Investigation of thin film deposition on stainless steel 304 substrates under different operating conditions

International Nuclear Information System (INIS)

Chowdhury, M A; Nuruzzaman, D M

2016-01-01

In recent times, friction and wear in relation to the deposited carbon films on the steel substrates are important issues for industrial applications. In this research study, solid thin films were deposited on the stainless steel 304 (SS 304) substrates under different operating conditions. In the experiments, natural gas (97.14% methane) was used as a precursor gas in a hot filament thermal chemical vapor deposition (CVD) reactor. Deposition rates on SS 304 substrates were investigated under gas flow rates 0.5 - 3.0 l/min, pressure 20 - 50 torr, gap between activation heater and substrate 3.0 - 6.0 mm and deposition duration 30 - 120 minutes. The obtained results show that there are significant effects of these parameters on the deposition rates on SS 304 within the observed range. Friction coefficient of SS 304 sliding against SS 314 was also investigated under normal loads 5 - 10 N and sliding velocities 0.5 - m/s before and after deposition. The experimental results reveal that in general, frictional values are lower after deposition than that of before deposition. (paper)

Irradiation behavior of German PWR RPV steels under operating conditions

Energy Technology Data Exchange (ETDEWEB)

May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

2011-07-01

In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

Ultrahigh strength martensite–austenite dual-phase steels with ultrafine structure: The response to indentation experiments

International Nuclear Information System (INIS)

Misra, R.D.K.; Venkatsurya, P.; Wu, K.M.; Karjalainen, L.P.

2013-01-01

In medium to high carbon steels, characterized by martensite–austenite microstructure processed by quenching and partitioning process, martensite potentially provides high strength, while austenite provides work hardening [Fu, Wu, and Misra, DOI: 10.1179/1743284712/068]. Given the significant interest in these steels in the steel community, the paper reports for the first time the nanoscale deformation experiments and accompanying microstructural evolution to obtain micromechanical insights into the deformation behavior of ultrahigh strength-high ductility dual-phase steels with significant retained austenite fraction of ∼0.35. During deformation experiments with nanoindenter, dislocations were distributed on several slip systems, whereas strain-induced twinned martensite and twinning were the deformation mechanisms in carbon-enriched and thermally stabilized retained austenite. Furthermore, ultrafine dual-phase steels exhibited high strain rate sensitivity.

Ultrahigh strength martensite-austenite dual-phase steels with ultrafine structure: The response to indentation experiments

Energy Technology Data Exchange (ETDEWEB)

Misra, R.D.K., E-mail: dmisra@louisiana.edu [Laboratory for Excellence in Advanced Steel Research, Center for Structural and Functional Materials, University of Louisiana at Lafayette, P.O. Box 44130, Lafayette, LA 70504 (United States); Venkatsurya, P. [Laboratory for Excellence in Advanced Steel Research, Center for Structural and Functional Materials, University of Louisiana at Lafayette, P.O. Box 44130, Lafayette, LA 70504 (United States); Wu, K.M. [International Research Institute for Steel Technolgy, Wuhan University of Science and Technology, Wuhan 430081 (China); Karjalainen, L.P. [Centre for Advanced Steels Research, University of Oulu, P.O. Box 4200, 90014 Oulu (Finland)

2013-01-10

In medium to high carbon steels, characterized by martensite-austenite microstructure processed by quenching and partitioning process, martensite potentially provides high strength, while austenite provides work hardening [Fu, Wu, and Misra, DOI: 10.1179/1743284712/068]. Given the significant interest in these steels in the steel community, the paper reports for the first time the nanoscale deformation experiments and accompanying microstructural evolution to obtain micromechanical insights into the deformation behavior of ultrahigh strength-high ductility dual-phase steels with significant retained austenite fraction of {approx}0.35. During deformation experiments with nanoindenter, dislocations were distributed on several slip systems, whereas strain-induced twinned martensite and twinning were the deformation mechanisms in carbon-enriched and thermally stabilized retained austenite. Furthermore, ultrafine dual-phase steels exhibited high strain rate sensitivity.

Corrosion of steels in sour gas environments

International Nuclear Information System (INIS)

Twigg, R.J.

1984-03-01

This report presents a study on the effects of sour gas environments on steels. Emphasis is placed on alloys commonly used in the heavy water, sour gas and refining industries. In addition, 'high strength, low alloy' steels, known as 'oil country tubular goods', are included. Reference is made to the effects of hydrogen sulphide environments on austenitic steels and on certain specialty steels. Theories of hydrogen-related cracking mechanisms are outlined with emphasis placed on sulphide stress cracking and hydrogen induced cracking in carbon and low alloy steels. Methods of controlling sulphide stress cracking and hydrogen induced cracking are addressed separately. Case histories from the heavy water, refining, and sour gas industries are used to illustrate operating experience and failure mechanisms. Finally, recommendations, based largely on the author's industrial experience, are made with respect to quality assurance and inspection requirements for sour service components. Only published literature was surveyed. Abstracts were made of all references, reviewing the major sources in detail

Operating Experience Insights into Pipe Failures for Electro-Hydraulic Control and Instrument Air Systems in Nuclear Power Plant. A Topical Report from the Component Operational Experience, Degradation and Ageing Programme

International Nuclear Information System (INIS)

2015-01-01

2010. The majority of the member organisations of the two projects were the same, often being represented by the same person. In May 2011, thirteen countries signed the CODAP 1. Term Agreement (Canada, Chinese Taipei, Czech Republic, Finland, France, Germany, Japan, Korea (Republic of), Slovak Republic, Spain, Sweden, Switzerland and the United States). The 1. Term work plan includes the preparation of Topical Reports to foster technical co-operation and to deepen the understanding of national differences in ageing management. The Topical Reports constitute CODAP Event Database and Knowledge Base insights reports and as such act as portals for future database application projects and in-depth studies of selected degradation mechanisms. Prepared in 2013 and published as NEA/CSNI/R(2014)6, a first Topical Report addressed flow accelerated corrosion (FAC) of carbon steel and low alloy steel piping. This, the second Topical Report addresses operating experience with electro-hydraulic control (EHC) and instrument air (IA) system piping. Degradation and failure of EHC or IA piping can adversely affect plant operability, and under certain circumstances lead to safety challenges. Both systems consist of significant lengths of small-diameter piping. The typical EHC system piping material is stainless steel; Type 304 or 316. Plants generally use carbon steel, copper, stainless steel, galvanised steel or combinations of two or more material types for IA system piping. The CODAP Topical Report on 'EHC and IA Piping Systems' includes a primer on the environmental and operational factors affecting the structural integrity of respective system, and evaluates service experience data as recorded in the CODAP Event Database. Also included in the report are descriptions of the national EHC and IA ageing management programme approaches and a summary of other information collected in the CODAP Knowledge Base. The report has been prepared by the CODAP Project Review Group, with

78 FR 37584 - U.S. Steel Tubular Products, Inc., Mckeesport Tubular Operations Division, Subsidiary of United...

Science.gov (United States)

2013-06-21

... make the following certification: All workers of U.S. Steel Tubular Products, McKeesport Tubular... Products, Inc., Mckeesport Tubular Operations Division, Subsidiary of United States Steel Corporation, Mckeesport, Pennsylvania; Notice of Amended Certification Pursuant to Section 221 of the Trade Act of 1974...

Accelerating the transfer and diffusion of energy saving technologies steel sector experience-Lessons learned

International Nuclear Information System (INIS)

Okazaki, Teruo; Yamaguchi, Mitsutsune

2011-01-01

It is imperative to tackle the issue globally mobilizing all available policies and measures. One of the important ones among them is technology transfer and diffusion. By utilizing international co-operation, industry can promote such measures in two ways: through government policy and through industry's own voluntary initiative. Needless to say, various government policies and measures play essential role. By the same token, industry initiative can complement them. There is much literature documenting the former. On the contrary there are few on the latter. This paper sheds light on the latter. The purpose of this paper is to explore the effectiveness of global voluntary sectoral approach for technology diffusion and transfer based on steel sector experience. The goal is to contribute toward building a worldwide low-carbon society by manufacturing goods with less energy through international cooperation of each sector. The authors believe that the voluntary sectoral approach is an effective method with political and practical feasibilities, and hope to see the continued growth of more initiatives based on this approach. - Highlights: → There exist huge reduction potentials in steel industries globally. → Technology transfer and diffusion are keys to achieve reductions. → Main barriers are economic, technological and policy-related. → Case studies in overcoming barriers are discussed. → In steel industry, a voluntary sectoral approach has shown to be effective.

CANDU operating experience

International Nuclear Information System (INIS)

McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.; Thurygill, E.W.

1980-05-01

The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

Vibration and Operational Characteristics of a Composite-Steel (Hybrid) Gear

Science.gov (United States)

Handschuh, Robert F.; LaBerge, Kelsen E.; DeLuca, Samuel; Pelagalli, Ryan

2014-01-01

Hybrid gears have been tested consisting of metallic gear teeth and shafting connected by composite web. Both free vibration and dynamic operation tests were completed at the NASA Glenn Spur Gear Fatigue Test Facility, comparing these hybrid gears to their steel counterparts. The free vibration tests indicated that the natural frequency of the hybrid gear was approximately 800 Hz lower than the steel test gear. The dynamic vibration tests were conducted at five different rotational speeds and three levels of torque in a four square test configuration. The hybrid gears were tested both as fabricated (machined, composite layup, then composite cure) and after regrinding the gear teeth to the required aerospace tolerance. The dynamic vibration tests indicated that the level of vibration for either type of gearing was sensitive to the level of load and rotational speed.

Aerosol measurements from plasma torch cuts on stainless steel, carbon steel, and aluminum

International Nuclear Information System (INIS)

Novick, V.J.; Brodrick, C.J.; Crawford, S.; Nasiatka, J.; Pierucci, K.; Reyes, V.; Sambrook, J.; Wrobel, S.; Yeary, J.

1996-01-01

The main purpose of this project is to quantify aerosol particle size and generation rates produced by a plasma torch whencutting stainless steel, carbon steel and aluminum. the plasma torch is a common cutting tool used in the dismantling of nuclear facilities. Eventually, other cutting tools will be characterized and the information will be compiled in a user guide to aid in theplanning of both D ampersand D and other cutting operations. The data will be taken from controlled laboratory experiments on uncontaminated metals and field samples taken during D ampersand D operations at ANL nuclear facilities. The plasma torch data was collected from laboratory cutting tests conducted inside of a closed, filtered chamber. The particle size distributions were determined by isokinetically sampling the exhaust duct using a cascade impactor. Cuts on different thicknesses showed there was no observable dependence of the aerosol quantity produced as a function of material thickness for carbon steel. However, data for both stainless steel and aluminum revealed that the aerosol mass produced for these materials appear to have some dependance on thickness, with thinner materials producing tmore aerosols. The results of the laboratory cutting tests show that most measured particle size distributions are bimodal with one mode at about 0.2 μm and the other at about 10 μm. The average Mass Median Aerodynamic Diameters (MMAD's) for these tests are 0.36 ±0.08 μm for stainless steel, 0.48 ±0.17μm for aluminum and 0.52±0.12 μm for carbon steel

Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

International Nuclear Information System (INIS)

Gieseler, Walter

1970-01-01

The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

International Nuclear Information System (INIS)

Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

1980-01-01

The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Experimental investigations on fiber laser color marking of steels

Energy Technology Data Exchange (ETDEWEB)

Amara, E.H., E-mail: amara@cdta.dz; Haïd, F.; Noukaz, A.

2015-10-01

Highlights: • We develop an experimental approach with the aim to bring a contribution to the comprehension of the occurring phenomena during laser color marking of steels. • We have used a home-made marking device composed of a pulsed fiber laser and galvanometric mirrors. • Both commercial and elaborated in laboratory steels have been used as samples. • The experiments have been performed for different laser beam operating parameters, under normal atmospheric conditions. • The treated samples were analyzed either by optical and scanning electronic microscopy, as well as by energy dispersion spectroscopy. - Abstract: We develop an experimental approach with the aim to bring a contribution to the comprehension of the occurring phenomena during laser color marking of steels. A home-made marking device using a pulsed fiber laser has been used to treat steel samples under different laser beam operating parameters, for different compositions of the processed steel, and at normal atmospheric conditions. The treated samples were analyzed either by optical and scanning electronic microscopy, as well as by energy dispersion spectroscopy. The results show the influence of the operating parameters on the obtained colors.

Experimental investigations on fiber laser color marking of steels

International Nuclear Information System (INIS)

Amara, E.H.; Haïd, F.; Noukaz, A.

2015-01-01

Highlights: • We develop an experimental approach with the aim to bring a contribution to the comprehension of the occurring phenomena during laser color marking of steels. • We have used a home-made marking device composed of a pulsed fiber laser and galvanometric mirrors. • Both commercial and elaborated in laboratory steels have been used as samples. • The experiments have been performed for different laser beam operating parameters, under normal atmospheric conditions. • The treated samples were analyzed either by optical and scanning electronic microscopy, as well as by energy dispersion spectroscopy. - Abstract: We develop an experimental approach with the aim to bring a contribution to the comprehension of the occurring phenomena during laser color marking of steels. A home-made marking device using a pulsed fiber laser has been used to treat steel samples under different laser beam operating parameters, for different compositions of the processed steel, and at normal atmospheric conditions. The treated samples were analyzed either by optical and scanning electronic microscopy, as well as by energy dispersion spectroscopy. The results show the influence of the operating parameters on the obtained colors

Operating experience feedback

International Nuclear Information System (INIS)

Cimesa, S.

2007-01-01

Slovenian Nuclear Safety Administration (SNSA) has developed its own system for tracking, screening and evaluating the operating experiences of the nuclear installations. The SNSA staff regularly tracks the operating experiences throughout the world and screens them on the bases of applicability for the Slovenian nuclear facilities. The operating experiences, which pass the screening, are thoroughly evaluated and also recent operational events in these facilities are taken into account. If needed, more information is gathered to evaluate the conditions of the Slovenian facilities and appropriate corrective actions are considered. The result might be the identification of the need for modification at the licensee, the need for modification of internal procedures in the SNSA or even the proposal for the modification of regulations. Information system helps everybody to track the process of evaluation and proper logging of activities. (author)

Operating Experience at NPP Krsko

International Nuclear Information System (INIS)

Kavsek, D.; Bach, B.

1998-01-01

Systematic analysis of operational experience by assessment of internal and industry events and the feedback of lessons learned is one of the essential activities in the improvement of the operational safety and reliability of the nuclear power plant. At NPP Krsko we have developed a document called ''Operating Experience Assessment Program''. Its purpose is to establish administrative guidance for the processing of operating events including on-site and industry events. Assessment of internal events is based on the following methods: Event and Causal Factor Charting, Change Analysis, Barrier Analysis, MORT (Management Oversight and Risk Tree Analysis) and Human Performance Evaluation. The operating experience group has developed a sophisticated program entitled ''Operating experience tracking system'' (OETS) in response to the need for a more efficient way of processing internal and industry operating experience information. The Operating Experience Tracking System is used to initiate and track operational events including recommended actions follow up. Six screens of the system contain diverse essential information which allows tracking of operational events and enables different kinds of browsing. OETS is a part of the NPP Krsko nuclear network system and can be easily accessed by all plant personnel. (author)

Effect of operational conditions of electroerosion machining on the surface microgeometry parameters of steels and alloys

International Nuclear Information System (INIS)

Foteev, N.K.

1976-01-01

Studies the influence of pulse duration and a series of operating conditions of a ShGI-40-440 spark-machining generator on changes in the basic surface microgeometry characteristics of components of stainless steel 1Kh18N10T, steel St 45 and hard alloy T14K8. The microgeometry characteristics of spark-machined surfaces differ significantly from the corresponding characteristics of surfaces machined by cutting and vibro-rolling

Testing of methods for decontamination of stainless steels and carbon steels conformably to demountable equipment of nuclear power plant with WWR type reactor

International Nuclear Information System (INIS)

Dergunova, G.M.; Nazarov, V.K.; Ozolin, A.B.; Smirnov, L.M.; Stel'mashuk, V.P.; Yulikov, E.I.; Vlasov, I.N.

1978-01-01

Results are given of experiments on decontamination of stainless steel by the oxidation-reduction method and also results of decontamination of carbon steel by means of solutions based on oxalic acid, citric acid and phosphoric acid. Investigations of efficiency of oxidation-reduction treatment were done on samples of stainless steel cut from the pipeline of the primary coolant circuit of reactor. Comparison is given of efficiency of oxidation-reduction methods of contamination of stainless steel in the case of application of different compositions of decontaminating solutions. Dependences are given for decontamination completeness on duration of operations, on temperature and on ratio of volume of decontaminating solutions to surface are of the sample. For carbon steels parameters are given for decontamination process by means of oxalic, citric and phosphoric acid solutions. (I.T.) [ru

78 FR 24803 - Hilco SP Rail, LLC-Acquisition and Operation Exemption-RG Steel Railroad Holding, LLC

Science.gov (United States)

2013-04-26

... DEPARTMENT OF TRANSPORTATION Surface Transportation Board [Docket No. FD 35734] Hilco SP Rail, LLC--Acquisition and Operation Exemption--RG Steel Railroad Holding, LLC Hilco SP Rail, LLC (Hilco), a noncarrier... Holding, LLC, and operate as a common carrier over an approximately 12-mile line of railroad in Sparrows...

Volatilization from PCA steel alloy

Energy Technology Data Exchange (ETDEWEB)

Hagrman, D.L.; Smolik, G.R.; McCarthy, K.A.; Petti, D.A.

1996-08-01

The mobilizations of key components from Primary Candidate Alloy (PCA) steel alloy have been measured with laboratory-scale experiments. The experiments indicate most of the mobilization from PCA steel is due to oxide formation and spalling but that the spalled particles are large enough to settle rapidly. Based on the experiments, models for the volatization of iron, manganese, and cobalt from PCA steel in steam and molybdenum from PCA steel in air have been derived.

Simulator experiments: effects of NPP operator experience on performance

International Nuclear Information System (INIS)

Beare, A.N.; Gray, L.H.

