow Accelerated Corrosion Program in the Belgian Nuclear Power Plants
Authors: Bogaert Anne-Sophie , Desmet Michel , Gendebien Arnaud
Published: February 2010

Since the Surry-accident of 1986, Electrabel and Tractebel Engineering have performed extensive ultrasonic inspection campaigns to detect pipe wall thinning due to Flow Accelerated Corrosion (FAC) in the Balance-of-Plant systems of the seven Belgian nuclear power plants. Since 2000 EPRI’s predictive software CHECWORKS is used as a means to focus future inspections on the most susceptible components. In 2005, Tractebel Engineering participated in a benchmark set-up by the Framatome Owners Group (FROG) that compared the different FAC predictive models used by the FROG members. In 2006, Electrabel and Tractebel Engineering decided to perform an assessment of the way in which the follow-up of Flow Accelerated Corrosion (FAC) is done in the Belgian nuclear plants. This paper summarizes the Flow Accelerated Corrosion program in the Belgian nuclear plants as well as some of the main aspects of the Flow Accelerated Corrosion management, including the use of a predictive software, the method of inspections and the actions taken to keep the FAC program up to date.
European Nuclear Pressure Components: What Kind of Rules Do We Need?
Authors: Daoust Ph. , Detroux P. , Weverbergh J.
Published: February 2010

In Europe, Nuclear Pressure Components are subject to different sets of rules in different countries. In the 1970s, small and medium-sized countries generally adopted a complete American set of rules transposed into national legislation. Larger nations that already had more detailed legislation preferred to introduce a full set of their own national regulations. In the early 2000s, the Pressure Equipment Directive was introduced (PED) in each EU country. Its main aims were to standardize the approaches, to strengthen the evaluation of safety risks and to open up the market. This paper summarizes the regulatory developments. In connection with potential nuclear investments by the GDF SUEZ group in France, different studies and comparisons have been carried out within TE. The paper gives the principal conclusions and recommendations from the ongoing evaluation of the rules.
Fatigue Crack Growth Life Prediction of an Extruded Thick-Walled Cylinder With Internal Axial Crack
Authors: Malik M. A. , Salam I. , Muhammad W.
Published: February 2010

The integrity of nearly all engineering structures is threatened by the presence of cracks. Structural failure occurs if a crack larger than a critical size exists. Although most well designed structures initially contain no critical cracks, subcritical cracks can grow to failure under fatigue loading, called fatigue crack propagation. Fatigue failure, that is, failure under repeated or cyclic loading, is a serious concern of engineering design. Under fatigue loading, the component may fail at a stress level that is far below its yield strength. The extruded materials are extensively used in chemical, food and nuclear industry and generally offer a unique combination of strength and freedom with regard to design solutions. During extrusion, material flow occurs in the direction of applied force and as a result microstructure changes. The process ultimately induces variation in the mechanical properties when tested along or across the extrusion direction. In present study, the fatigue crack propagation in a thick-walled cylinder is analyzed through detailed experimental work and finite element analysis and the fatigue crack growth life of the cylinder, with internal axial crack, has been predicted. Fatigue crack growth tests were conducted on middle tension, M(T), samples. The fatigue crack growth data of the cylinder is obtained in the transverse direction which experiences the maximum tensile stress (hoop stress). The tests were conducted at ambient temperature in air atmosphere. The data thus obtained include the fatigue crack growth and the SN curves. The fatigue crack growth life of the thick-walled cylinder has been predicted with ANSYS structural software. An ANSYS Parametric Design Language (APDL) code has been written to simulate the crack growth process. The actual data obtained from the experimental work has been used for the simulation work.
Next Generation Safety Analysis Methods for SFRs—(3) Thermal Hydraulics Models of COMPASS Code and Experimental Analyses
Authors: Yamamoto Yuichi , Hirano Etsujo , Oue Masaya , Shimizu Sensuke , Shirakawa Noriyuki , Koshizuka Seiichi , Morita Koji , Yamano Hidemasa , Tobita Yoshiharu
Published: February 2010