1985-01-01

Experiments are being conducted on nuclear power plant (NPP) control room training simulators by the Oak Ridge National Laboratory, its subcontractor, General Physics Corporation, and participating utilities. The experiments are sponsored by the Nuclear Regulatory Commission's (NRC) Human Factors and Safeguards Branch, Division of Risk Analysis and Operations, and are a continuation of prior research using simulators, supported by field data collection, to provide a technical basis for NRC human factors regulatory issues concerned with the operational safety of nuclear power plants. During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE boiling water reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of (a) senior reactor operator (SRO) experience, (b) operating crew augmentation with an STA and (c) practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator. Methodology and results to date are reported

Experimental study and calculation of boiling heat transfer on steel plates during runout table operation

International Nuclear Information System (INIS)

Liu, Z.D.; Fraser, D.; Samarasekera, I.V.

2002-01-01

Within a hot strip steel mill, red hot steel is hot rolled into a long continuous slab that is led onto what is called the runout table. Temperatures of the steel at the beginning of this table are around 900 o C. Above and below the runout table are banks of water jets, sprays or water curtains that rapidly cool the steel slab. The heat transfer process itself may be considered one of the most complicated in the industrial world. The cooling process that occurs on the runout table is crucial and governs the final mechanical properties and flatness of a steel strip. However, very limited data of industrial conditions has been available and that which is available is poorly understood. To study heat transfer during runout table cooling, an industrial scale pilot runout table facility was constructed at the University of British Columbia (UBC). This paper describes the experimental details, data acquisition and data handling techniques for steel plates during water jet impingement cooling by one circular water jet from industrial headers. The effect of cooling water temperature and initial steel plate temperature as well as varying water jet diameters on heat transfer was systematically investigated. A two-dimensional finite element scheme based inverse heat conduction model was developed to calculate surface heat transfer coefficients along the impinging surface. Heat flux curves at the stagnation area were obtained for selected tests. A quantitative relationship between adjustable processing parameters and heat transfer coefficients along the impinging surface during runout table operation is discussed. The results of the study were used to upgrade an extensive process model developed at UBC. The model ties in the cooling rate and hence two dimensional temperature gradients to the resulting microstructure and final mechanical properties of the steel. This process model is widely used by major steel industries in Canada and the United States. (author)

Niobium Application, Metallurgy and Global Trends in Pressure Vessel Steels

Science.gov (United States)

Jansto, Steven G.

Niobium-containing high strength steel materials have been developed for a variety of pressure vessel applications. Through the application of these Nb-bearing steels in demanding applications, the designer and end user experience improved toughness at low temperature, excellent fatigue resistance and fracture toughness and excellent weldability. These enhancements provide structural engineers the opportunity to further improve the pressure vessel design and performance. The Nb-microalloy alloy designs also result in reduced operational production cost at the steel operation, thereby embracing the value-added attribute Nb provides to both the producer and the end user throughout the supply chain. For example, through the adoption of these Nb-containing structural materials, several design-manufacturing companies are considering improved designs which offer improved manufacturability, lower overall cost and better life cycle performance.

Benchmark experiments to test plutonium and stainless steel cross sections. Topical report

International Nuclear Information System (INIS)

Jenquin, U.P.; Bierman, S.R.

1978-06-01

The Nuclear Regulatory Commission (NRC) commissioned Battelle, Pacific Northwest Laboratory (PNL) to ascertain the accuracy of the neutron cross sections for the isotopes of plutonium and the constituents of stainless steel and determine if improvements can be made in application to criticality safety analysis. NRC's particular area of interest is in the transportation of light-water reactor spent fuel assemblies. The project was divided into two tasks. The first task was to define a set of integral experimental measurements (benchmarks). The second task is to use these benchmarks in neutronics calculations such that the accuracy of ENDF/B-IV plutonium and stainless steel cross sections can be assessed. The results of the first task are given in this report. A set of integral experiments most pertinent to testing the cross sections has been identified and the code input data for calculating each experiment has been developed

Exploring the Use of Design of Experiments in Industrial Processes Operating Under Closed-Loop Control

DEFF Research Database (Denmark)

Capaci, Francesca; Kulahci, Murat; Vanhatalo, Erik

2017-01-01

Industrial manufacturing processes often operate under closed-loop control, where automation aims to keep important process variables at their set-points. In process industries such as pulp, paper, chemical and steel plants, it is often hard to find production processes operating in open loop....... Instead, closed-loop control systems will actively attempt to minimize the impact of process disturbances. However, we argue that an implicit assumption in most experimental investigations is that the studied system is open loop, allowing the experimental factors to freely affect the important system...... responses. This scenario is typically not found in process industries. The purpose of this article is therefore to explore issues of experimental design and analysis in processes operating under closed-loop control and to illustrate how Design of Experiments can help in improving and optimizing...

Kinetics of fatigue crack growth and crack paths in the old puddled steel after 100-years operating time

Directory of Open Access Journals (Sweden)

G. Lesiuk

2015-10-01

Full Text Available The goal of the authors’ investigations was determination of the fatigue crack growth in fragments of steel structures (of the puddled steel and its cyclic behavior. Tested steel elements coming from the turn of the 19th and 20th were gained from still operating ancient steel construction (a main hall of Railway Station, bridges etc.. This work is a part of investigations devoted to the phenomenon of microstructural degradation and its potential influence on their strength properties. The analysis of the obtained results indicated that those long operating steels subject to microstructure degradation processes consisting mainly in precipitation of carbides and nitrides inside ferrite grains, precipitation of carbides at ferrite grain boundaries and degeneration of pearlite areas [1, 2]. It is worth noticing that resistance of the puddled steel to fatigue crack propagation in the normalized state was higher. The authors proposed the new kinetic equation of fatigue crack growth rate in such a steel. Thus the relationship between the kinetics of degradation processes and the fatigue crack growth rate also have been shown. It is also confirmed by the materials research of the viaduct from 1885, which has not shown any significant changes in microstructure. The non-classical kinetic fatigue fracture diagrams (KFFD based on deformation ( or energy (W approach was also considered. In conjunction with the results of low- and high-cycle fatigue and gradual loss of ductility as a consequence (due to the microstructural degradation processes - it seems to be a promising construction of the new kinetics fatigue fracture diagrams with the energy approach.

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

Operating experience feedback in TVO

Energy Technology Data Exchange (ETDEWEB)

Piirto, A [Teollisuuden Voima Oy (Finland)

1997-12-31

TVO is a power company operating with two 710 MW BWR units at Olkiluoto. For operating experience feedback TVO has not established a separate organizational unit but rather relies on a group of persons representing various technical disciplines. The ``Operating Experience Group`` meets at about three-week intervals to handle the reports of events (in plant and external) which have been selected for handling by an engineer responsible for experience feedback. 7 charts.

Nuclear units operating improvement by using operating experience

International Nuclear Information System (INIS)

Rotaru, I.; Bilegan, I.C.

1997-01-01

The paper presents how the information experience can be used to improve the operation of nuclear units. This areas include the following items: conservative decision making; supervisory oversight; teamwork; control room distraction; communications; expectations and standards; operator training and fundamental knowledge, procedure quality and adherence; plant status awareness. For each of these topics, the information illustrate which are the principles, the lessons learned from operating experience and the most appropriate exemplifying documents. (authors)

Steel making

CERN Document Server

Chakrabarti, A K

2014-01-01

"Steel Making" is designed to give students a strong grounding in the theory and state-of-the-art practice of production of steels. This book is primarily focused to meet the needs of undergraduate metallurgical students and candidates for associate membership examinations of professional bodies (AMIIM, AMIE). Besides, for all engineering professionals working in steel plants who need to understand the basic principles of steel making, the text provides a sound introduction to the subject.Beginning with a brief introduction to the historical perspective and current status of steel making together with the reasons for obsolescence of Bessemer converter and open hearth processes, the book moves on to: elaborate the physiochemical principles involved in steel making; explain the operational principles and practices of the modern processes of primary steel making (LD converter, Q-BOP process, and electric furnace process); provide a summary of the developments in secondary refining of steels; discuss principles a...

Evaluation criteria of structural steel reliability

International Nuclear Information System (INIS)

Zav'yalov, A.S.

1980-01-01

Different low-carbon and medium-carbon structural steels are investigated. It is stated that steel reliability evaluation criteria depend on the fracture mode, steel suffering from the brittle fracture under the influence of the stresses (despite their great variety) arising in articles during the production and operation. Fibrous steel fracture at the given temperature and article thickness says about its high ductility and toughness and brittle fractures are impossible. Brittle fractures take place in case of a crystalline and mixed fracture with a predominant crystalline component. Evaluation methods of article and sample steel structural strength differing greatly from real articles in a thickness (diameter) or used at temperatures higher than possible operation temperatures cannot be reliability evaluation criteria because at a great thickness (diameter) and lower operation temperatures steel fracture and its strain mode can change resulting in a sharp reliability degradation

Regulatory challenges in using nuclear operating experience

International Nuclear Information System (INIS)

2006-01-01

There can be no doubt that the systematic evaluation of operating experience by the operator and the regulator is essential for continued safe operation of nuclear power plants. Recent concerns have been voiced that the operating experience information and insights are not being used effectively to promote safety. If these concerns foreshadow a real trend in OECD countries toward complacency in reporting and analysing operating events and taking corrective actions, then past experience suggests that similar or even more serious events will recur. This report discusses how the regulator can take actions to assure that operators have effective programmes to collect and analyse operating experience and, just as important, for taking steps to follow up with actions to prevent the events and conditions from recurring. These regulatory actions include special inspections of an operator operating experience programme and discussion with senior plant managers to emphasize the importance of having an effective operating experience programme. In addition to overseeing the operator programmes, the regulator has the broader responsibility for assuring that industry-wide trends, both national and international are monitored. To meet these responsibilities, the regulatory body must have its own operating experience programme, and this report discusses the important attributes of such regulatory programmes. It is especially important for the regulator to have the capability for assessing the full scope of operating experience issues, including those that may not be included in an operator operating experience programme, such as new research results, international operating experience, and broad industry trend information. (author)

Simulator experiments: effects of NPP operator experience on performance

International Nuclear Information System (INIS)

Beare, A.N.; Gray, L.H.

1984-01-01

During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE Boiling Water Reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of senior reactor operator (SRO) experience, operating crew augmentation with an STA and practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. Sixteen two-man crews of licensed operators were employed in a 2 x 2 factorial design. The SROs leading the crews were split into high and low experience groups on the basis of their years of experience as an SRO. One half of the high- and low-SRO experience groups were assisted by an STA. The crews responded to four simulated plant casualties. A five-variable set of content-referenced performance measures was derived from task analyses of the procedurally correct responses to the four casualties. System parameters and control manipulations were recorded by the computer controlling the simulator. Data on communications and procedure use were obtained from analysis of videotapes of the exercises. Questionnaires were used to collect subject biographical information and data on subjective workload during each simulated casualty. For four of the five performance measures, no significant differences were found between groups led by high (25 to 114 months) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have significantly shorter task performance times than crews led by high experience SROs. The presence of the STA had no significant effect on overall team performance in responding to the four simulated casualties. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator

Temporal sealing material of tritium-contaminated stainless steel

International Nuclear Information System (INIS)

Wen Wei; Dan Guiping; Zhang Dong; Qiu Yongmei; Zhang Li

2010-01-01

Tritium can be released from the exterior of tritium-contaminated stainless steel by slight stirring while decontaminating and disassembling. In order to avoid secondary tritium contamination to environment and operators, it is necessary to cover with an effective coating to tritium on the exterior of tritium-contaminated stainless steel and fill an effective substance to tritium inside. The results of tritium sealed experiments show that sealing efficiency of neutral silicone rubber is more than 85% for condition of static state and more than 99% for foam concrete condition of dynamic state. Neutral silicone rubber and foam concrete which have finer sealing efficiency can be used as temporal sealed material for the decontamination and disassembly of tritium-contaminated stainless steel. (authors)

Microbial-Influenced Corrosion of Corten Steel Compared with Carbon Steel and Stainless Steel in Oily Wastewater by Pseudomonas aeruginosa

Science.gov (United States)

Mansouri, Hamidreza; Alavi, Seyed Abolhasan; Fotovat, Meysam

2015-07-01

The microbial corrosion behavior of three important steels (carbon steel, stainless steel, and Corten steel) was investigated in semi petroleum medium. This work was done in modified nutrient broth (2 g nutrient broth in 1 L oily wastewater) in the presence of Pseudomonas aeruginosa and mixed culture (as a biotic media) and an abiotic medium for 2 weeks. The behavior of corrosion was analyzed by spectrophotometric and electrochemical methods and at the end was confirmed by scanning electron microscopy. The results show that the degree of corrosion of Corten steel in mixed culture, unlike carbon steel and stainless steel, is less than P. aeruginosa inoculated medium because some bacteria affect Corten steel less than other steels. According to the experiments, carbon steel had less resistance than Corten steel and stainless steel. Furthermore, biofilm inhibits separated particles of those steels to spread to the medium; in other words, particles get trapped between biofilm and steel.

The FSM technology -- Operational experience and improvements in local corrosion analysis

International Nuclear Information System (INIS)

Stroemmen, R.; Horn, H.; Gartland, P.O.; Wold, K.

1996-01-01

FSM (Field Signature Method) is a non-intrusive monitoring technique based on a patented principle, developed for the purpose of detection and monitoring of both general and localized corrosion, erosion and cracking in steel and metal structures, piping systems and vessels. Since 1991 FSM has been used for a wide range of applications e.g. for buried and open pipelines, process piping offshore, subsea pipelines and flowlines, applications in the nuclear power industry and in materials research in general. This paper describes typical applications of the FSM technology, and presents operational experience from some of the landbased and subsea installations. The paper also describes recent enhancements in the FSM technology and in the analysis of FSM readings, allowing for monitoring and detailed quantification of pitting and mesa corrosion, and of corrosion in welds

Steel Industry Wastes.

Science.gov (United States)

Schmidtke, N. W.; Averill, D. W.

1978-01-01

Presents a literature review of wastes from steel industry, covering publications of 1976-77. This review covers: (1) coke production; (2) iron and steel production; (3) rolling operations; and (4) surface treatment. A list of 133 references is also presented. (NM)

Steels for nuclear power. I

International Nuclear Information System (INIS)

Bohusova, O.; Brumovsky, M.; Cukr, B.; Hatle, Z.; Protiva, K.; Stefec, R.; Urban, A.; Zidek, M.

1976-01-01

The principles are listed of nuclear reactor operation and the reactors are classified by neutron energy, fuel and moderator designs, purpose and type of moderator. The trend and the development of light-water reactor applications are described. The fundamental operating parameters of the WWER type reactors are indicated. The effect is discussed of neutron radiation on reactor structural materials. The characteristics are described of steel corrosion due to the contact of the steel with steam or sodium in the primary coolant circuit. The reasons for stress corrosion are given and the effects of radiation on corrosion are listed. The requirements and criteria are given for the choice of low-alloy steel for the manufacture of pressure vessels, volume compensators, steam generators, cooling conduits and containment. A survey is given of most frequently used steels for pressure vessels and of the mechanical and structural properties thereof. The basic requirements for the properties of steel used in the primary coolant circuit are as follows: sufficient strength in operating temperature, toughness, good weldability, resistance to corrosion and low brittleness following neutron irradiation. The materials are listed used for the components of light-water and breeder reactors. The production of corrosion-resistant steels is discussed with a view to raw materials, technology, steel-making processes, melting processes, induction furnace steel-making, and to selected special problems of the chemical composition of steels. The effects are mainly discussed of lead, bismuth and tin as well as of some other elements on hot working of high-alloy steels and on their structure. The problems of corrosion-resistant steel welding and of pressure vessel cladding are summed up. Also discussed is the question of the concept and safeguards of the safety of nuclear installation operation and a list is presented of most commonly used nondestructive materials testing methods. The current

CANDU operating experience

International Nuclear Information System (INIS)

McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

1982-03-01

The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

Recent operating experience issues with 17-4 PH in LWRs

International Nuclear Information System (INIS)

Olender, A.; Gorman, J.; Marks, C.; Ilevbare, G.

2015-01-01

The stainless steel 17-4 PH has been used successfully in LWRs for quite some time. Its properties as a precipitation hardening martensitic stainless steel are desirable for high strength, high hardness applications, such as valve stems, bolting, and turbine blades. However, some in-service issues with this material have arisen over time and are addressed in this paper. Although the use of 17-4 PH in high hardness conditions is discouraged as it offers less resistance to SCC than lower hardness conditions, the utilization, and subsequent failure, of such components has occurred as recently as 2007. Thermal embrittlement can increase susceptibility to SCC, another major issue affecting 17-4 PH. The recommended maximum operating temperature is, therefore, 243 C. degrees (470 F. degrees) due to recent failures attributed to this cause. Connections with dissimilar metals have also recently been a problem, as inadequate tolerances have been left between components to allow for differences in coefficients of thermal expansion as the temperature is ramped up to operating conditions. However, with the proper precautions, 17-4 PH can, and has, performed successfully during long-term plant operation. (authors)

Industry Operating Experience Process at Krsko NPP

International Nuclear Information System (INIS)

Bach, B.; Bozin, B.; Cizmek, R.

2012-01-01

Experience has shown that number of minor events and near misses, usually without immediate or significant impact to plant safety and reliability, are precursors of significant or severe events due to the same or similar root or apparent cause(s). It is therefore desirable to identify and analyze weaknesses of the precursor problems (events) in order to prevent occurrence of significant events. Theoretically, significant events could be prevented from occurring if the root cause(s) of these precursor problems could be identified and eliminated. The Operating Experience Program identifies such event precursors and by reporting them to the industry, plant specific corrective actions can be taken to prevent events at other operational plants. The intent of the Operating Experience Program is therefore to improve nuclear power plant safety and reliability of the operating nuclear power plants. Each plant develops its own Operating Experience Program in order to learn from the in-house operating experience as well as from the world community of nuclear plants. The effective use of operating experience includes analyzing both plant and industry events in order to identify fundamental weaknesses and then determining appropriate plant-specific actions that will minimize the likelihood of similar events. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures. Krsko NPP is developed it own Operating Experience Program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The Operating Experience Program is a part of the Corrective Action Program, which is among top management programs, thus program is strongly encouraged by top management. The purpose of Operating Experience Program is to provide guidance for using, sharing, and evaluating operating experience information

Combining risk analysis and operating experience

International Nuclear Information System (INIS)

1986-10-01

In recent years there has been an increasing interest in the systematic utilization of operating experience in the decision making process concerning large industrial facilities. Even before the advent of Probabilistic Safety Assessment (PSA), operating experience had always played an important role in such decisions. Of course, operating experience has always been an input to PSA also; however, as PSA becomes more mature and the quality and quantity of operating experience improve, greater emphasis is now being placed on the use of operating experience to update and validate PSA and thereby provide a more rational basis for decision making. This report outlines the ways in which data are collected, processed using mathematical techniques and utilized in decision making. It is not intended to provide details of the methods and procedures to be used in these areas, but is rather intended as an introduction to these topics and some of the relevant literature. The meeting presentations were divided into three sessions devoted to the following topics: evaluation of nuclear power plants operational experience (5 papers); uncertainties (2 papers); probabilistic safety assessment studies in Member States (7 papers). A separate abstract was prepared for each of these papers

ERB-II operating experience

International Nuclear Information System (INIS)

Smith, R.N.; Cissel, D.W.; Smith, R.R.

1977-01-01

As originally designed and operated, EBR-II successfully demonstrated the concept of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle (mini-nuclear park). Subsequent operation has been as an irradiation facility, a role which will continue into the foreseeable future. Since the beginning of operation in 1961, operating experience of EBR-II has been very satisfactory. Most of the components and systems have performed well. In particular, the mechanical performance of heat-removal systems has been excellent. A review of the operating experience reveals that all the original design objectives have been successfully demonstrated. To date, no failures or incidents resulting in serious in-core or out-of-core consequences have occurred. No water-to-sodium leaks have been detected over the life of the plant. At the present time, the facility is operating very well and continuously except for short shutdowns required by maintenance, refueling, modification, and minor repair. A plant factor of 76.9% was achieved for the calendar year 1976

Study of the performances of nano-case treatment cutting tools on carbon steel work material during turning operation

Science.gov (United States)

Afolalu, S. A.; Okokpujie, I. P.; Salawu, E. Y.; Abioye, A. A.; Abioye, O. P.; Ikumapayi, O. M.

2018-04-01

The degree of holding temperature and time play a major role in nano-case treatment of cutting tools which immensely contributed to its performance during machining operation. The objective of this research work is to carryout comparative study of performance of nano-case treatment tools developed using low and medium carbon steel as work piece. Turning operation was carried out under two different categories with specific work piece on universal lathe machine using HSS cutting tools 100 mm × 12mm × 12mm that has been nano-case treated under varying conditions of temperatures and timeof 800,850, 900, 950°C and 60, 90, 120 mins respectively. The turning parameters used in evaluating this experiment were cutting speed of 270, 380 and 560mm/min, feed rate of 0.15, 0.20 and 0.25 mm/min, depth of cut of 2mm, work piece diameter of 25mm and rake angle of 7° each at three levels. The results of comparative study of their performances revealed that the timespent in the machining of low carbon steel material at a minimum temperature and time of 800°C, 60 mins were1.50, 2.17 mins while at maximum temperature and time of 950°C, 120 mins were 1.19, 2.02 mins. It was also observed that at a corresponding constant speed of 270,380 and 560mm/min at higher temperature and time, a relative increased in the length of cut were observed. Critical observation of the result showed that at higher case hardening temperature and time (950°C/120mins), the HSS cutting tool gave a better performance as lesser time was consumed during the turning operation.

Operation and experience of a 2 km coated conductor REEL – to – REEL copper pulse plating facility

International Nuclear Information System (INIS)

Floegel-Delor, U; Riedel, T; Wippich, D; Rothfeld, R; Schirrmeister, P; Koenig, R; Werfel, F N; Usoskin, A; Rutt, A

2014-01-01

Bruker HTS manufactures YBCO based superconducting wires of the second generation on low- cost Stainless Steel substrate (100 μm thick). With 250 – 500 A/cm@77 K, SF, 650 MPa tensile strength and 6 mm bending radius excellent electrical and mechanical properties are achieved. As complementation of the 2G fabrication technology an automated 2 km copper pulse plating facility has been installed in 2012. We report here the operation requirements and the experiences of the copper plating technique.

Prospects of structural steels

International Nuclear Information System (INIS)

Bannykh, O.A.

2012-01-01

The current state of world steel production is considered as well as the development strategy of metallurgy industry in the Russian Federation through to 2020. The main factors determining the conservation of steel as perspective material for industry are given: energy expenses on production, the well-proven recirculation technology, the capability of changing steel properties in wide range, temperature range of operation. The conclusion is made that in the immediate future steel will not lose its importance [ru

High-strength maraging steels

International Nuclear Information System (INIS)

Grachev, S.V.; Shejn, A.S.

1989-01-01

Analysis of data on technological and operation properties of maraging steels on Fe-Cr-Ni, Fe-Ni, Fe-Cr-Co-Mo bases is given. Their advantages and drawbacks are pointed out. The scheme of strengthening heat treatment is considered. The fields of the most effective application of maraging steels for instance, for products operating under conditions of low-cycle and shock cyclic loading are mentioned

Operating experience: safety perspective

International Nuclear Information System (INIS)

Piplani, Vivek; Krishnamurthy, P.R.; Kumar, Neeraj; Upadhyay, Devendra

2015-01-01

Operating Experience (OE) provides valuable information for improving NPP safety. This may include events, precursors, deviations, deficiencies, problems, new insights to safety, good practices, lessons and corrective actions. As per INSAG-10, an OE program caters as a fundamental means for enhancing the defence-in-depth at NPPs and hence should be viewed as ‘Continuous Safety Performance Improvement Tool’. The ‘Convention on Nuclear Safety’ also recognizes the OE as a tool of high importance for enhancing the NPP safety and its Article 19 mandates each contracting party to establish an effective OE program at operating NPPs. The lessons drawn from major accidents at Three Mile Island, Chernobyl and Fukushima Daiichi NPPs had prompted nuclear stalwarts to change their safety perspective towards NPPs and to frame sound policies on issues like safety culture, severe accident prevention and mitigation. An effective OE program, besides correcting current/potential problems, help in proactively improving the NPP design, operating and maintenance procedures, practices, training, etc., and thus plays vital role in ensuring safe and efficient operation of NPPs. Further it enhances knowledge with regard to equipment operating characteristics, system performance trends and provides data for quantitative and qualitative safety analysis. Besides all above, an OE program inculcates a learning culture in the organisation and thus helps in continuously enhancing the expertise, technical competency and knowledge base of its staff. Nuclear and Radiation Facilities in India are regulated by Atomic Energy Regulatory Board (AERB). Operating Plants Safety Division (OPSD) of AERB is involved in managing operating experience activities. This paper provides insights about the operating experience program of OPSD, AERB (including its on-line data base namely OPSD STAR) and its utilisation in improving the regulations and safety at Indian NPPs/projects. (author)

Operating experience feedback program at Olkiluoto NPP

International Nuclear Information System (INIS)

Kosonen, Mikko

2002-01-01

Recent review and development of the operating experience feedback program will be described. The development of the program has been based on several reviews by outside organizations. Main conclusions from these review reports and from the self assessment of safety performance, safety problems and safety culture on the basis of the operational events made by ASSET-method will be described. An approach to gather and analyze small events - so-called near misses - will be described. The operating experience program has been divided into internal and external operating experience. ASSET-methodology and a computer program assisting the analysis are used for the internal operating experience events. Noteworthy incidents occurred during outage are analyzed also by ASSET-method. Screening and pre analysis of the external operating experience relies on co-operation with ERFATOM, an organization of Nordic utilities for the exchange of nuclear industry experience. A short presentation on the performance of the Olkiluoto units will conclude the presentation. (author)

Pitting Corrosion of the Resistance Welding Joints of Stainless Steel Ventilation Grille Operated in Swimming Pool Environment

Directory of Open Access Journals (Sweden)

Mirosław Szala

2018-01-01

Full Text Available This work focuses on the pitting corrosion of ventilation grilles operated in swimming pool environments. The ventilation grille was made by resistance welding of stainless steel rods. Based on the macroscopic and microscopic examinations, the mechanism of the pitting corrosion was confirmed. Chemical composition microanalysis of sediments as well as base metal using scanning electron microscopy and energy-dispersive spectroscopy (SEM-EDS method was carried out. The weldments did not meet the operating conditions of the swimming pool environment. The wear due to the pitting corrosion was identified in heat affected zones of stainless steel weldment and was more severe than the corrosion of base metal. The low quality finish of the joints and influence of the welding process on the weld metal microstructure lead to accelerated deposition of corrosion effecting elements such as chlorine.

Research on key technology of the verification system of steel rule based on vision measurement

Science.gov (United States)

Jia, Siyuan; Wang, Zhong; Liu, Changjie; Fu, Luhua; Li, Yiming; Lu, Ruijun

2018-01-01

The steel rule plays an important role in quantity transmission. However, the traditional verification method of steel rule based on manual operation and reading brings about low precision and low efficiency. A machine vison based verification system of steel rule is designed referring to JJG1-1999-Verificaiton Regulation of Steel Rule [1]. What differentiates this system is that it uses a new calibration method of pixel equivalent and decontaminates the surface of steel rule. Experiments show that these two methods fully meet the requirements of the verification system. Measuring results strongly prove that these methods not only meet the precision of verification regulation, but also improve the reliability and efficiency of the verification system.

Operating experience

International Nuclear Information System (INIS)

McRae, L.P.; Six, D.E.

1991-01-01

In 1987, Westinghouse Hanford Company began operating a first-generation integrated safeguards system in the Plutonium Finishing Plant storage vaults. This Vault Safety and Inventory System is designed to integrate data into a computer-based nuclear material inventory monitoring system. The system gathers, in real time, measured physical parameters that generate nuclear material inventory status data for thousands of stored items and sends tailored report to the appropriate users. These data include canister temperature an bulge data reported to Plant Operations and Material Control and Accountability personnel, item presence and identification data reported to Material Control and Accountability personnel, and unauthorized item movement data reported to Security response forces and Material Control and Accountability personnel. The Westinghouse Hanford Company's experience and operational benefits in using this system for reduce radiation exposure, increase protection against insider threat, and real-time inventory control are discussed in this paper

Dome style heavy wall steel casting manufactured by metallic core mould system

International Nuclear Information System (INIS)

Yamamoto, Shiro; Saeki, Keiji; Hirose, Yutaka; Takebayashi, Kazunari; Kawasaki, Masatoshi

1986-01-01

Semi-spherical thick walled steel castings are one of the main products of Nippon Chutanko K.K., but there have been the problems of internal defects peculiar to large thick walled steel castings, and the various improvements have been carried out so far for the manufacturing method, but still some of those remains. Based on the anxiety about the reliability of large steel castings, the conversion to forging has been studied. For the purpose of thoroughly improving the internal quality of thick walled steel castings to compete with forgings, on the basis of the operating experience of chills, the development of the casting techniques changing cores completely to metallic cores has been advanced. After the preliminary experiment using models, a semi-spherical thick walled steel casting mentioned before was manufactured by this metallic core casting method for trial, and the detailed investigation was carried out. As the result, the excellent internal quality was confirmed, accordingly at present, the production is made by this method. The form, dimensions and specification of the semi-spherical thick walled steel castings, the conventional casting plan, the metallic core casting plan, the design of metallic cores, molding and casting, and the examination of the castings made for trial are reported. (Kako, I.)

Operating experience with snubbers

International Nuclear Information System (INIS)

Levin, H.; Cudlin, R.

1978-06-01

Recent operating experience with hydraulic and mechanical snubbers has indicated that there is a need to evaluate current practice in the industry associated with snubber qualification testing programs, design and analysis procedures, selection and specification criteria, and the preservice inspection and inservice surveillance programs. The report provides a summary of operational experiences that represent problems that are generic throughout the industry. Generic Task A-13 is part of the NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants described in NUREG-0410. The report is based upon a rather large amount of data that have become available in the past four years. These data have been evaluated by the Division of Operating Reactors to develop a data base for use in connection with several NRC activities including Category A, Technical Activity A-13 (Snubbers); the Standard Review Plan; future Regulatory Guides; ASME Code Provisions; and various technical specifications of operating nuclear power plants

Research into the melting/refining of contaminated steel scrap arising in the dismantling of nuclear installations

International Nuclear Information System (INIS)

Harvey, D.S.

1990-01-01

The main part of this report is concerned with the steel-making behaviour of various radioisotopes encountered in steel from decommissioning of nuclear installations (e.g. cobalt 60, caesium 134 and europium 154). Under a wide range of conditions cobalt is largely absorbed by the steel, europium is absorbed by the slag, whereas caesium may be largely volatized, or largely absorbed by the slag. Radiation exposures which might occur during a large-scale recycling operation, during routine operations and accidents would not be significant according to published criteria in the UK. The second part of the report concerns the detection of radioactive materials which may be accidentally delivered to steelworks in scrap steel and used in steel-making. Detectors have been developed which would indicate the presence of radioactivity in scrap. A survey of the steelworks revealed areas where detection might be performed. Experiments have shown that a gamma ray detector of large volume could provide useful sensitivity of detection

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

1996-09-01

Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

International Nuclear Information System (INIS)

Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

1996-09-01

Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments

14 CFR 121.434 - Operating experience, operating cycles, and consolidation of knowledge and skills.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Operating experience, operating cycles, and... Qualifications § 121.434 Operating experience, operating cycles, and consolidation of knowledge and skills. (a... position, the operating experience, operating cycles, and the line operating flight time for consolidation...

Operating experience review for the AP1000 plant

International Nuclear Information System (INIS)

Chaney, T. E.; Lipner, M. H.

2006-01-01

Westinghouse is performing an update to the Operating Experience Review (OER) Report for the AP1000 project to account for operating experience since December 1996. Significant Operating Experience Reports, Significant Event Reports, Significant Event Notifications, Operations and Maintenance Reminders, Topical Reports, Event Analysis Reports and Licensee Event Reports were researched for pertinent input to the update. As a part of the OER, Westinghouse has also conducted operator interviews and observations during simulated plant operations and after operating events. The main purpose of the OER is to identify Human Factors Engineering (HFE) related safety issues from existing operating plant experience and to ensure that these issues are addressed in the new design. The issues and lessons learned regarding operating experience provide a basis for improving the plant design. (authors)

Future directions for ferritic/martensitic steels for nuclear applications

International Nuclear Information System (INIS)

Klueh, R.L.; Swindeman, R.W.

2000-01-01

High-chromium (7-12% Cr) ferritic/martensitic steels are being considered for nuclear applications for both fission and fusion reactors. Conventional 9-12Cr Cr-Mo steels were the first candidates for these applications. For fusion reactors, reduced-activation steels were developed that were patterned on the conventional steels but with molybdenum replaced by tungsten and niobium replaced by tantalum. Both the conventional and reduced-activation steels are considered to have an upper operating temperature limit of about 550degC. For improved reactor efficiency, higher operating temperatures are required. For ferritic/martensitic steels that could meet such requirements, oxide dispersion-strengthened (ODS) steels are being considered. In this paper, the ferritic/martensitic steels that are candidate steels for nuclear applications will be reviewed, the prospect for ODS steel development and the development of steels produced by conventional processes will be discussed. (author)

Corrosion fatigue crack growth in clad low-alloy steels: Part 1, medium-sulfur forging steel

International Nuclear Information System (INIS)

James, L.A.; Poskie, T.J.; Auten, T.A.; Cullen, W.H.

1996-01-01

Corrosion fatigue crack propagation tests were conducted on a medium- sulfur ASTM A508-2 forging steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 30.3--38.3 mm, and depths of 13.1--16.8 mm. The experiments were conducted in a quasi-stagnant low-oxygen (O 2 < 10 ppb) aqueous environment at 243 degrees C, under loading conditions (ΔK, R, and cyclic frequency) conductive to environmentally-assisted cracking (EAC) in higher-sulfur steels under quasi-stagnant conditions. Earlier experiments on unclad compact tension specimens of this heat of steel did not exhibit EAC, and the present experiments on semi-elliptical surface cracks penetrating cladding also did not exhibit EAC

The magnetized steel and scintillator calorimeters of the MINOS experiment

Energy Technology Data Exchange (ETDEWEB)

Michael, : D.G.

2008-05-01

The Main Injector Neutrino Oscillation Search (MINOS) experiment uses an accelerator-produced neutrino beam to perform precision measurements of the neutrino oscillation parameters in the 'atmospheric neutrino' sector associated with muon neutrino disappearance. This long-baseline experiment measures neutrino interactions in Fermilab's NuMI neutrino beam with a near detector at Fermilab and again 735 km downstream with a far detector in the Soudan Underground Laboratory in northern Minnesota. The two detectors are magnetized steel-scintillator tracking calorimeters. They are designed to be as similar as possible in order to ensure that differences in detector response have minimal impact on the comparisons of event rates, energy spectra and topologies that are essential to MINOS measurements of oscillation parameters. The design, construction, calibration and performance of the far and near detectors are described in this paper.

ATLAS IBL operational experience

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00237659; The ATLAS collaboration

2017-01-01

The Insertable B-Layer (IBL) is the inner most pixel layer in the ATLAS experiment, which was installed at 3.3 cm radius from the beam axis in 2014 to improve the tracking performance. To cope with the high radiation and hit occupancy due to proximity to the interaction point, a new read-out chip and two different silicon sensor technologies (planar and 3D) have been developed for the IBL. After the long shut-down period over 2013 and 2014, the ATLAS experiment started data-taking in May 2015 for Run-2 of the Large Hadron Collider (LHC). The IBL has been operated successfully since the beginning of Run-2 and shows excellent performance with the low dead module fraction, high data-taking efficiency and improved tracking capability. The experience and challenges in the operation of the IBL is described as well as its performance.

Survival of flexible, braided, bonded stainless steel lingual retainers : a historic cohort study

NARCIS (Netherlands)

Foek, D. J. Lie Sam; Ozcan, M.; Verkerke, G. J.; Sandham, John; Dijkstra, P. U.