A computer code, named COMPASS, is being developed employing the Moving Particle Semi-implicit (MPS) method for various complex phenomena of core disruptive accidents (CDAs) in the sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In FYs2006 and 2007 (Japanese Fiscal Year, hereafter), the development of basic functions of COMPASS was completed and fundamental verification calculations were carried out. In FY2007, the integrated verification program using available experimental data for key phenomena in CDAs was also started. In this paper, we show the basic verification calculations for the phase change model of COMPASS and the results of experimental analyses, together with the outline of the formulation of MPS method and the conceptual design of the COMPASS code.
Implications of Nuclear Power Plant Regulation for Future Nuclear Plants Based on Risk and Performance Information
Authors: Modarres Mohammad
Published: February 2010

A largely probabilistic regulatory framework using best estimate, goal-driven, risk-informed, and performance-based methods is proposed. This framework relies on continuous probabilistic assessment of performance of plant systems, structures, components and operators that assure attainment of a broad set of overarching technology-neutral protective, mitigative, and preventive goals in all phases of plant operations. Required levels of performance are set through formal apportionment such that they are consistent with the overarching goals. Regulatory acceptance would be based on the probability bounds or confidence levels with which the required levels of performance have been attained.
A Study of a Small Nuclear Power Plant System for District Heating
Authors: Imamura Mitsuru , Sato Kotaro , Narabayashi Tadashi , Shimazu Yoichiro , Tsuji Masashi
Published: February 2010

We have studied nuclear power plant for district heating. Already some towns and villages in Hokkaido have requested small reactor for district heating. Using existing technology allows us to shorten development period and to keep a lid on development cost. We decided to develop new reactor based on “MUTSU” reactor technology because “MUTSU” had already proved its safety. And this reactor was boron free reactor. It allows plant system to reduce the chemical control system. And moderator temperature coefficient is deeply negative. It means to improve its operability and leads to dependability enhancement. We calculated burn-up calculation of erbium addition fuel. In the result, the core life became about 10 years. And we adapt the cassette type refueling during outage in order to maintain nonproliferation. In the district heating system, a double heat exchanger system enables to response to load change in season. To obtain the acceptance of public, this system has a leak prevention system of radioactive materials to public. And road heating system of low grade heat utilization from turbine condenser leads to improve the heat utilization efficiency. We carried out performance evaluation test of district heating pipeline. Then the heat loss of pipeline is estimated at about 0.440 °C/km. This result meets general condition, which is about 1 °C/km. This small plant has passive safety system. It is natural cooling of containment vessel. In case of loss of coolant accident, decay heat can remove by natural convection air cooling after 6 hours. Decay heat within 6 hours can remove by evaporative heat transfer of pool on containment vessel.
Boron Transient Analysis by the GRS Coupled Code Systems QUABOX-CUBBOX/ATHLET and TORT-TD/ATHLET
Authors: Velkov K. , Seubert A. , Pasichnyk I. , Pautz A.
Published: February 2010

The application of modern coupled thermal-hydraulic neutron-kinetic code systems is state of the art for performing safety analyses. In this paper, a radially asymmetric boron dilution transient in a PWR MOX/UO2 core is defined. The transient is analysed using the coupled code system QUABOX-CUBBOX/ATHLET which is based on fuel-assembly coarse mesh diffusion theory, and the results are compared to the solution obtained with the transport theory-based coupled code system TORT-TD/ATHLET with a pin-wise representation of the core. The aim of this study is to investigate the impact of the transport approach and the diffusion approximation on the simulation results.
CFD Investigation of Experimental Data Proposed to be a Validation Data Set
Authors: Johnson Richard W.
Published: February 2010