The objectives of this study were to retrospectively evaluate the clinical survival rate of flexible, braided, rectangular bonded stainless steel lingual retainers, and to investigate the influence of gender, age of the patient, and operator experience on survival after orthodontic treatment at the

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Regulatory challenges in using nuclear operating experience

International Nuclear Information System (INIS)

2006-01-01

The fundamental objective of all nuclear safety regulatory bodies is to ensure that nuclear utilities operate their plants in an acceptably safe manner at all times. Learning from experience has been a key element in meeting this objective. It is therefore very important for nuclear power plant operators to have an active programme for collecting, analysing and acting on the lessons of operating experience that could affect the safety of their plants. NEA experts have noted that almost all of the recent, significant events reported at international meetings have occurred earlier in one form or another. Counteractions are usually well-known, but information does not always seem to reach end users, or corrective action programmes are not always rigorously applied. Thus, one of the challenges that needs to be met in order to maintain good operational safety performance is to ensure that operating experience is promptly reported to established reporting systems, preferably international in order to benefit from a larger base of experience, and that the lessons from operating experience are actually used to promote safety. This report focuses on how regulatory bodies can ensure that operating experience is used effectively to promote the safety of nuclear power plants. While directed at nuclear power plants, the principles in this report may apply to other nuclear facilities as well. (author)

Experimental Investigation on Friction and Wear Properties of Different Steel Materials

Directory of Open Access Journals (Sweden)

M.A. Chowdhury

2013-03-01

Full Text Available Friction coefficient and wear rate of different steel materials are investigated and compared in this study. In order to do so, a pin on disc apparatus is designed and fabricated. Experiments are carried out when different types of disc materials such as stainless steel 314 (SS 314, stainless steel 202 (SS 202 and mild steel slide against stainless steel 314 (SS 314 pin. Experiments are conducted at normal load 10, 15 and 20 N, sliding velocity 1, 1.5 and 2 m/s and relative humidity 70%. At different normal loads and sliding velocities, variations of friction coefficient with the duration of rubbing are investigated. The obtained results show that friction coefficient varies with duration of rubbing, normal load and sliding velocity. In general, friction coefficient increases for a certain duration of rubbing and after that it remains constant for the rest of the experimental time. The obtained results reveal that friction coefficient decreases with the increase in normal load for all the tested materials. It is also found that friction coefficient increases with the increase in sliding velocity for all the materials investigated. Moreover, wear rate increases with the increase in normal load and sliding velocity for SS 314, SS 202 and mild steel. In addition, at identical operating condition, the magnitudes of friction coefficient and wear rate are different for different materials depending on sliding velocity and normal load.

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

Fragmentation of armor piercing steel projectiles upon oblique perforation of steel plates

Directory of Open Access Journals (Sweden)

Aizik F.

2012-08-01

Full Text Available In this study, a constitutive strength and failure model for a steel core of a14.5 mm API projectile was developed. Dynamic response of a projectile steel core was described by the Johnson-Cook constitutive model combined with principal tensile stress spall model. In order to obtain the parameters required for numerical description of projectile core material behavior, a series of planar impact experiments was done. The parameters of the Johnson-Cook constitutive model were extracted by matching simulated and experimental velocity profiles of planar impact. A series of oblique ballistic experiments with x-ray monitoring was carried out to study the effect of obliquity angle and armor steel plate thickness on shattering behavior of the 14.5 mm API projectile. According to analysis of x-ray images the fragmentation level increases with both steel plate thickness and angle of inclination. The numerical modeling of the ballistic experiments was done using commercial finite element code, LS-DYNA. Dynamic response of high hardness (HH armor steel was described using a modified Johnson-Cook strength and failure model. A series of simulations with various values of maximal principal tensile stress was run in order to capture the overall fracture behavior of the projectile’s core. Reasonable agreement between simulated and x-ray failure pattern of projectile core has been observed.

Study on antioxidant experiment on forged steel tube sheet and tube hole for steam generator

International Nuclear Information System (INIS)

Zong Hai; Wang Detai; Ding Yang

2012-01-01

Antioxidant experiment on forged steel tube sheet and tube hole for steam generator was studied and the influence of different simulated heat treatments on the antioxidant performance of tube sheet and tube hole was made. The influence of different antioxidant methods on the size of tube hole was drawn. Furthermore, the change of size and weight of 18MnD5 forged steel tube sheet on the condition of different simulated heat treatments was also studied. The analytical results have proved reference information for the use of 18MnD5 material and for key processes of processing tube hole and wearing and expanding U-style tube. (authors)

A Simple Experiment To Measure the Content of Oxygen in the Air Using Heated Steel Wool

Science.gov (United States)

Vera, Francisco; Rivera, Rodrigo; Nunez, Cesar

2011-01-01

The typical experiment to measure the oxygen content in the atmosphere uses the rusting of steel wool inside a closed volume of air. Two key aspects of this experiment that make possible a successful measurement of the content of oxygen in the air are the use of a closed atmosphere and the use of a chemical reaction that involves the oxidation of…

Corrosion resistance of stainless steel pipes in soil

Energy Technology Data Exchange (ETDEWEB)

Sjoegren, L.; Camitz, G. [Swerea KIMAB AB, Box 55970, SE-102 16 Stockholm (Sweden); Peultier, J.; Jacques, S.; Baudu, V.; Barrau, F.; Chareyre, B. [Industeel and ArcelorMittal R and D, 56 rue Clemenceau, BP19, FR-71201 le Creusot, Cedex (France); Bergquist, A. [Outokumpu Stainless AB, P.O. Box 74, SE-774 22 Avesta (Sweden); Pourbaix, A.; Carpentiers, P. [Belgian Centre for Corrosion Study, Avenue des Petits-Champs 4A, BE 1410 Waterloo (Belgium)

2011-04-15

To be able to give safe recommendations concerning the choice of suitable stainless steel grades for pipelines to be buried in various soil environments, a large research programme, including field exposures of test specimens buried in soil in Sweden and in France, has been performed. Resistance against external corrosion of austenitic, super austenitic, lean duplex, duplex and super duplex steel grades in soil has been investigated by laboratory tests and field exposures. The grades included have been screened according to their critical pitting-corrosion temperature and according to their time-to-re-passivation after the passive layer has been destroyed locally by scratching. The field exposures programme, being the core of the investigation, uses large specimens: 2 m pipes and plates, of different grades. The exposure has been performed to reveal effects of aeration cells, deposits or confined areas, welds and burial depth. Additionally, investigations of the tendency of stainless steel to corrode under the influence of alternating current (AC) have been performed, both in the laboratory and in the field. Recommendations for use of stainless steels under different soil conditions are given based on experimental results and on operating experiences of existing stainless steel pipelines in soil. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

OPTIMIZATION OF SURFACE ROUGHNESS OF AISI 304 AUSTENITIC STAINLESS STEEL IN DRY TURNING OPERATION USING TAGUCHI DESIGN METHOD

Directory of Open Access Journals (Sweden)

D. PHILIP SELVARAJ

2010-09-01

Full Text Available The present work is concentrated with the dry turning of AISI 304 Austenitic Stainless Steel (ASS. This paper presents the influence of cutting parameters like cutting speed, feed rate and depth of cut on the surface roughness of austenitic stainless steel during dry turning. A plan of experiments based on Taguchi’s technique has been used to acquire the data. An orthogonal array, the signal to noise (S/N ratio and the analysis of variance (ANOVA are employed to investigate the cutting characteristics of AISI 304 austenitic stainless steel bars using TiC and TiCN coated tungsten carbide cutting tool. Finally the confirmation tests that have been carried out to compare the predicted values with the experimental values confirm its effectiveness in the analysis of surface roughness.

Prevention of radioactivity in steel. Necessity of an international co-operation between industry and government

International Nuclear Information System (INIS)

Amedro, G.

1999-01-01

European steel companies will not melt radioactive contaminated scrap, even if it would be allowed to do so by legislation and proposed clearance levels. Scrap delivering companies as well as steel producing companies are well armed with scrap charge control. Additional control during steelmaking is given by analysing a crude steel sample of each heat. Regulations regarding health criteria are available on national and European level (RP43 and RP89) but are not usable in practice for the control of scrap. Problems which are at present left to be solved by each site equipped with a means of detection cover principally the following areas: Definition and detection of abnormal radioactivity without identification of the radioactive element; Common definition of alarm thresholds by suppliers and customers; Emergency isolation measures for suspect vehicles (e.g. immediate return or quarantine), notably involving illicit material from a third country; Identification, isolation, handling and destination of discovered radioactive products; Financing of associated costs; Public relations; Preventive actions. It seems to be necessary to produce a benchmark for the activity level in finished steel to determine the actual level of 'normal radioactivity' in the European steel pool. A comparison of the actual state of the art in the European steel industry due to the measures already taken would be useful. The present situation can only be efficiently improved by action in common. Industry has made large investments in detection equipment and in staff training. It has now practical experience. Additional regulatory provisions and appropriate logistical means are now awaited from government, i.e. concerning technical know-how and expertise, properly adapted equipment and in certain cases financial intervention. European directives should not constitute an impediment and should, if necessary, be adapted to needs. When the situation is clarified, future regular checks will be an

Magnet operating experience review for fusion applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1991-11-01

This report presents a review of magnet operating experiences for normal-conducting and superconducting magnets from fusion, particle accelerator, medical technology, and magnetohydrodynamics research areas. Safety relevant magnet operating experiences are presented to provide feedback on field performance of existing designs and to point out the operational safety concerns. Quantitative estimates of magnet component failure rates and accident event frequencies are also presented, based on field experience and on performance of similar components in other industries

The Wonderland of Operating the ALICE Experiment

CERN Document Server

Augustinus, A; Pinazza, O; Rosinský, P; Lechman, M; Jirdén, L; Chochula, P

2011-01-01

ALICE is one of the experiments at the Large Hadron Collider (LHC), CERN, Geneva, Switzerland. Composed of 18 sub-detectors each with numerous subsystems that need to be controlled and operated in a safe and efficient way. The Detector Control System (DCS) is the key to this and has been used by detector experts with success during the commissioning of the individual detectors. During the transition from commissioning to operation, more and more tasks were transferred from detector experts to central operators. By the end of the 2010 datataking campaign, the ALICE experiment was run by a small crew of central operators, with only a single controls operator. The transition from expert to non-expert operation constituted a real challenge in terms of tools, documentation and training. A relatively high turnover and diversity in the operator crew that is specific to the HEP experiment environment (as opposed to the more stable operation crews for accelerators) made this challenge even bigger. Thi...

Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

International Nuclear Information System (INIS)

Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

1995-01-01

In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

Operating experience in reprocessing

International Nuclear Information System (INIS)

Schueller, W.

1983-01-01

Since 1953, reprocessing has accumulated 180 years of operating experience in ten plants, six of them with 41 years of operation in reprocessing oxide fuel from light water reactors. After abortive, premature attempts at what is called commercial reprocessing, which had been oriented towards the market value of recoverable uranium and plutonium, non-military reprocessing technologies have proved their technical feasibility, since 1966 on a pilot scale and since 1976 on an industrial scale. Reprocessing experience obtained on uranium metal fuel with low and medium burnups can now certainly be extrapolated to oxide fuel with high burnup and from pilot plants to industrial scale plants using the same technologies. The perspectives of waste management of the nuclear power plants operated in the Federal Republic of Germany should be viewed realistically. The technical problems still to be solved are in a balanced relationship to the benefit arising to the national economy out of nuclear power generation and can be solved in time, provided there are clearcut political boundary conditions. (orig.) [de

Examination of the X-ray piping diagnostic system using EGS4 (measuring the thickness of a steel pipe with rust)

International Nuclear Information System (INIS)

Kajiwara, G.

2001-01-01

In a series of papers entitled 'Examination of the X-ray piping diagnostic system using EGS4' presented the proceedings of the EGS4 users' meetings, I discussed the possibility of measuring the thickness of piping walls with rust. In the present paper, I describe, based on our earlier results, how the thickness of steel pipes with rust can be measured. I conducted EGS4 simulation to measure the thickness of a combination of steel and rust and made an energy absorption diagram for this combination. The equivalent thickness of steel was obtained through experiments and the system operation. The thickness of the steel determined by using the diagram agreed well with the actual steel thickness obtained by the experiments. In the future, we will focus on how to automate this measurement procedure and how to use the same procedure to measure the thickness of pipes filled with water. (author)

Interfacial interactions between some La-based perovskite thick films and ferritic steel substrate with regard to the operating conditions of SOFC

International Nuclear Information System (INIS)

Przybylski, K.; Brylewski, T.; Morgiel, J.

2004-01-01

An overview is presented on the oxidation kinetics, electrical properties and microstructure investigations of the oxide products formed on Fe-25 wt.-%Cr steel uncoated and coated with electrical conducting films of (La,Ca)CrO 3 or (La,Sr)CoO 3 in air and H 2 /H 2 O gas mixture at 1023-1173 K for up to 480 hrs with regard to their application as the SOFC metallic interconnect. The application of the Fe-25Cr steel in SOFC operating at 1073 K requires its surface modification to improve the electrical conductivity of chromia scale forming on the uncoated steel surface. The thick films of (La,Ca)CrO 3 and (La,Sr)CoO 3 with the thickness range of 20-100 μm, coated on the Fe-25Cr steel by screen-printing method helped solve this problem. TEM-SAD, SEM-EDS and impedance spectroscopy investigations have shown significant influence of the multilayer products formation at the substrate steel/coating films interfacial zone on the electrical properties of the metallic interconnect. (orig.)

A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

Science.gov (United States)

Şahin, Sümer; Übeyli, Mustafa

2008-12-01

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

Operational reliability of high pressure steam lines of pearlitic steels after 150-200 thousand h service

International Nuclear Information System (INIS)

Veksler, E.Ya.; Chajkovskij, V.M.; Osasyuk, V.V.

1980-01-01

Usage of both calculational and physical methods is recommended to estimate a service operating life of long-term working steam line materials. Application of these methods is demonstrated when studying steam line bends made of 12MKh and 12Kh1MF pearlitic steels. Good coincidence of results for the determination of residual durability of steam lines is obtained using these two methods [ru

Static behavior of the weld in the joint of the steel support element using experiment and numerical modeling

Science.gov (United States)

Krejsa, M.; Brozovsky, J.; Mikolasek, D.; Parenica, P.; Koubova, L.

2018-04-01

The paper is focused on the numerical modeling of welded steel bearing elements using commercial software system ANSYS, which is based on the finite element method - FEM. It is important to check and compare the results of FEM analysis with the results of physical verification test, in which the real behavior of the bearing element can be observed. The results of the comparison can be used for calibration of the computational model. The article deals with the physical test of steel supporting elements, whose main purpose is obtaining of material, geometry and strength characteristics of the fillet and butt welds including heat affected zone in the basic material of welded steel bearing element. The pressure test was performed during the experiment, wherein the total load value and the corresponding deformation of the specimens under the load was monitored. Obtained data were used for the calibration of numerical models of test samples and they are necessary for further stress and strain analysis of steel supporting elements.

Heavy reflector experiments composed of carbon steel and nickel in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Silva, Graciete Simoes de Andrade e; Mura, Luis Felipe; Jerez, Rogerio; Mendonca, Arlindo Gilson; Fuga, Rinaldo

2013-01-01

The heavy reflector experiments performed in the IPEN/Mb-01 research reactor facility comprise a set of critical configurations employing the standard 28x26-fuel-rod configuration. The heavy reflector either, carbon steel or nickel plates was placed at one of the faces of the IPEN/MB-01 reactor. Criticality is achieved by inserting the control banks BC1 and BC2 to the critical position. 32 plates around 0.3 mm thick were used in all the experiment. The chosen distance between last fuel rod row and the first laminate for all types of laminates was 5.5 mm. Considering initially the carbon steel case, the experimental data reveal that the reactivity decreases up to the fifth plate and after that it increases, becomes nearly zero (which was equivalent to initial zero excess reactivity with zero plates) for the 28 plates case and reaches a value of 42.73 pcm when the whole set of 32 plates are inserted in the reflector. This is a very striking result because it demonstrates that when all 32 plates are inserted in the reflector there is a net gain of reactivity. The reactivity behavior demonstrates all the physics events already mentioned in this work. When the number of plates are small (around 5), the neutron absorption in the plates is more important than the neutron reflection and the reactivity decreases. This condition holds up to a point where the neutron reflection becomes more important than the neutron absorption in the plates and the reactivity increases. The experimental data for the nickel case shows the main features of the carbon steel case, but for the carbon steel case the reactivity gain is small, thus demonstrating that carbon steel or essentially iron has not the reflector capability as the nickel laminates do. The measured data of nickel plates show a higher reactivity gain, thus demonstrating that nickel is a better reflector than iron. The theoretical analysis employing MCNP5 and ENDF/B-VII.0 show that the calculated results have good results up to

Kayenta advanced series compensation operational experience

Energy Technology Data Exchange (ETDEWEB)

1994-01-01

The world's first three-phase, thyristor-controlled series compensation scheme with continuously variable impedance has been introduced into a transmission system. Energized and dedicated in September 1992, the installation was placed into commercial operation in January 1993 and has provided over one year of operating experience. This paper describes the 230 kV, 330 MVAr (60 Hz) advanced series compensation (ASC) project, located in north-eastern Arizona at Kayenta Substation on the 320 km Glen Canyon-Shiprock transmission line. The paper describes operating experiences, coordination with phase shifting transformer, phase shifter failure, platform power, system disturbances, and future plans.

Evaluation of Radiation Exposure during Construction and Operation of Concrete Bridge Reinforced with Very Low Level Radioactive Steel

International Nuclear Information System (INIS)

Panik, M.; Necas, V.

2012-01-01

A lot of nuclear power plants are approaching the end of their lifetime and they will be phased out. Decommissioning of these nuclear power plants involve complete dismantling of technologies and demolition of buildings. During this process it is produced plenty of waste material of different categories. Significant portion of decommissioning materials comprise radionuclides what is caused by contamination and activation processes mostly from the operational period of nuclear power plant. Attention in this paper is paid to waste steel from the decommissioning of nuclear power plants with the specific activity just slightly exceeding legislation limits for the unconditional release into the environment. From the traditional point of view this material should be treated, conditioned and disposed on the radioactive waste repository. Second possibility is to release this material conditionally and reuse it in chosen industrial application. Very low level radioactive steel scrap should be melted and melting products should be processed into products that can be applied in industry. First option requires considerable financial investment, human resources and repository capacity. Second option saves some financial funds and it enables to reuse and save potentially valuable material for the future. Paper comprises evaluation of external and internal exposure during construction and operation of concrete bridges that utilize very low level radioactive steel as part of their reinforcement. Two models of representative concrete bridges were created. External gamma exposure and exposure from inhalation and ingestion of radionuclides were calculated using suitable computational tools. VISIPLAN 3D ALARA planning tool was chosen for the calculation of external gamma exposure. Software GOLDSIM enables to calculate transport of radionuclides initially contained in conditionally released reinforcement steel through subsoil and sequential exposure of people caused by inhalation of

TSTA Piping and Flame Arrestor Operating Experience Data

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee C.; Willms, R. Scott

2014-10-01

The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility operated from 1984 to 2001, running a prototype fusion fuel processing loop with ~100 grams of tritium as well as small experiments. There have been several operating experience reports written on this facility’s operation and maintenance experience. This paper describes analysis of two additional components from TSTA, small diameter gas piping that handled small amounts of tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The operating experiences and the component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.

Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

International Nuclear Information System (INIS)

Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

1999-01-01

A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

Steel structures for nuclear facilities

International Nuclear Information System (INIS)

1993-01-01

In the guide the requirements concerning design and fabrication of steel structures for nuclear facilities and documents to be submitted to the Finnish Centre for Radiation and Nuclear Safety (STUK) are presented. Furthermore, regulations concerning inspection of steel structures during construction of nuclear facilities and during their operation are set forth

Status Summary of FY16 Atom Probe Tomography Studies on UCSB ATR-2 Irradiated RPV Steels

Energy Technology Data Exchange (ETDEWEB)

Wells, Peter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Odette, G. Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-05-01

The University of California Santa Barbara-2 RPV Steel Irradiation experiment was awarded in 2010 by the Nuclear Science User Facility (formerly ATR NSUF) through a competitive peer review proposal process. The experiment involved irradiation of nearly 1300 samples distributed over 13 capsules. The major objective of this experiment was to better understand embrittlement behavior of reactor pressure steels at doses beyond which available data exists yet may be achieved if reactor operating licenses are extended beyond 60 years. The experiment was instrumented during irradiation and active temperature control was used to maintain the temperature at the design temperature. Six samples were selected from a large matrix of materials to perform atom probe tomography (APT) to look at formation of high dose phases. The nature and formation behavior of these phases is discussed.

Nuclear Power Plant Operating Experience from the IAEA/NEA International Reporting System for Operating Experience 2012-2014

International Nuclear Information System (INIS)

2018-03-01

The International Reporting System for Operating Experience (IRS) is an essential element of the international operating experience feedback system for nuclear power plants. Its fundamental objective is to contribute to improving safety of commercial nuclear power plants which are operated worldwide. IRS reports contain information on events of safety significance with important lessons learned which assist in reducing recurrence of events at other plants. This sixth publication, covering the period 2012 - 2014, follows the structure of the previous editions. It highlights important lessons based on a review of the approximately 240 event reports received from the participating countries over this period.

Ontario Hydro CANDU operating experience

International Nuclear Information System (INIS)

Bartholomew, R.W.; Woodhead, L.W.; Horton, E.P.; Nichols, M.J.; Daly, I.N.

1987-01-01

The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

Challenges in Special Steel Making

Science.gov (United States)

Balachandran, G.

2018-02-01

Special bar quality [SBQ] is a long steel product where an assured quality is delivered by the steel mill to its customer. The bars have enhanced tolerance to higher stress application and it is demanded for specialised component making. The SBQ bars are sought for component making processing units such as closed die hot forging, hot extrusion, cold forging, machining, heat treatment, welding operations. The final component quality of the secondary processing units depends on the quality maintained at the steel maker end along with quality maintained at the fabricator end. Thus, quality control is ensured at every unit process stages. The various market segments catered to by SBQ steel segment is ever growing and is reviewed. Steel mills need adequate infrastructure and technological capability to make these higher quality steels. Some of the critical stages of processing SBQ and the critical quality maintenance parameters at the steel mill in the manufacture has been brought out.

Recent Niobium Developments for High Strength Steel Energy Applications

Science.gov (United States)

Jansto, Steven G.

Niobium-containing high strength steel materials have been developed for oil and gas pipelines, offshore platforms, nuclear plants, boilers and alternative energy applications. Recent research and the commercialization of alternative energy applications such as windtower structural supports and power transmission gear components provide enhanced performance. Through the application of these Nb-bearing steels in demanding energy-related applications, the designer and end user experience improved toughness at low temperature, excellent fatigue resistance and fracture toughness and excellent weldability. These enhancements provide structural engineers the opportunity to further improve the structural design and performance. For example, through the adoption of these Nb-containing structural materials, several design-manufacturing companies are initiating new windtower designs operating at higher energy efficiency, lower cost, and improved overall material design performance.

Operational experience with superconducting synchrotron magnets

International Nuclear Information System (INIS)

Martin, P.S.

1987-01-01

The operational experience with the Fermilab Tevatron is presented, with emphasis on reliability and failure modes. Comparisons are made between the operating efficiencies for the superconducting machine and for the conventional Main Ring

Operational experience with superconducting synchrotron magnets

International Nuclear Information System (INIS)

Martin, P.S.

1987-03-01

The operational experience with the Fermilab Tevatron is presented, with emphasis on reliability and failure modes. Comprisons are made between the operating efficiencies for the superconducting machine and for he conventional Main Ring

Development of PWR pressure vessel steels

International Nuclear Information System (INIS)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

Development of PWR pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

Characterization Of Oxide Layers Formed On 13CrMo4-5 Steel Operated For A Long Time At An Elevated Temperature

Directory of Open Access Journals (Sweden)

Gwoździk M.

2015-09-01

Full Text Available The paper contains results of studies into the formation of oxide layers on 13CrMo4-5 (15HM steel long-term operated at an elevated temperature. The oxide layer was studied on a surface and a cross-section at the inner and outer surface of the tube wall. The 13CrMo4-5 steel operated at the temperature of 470°C during 190,000 hours was investigated. X-ray structural examinations (XRD were carried out, microscope observation s using an optical, scanning microscope were performed. The native material chemical composition was analysed by means of emission spark spectroscopy, while that of oxide layers on a scanning microscope (EDS. The studies on the topography of the oxide layers comprised studies on the roughness plane, which were carried out using a AFM microscope designed for 2D and 3D studies on the surface. Mechanical properties of the oxide layer – steel (substrate were characterised on the basis of scratch test. The adhesion of oxide layers, friction force, friction coefficient, scratching depth were determined as well as the force at which the layer was delaminated.

Operating experience insights supporting ageing assessments

International Nuclear Information System (INIS)

Nitoi, M.

2013-01-01

Be effective in ageing management means looking at the right aspects, with the right techniques, and one of the most effective tool which could be used for that purpose is the analysis of operating experience. The paper has as objective to perform a review of available operating experience, with the aim to provide a better picture about the impact of ageing effects. The IAEA International Reporting System and NRC Licensee Event Reports were chosen as reference databases, both databases being internationally recognized as important sources of information about events occurrences in the nuclear power plants. The ageing related events identified in the selected time window were analyzed in detail, and the contributions of each major degradation mechanisms that have induced the ageing related events (specific to each defined group of components) was represented and discussed. The paper demonstrates the possibility to use operating experience insights in highlighting the ageing effects. (authors)

Research on common methods for evaluating the operation effect of integrated wastewater treatment facilities of iron and steel enterprises

Science.gov (United States)

Bingsheng, Xu

2017-04-01

Considering the large quantities of wastewater generated from iron and steel enterprises in China, this paper is aimed to research the common methods applied for evaluating the integrated wastewater treatment effect of iron and steel enterprises. Based on survey results on environmental protection performance, technological economy, resource & energy consumption, services and management, an indicator system for evaluating the operation effect of integrated wastewater treatment facilities is set up. By discussing the standards and industrial policies in and out of China, 27 key secondary indicators are further defined on the basis of investigation on main equipment and key processes for wastewater treatment, so as to determine the method for setting key quantitative and qualitative indicators for evaluation indicator system. It is also expected to satisfy the basic requirements of reasonable resource allocation, environmental protection and sustainable economic development, further improve the integrated wastewater treatment effect of iron and steel enterprises, and reduce the emission of hazardous substances and environmental impact.

High temperature oxidation behavior of ODS steels

Science.gov (United States)

Kaito, T.; Narita, T.; Ukai, S.; Matsuda, Y.

2004-08-01

Oxide dispersion strengthened (ODS) steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Oxidation testing of ODS steel was conducted under a controlled dry air atmosphere to evaluate the high temperature oxidation behavior. This showed that 9Cr-ODS martensitic steels and 12Cr-ODS ferritic steels have superior high temperature oxidation resistance compared to 11 mass% Cr PNC-FMS and 17 mass% Cr ferritic stainless steel. This high temperature resistance is attributed to earlier formation of the protective α-Cr 2O 3 on the outer surface of ODS steels.

ETSON proposal on the European operational experience feedback system

International Nuclear Information System (INIS)

Maqua, Michael; Bertrand, Remy; Gelder, Pieter de

2007-01-01

The new IAEA Safety Fundamentals states regarding the operating experience feedback: The feedback of operating experience from facilities and activities - and, where relevant, from elsewhere - is a key means of enhancing safety. Processes must be put in place for the feedback and analysis of operating experience, including initiating events, accident precursors, near misses, accidents and unauthorized acts, so that lessons may be learned, shared and acted upon. This presentation deals with the proposal of the ETSON (European TSO Network) to optimize the European operating experiences feedback (OEF). It is generally recognized that the efficiency of nuclear safety supervision by public authorities is based on two key requirements: - the existence of a competent authority at national level, benefiting from an appropriate legislative and regulatory basis, from adequate (quantitatively and qualitatively) human resources, particularly for inspection purposes, - the availability of resources devoted to highly specialised independent technical expertise, in order to provide competent authorities with pertinent technical opinions on: -- the safety files provided by operators, for the purpose of licensing corresponding activities, -- the exploitation for regulatory purposes of the operating experience feed back from licensed nuclear installations. There are two worldwide systems intended to learn lessons from experience: the WANO (World Association of Nuclear Operators) system established by the licensees with access restricted to operating organizations and the IRS system jointly operated by IAEA and OECD/NEA accessible to regulators and to some other users nominated by the regulators in their countries. The IRS itself is dedicated to the analysis of safety significant operating events. NEA/CNRA runs a permanent working group on operating experience (WGOE). WGOE provides among other things also generic reports on safety concerns related to operating experiences and

Ontario Hydro CANDU operating experience

International Nuclear Information System (INIS)

Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

1984-03-01

The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

Operating experience and TPA: the Italian perspective

International Nuclear Information System (INIS)

Grimaldi, G.

1990-01-01

Collection and analysis of operating experience from the Italian plants and utilization of abroad data both to plants in operation and in construction are presented. Some results are also referred, aimed to evidence the role of the international cooperation to safe operation of nuclear plants. The approach to the Trend and Pattern analyses is described as well, and the use of computerized techniques of analysis on personal computer. Finally on going activities are introduced, specifically application of operating experience of plants in operation to small sized reactors and to ones with more intrinsic safety characteristics; review of the reporting system for future application and comparative analysis of the different realization of selected safety systems

Corrosion behavior of steels in flowing lead-bismuth under abnormal conditions

International Nuclear Information System (INIS)

Doubkova, A.; Di Gabriele, F.; Brabec, P.; Keilova, E.

2008-01-01

The project IP EUROTRANS, domain DEMETRA, is primary focused on the study of the technology of the interaction between steels and heavy liquid metals. The characterization of the metal response to sudden changes, simulating accidental conditions in liquid lead-bismuth eutectic was carried out. This paper reports the results of two hot-spot simulations with two different oxygen concentrations (10 -8 wt%, 10 -6 wt%). Each experiment was divided in two main periods: the initial, long period at the standard operating temperature 550 deg. C; the second, short period, at higher temperature, 650 deg. C. The damage that occurs on the austenitic steel AISI 316L and the ferritic-martensitic steel T91 was investigated. The amount of damage for both steels was higher at lower oxygen contents and the short, hot spot simulation, markedly affected the T91. At higher oxygen content the amount of damage decreased. A few, localized pits, were observed; however, there was no visible increment in the amount of damage after the hot spot simulation

FFTF operational experience

International Nuclear Information System (INIS)

Newland, D.J.; Krupar, J.J.

1984-01-01

In April 1982, the FFTF began its first nominally 100 day irradiation cycle. Since that time the plant has operated very well with steadily increasing plant capacity factors during its first four cycles. One hundred fifty fuel assemblies (eighty of which are experiments) and over 32,000 individual fuel pins have been irradiated, some in excess of 100 MWd/Kg burnup. Specialized equipment and systems unique to sodium cooled reactor plants have performed well

Steam generator operating experience update, 1982-1983

International Nuclear Information System (INIS)

Frank, L.

1984-06-01

This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed

Impact of LMFBR operating experience on PFBR design

International Nuclear Information System (INIS)

Bhoje, S.B.; Chetal, S.C.; Chellapandi, P.; Govindarajan, S.; Lee, S.M.; Kameswara Rao, A.S.L.; Prabhakar, R.; Raghupathy, S.; Sodhi, B.S.; Sundaramoorthy, T.R.; Vaidyanathan, G.

2000-01-01

PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

BEHAVIOR OF STEEL DP 600 UNDER DYNAMIC CONDITIONS

Directory of Open Access Journals (Sweden)

Miroslav Német

2014-01-01

Full Text Available Normal 0 21 false false false MicrosoftInternetExplorer4 Dynamic tensile testing of sheet steels is becoming more important. Experimental dynamic tensile technique is depending on the strain rate. For experiments was used two testing method servo hydraulic and single bar method. Experiments was realized on steel grade DP 600. Steel were performed and evaluated static and dynamic tests. Was investigated substructure in static and dynamic loading conditions.

Emergency planning and operating experience

International Nuclear Information System (INIS)

Halpern, O.; Breniere, J.

1984-01-01

The purpose of this paper is to derive lessons from operating experience for the planning of emergency measures. This operating experience has two facets: it is obtained not only from the various incidents and accidents which have occurred in countries with nuclear power programmes and from the resulting application of emergency plans but also from the different exercises and simulations carried out in France and in other countries. Experience generally confirms the main approaches selected for emergency plans. The lessons to be derived are of three types: first, it appears necessary to set forth precisely the responsibilities of each person involved in order to prevent a watering-down of decisions in the event of an accident; secondly, considerable improvements need to be made in the different communication networks to be used; and thirdly, small accidents with minor radiological consequences deserve as systematic and thorough an approach as large and more improbable accidents. (author)

Note: Electrostatic detection of stainless steel dust particles for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Landy, P. [Mechanical and Aerospace Engineering Department, Cornell University, Ithaca, New York 14853 (United States); Skinner, C. H.; Schneider, H. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2014-03-15

Dust accumulation inside next-step fusion devices poses a significant safety concern and dust diagnostics will be needed to assure safe operations. An electrostatic dust detection device has been successfully demonstrated in the National Spherical Torus Experiment, Tore Supra, and the Large Helical Device, and the detector's response to carbon particles was previously characterized in laboratory experiments. This paper presents laboratory results showing that detection of stainless steel particles at levels as low as several μg/cm{sup 2} is also possible.

High - speed steel for precise cased tools

International Nuclear Information System (INIS)

Karwiarz, J.; Mazur, A.

2001-01-01

The test results of high-vanadium high - speed steel (SWV9) for precise casted tools are presented. The face -milling cutters of NFCa80A type have been tested in industrial operating conditions. An average life - time of SWV9 steel tools was 3-10 times longer compare to the conventional high - speed milling cutters. Metallography of SWB9 precise casted steel revealed beneficial for tool properties distribution of primary vanadium carbides in the steel matrix. Presented results should be a good argument for wide application of high - vanadium high - speed steel for precise casted tools. (author)

Operating practical experience at Argentina

International Nuclear Information System (INIS)

Quihillalt, Oscar

1997-01-01

Operating experiences of Atucha-1 and Embalse Nuclear Power Plants were discussed in this work. The technical and economic aspects, such as reliability, availability, personnel training, operating costs, prices and market, which exercise influence upon Argentina nuclear energy policy, mainly on the power electric generation by nuclear power plants were considered. Finally the current status of the nucleoelectric sector in Argentina and forecasting were analysed

Aging in PWR conditions of martensitic stainless steels

International Nuclear Information System (INIS)

Boursier, J.M.; Buisine, D.; Fronteau, M.; Michel, D.; Rouillon, Y.; Yrieix, B.; Meyzaud, Y.

1998-01-01

Martensitic stainless steels are largely used in Nuclear Power Plant (pump impeller, valve stem...) because of their high mechanical characteristics and their good resistance to corrosion. Nevertheless some of those components could operate at temperature higher than 250 deg.C, which could embrittle the material by the precipitation of a chromium-rich phase during aging. In collaboration with Framatome, Electricite de France has undertaken numerous studies in order to understand this process of embrittlement. This paper presents a review of the metallurgical investigations on martensitic stainless steels components which were performed in the EDF hot laboratory. In peculiar, it should be noted the good correlation between inservice experience and the modelling developed by EDF R and D division. Finally and in association with safety analysis, these results will allow to establish the maintenance strategy of the French Nuclear Power Plants. (authors)

Experiment on electrolysis decontamination of stainless steel pipes

International Nuclear Information System (INIS)

Wang Dongwen; Dou Tianjun; Zhao Yujie

2004-01-01

A new electrolytic decontamination method used metal balls as conducting anode was investigated. The influences of current density, solution property and diameter of pipes on efficiency of electrolytic decontamination were examined and the efficiency of this method was compared with that of common electrolytic method under the same experimental conditions. Decontamination of samples of stainless steel pipes contaminated by plutonium was performed. Experimental results indicate that decontamination of stainless steel pipes contaminated by plutonium can be achieved at the optimum conditions of greater than 0.2 A·cm -2 current density, 5% sulfuric acid electrolyte and 5 min electrolysis. This method can be used in the decontamination of a wide variety of decommissioned metal materials. (author)

Accelerator/Experiment Operations - FY 2016

International Nuclear Information System (INIS)

Blake, A.; Convery, M.; Geer, S.; Geesaman, D.; Harris, D.; Johnson, D.; Lang, K.; McFarland, K.; Messier, M.; Moore, C. D.; Newhart, D.; Reimer, P. E.; Plunkett, R.; Rominsky, M.; Sanchez, M.; Schmidt, J. J.; Shanahan, P.; Tate, C.; Thomas, J.; Donatella Torretta, Donatella Torretta; Matthew Wetstein, Matthew Wetstein

2016-01-01

This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2016. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2016 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the MicroBooNE experiment and the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment, LArIAT experiment and Meson Test Beam activities in the 120 GeV external switchyard beam (SY120). Each section was prepared by the relevant authors, and was then edited for inclusion in this summary.