The U. S. Department of Energy (DOE) is currently supporting the development of a next generation nuclear plant (NGNP). The NGNP is based on the very high temperature reactor (VHTR), which is a Generation IV gas-cooled reactor concept that will use helium as the coolant. Computational fluid dynamics (CFD) calculations are to be employed to estimate the details of the flow and heat transfer in the lower plenum where the heated coolant empties before exiting the reactor vessel. While it is expected that CFD will be able to provide detailed information about the flow, it must be validated using experimental data. Detailed experimental data have been taken in the INL’s matched index of refraction (MIR) facility of a scaled model of a section of the prismatic VHTR lower plenum. The present article examines the data that were taken to investigate the suitability of such data to be a validation data set for CFD calculations. CFD calculations were made to compare with the MIR data to explore potential issues and make recommendations regarding the experiment.
Crack Growth Model for the Probabilistic Assessment of Inspection Strategies for Steam Generator Tubes
Authors: Wolf Brian , Revankar Shripad T. , Riznic Jovica R.
Published: February 2010

Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits as an approach used to steam generator tube fitness-for-service assessment has begun to be used increasingly in recent years throughout the nuclear power industry. One of the important topics in the assessment is the crack growth model that predicts the growth of crack size with time. The plot of the log of change in crack length per cycle, against the log of stress intensity factor range shows three distinct regions where in Regions I and III the change in crack length per cycle increases faster with stress intensity factor. On the other hand Region II shows linear variation. While effective for predicting Region II of fatigue crack growth, “Paris law” is often not properly used in making life predictions because a majority of the fatigue life of a specimen occurs in Region I. Physical models for Region I growth have not yet been successful. Recently substantial effort is made in modeling crack growth model with experimental data for Regions I and II. The present work reviews recent work in predicting the fatigue crack growth rate in Region I and II. The characteristics of such models are examined and their limitations are discussed in context with steam generator tube crack growth. Recommendations are made on crack growth models for region I and II.
Research on Flow Characteristics of Supercritical CO2 Axial Compressor Blades by CFD Analysis
Authors: Takagi Kazuhisa , Muto Yasushi , Ishizuka Takao , Aritomi Masanori , Kikura Hiroshige
Published: February 2010

A supercritical CO2 gas turbine of 20MPa is suitable to couple with the Na-cooled fast reactor since Na - CO2 reaction is mild at the outlet temperature of 800K, the cycle thermal efficiency is relatively high and the size of CO2 gas turbine is very compact. In this gas turbine cycle, a compressor operates near the critical point. The property of CO2 and then the behavior of compressible flow near the critical point changes very sharply. So far, such a behavior is not examined sufficiently. Then, it is important to clarify compressible flow near the critical point. In this paper, an aerodynamic design of the axial supercritical CO2 compressor for this system has been carried out based on the aerodynamic design method of Cohen. The compressor design point was selected to achieve the maximum cycle thermal efficiency of 43.8%. For this point, the compressor design conditions were determined. They are a mass flow rate of 2030kg/s, an inlet temperature of 308K, an inlet static pressure of 8.26MPa, an outlet static pressure of 20.6MPa and a rotational speed of 3600rpm. The mean radius was constant through axial direction. The design point was determined so as to keep the diffusion factor and blade stress within the allowable limits. Number of stages and an expected adiabatic efficiency was 10 and 87%, respectively. CFD analyses by FLUENT have been done for this compressor blade. The blade model consists of one set of a guide vane, a rotor blade and a stator blade. First, analysis was conducted for the blade, whose condition is remote from the critical point and the possibility of divergence is very small. Secondly, the analyses were conducted for the conditions nearer to the critical point. Gradually, divergence of calculation was encountered. The analyses were conducted under the assumption both of the real gas properties and also of the ideal gas properties. Convergence was relatively easy for the latter. Main output of calculation is a value of the mass flow rate, which agreed very well with those of the aerodynamic design. Absolute velocity distributions, relative velocity distributions and static pressure distributions surrounding rotor blade and stator blade were obtained. The characteristics of these distributions were consistent with those of the fundamental theory and these analyses were justified.
Aspects of Radiation Damage Effects in Iron-Chromium Alloys From the Point of View of Atomistic Modelling
Authors: Terentyev Dmitry , Bonny Giovanni , Castin Nicolas
Published: February 2010