Accelerator/Experiment Operations - FY 2016

Energy Technology Data Exchange (ETDEWEB)

Blake, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Convery, M. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Geer, S. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Geesaman, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Harris, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Johnson, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Lang, K. [Argonne National Lab. (ANL), Argonne, IL (United States); McFarland, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Messier, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Moore, C. D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Newhart, D. [Fermilab; Reimer, P. E. [Argonne; Plunkett, R. [Fermilab; Rominsky, M. [Fermilab; Sanchez, M. [Iowa State U.; Schmidt, J. J. [Fermilab; Shanahan, P. [Fermilab; Tate, C. [Fermilab; Thomas, J. [University Coll. London; Donatella Torretta, Donatella Torretta [Fermilab; Matthew Wetstein, Matthew Wetstein [Iowa State University

2016-10-01

This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2016. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2016 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the MicroBooNE experiment and the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment, LArIAT experiment and Meson Test Beam activities in the 120 GeV external switchyard beam (SY120). Each section was prepared by the relevant authors, and was then edited for inclusion in this summary.

Properties and application of new bainitic and martensitic creep resistance steels

International Nuclear Information System (INIS)

Pasternak, J.; Dobrzanski, J.

2008-01-01

Supercritical operating parameters of lower emission power units, require novel creep resisting steels to be applied for boiler and pipe systems. Among them are T23 bainitic steels for water walls of boiler combustion chamber and martensitic VM12 steels for superheater coils were tested. RAFAKO S.A. has been co-operating with the Silesian Technical University in Katowice, the Institute of Welding and the Institute for Ferrous Metallurgy in Gliwice for several years now, initiating research and development programmes, implementing the new creep-resistant steels and actively participating in European programmes COST522 and COST536. This paper contains selected information and test results before implementation of the new creep-resistant steels, including: evaluation of working parameters, temperature conditions of main boiler components, which influence reliability and safety, selection of steels for furnace chamber components (approx. 2.5 % Cr) and steam superheater components (9-12 % Cr) destination, evaluation of the requested level of welded joints technological and strength properties, measurements and non-destructive examinations, evaluation of welded joints and HAZ structure by means of LM, TEM and SEM methods in the welding technology implementation process, evaluation of corrosion mechanisms and creep-resistance results - loss of service life - for selected evaporator and steam superheater components, as crucial elements in evaluation of reliability and safety of boiler equipment. Such an examination program includes assessment of steel structure stability during operation period in actual operational conditions. It was clearly shown that operation period have little impact on changes occurring in microstructure and other properties of examined steel grades. (author)

Corrosion of carbon steel in contact with bentonite

International Nuclear Information System (INIS)

Dobrev, D.; Vokal, A.; Bruha, P.

2010-01-01

Document available in extended abstract form only. Carbon steel canisters were chosen in a number of disposal concepts as reference material for disposal canisters. The corrosion rates of carbon steels in water solution both in aerobic and anaerobic conditions are well known, but only scarce data are available for corrosion behaviour of carbon steels in contact with bentonite. A special apparatus, which enables to measure corrosion rate of carbon steels under conditions simulating conditions in a repository, namely in contact with bentonite under high pressure and elevated temperatures was therefore prepared to study: - Corrosion rate of carbon steels in direct contact with bentonite in comparison with corrosion rate of carbon steels in synthetic bentonite pore water. - Influence of corrosion products on bentonite. The apparatus is composed of corrosion chamber containing a carbon steel disc in direct contact with compacted bentonite. Synthetic granitic water is above compacted bentonite under high pressure (50 - 100 bar) to simulate hydrostatic pressure in a repository. The experiments can be carried out under various temperatures. Bentonites used for experiments were Na-type of bentonite Volclay KWK 80 - 20 and Ca-Mg Czech bentonite from deposit Rokle. Before adding water into corrosion system the corrosion chamber was purged by nitrogen gas. The saturation of bentonite and corrosion rate were monitored by measuring consumption of water, pressure increase caused by swelling pressure of bentonite and by generation of hydrogen. Corrosion rate was also determined after corrosion experiments from weight loss of samples. The results of experiments show that the corrosion behaviour of carbon steels in contact with bentonite is very different from corrosion of carbon steels in water simulating bentonite pore water solution. The corrosion rates of carbon steel in contact with bentonite reached after 30 days of corrosion the values approaching 40 mm/yr contrary to values

MIT January Operational Internship Experience 2011

Science.gov (United States)

DeLatte, Danielle; Furhmann, Adam; Habib, Manal; Joujon-Roche, Cecily; Opara, Nnaemeka; Pasterski, Sabrina Gonzalez; Powell, Christina; Wimmer, Andrew

2011-01-01

This slide presentation reviews the 2011 January Operational Internship experience (JOIE) program which allows students to study operational aspects of spaceflight, how design affects operations and systems engineering in practice for 3 weeks. Topics include: (1) Systems Engineering (2) NASA Organization (3) Workforce Core Values (4) Human Factors (5) Safety (6) Lean Engineering (7) NASA Now (8) Press, Media, and Outreach and (9) Future of Spaceflight.

EBR-II: twenty years of operating experience

International Nuclear Information System (INIS)

Lentz, G.L.; Buschman, H.W.; Smith, R.N.

1985-01-01

Experimental Breeder Reactor No. 2 (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. For the last 20 years EBR-II has operated safely, has demonstrated stable operating characteristics, has shown excellent performance of its sodium components, and has had an excellent plant factor. These years of operating experience provide a valuable resource to the nuclear community for the development and design of future liquid metal fast reactors. This report provides a brief description of the EBR-II plant and its early operating experience, describes some recent problems of interest to the nuclear community, and also mentions some of the significant operating achievements of EBR-II. Finally, a few words and speculations on EBR-II's future are offered. 4 figs., 1 tab

Accelerator/Experiment Operations - FY 2015

Energy Technology Data Exchange (ETDEWEB)

Czarapata, P. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); et al.

2015-10-01

This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2015. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2015 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment and Meson Test Beam (MTest) activities in the 120 GeV external Switchyard beam (SY120).

Steel Creek primary producers: Periphyton and seston, L-Lake/Steel Creek Biological Monitoring Program, January 1986--December 1991

Energy Technology Data Exchange (ETDEWEB)

Bowers, J.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Toole, M.A.; van Duyn, Y. [Normandeau Associates Inc., New Ellenton, SC (United States)

1992-02-01

The Savannah River Site (SRS) encompasses 300 sq mi of the Atlantic Coastal Plain in west-central South Carolina. Five major tributaries of the Savannah River -- Upper Three Runs Creek, Four Mile Creek, Pen Branch, Steel Creek, and Lower Three Runs Creek -- drain the site. In 1985, L Lake, a 400-hectare cooling reservoir, was built on the upper reaches of Steel Creek to receive effluent from the restart of L-Reactor and to protect the lower reaches from thermal impacts. The Steel Creek Biological Monitoring Program was designed to assess various components of the system and identify and changes due to the operation of L-Reactor or discharge from L Lake. An intensive ecological assessment program prior to the construction of the lake provided baseline data with which to compare data accumulated after the lake was filled and began discharging into the creek. The Department of Energy must demonstrate that the operation of L-Reactor will not significantly alter the established aquatic ecosystems. This report summarizes the results of six years` data from Steel Creek under the L-Lake/Steel Creek Monitoring Program. L Lake is discussed separately from Steel Creek in Volumes NAI-SR-138 through NAI-SR-143.

Steel Creek primary producers: Periphyton and seston, L-Lake/Steel Creek Biological Monitoring Program, January 1986--December 1991

International Nuclear Information System (INIS)

Bowers, J.A.; Toole, M.A.; van Duyn, Y.

1992-02-01

The Savannah River Site (SRS) encompasses 300 sq mi of the Atlantic Coastal Plain in west-central South Carolina. Five major tributaries of the Savannah River -- Upper Three Runs Creek, Four Mile Creek, Pen Branch, Steel Creek, and Lower Three Runs Creek -- drain the site. In 1985, L Lake, a 400-hectare cooling reservoir, was built on the upper reaches of Steel Creek to receive effluent from the restart of L-Reactor and to protect the lower reaches from thermal impacts. The Steel Creek Biological Monitoring Program was designed to assess various components of the system and identify and changes due to the operation of L-Reactor or discharge from L Lake. An intensive ecological assessment program prior to the construction of the lake provided baseline data with which to compare data accumulated after the lake was filled and began discharging into the creek. The Department of Energy must demonstrate that the operation of L-Reactor will not significantly alter the established aquatic ecosystems. This report summarizes the results of six years' data from Steel Creek under the L-Lake/Steel Creek Monitoring Program. L Lake is discussed separately from Steel Creek in Volumes NAI-SR-138 through NAI-SR-143

Continuous Air Monitor Operating Experience Review

International Nuclear Information System (INIS)

Cadwallader, L.C.; Bruyere, S.A.

2008-01-01

Continuous air monitors (CAMs) are used to sense radioactive particulates in room air of nuclear facilities. CAMs alert personnel of potential inhalation exposures to radionuclides and can also actuate room ventilation isolation for public and environmental protection. This paper presents the results of a CAM operating experience review of the DOE Occurrence Reporting and Processing System (ORPS) database from the past 18 years. Regulations regarding these monitors are briefly reviewed. CAM location selection and operation are briefly discussed. Operating experiences reported by the U.S. Department of Energy and in other literature sources were reviewed to determine the strengths and weaknesses of these monitors. Power losses, human errors, and mechanical issues cause the majority of failures. The average 'all modes' failure rate is 2.65E-05/hr. Repair time estimates vary from an average repair time of 9 hours (with spare parts on hand) to 252 hours (without spare parts on hand). These data should support the use of CAMs in any nuclear facility, including the National Ignition Facility and the international ITER experiment

A comparison of the tribological behaviour of steel/steel, steel/DLC and DLC/DLC contact when lubricated with mineral and biodegradable oils

OpenAIRE

Kalin, Mitjan; Vižintin, Jože

2015-01-01

Diamond-like carbon (DLC) coatings, which can nowadays be applied to many highly loaded mechanical components, sometimes need to operate under lubricated conditions. It is reasonable to expect that in steel/DLC contacts, at least the steel counter body will behave according to conventional lubrication mechanisms and will interact with lubricants and additives in the contact. However, in DLC/DLC contacts, such mechanisms are still unclear. For example, the "inertness" of DLC coatings raises se...

Operation experience with the TRIGA reactor Wien 2004

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-01-01

The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

Tem study of thermal ageing of ferrite in cast duplex stainless steel

International Nuclear Information System (INIS)

Nenonen, P.; Massoud, J.P.; Timofeev, B.T.

2002-01-01

The changes in the microstructure and composition of ferrite in two types of cast duplex stainless steels and in an austenitic-ferritic weld metal after long term thermal ageing has been studied using analytical transmission electron microscope (FEGTEM). A cast test steel containing Mo was investigated first as a reference material in three different conditions: as solution annealed, aged at 300 C and aged at 400 C. This investigation was carried out to gain experience of how EDS (X-ray analyser) analyser and TEM (transmission electron microscope) can be used to study elemental inhomogeneity, which is usually investigated with an atom probe (APFIM). The two other materials, an austenitic-ferritic weld metal and a cast duplex Ti-stabilised stainless steel used for long time at NPP operation temperature were investigated using the experience obtained with the test steel. The results showed that analytical TEM can be used to investigate elemental inhomogeneity of ferrite, but there are several important things to be taken into account when the spectra for this purpose are collected. These things are, such as the thickness of the specimen, probe size, contamination rate, 'elemental background' of the spectrum and possible enrichment of certain alloying elements in the surface oxide layer of the TEM-specimens. If minor elements are also analysed, it may increase the scattering of the results. (authors)

Large-scale thermal-shock experiments with clad and unclad steel cylinders

International Nuclear Information System (INIS)

Cheverton, R.D.

1992-01-01

Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of pressurized-water-reactor pressure vessels have been under investigation at the Oak Ridge National Laboratory for nearly 20 years. During that time, twelve thermal-shock experiments with thick-walled (152 mm) steel cylinders were conducted as a part of the investigations. The first eight experiments were conducted with unclad cylinders initially containing shallow (8--19 mm) two-dimensional and semicircular inner-surface flaws. These experiments demonstrated, in good agreement with linear elastic fracture mechanics, crack initiation and arrest, a series of initiation/arrest events with deep penetration of the wall, long crack jumps, arrest with the stress intensity factor (K I ) increasing with crack depth, extensive surface extension of an initially short and shallow (semicircular) flaw, and warm prestressing with K I ≤ 0. The remaining four experiments were conducted with clad cylinders containing initially shallow (19--24 mm) semielliptical subclad and surface flaws at the inner surface. In the first of these experiments one of six equally spaced (60 degrees) open-quotes identicalclose quotes subclad flaws extended nearly the length of the cylinder (1,220 mm) beneath the cladding (no crack extension into the cladding) and nearly 50% of the wall, radially. For the final experiment, four of the semielliptical subclad flaws that had not propagated previously were converted to surface flaws, and they experienced extensive extension beneath the cladding with no cracking of the cladding. Information from this series of thermal-shock experiments is being used in the evaluation of the PTS issue

Cold formability of steels

International Nuclear Information System (INIS)

Lafond, G.; Leclerq, G.; Moliexe, F.; Namdar, R.; Roesch, L.; Sanz, G.

1977-01-01

This work was essentially aimed to the study of the following three questions. Is it possible to assess the cold formability of steels using simple material properties as criteria. What values of mechanical properties can one expect to reach in cold formed parts. Are there simple ways of characterizing the speroidization treatments carried out on steels before cold forming operations. The present report describes the results obtained during this investigation. It is logically divided into three separate parts. Experimental study of cold formability in wire drawing. Influence of metallurgical variables on mechanical properties of high carbon cold drawn wires. Contribution to the study of characterization methods of cold forming steels subjected to a spheroidization heat treatment

Ductility of high chromium stainless steels

International Nuclear Information System (INIS)

Peretyat'ko, V.N.; Kazantsev, A.A.

1997-01-01

Aimed to optimize the hot working conditions for high chromium stainless steels the experiments were carried in the temperature range of 800-1300 deg C using hot torsion tests and cylindrical specimens of ferritic and ferritic-martensitic steels 08Kh13, 12Kh13, 20Kh13, 30Kh13 and 40Kh13. Testing results showed that steel plasticity varies in a wide range depending on carbon content. Steels of lesser carbon concentration (08Kh13 and 12Kh13) exhibit a sharp increase in plasticity with a temperature rise, especially in the interval of 1200-1250 deg C. Steels 20Kh13 and 30Kh13 display insignificant plasticity increasing, whereas plastic properties of steel 40Kh13 increase noticeably in the range of 1000-1300 deg C. It is shown that optimal hot working conditions for specific steel must be selected with account of steel phase composition at high temperatures

Influence of tempering on mechanical properties of ferritic martensitic steels

International Nuclear Information System (INIS)

Chun, Y. B.; Han, C. H.; Choi, B. K.; Lee, D. W.; Kim, T. K.; Jeong, Y. H.; Cho, S.

2012-01-01

In the mid-1980s research programs for development of low activation materials began. This is based on the US Nuclear Regulatory Commission Guidelines (10CFR part 61) that were developed to reduce long-lived radioactive isotopes, which allows nuclear reactor waste to be disposed of by shallow land burial when removed from service. Development of low activation materials is also key issue in nuclear fusion systems, as the structural components can became radioactive due to nuclear transmutation caused by exposure to high dose neutron irradiation. Reduced-activation ferritic martensitic (RAFM) steels have been developed in the leading countries in nuclear fusion technology, and are now being considered as primary candidate material for the test blanket module (TBM) in the international thermonuclear experiment reactor (ITER). RAFM steels developed so far (e.g., EUROFER 97 and F82H) meet the requirement for structural application in the ITER. However, if such alloys are used in the DEMO or commercial fusion reactor is still unclear, as the reactors are designed to operate under much severe conditions (i.e., higher outlet coolant temperature and neutron fluences). Such harsh operating conditions lead to development of RAFM steels with better creep and irradiation resistances. Mechanical properties of RAFM steels are strongly affected by microstructural features including the distribution, size and type of precipitates, dislocation density and grain size. For a given composition, such microstructural characteristics are determined mainly by thermo-mechanical process employed to fabricate the final product, and accordingly a final heat treatment, i.e., tempering is the key step to control the microstructure and mechanical properties. In the present work, we investigated mechanical properties of the RAFM steels with a particular attention being paid to effects of tempering on impact and creep properties

Research to sustain cases for Magnox-reactor steel pressure vessels

International Nuclear Information System (INIS)

Graham, W.J.

1997-01-01

Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

Operating manual for the critical experiments facility

International Nuclear Information System (INIS)

1986-01-01

The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written

Operating manual for the critical experiments facility

Energy Technology Data Exchange (ETDEWEB)

1986-01-01

The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written.

Accelerator/Experiment operations - FY 2006

Energy Technology Data Exchange (ETDEWEB)

Brice, S.; Conrad, J.; Denisov, D.; Ginther, G.; Holmes, S.; James, C.; Lee, W.; Louis, W.; Moore, C.; Plunkett, R.; Raja, R.; /Fermilab

2006-10-01

This Technical Memorandum (TM) summarizes the Fermilab accelerator and experiment operations for FY 2006. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2006 Run II at the Tevatron Collider, the MiniBooNE experiments running in the Booster Neutrino Beam in neutrino and antineutrino modes, MINOS using the Main Injector Neutrino Beam (NuMI), and SY 120 activities.

Passivation condition of carbon steel in bentonite/sand mixture

International Nuclear Information System (INIS)

Taniguchi, Naoki; Kawakami, Susumu

2002-03-01

It is essential to understand the corrosion type of carbon steel under the repository conditions for the lifetime assessment of carbon steel overpack used for geological isolation of high-level radioactive waste. According to the previous study, carbon steel is hard to passivate in buffer material assuming a chemical condition range of groundwater in Japan. However, concrete support will be constructed around the overpack in the case of repository in the soft rock system and groundwater having a higher pH may infiltrate to buffer material. There is a possibility that the corrosion type of carbon steel will be influenced by the rise of the pH in groundwater. In this study, anodic polarization experiments were performed to understand the passivation condition of carbon steel in buffer material saturated with water contacted with concrete. An ordinary concrete an a low-alkalinity concrete were used in the experiment. The results of the experiments showed that the carbon steel can passivate under the condition that water having pH > 13 infiltrate to the buffer material assuming present property of buffer material. If the low-alkalinity concrete is selected as the support material, passivation can not occur on carbon steel overpack. The effect of the factors of buffer material such as dry density and mixing ratio of sand on the passivation of carbon steel was also studied. The results of the study showed that the present property of buffer material is enough to prevent passivation of carbon steel. (author)

Experimental Investigation on Friction and Wear Properties of Different Steel Materials

OpenAIRE

M.A. Chowdhury; D.M. Nuruzzaman

2013-01-01

Friction coefficient and wear rate of different steel materials are investigated and compared in this study. In order to do so, a pin on disc apparatus is designed and fabricated. Experiments are carried out when different types of disc materials such as stainless steel 314 (SS 314), stainless steel 202 (SS 202) and mild steel slide against stainless steel 314 (SS 314) pin. Experiments are conducted at normal load 10, 15 and 20 N, sliding velocity 1, 1.5 and 2 m/s and relative humidity 70%. A...

An Experiment Study on Surface Roughness in High Speed Milling NAK80 Die Steel

Directory of Open Access Journals (Sweden)

Su Fa

2016-01-01

Full Text Available The paper introduces that the high speed milling experiments on NAK80 die steel was carried out on the DMU 60 mono BLOCK five axis linkage high speed CNC machining center tool by the TiAlN coated tools, in order to research the effect of milling parameters on surface roughness Ra. The results showed that the Ra value increased with the decrease of milling speed vc, increased with the axial depth of milling ap, and feed per tooth fz and radial depth of milling ae. On the basis of the single factor experiment results, the mathematics model for between surface roughness and milling parameters were established by linear regression analysis.

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

1995-01-01

As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

Incorporating operational experience and design changes in availability forecasts

International Nuclear Information System (INIS)

Norman, D.

1988-01-01

Reliability or availability forecasts which are based solely on past operating experience will be precise if the sample is large enough, and unbiased if nothing in the future design, environment, operating region or anything else changes. Unfortunately, life is never like that. This paper considers the methodology and philosophy of modifying forecasts based on past experience to take account also of changes in design, construction methods, operating philosophy, environments, operator training and so on, between the plants which provided the operating experience and the plant for which the forecast is being made. This emphasises the importance of collecting, assessing, and learning from past data and of a thorough knowledge of future designs, and procurement, operation, and maintenance policies. The difference between targets and central estimates is also discussed. The paper concludes that improvements in future availability can be made by learning from past experience, but that certain conditions must be fulfilled in order to do so. (author)

BN-600 power unit 15-year operating experience

International Nuclear Information System (INIS)

Saraev, O.M.; Oshkanov, N.N.; Vylomov, V.V.

1996-01-01

Comprehensive experience has been gained with the operating fast reactor BN-600 with a power out of 600 MWe. This paper includes important performance results and gives also an overview of the experience gained from BN-600 NPP commercial operation during 15 years. (author). 2 figs, 1 tab

Recent operating experiences and programs at EBR-II

International Nuclear Information System (INIS)

Lentz, G.L.

1984-01-01

Experimental Breeder Reactor No. II (EBR-II) is a pool-type, unmoderated, sodium-cooled reactor with a design power of 62.5 MWt and an electrical generation capability of 20 MW. It has been operated by Argonne National Laboratory for the US government for almost 20 years. During that time, it has operated safely and has demonstrated stable operating characteristics, high availability, and excellent performance of its sodium components. The 20 years of operating experience of EBR-II is a valuable resource to the nuclear community for the development and design of future LMFBR's. Since past operating experience has been extensively reported, this report will focus on recent programs and events

Review on experience in operation of steels for use as fast reactor fuel element cladding and wrappers

International Nuclear Information System (INIS)

Weisz, Michel.

1978-01-01

The informations on the behavior of steels which can be gathered from routine destructive or non-destructive examination of fast reactor cans and wrapper tubes are presented. The relative merits of swelling measurements made on specimens or on real cans are compared. The diametral deformations of all the cans of a bundle and the immersion density measurements on industrial fabrication of wrapper tubes are discussed. The swelling temperature relationships and the double peak swelling of SA316 in relation with microstructural evolution are studied. Irradiation creep is also investigated, particularly the bulging of the wrapper tubes allowing to derive mean creep rate measurements for the high dose region [fr

Microstructural evolution of austenitic stainless steels irradiated to 17 dpa in spectrally tailored experiment of the ORR and HFIR at 400{degrees}C

Energy Technology Data Exchange (ETDEWEB)

Wakai, E.; Hashimoto, N.; Gibson, L.T. [Oak Ridge National Lab., TN (United States)] [and others

1997-08-01

The microstructural evolution of austenitic JPCA aged and solution annealed JPCA, 316R, C, K, and HP steels irradiated at 400{degrees}C in spectrally tailored experiments of the ORR and HFIR has been investigated. The helium generation rates were about 12-16 appm He/dpa on the average up to 17.3 dpa. The number densities and average diameters of dislocation loops in the steels have ranges of 3.3 x 10{sup 21} m{sup -3} and 15.2-26.3 nm, respectively, except for HP steel for which they are 1.1 x 10{sup 23} m{sup -3} and 8.0 nm. Precipitates are formed in all steels except for HP steel, and the number densities and average diameters have ranges of 5.2 x 10{sup 20} - 7.7 x 10{sup 21} m{sup -3} and 3.4- 19.3 nm, respectively. In the 216R, C, and K steels, the precipitates are also formed at grain boundaries, and the mean sizes of these are about 110, 50, and 50 nm, respectively. The number densities of cavities are about 1 x 10{sup 22} m{sup -3} in all the steels. The swelling is low in the steels which form the precipitates.

Operational experience with Dragon reactor experiment of relevance to commercial reactors

International Nuclear Information System (INIS)

Capp, P.D.; Simon, R.A.

1976-01-01

An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

International Nuclear Information System (INIS)

Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

1994-01-01

In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

Austenitic stainless steels and high strength copper alloys for fusion components

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Zinkle, S.J.; Alexander, D.J.; Stubbins, J.F.

1998-01-01

An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop A125), and a precipitation-hardened copper alloy (Cu-Cr-Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop A125 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface. (orig.)

Treatment of operational experience of nuclear power plants in WANO

International Nuclear Information System (INIS)

Ibanez, M.

2013-01-01

The article describes the activities associated to the Operating Experience Programme of the World Association of Nuclear Operators. The programme manages the event reports submitted by the nuclear power plants to the WANO database for the preparation by the Operating Experience Central Team of some documents like the significant Operating Experience Reports and Significant Event Reports that help the stations to avoid similar events. (Author)

Assessment of precipitates of isothermal aged austenitic stainless steel using measurement techniques of ultrasonic attenuation

International Nuclear Information System (INIS)

Kim, Hun Hee; Kim, Hak Joon; Song, Sung Jin; Lim, Byeong Soo; Kim, Kyung Cho

2014-01-01

AISI 316L stainless steel is widely used as a structural material of high temperature thermoelectric power plants, since austenitic stainless steel has excellent mechanical properties. However, creep damage is generated in these components, which are operated under a high temperature and high pressure environment. Several researches have been done on how microstructural changes of precipitates affect to the macroscopic mechanical properties. And they investigate the relation between ultrasonic parameters and metallurgical results. But, these studies are limited by experiment results only. In this paper, attenuations of ultrasonic with isothermal damaged AISI 316L stainless steel were measured. Also, simulation of ultrasonic attenuation with variation of area fraction and size of precipitates were performed. And, from the measured attenuations, metallographic data and simulation results, we investigate the relations between the ultrasonic attenuations and the material properties which is area fraction of precipitates for the isothermal damaged austenitic stainless steel specimens. And, we studied parametric study for investigation of the relation between ultrasonic parameters and metallurgical results of the isothermal damaged AISI 316L stainless steel specimens using numerical methods.

Austenitic stainless steel-to-ferritic steel transition joint welding for elevated temperature service

International Nuclear Information System (INIS)

King, J.F.; Goodwin, G.M.; Slaughter, G.M.

1978-01-01

Transition weld joints between ferritic steels and austenitic stainless steels are required for fossil-fired power plants and proposed nuclear plants. The experience with these dissimilar-metal transition joints has been generally satisfactory, but an increasing number of failures of these joints is occurring prematurely in service. These concerns with transition joint service history prompted a program to develop more reliable joints for application in proposed nuclear power plants

The Study for Recycling NORM - Contaminated Steel Scraps from Steel Industry

International Nuclear Information System (INIS)

Tsai, K. F.; Lee, Y. S.; Chao, H. E.

2003-01-01

Since 1994, most of the major steel industries in Taiwan have installed portal monitor to detect the abnormal radiation in metal scrap feed. As a result, the discovery of NORM (Naturally Occurring Radioactive Material) has increased in recent years. In order to save the natural resources and promote radiation protection, an experimental melting process for the NORM contaminated steel scraps was carried out by the Institute of Nuclear Energy Research (INER) Taiwan, ROC. The experimental melting process has a pretreatment step that includes a series of cutting and removal of scales, sludge, as well as combustible and volatile materials on/in the steel scraps. After pretreatment the surface of the steel scraps are relatively clean. Then the scraps are melted by a pilot-type induction furnace. This experiment finally produced seven ingots with a total weight of 2,849 kg and 96.8% recovery. All of the surface dose rates are of the background values. The activity concentrations of these ingots are also below the regulatory criteria. Thus, these NORM-bearing steel scraps are ready for recycling. This study has been granted by the regulatory authority

Examining the temperature behavior of stainless steel surfaces exposed to hydrogen plasmas in the Lithium Tokamak eXperiment (LTX)

Science.gov (United States)

Bedoya, Felipe; Allain, Jean Paul; Kaita, Robert; Lucia, Matthew; St-Onge, Denis; Ellis, Robert; Majeski, Richard

2014-10-01

The Materials Analysis Particle Probe (MAPP) is an in-situ diagnostic designed to characterize plasma-facing components (PFCs) in tokamak devices. MAPP is installed in LTX at Princeton Plasma Physics Laboratory. MAPP's capabilities include remotely operated XPS acquisition and temperature control of four samples. The recent addition of a focused ion beam allows XPS depth profiling analysis. Recent published results show an apparent correlation between hydrogen retention and temperature of Li coated stainless steel (SS) PFCs exposed to plasmas like those of LTX. According to XPS data, the retention of hydrogen by the coated surfaces decreases at above 180 °C. In the present study MAPP will be used to study the oxidation of Li coatings as a function of time and temperature of the walls when Li coatings are applied. Experiments in the ion-surface interaction experiment (IIAX) varying the hydrogen fluence on the SS samples will be also performed. Conclusions resulting from this study will be key to explain the PFC temperature-dependent variation of plasma performance observed in LTX. This work was supported by U.S. DOE Contracts DE-AC02-09CH11466, DE-AC52-07NA27344 and DE-SC0010717.

Operating experience with high beta superconducting RF cavities

International Nuclear Information System (INIS)

Dylla, H.F.; Doolittle, L.R.; Benesch, J.F.

1993-01-01

The number of installed and operational β=1 superconducting rf cavities has grown significantly over the last two years in accelerator laboratories in Europe, Japan and the U.S. The total installed acceleration capability as of mid-1993 is approximately 1 GeV at nominal gradients. Major installations at CERN, DESY, KEK and CEBAF have provided large increments to the installed base and valuable operational experience. A selection of test data and operational experience gathered to date is reviewed

Operating experience with high beta superconducting rf cavities

International Nuclear Information System (INIS)

Dylla, H.F.; Doolittle, L.R.; Benesch, J.F.

1993-06-01

The number of installed and operational β = 1 superconducting rf cavities has grown significantly over the last two years in accelerator laboratories in Europe, Japan and the US. The total installed acceleration capability as of mid-1993 is approximately 1 GeV at nominal gradients. Major installations at CERN, DESY, KEK and CEBAF have provided large increments to the installed base and valuable operational experience. A selection of test data and operational experience gathered to date is reviewed

Notch aspects of RSP steel microstructure

Directory of Open Access Journals (Sweden)

Michal Černý

2012-01-01

Full Text Available For a rather long time, basic research projects have been focused on examinations of mechanical properties for Rapid Solidification Powder (RSP steels. These state-of-art steels are commonly known as “powdered steels“. In fact, they combine distinctive attributes of conventional steel alloys with unusual resistance of construction material manufactured by so called “pseudo-powdered” metallurgy.Choice of suitable materials for experimental verification was carried out based on characteristic application of so called “modern steel”. First, groups of stainless and tool steel types (steel grades ČSN 17 and 19 were selected. These provided representative specimens for the actual comparison experiment. For stainless steel type, two steel types were chosen: hardenable X47Cr14 (ČSN 17 029 stainless steel and non-hardenable X2CrNiMo18-14-3 (ČSN 17 350 steel. They are suitable e.g. for surgical tools and replacements (respectively. For tooling materials, C80U (ČSN 19 152 carbon steel and American D2 highly-alloyed steel (ČSN “equivalent” being 19 572 steel were chosen for the project. Finally, the M390 Böhler steel was chosen as representative of powdered (atomized steels. The goal of this paper is to discuss structural aspects of modern stainless and tool steel types and to compare them against the steel made by the RSP method. Based on the paper's results, impact of powdered steel structural characteristics on the resistance to crack initiation shall be evaluated.

Entering 'A NEW REALM' of KIBO Payload Operations - Continuous efforts for microgravity experiment environment and lessons learned from real time experiment operations in KIBO -

International Nuclear Information System (INIS)

Sakagami, K; Goto, M; Matsumoto, S; Ohkuma, H

2011-01-01

On January 22nd, 2011(JST), KOUNOTORI2 (H-II Transfer Vehicle: HTV2) was successfully launched from Tanegashima Space Center toward the International Space Station (ISS) and two new JAXA payload racks, Kobairo rack and MSPR (Multi-purpose Small Payload Rack) were transferred to ISS/KIBO (Japanese Experiment Module: JEM). In addition to Saibo rack and Ryutai rack which are already in operation in KIBO, in total 4 Japanese experiment payload racks start operations in KIBO. Then KIBO payload operations embark on a new realm, full utilization phase. While the number and variety of microgravity experiments become increasing, simultaneous operation constraints should be considered to achieve multitask payload operations in ISS/KIBO and ever more complicated cooperative operations between crewmember and flight control team/science team are required. Especially for g-jitter improvement in ISS/KIBO, we have greatly advanced cooperative operations with crewmember in the recent increment based on the microgravity data analysis results. In this paper, newly operating Japanese experiment payloads characteristics and some methods to improve g-jitter environment are introduced from the front line of KIBO payload operations.

Practical experience with welding new generation steel PB2 assigned for power industry

Energy Technology Data Exchange (ETDEWEB)

Kwiecinski, Krzysztof; Lomozik, Miroslaw [Instytut Spawalnictwa, Gliwice (Poland); Urzynicok, Michal [Boiler Elements Factory ' ZELKOT' , Koszecin (Poland)

2010-07-01

This paper presents a new generation steel PB2 assigned for the power industry. In this article the authors present the results of non-destructive (VT, PT, RT) and destructive (tensile test, bending test, hardness measurements, impact strength, macro- and micrograph, fractography) tests. The major objective of the examinations was to verify properties of welded joints made of PB2 steel. Investigation of welded joints made of PB2 steel was performed in Instytut Spawalnictwa in Gliwice and it brings one of the first positive results for this type of steel in the world. (orig.)

USA/FBR program status FFTF operations startup experience

International Nuclear Information System (INIS)

Moffitt, W.C.; Izatt, R.D.

1981-06-01

This paper gives highlights of the major Operations evaluations and operational results of the startup acceptance testing program and initiation of normal operating cycles for experiment irradiation in the FFTF. 33 figures

Energy and materials flows in the fabrication of iron and steel semifinished products

Energy Technology Data Exchange (ETDEWEB)

Darby, J.B. Jr.; Arons, R.M.

1979-08-01

The flow of energy and materials in the fabrication of iron and steel semifinished products from molten metal is discussed. The focus is on techniques to reduce the amount of energy required to produce the typical products of integrated steel plants and iron and steel foundries. In integrated steel plants, if only 50% of the steel being cast were continuously cast, industry-wide energy consumption would be reduced by 6 to 15%. Further major energy savings could be achieved by increased use of by-product gases and regenerators in the various reheat operations. Finally, systems optimization studies to maintain the even flow of materials at full capacity should yield further improvements in energy efficiency. In foundry operations, alternate heating methods in forging operations and the use of no-bake molding and core materials should result in substantial energy savings. Studies of specific operations will suggest housekeeping changes to minimize wasted energy. These changes might include fixing heat leaks, reducing floor space requirements, improving temperature regulation, lowering working temperatures in some steel-forming operations, redesigning products, and minimizing scrap generation. There is also a need for new, energy conserving technologies. A good example would be the development of nondestructive testing to determine the existence, location, and size of defects in ingots at elevated temperatures. A second example is the need to reduce, through system studies, the large amount of scrap typical of foundry operations. Finally, computer control of steel mill operations (materials flow, furnace residence times, excessive heating or overheating, and full capacity utilization of all facilities at all times) deserves further study.