Fe-Cr alloys are the basis of high-Cr ferritic steels, which are the candidate structural materials for near future power plants. Recently, a significant effort has been put in the development of theoretical models dealing with the response of Fe-Cr alloys to irradiation. Here, we give a brief overview of the current level of understanding of radiation damage in Fe-Cr alloys, based on the most recent results. In particular, we review and summarize data obtained using different atomistic modelling techniques in order to refine the most important findings achieved over the past few years.
The Vanishing Influence of Microstructure on Growing Short Cracks
Authors: Simonovski I. , Cizelj L.
Published: February 2010

Microstructural features play a prominent role in the initialization and propagation of microstructurally short cracks. These features include crystallographic orientations of grains, grain boundaries, inclusions, voids, material phases, etc. Their influence is expected to vanish both with increased distance from the crack tip as well as with increased crack length. Current engineering fracture mechanics techniques do not account for these features and can therefore only be applied to sufficiently long cracks. From the engineering point of view it is therefore quite useful to have an estimate of this length. In this paper we build upon the authors’ previous work to estimate decreasing influence from the surrounding microstructural features with increasing crack length. To achieve this, a model containing a large number of randomly sized, shaped and oriented grains is employed. The grain structure is modeled using a Voronoi tessellation. A series of cracks with lengths from one to several grain lengths are inserted into the model, extending from a grain at the surface towards the interior of the model. The crack tip opening displacements (CTOD) are estimated and statistically analyzed for a series of random crystallographic orientation sets assigned to the grains adjacent to the crack. Anisotropic elasticity and crystal plasticity constitutive models are employed at the grain size scale. A rapid decrease of the CTOD scatter is observed with increasing crack length showing a clear tendency to stabilize at standard deviation of about 5% at the effective crack length of about 6 average grain sizes (0.3 mm in AISI 316 stainless steel. A crack with the length exceeding about 10 average grain sizes can be safely taken as a macroscopic crack.
Construction Aspects of the Doel 1 and 2 SGR Projects
Authors: Kroes Bert , Gobert Edmond , Delhaye Xavier , Devolder Peter , Sonville Michel
Published: February 2010

The Doel 1 and 2 PWR Nuclear Power Stations are the oldest commercially operating units in Belgium and the last to replace their two Steam Generators. The Doel 2 Steam Generators were replaced in 2004 and those of Doel 1 will be replaced late 2009. The replacement poses a particular challenge as these are the only stations in Belgium requiring the creation of primary and secondary containment opening for the SG exchange operation. Other construction challenges result from the a-typical SG support configuration which dates from the period well before the more or less standardized support configuration as used for later PWR units. The current paper discusses the construction approaches selected to facilitate the exchange operation and to minimize the outage duration and radiation worker exposure. The main particularities of the construction effort concern the secondary containment opening and closing using a structural formwork assembly, the use of containment platforms hanging inside the primary containment allowing for parallel primary and secondary containment reconstruction and the de-activation of some of the primary coolant piping and SG restraints following the licensing acceptance of the Leak Before Break concept for the primary piping. The specific construction options that made the Doel 2 replacement a success will be presented in this paper.
Advanced Recycling Reactor Design Study
Authors: Mito Makoto , Kunishima Shigeru , Stein Kim O. , Ikeda Kazumi , Sato Horoyuki
Published: February 2010

The Advanced Recycling Reactor (ARR) design study sponsored by DOE of USA has been conducted [1]. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. The International Nuclear Recycling Alliance (INRA) takes advantage of international experience and agreed to use Japan Sodium-cooled Fast Reactor (JSFR) [2] as reference for Funding Opportunity Announcement (FOA) studies [1]. Since the scale-up factor of two is acceptable increase from manufacturing and licensing points of view, INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. Japan has conducted R&Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The ARR design is based on such the technology enhancements that it can benefit from this development effort and the ARR3 can become cost competitive with the similar sized LWRs. Major features of key technology enhancements are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45–51 kg/TWeh. The ARR1 consists of a reactor building (including reactor auxiliary facilities and electrical / control systems), a turbine building, and a reprocessing building. The dimensions of the overall reactor building will be 46.1 m (W) × 72.8 m (L) × 70.3 m (H), and the volume of the building will be approximately 180,000 m3.
Assessment of TRANSURANUS Fuel Performance Code Against Studsvik Inter-Ramp BWR Database
Authors: Adorni Martina , Del Nevo Alessandro , Van Uffelen Paul , Oriolo Francesco , D’Auria Francesco
Published: February 2010

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation - International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions “v1m1j07” and “v1m1j08”. The starting point of the activity is the availability of a “new” transient fission gas release model, the “TFGR model”, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.
Ultimate Heat Sink Sizing Requirements on a Month-to-Month Basis
Authors: Zheng Dong , Vieira Allen T. , Jarvis Julie M. , Emsurak George P.
Published: February 2010

The Ultimate Heat Sink (UHS) of a nuclear power plant is a complex cooling water system which serves the plant during normal and accident conditions. For some next generation nuclear plants, the UHS sizing is a major design and licensing analysis task. The analysis involves detailed modeling of the transient heat loads and the selection of worst-case meteorological data for the plant site. The UHS sizing requirements for a representative next generation nuclear power plant are evaluated on a month-to-month basis. This paper assesses the UHS water requirement for each month of year. The UHS analysis for a representative next generation nuclear plant with mechanical draft cooling towers and a water basin is used to determine the maximum evaporation of the basin for the worst-case meteorological data on a month-to-month basis. To size the cooling tower basin, automated methods have been developed which determine the highest evaporative losses from the basin and highest basin temperature over a 30-day design basis accident period. This paper also evaluates the month-to-month basin temperature changes. This assessment is done for a representative next generation nuclear power plant and considers the monthly historical meteorological data over 45 years. The result of this assessment of monthly UHS water requirement is of interest in assessing the margin in the UHS design. This monthly assessment is also useful in demonstrating that the automated methods used to establish the limiting 30-day meteorological condition are indeed accurate. In addition, these results may be useful in helping to plan plant maintenance activities.
Temperature Fluctuations in a CANDU Moderator Test Facility
Authors: Farhadi F. , Ashgriz N. , Kwee M. , Girard R. , Parlatan Y. , Ali M.
Published: February 2010

A two-dimensional numerical model of the CANDU Moderator Test Facility (MTF) is presented. The results show fluctuations in velocity and temperature. The thermal solution of the MTF model indicates that the buoyancy forces dominate in the inner core of the tank, whereas, the inlet jet induced inertial forces dominate the outer edges of the tank. The interaction between these flows forms a complex and unstable flow structure within the tank. The largest flow fluctuations occur outside the tube bank where the inlet jets continue to flow, and around the top of the tank where the two inlet jets impinge on each other. The high velocity gradients between the inlet jets and the surrounding slower moving moderator fluid generate small vortices with low fluctuation amplitude but high frequencies.
Updated Probabilistic Analysis of External Events Data and Safety of the Nuclear Power Plant in Lithuania
Authors: Kupciuniene Kristina , Alzbutas Robertas
Published: February 2010

The main purpose of performed work is the probabilistic analysis of extreme external events, which have a potential influence on safety of the present nuclear power plant (NPP) in Lithuania. This analysis can also be related to the future NPP in Lithuania at the same site. At first, the methodology was established for screening out hazardous events, which impact on the present Ignalina NPP safety is not significant. For risk estimation, the following external events were considered in detail: forest fire, external flood, airplane crash, extreme wind. In order to estimate probabilities of hazards occurrence the statistical data related to various external events were collected, mathematical models were constructed and probabilities of these events occurrence were determined. Statistical, meteorological and other updated data from the Republic of Lithuania has been used to estimate probability of the most important hazardous events. Due to many factors affecting the inaccuracy of any result it is not enough to calculate only the estimate of the probability of the event. There is a need also to evaluate errors or variations of result made by such estimation. For such purpose uncertainty and sensitivity analysis was performed for a probability estimate, i.e. frequency of the event. This analysis determines the parameters that have the greatest influence on the probabilistic model results. The modelling of external events occurrence and its impact on the NPP is significant in order to evaluate the safety of NPP and to prevent failure of the systems important to the safety.
Design and Development of Diverse Safety Rod and Its Drive Mechanism for PFBR
Authors: Vijayashree R. , ChellaPandi P. , Natesan K. , Jalaldeen S. , Chetal S. C. , Raj Baldev
Published: February 2010

Prototype Fast Breeder Reactor (PFBR) is U-PuO2 fuelled sodium cooled Pool type Fast Reactor and it is currently under advanced stage of construction at Kalpakkam, India. The Fast Breeder Test Reactor (FBTR) which is the only fast reactor currently operational in India is having only one shutdown system. However the IAEA and Atomic Energy Regulatory Board (AERB) Guide Lines call for two independent fast acting diverse shutdown systems for the present generation reactors. Hence PFBR is equipped with two independent, fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms and neutron absorbing rods. The two shutdown systems of PFBR are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of PFBR are called as Diverse Safety rods (DSR) and their drive mechanisms are called as Diverse Safety Rod Drive Mechanisms (DSRDM). DSR are normally parked above active core by DSRDM. On receiving scram signal, Electromagnet of DSRDM is de-energised and it facilitates fast shutdown of the reactor by dropping the DSR in to the active core. This paper presents chronological design and development of the prototype DSR and DSRDM starting from the design specifications. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual & detailed design features are explained with the help of figures. Various important design options considered in the initial design stage, choice of final design along with brief explanation for the particular choice are also given for some of the important components. Details on material of construction are given at appropriate places. Details on various analysis such as large displacement analysis for buckling, bending analysis for determining reactive forces and friction in the mechanism, thermal stress analysis of electromagnet during scram, flow induced vibration analysis of DSRDM and DSR and hydraulic analysis for estimating the pressure drop and drop time of DSR are also given. Test plans for design verification, manufacturing and shop testing experience of prototype systems, and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are also briefed.
Ultrasonic Phased Array Examination of Butt-Fusion Joints in High-Density Polyethylene
Authors: Frederick Caleb , Porter Allen , Zimmerman David
Published: February 2010

With the increasing use of High-Density Polyethylene (HDPE) piping for nuclear applications, nondestructive evaluation is an important tool for evaluation of the integrity in fused joints. This paper will discuss the method of using Ultrasonic Phased Array for inspecting Butt-Fusion (BF) joints in HDPE piping. The benefit of Phased Array is the ability to perform a volumetric inspection using multiple angles which greatly increases the probability of detection of defects, and allows the data to be analyzed using a representative 2-dimentional image of the joint [1]. It has been determined that successfully producing BF joints is highly dependent on environmental and mating-surface conditions. The primary defects of concern are lack-of-fusion (LOF), an area of the joint where there is no bond [2], cold fusion (CF), an area of partial bond, and inclusion. Phased Array has successfully demonstrated the ability of detecting and characterizing these defects using low frequency ultrasound. Factors addressed include joint location, wall thickness, material temperature, transducer wedge material, and manual vs. automated data acquisition.
Application of the Relative Power Contribution Methodology to the Analysis of a Control System Failure
Authors: Castillo-Dura´n Rogelio , Ortiz-Villafuerte Javier , Amador-Garci´a Rodolfo , del-Valle-Gallegos Edmundo , Palacios-Herna´ndez Javier C. , Calleros-Micheland Gabriel
Published: February 2010

The Relative Power Contribution methodology has been applied to delineate the initiating event leading to a BWR transient. Diverse reactor signals were analyzed to calculate the coefficients required on the relative power contribution method. Those coefficients were computed from an autoregressive multivariable model. Among the signals used in the analysis of the transient event are total flow through the core, pressure drop across the core, feedwater flow, and reactor power. Analyses of the same type of transient event showed a resonance of the main event frequency on the range within which it has been considered and observed frequencies related to some failures of certain control systems of a nuclear power plant. Those analyses employed the short-time Fourier transform or the power spectral density, for time-frequency and frequency-only domains, respectively. In this work, the same value of the frequency of the resonance mentioned above was obtained through the relative power contribution analysis, but, furthermore it was found that the feedwater flow behavior had an important impact on the transient event, and also that the transient event was not initiated by a reactivity-related instability.
Analyses of the Failed Impulsion Lines of NPP Dukovany Steam Generator
Authors: Postler M. , Burda J. , Tkadlcˇi´k P.
Published: February 2010

Over the years the cracks have been detected on the impulsion lines of all steam generators of the NPP Dukovany. The tubing is made from the austenitic stainless steel. These lines are designed for purposes of measurement and also containment of possible leaks, e.g. between the primary and secondary steam generator seals. Their integrity is periodically tested during each outage and if the leak is detected the tube segment must be cut out and the new one is welded in place. Most of the time the tubes are “dry”, i.e. no medium is flowing inside. Due to confined space the movement of persons around the steam generator is difficult and often the relatively subtle impulsion lines have been subjected to stresses leading to their bending. This deformation has then led to evolution of the failure. To address the issue the original manufacturing, test and control procedures for the impulsion tubing have been studied. Several cracked tubes have been analyzed thoroughly to find the root cause and offer possible remedies.
Bayesian Determination of Sample Sizes for Inspections of Feeder Hangers
Authors: Hoppe Fred M. , Riznic Jovica
Published: February 2010

The purpose of this paper is to show how a Bayesian approach may be used to determine sample sizes for future inspections of hangers in feeder tubes at nuclear plants. Predictive distributions are obtained and tables given for the number of fretting locations in the uninspected feeder population. If, upon inspection, the number of frets observed is too large for a specified confidence level, then the tables can be used to determine how many additional feeders to inspect.
Supercritical Water Heat-Transfer Correlation for Vertical Bare Tubes
Authors: Mokry Sarah , Gospodinov Yevgeniy , Pioro Igor , Kirillov Pavel
Published: February 2010

This paper presents an analysis of heat transfer to supercritical water in bare vertical tubes. A large set of experimental data, obtained in Russia, was analyzed and an updated heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, the new correlation presented in this paper can be used for preliminary heat-transfer calculations in SCWR fuel channels. The experimental dataset was obtained for supercritical water flowing upward in a 4-m-long vertical bare tube. The data was collected at pressures of about 24 MPa for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values for mass flux ranged from 200–1500 kg/m2s, for heat flux up to 1250 kW/m2 and inlet temperatures from 320 to 350°C. Previous study (Pioro et al., 2008) confirmed that there are three heat-transfer regimes for forced convective heat transfer to water flowing inside tubes at supercritical pressures: (1) Normal heat-transfer regime; (2) Deteriorated heat-transfer regime, characterized by lower than expected heat transfer coefficients (HTCs) (i.e., higher than expected wall temperatures) than in the normal heat-transfer regime; and (3) Improved heat-transfer regime with higher-than-expected HTC values, and thus lower values of wall temperature within some part of a test section compared to those of the normal heat-transfer regime. The HTC data were compared to those values calculated with the Dittus-Boelter and Bishop et al. correlations. The comparison showed that the Bishop et al. correlation represents more closely HTC profiles along the heated length of the tube than the Dittus-Boelter correlation. The latter correlation deviates significantly from experimental data within the pseudocritical range. However, outside the pseudocritical region, the Dittus-Boelter correlation can predict closely experimental HTCs. It should be noted that neither of these correlations can be used for prediction of HTCs within the deteriorated heat-transfer regime. An updated heat-transfer correlation is presented in this paper for forced convective heat transfer in the normal heat-transfer regime to supercritical water in a bare vertical tube. It has demonstrated a good fit (±25%) for the analyzed dataset. This correlation can be used for future comparisons with other independent datasets, with bundled data, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids.
Corrosion Behaviour of Steels in Lead-Bismuth Eutectic (LBE) in the LINCE Loop Up to 5000 h
Authors: Marti´n-Mun˜oz F. J. , Soler-Crespo L. , Go´mez-Bricen˜o D.
Published: February 2010

Lead bismuth eutectic (LBE) has been proposed as a candidate for coolant and/or as spallation target material of an Accelerator Driven Systems (ADS) and as coolant for a Generation IV reactor, the Lead-cooled Fast Reactor (LFR). Lead and lead alloys, such as lead-bismuth eutectic (LBE), are efficient heat transfer media because of high thermal conductivity and high heat capacity. However, one of the critical issues in the design of these systems is the compatibility of steels with LBE. Therefore, it is important to know the behaviour of steels under conditions similar to which can be expected during the course of reactor operation. Corrosion tests using a liquid metal loop provide result very useful. Forced convection loop in CIEMAT, LINCE, was designed for long-term corrosion tests in LBE. It consists in a centrifugal pump, an exchange heater, a refrigerator, a storage tank, two tests areas (hot and cold) and tubing to connect the previous components. The volume of LBE is 300 litres and the flow rate approximately 1m/s. The maximum temperature is 450°C and the minimum 300°C, which means a temperature gradient of 150°C, and it is controlled by 25 temperature regulators. The electrical power is 80kW. An oxygen control system (OCS) has been implemented in the loop. The corrosion behaviour of AISI 316L and T91 steels is being investigated in flowing lead-bismuth eutectic (LBE) at 300 and 450 °C for 2000 and 5000 h. The tests are performed in the LINCE loop, under controlled oxygen activity in LBE. The results obtained show that neither material exhibited a weight loss after exposure to the flowing LBE at 300°C instead 5000 h. However, probes exposure at 450°C show important weight loss after 5000 h.
The Effects of the Gravity Center on the Hydrodynamic Behavior of a Wave Rider Buoy
Authors: Goodarzi Mohsen
Published: February 2010

A numerical procedure was employed to investigate the dynamic performance of a wave rider buoy. The linear matrix form has been derived to describe the equations of the motion. The hydrostatic and hydrodynamic coefficients of the wave rider buoy have been computed. Three cases of the buoy center of gravity were considered. The numerical results showed that, the stability of the buoy was increased by lowering the buoy center of gravity, but the quality of the dynamic response should be reduced, which is a disadvantage regarding the mission of the buoy.
On the Selection of a Width of a Caisson
Authors: Tung C. C. , Lin J. G.
Published: February 2010

This study is concerned with the selection of the width of a caisson. At present, the width of a caisson is determined based on the consideration that the caisson, struck by a prescribed design impulsive breaking wave force, must remain stable against sliding and overturning. In this study, the problem is formulated in a different way. The equations of equilibrium of the caisson, together with the criteria for the caisson to remain at rest, are used to derive expressions for the selection of the width of the caisson. The criteria of the rest mode are presented graphically using force as the abscissa and the coefficient of friction between the caisson and the base as the ordinate, and the region corresponding to a rest mode is indicated. The present approach gives results equivalent to those of the existing approach and, by treating the coefficient of friction as random, the safety of the caisson is expressed in probability terms.