Decontamination of Stainless Steel SS 304 Type with Pressurized CO2 Solid

International Nuclear Information System (INIS)

Sutoto

2007-01-01

The abrasive decontamination of the stainless steel valve using 12 bar pressurized CO 2 solid has been done. Experiment activities was performed in the HOT CELL facility with variation of blasting time 15, 30, 45 and 60 seconds. The result of experiment shown that the operation of abrasive decontamination during 45 seconds gives the decreasing of the equipment radiation dose rate from 460 to 200 mRem/h and decontamination factor 1.35. The secondary waste from decontamination activities was treated by filtration method using HEPA filter and activated carbon filter. (author)

Effect of boron control of environment on corrosion and resistance to low-cycle corrosion fatigue in structural steels

International Nuclear Information System (INIS)

Babej, Yu.I.; Zhitkov, V.V.; Zvezdin, Yu.I.; Liskevich, I.Yu.; Nazarov, A.A.

1982-01-01

Tests of the specimens on total, contact and crevice corrosion, corrosion cracking and low-cycle fatigue are conducted for determination of corrosion and corrosion-fatigue characteristics in the 15Kh3NMFA, 10N3MFA, 10Kh16N4B, 05Kh13N6M2 structural steels, used in energetics. The environment is subjected to boron control and contacting with atmosphere for simulation of stop and operation modes of the facility. The experiments are carried out in the distilled water with 12g/l H 3 BO 3 and 10 mg/l Cl' at 25, 60, 100 deg C under contacting with atmosphere. It is established, that the pearlitic steels 15Kh3NMFA, 10N3MFA, as well as transition and martensitic 05Kh13N6M2 and 10Kh16N4B steels are highly stable to total, crevice and contact corrosion at the high parameters of aqueous boron-containing medium. Steel resistance to low-cycle fracture decreases slightly under the conditions similar to the operation ones, in the water with 12 g/l H 3 BO 3 . Durability of the pearlitic steels at the simulation of stop conditions decreases more noticeably, crack formation as a rule, initiating from corrosion spots

Behavior of stainless steels in pressurized water reactor primary circuits

International Nuclear Information System (INIS)

Féron, D.; Herms, E.; Tanguy, B.

2012-01-01

Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

The industrial ecology of steel

Energy Technology Data Exchange (ETDEWEB)

Considine, Timothy J.; Jablonowski, Christopher; Considine, Donita M.M.; Rao, Prasad G.

2001-03-26

This study performs an integrated assessment of new technology adoption in the steel industry. New coke, iron, and steel production technologies are discussed, and their economic and environmental characteristics are compared. Based upon detailed plant level data on cost and physical input-output relations by process, this study develops a simple mathematical optimization model of steel process choice. This model is then expanded to a life cycle context, accounting for environmental emissions generated during the production and transportation of energy and material inputs into steelmaking. This life-cycle optimization model provides a basis for evaluating the environmental impacts of existing and new iron and steel technologies. Five different plant configurations are examined, from conventional integrated steel production to completely scrap-based operations. Two cost criteria are used to evaluate technology choice: private and social cost, with the latter including the environmental damages associated with emissions. While scrap-based technologies clearly generate lower emissions in mass terms, their emissions of sulfur dioxide and nitrogen oxides are significantly higher. Using conventional damage cost estimates reported in the literature suggests that the social costs associated with scrap-based steel production are slightly higher than with integrated steel production. This suggests that adopting a life-cycle viewpoint can substantially affect environmental assessment of new technologies. Finally, this study also examines the impacts of carbon taxes on steel production costs and technology choice.

Plasticity induced by phase transformation in steel: experiment vs modeling

International Nuclear Information System (INIS)

Tahimi, Abdeladhim

2011-01-01

The objectives of this work are: (i) understand the mechanisms and phenomena involved in the plasticity of steels in the presence of a diffusive or martensitic phase transformation. (ii) develop tools for predicting TRIP, which are able to correctly reproduce the macroscopic deformation for cases of complex loading and could also provide information about local elasto-visco-plastic interactions between product and parent phases. To this purpose, new experimental tests are conducted on 35NCD16 steel for austenite to martensite transformation and on 100C6 steel for austenite to pearlite transformation. The elasto viscoplastic properties of austenite and pearlite of the 100C6 steel are characterized through tension compression and relaxation tests. The parameters of macro-homogeneous and crystal-based constitutive laws could then be identified such as to analyse different models with respect to the experimental TRIP: the analytical models of Leblond (1989) and Taleb and Sidoroff (2003) but also, above all, different numerical models which can be distinguished by the prevailing assumptions concerning the local kinetics and the constitutive laws. An extension of the single-grain model dedicated to martensitic transformations developed during the thesis of S. Meftah (2007) is proposed. It consists in introducing the polycrystalline character of the austenite through a process of homogenization based on a self-consistent scheme by calculating the properties of an Equivalent Homogeneous Medium environment (EHM). (author)

Experience on operational safety improvement of control and operation support systems

International Nuclear Information System (INIS)

Itoh, N.; Nakagawa, T.; Mano, K.

1988-01-01

Japanese nuclear industry started in 1956 and about 30 years have passed since that time. Through these years, we have made a lot of efforts and developments in the field of Control and Instrumentation (C and I) system. The above 30 years and following years can be divided into four major periods. The first one is the period of research, the second of domestic production, the third of improvement, and the fourth of advancement. Improvements of C and I system, which we have made in those periods have made a great contribution to enhancement of reliability, availability and operability of nuclear power plants. Fig. 1 shows TEPCO's nuclear power plant (BWR) construction experience and technical trend of C and I system in Japan. This paper is to introduce the efforts and operational experience on control and operation support systems

Effect of ausforming on nanobainite steel

International Nuclear Information System (INIS)

Gong, W.; Tomota, Y.; Koo, M.S.; Adachi, Y.

2010-01-01

The effect of ausforming on kinetics, morphology and crystallography of nanobainite steel was examined by electron backscattered diffraction and transmission electron microscopy. Ausforming has been found to accelerate bainite transformation at 573 K. A characteristic microstructure consisting of blocky bainitic laths and retained austenite is observed in the ausformed bainite steel, where strong variant selection takes place due to the operated slip systems.

Investigation of Mechanical Properties of Unidirectional Steel Fiber/Polyester Composites: Experiments and Micromechanical Predictions

DEFF Research Database (Denmark)

Raghavalu Thirumalai, Durai Prabhakaran; Løgstrup Andersen, Tom; Bech, Jakob Ilsted

2016-01-01

the role of material and process parameters on material properties. Two types of SFRP were studied: polyester resin reinforced by both steel fabric containing unidirectional fibers and steel fibers wound on a metal frame with 0° orientations. The effects of the fiber volume fraction and the role of polymer......The article introduces steel fiber reinforced polymer composites, which is considered new for composite product developments. These composites consist of steel fibers or filaments of 0.21 mm diameter embedded in a polyester resin. The goal of this investigation is to characterize the mechanical...... performance of steel fiber reinforced polyester composites at room temperature. The mechanical properties of unidirectional steel fiber reinforced polyester composites (SFRP) are evaluated experimentally and compared with the predicted values by micro-mechanical models. These predictions help to understand...

Thermal ageing of steels; from expertise and understanding of the ageing mechanisms to a maintenance strategy for operating nuclear power plants

International Nuclear Information System (INIS)

Bezdikian, G.; Ould, P.

2004-01-01

Some parts of reactor coolant circuit on Nuclear PWR power plants, elbows on primary circuit, are made in cast duplex stainless steel material. It is now identify that the mechanical characteristic of this material should be decrease under thermal ageing mainly after a long time in operation in at reactor coolant circuit temperature conditions. The sensitiveness to the thermal ageing of these components is in relation with chemical composition and the ferrite content, especially the grade of Chromium equivalent (Ceq %Cr + %Si + %Mo). In the context of justification to maintain in operation on the plants these cat duplex components, an important programme of expertises was carried out on cast elbows after removing on the plants during the Steam Generators replacements (SGR). Several expertises, performed in the objective to understand the thermal ageing phenomenon and mechanism on cast components in service on plants, were permit to validate the prediction formulas established from a large database and programme in laboratories. The expertises were based on a lot of metallurgical, mechanical and chemical characteristics of components in operation Small Angle Neutrons Scattering (SANS), Thermal Electric Power (TEP), micro hardness and toughness measurement on small specimens from boat sample (CT10-5) The expertise carried out on one SG inlet elbows from DAMPIERRE, removed a during SGR after 100000 h in operation is shown, the toughness values are very high compared to the prediction formulas. The TEP measurements performed on the specimen cut off on two elbows and the ingots of the same material aged in laboratory in furnace, are very coherent; it is confirmed that this methodology is a good indicator to follow the ageing characteristic of material. The results of expertises on aged material are a mean of validation of the methodology applied on the file of demonstration of maintaining in operation of cast duplex stainless steel sensitive to thermal ageing. So the

Development for the production of low phosphorus steel in operations at Arcelor Mittal Tubarao

Energy Technology Data Exchange (ETDEWEB)

Luiz-Correa, W.; Silva-Furtado, H.; Oliveira, J. R. de

2013-06-01

The growing demand to produce steel with lower phosphorus (P) content happens in tandem with the increase in the content of this element in the ores used in the production of pig iron, leading to a constant evolution in the process of steel dephosphorization in BOF converters. Arcelor Mittal Tubarao (AMT), located in the municipality of Serra, Brazil, currently produces 5 million t of steel, but has an installed capacity of 7.5 million. This work aims at showing the development carried out in the dephosphorization of silicon-aluminium killed steel produced in the Arcelor Mittal Tubarao converters. The analysis of process variables such as flux addition, oxygen lance position and temperature at the end of blow are based on classical phosphorus partition models. The results compare phosphorus values in liquid steel before and after modifications in the variables and the refractory wear caused by the new procedures applied to AMT steelmaking converters. (Author)

Review and updates of the risk assessment for advanced test reactor operations for operating events and experience

International Nuclear Information System (INIS)

Atkinson, S.A.

1996-01-01

Annual or biannual reviews of the operating history of the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) have been conducted for the purpose of reviewing and updating the ATR probabilistic safety assessment (PSA) for operating events and operating experience since the first compilation of plant- specific experience data for the ATR PSA which included data for operation from initial power operation in 1969 through 1988. This technical paper briefly discusses the means and some results of these periodic reviews of operating experience and their influence on the ATR PSA

Operational experience with SLAC's beam containment electronics

International Nuclear Information System (INIS)

Constant, T.N.; Crook, K.; Heggie, D.

1977-03-01

Considerable operating experience was accumulated at SLAC with an extensive electronic system for the containment of high power accelerated beams. Average beam power at SLAC can approach 900 kilowatts with the potential for burning through beam stoppers, protection collimators, and other power absorbers within a few seconds. Fast, reliable, and redundant electronic monitoring circuits have been employed to provide some of the safeguards necessary for minimizing the risk to personnel. The electronic systems are described, and the design philosophy and operating experience are discussed

CORROSION AND CHEMICAL WASTE IN SAWBLADES STEEL USED IN WOOD

Directory of Open Access Journals (Sweden)

Paulo Fernando Trugilho

2002-01-01

Full Text Available The objective this work was to evaluate the chemical waste provoked by the wood on the sheets of steel used in the making of the mountains and cut tools. It was certain the correlationbetween the chemical waste and the extractive soluble in cold water, hot water and in the sequencetoluene and ethanol content. Two types of steel and twenty-seven species different from wood wereused. The corrosive agent, constituted of 50 g of fresh sawdust (moist mixed to 50 ml of distilledwater, it was prepared and placed inside of the plastic box, hermetically closed, on the samples ofsteel, which were totally immersed. The box was placed in a water bath pre-heated to 75°C, that themedium temperature of reaction is considered, that affects the sheet of the sawblade in operation. Thisgroup was operated to 80 rotations per minute (rpm. The time of reaction was of four hours. Afterthat time the corrosive agent was discarded and the samples were washed, dried and weighed. At theend, each sample was processed by a total period of forty hours. The chemical waste was evaluated by the weight difference suffered from beginning at the end of the experiment. For theresults it was observed that the Eucalyptus tradryphloia and the Eucalyptus phaeotricha the speciesthat provoked were, respectively, the largest and smaller chemical waste for the two types of steelappraised. Great variation exists in the chemical waste due to the effect of the species. The corrosionand chemical waste are especially related with the quality of the material solved in ethanol. The 1070steel were more attached than the 6170 steel.

Processing and refinement of steel microstructure images for assisting in computerized heat treatment of plain carbon steel

Science.gov (United States)

Gupta, Shubhank; Panda, Aditi; Naskar, Ruchira; Mishra, Dinesh Kumar; Pal, Snehanshu

2017-11-01

Steels are alloys of iron and carbon, widely used in construction and other applications. The evolution of steel microstructure through various heat treatment processes is an important factor in controlling properties and performance of steel. Extensive experimentations have been performed to enhance the properties of steel by customizing heat treatment processes. However, experimental analyses are always associated with high resource requirements in terms of cost and time. As an alternative solution, we propose an image processing-based technique for refinement of raw plain carbon steel microstructure images, into a digital form, usable in experiments related to heat treatment processes of steel in diverse applications. The proposed work follows the conventional steps practiced by materials engineers in manual refinement of steel images; and it appropriately utilizes basic image processing techniques (including filtering, segmentation, opening, and clustering) to automate the whole process. The proposed refinement of steel microstructure images is aimed to enable computer-aided simulations of heat treatment of plain carbon steel, in a timely and cost-efficient manner; hence it is beneficial for the materials and metallurgy industry. Our experimental results prove the efficiency and effectiveness of the proposed technique.

Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

International Nuclear Information System (INIS)

Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

2007-01-01

Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to ∼ 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

Concluding from operating experience to instrumentation and control systems

International Nuclear Information System (INIS)

Pleger, H.; Heinsohn, H.

1997-01-01

Where conclusions are drawn from operating experience to instrumentation and control systems, two general statements should be made. First: There have been braekdowns, there have also been deficiencies, but in principle operating experience with the instrumentation and control systems of German nuclear power plants has been good. With respect to the debates about the use of modern digital instrumentation and control systems it is safe to say, secondly, that the instrumentation and control systems currently in use are working reliably. Hence, there is no need at present to replace existing systems for reasons of technical safety. However, that time will come. It is a good thing, therefore, that the use of modern digital instrumentation and control systems is to begin in the field of limiting devices. The operating experience which will thus be accumulated will benefit digital instrumentation and control systems in their qualification process for more demanding applications. This makes proper logging of operating experience an important function, even if it cannot be transferred in every respect. All parties involved therefore should see to it that this operating experience is collected in accordance with criteria agreed upon so as to prevent unwanted surprises later on. (orig.) [de

Tantalum-Addition Effect on Tensile and Creep Properties in 9Cr-0.5Mo-2W-V-Nb Steels

International Nuclear Information System (INIS)

Kim, Woo Gon; Kim, Sung Ho; Back, Jong Hyuk; Kim, Tae Kyu; Lee, Chan Bock

2011-01-01

Ferritic/martensitic steels (FMS) are being considered prospectively as cladding materials of a SFR fuel in Gen-IV nuclear systems. There are sound technical justifications for these material selections, and the adoption of the FM steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for a Gen-IV SFR. To develop an improved FM steel for the Gen-IV SFR fuel cladding in Korea, a R and D program has been progressed since 2007. Categories of materials considered in the program included 8∼12% Cr FM steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel (hereafter Gr. 92) to offset the difficulties encountered with commercial available high Cr (8∼12%) steels. Since the fuel cladding in a Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the developed cladding should thus sustain both superior irradiation and temperature stabilities during its operational life. The newly developed advanced steel should also overcome severe drawbacks: mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. The aim of this study is to investigate the effect of Ta addition on the tensile and creep properties of the three alloys which are designed, manufactured and tested. Their properties are obtained and compared for developing new FM fuel cladding materials

[Operating Room Nurses' Experiences of Securing for Patient Safety].

Science.gov (United States)

Park, Kwang Ok; Kim, Jong Kyung; Kim, Myoung Sook

2015-10-01

This study was done to evaluate the experience of securing patient safety in hospital operating rooms. Experiential data were collected from 15 operating room nurses through in-depth interviews. The main question was "Could you describe your experience with patient safety in the operating room?". Qualitative data from the field and transcribed notes were analyzed using Strauss and Corbin's grounded theory methodology. The core category of experience with patient safety in the operating room was 'trying to maintain principles of patient safety during high-risk surgical procedures'. The participants used two interactional strategies: 'attempt continuous improvement', 'immersion in operation with sharing issues of patient safety'. The results indicate that the important factors for ensuring the safety of patients in the operating room are manpower, education, and a system for patient safety. Successful and safe surgery requires communication, teamwork and recognition of the importance of patient safety by the surgical team.

Fire protection system operating experience review for fusion applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1995-12-01

This report presents a review of fire protection system operating experiences from particle accelerator, fusion experiment, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of fire protection system component failure rates and fire accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with these systems are discussed, including spurious operation. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor

Fire protection system operating experience review for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1995-12-01

This report presents a review of fire protection system operating experiences from particle accelerator, fusion experiment, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of fire protection system component failure rates and fire accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with these systems are discussed, including spurious operation. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor.

Improving Strength-Ductility Balance of High Strength Dual-Phase Steels by Addition of Vanadium

Science.gov (United States)

Gong, Yu; Hua, M.; Uusitalo, J.; DeArdo, A. J.

For galvanized or galvannealed steels to be commercially successful, they must exhibit several attributes: (i) easy and inexpensive processing in the hot mill, cold mill and on the coating line, (ii) high strength with good formability and spot weldability, and (iii) good corrosion resistance, especially after cold forming. For good corrosion resistance, the coating must have sufficient coverage, be of uniform thickness, and most importantly, the coating must survive the cold stamping or forming operation. The purpose of this paper is to present research aiming at improving the steel substrate, such that high strength can be obtained while maintaining good global formability (tensile ductility), local formability (sheared-edge ductility), and good spot weldability. It is well-known that the strength of DP steels is controlled by several factors, including the amount of martensite found in the final microstructure. Recent research has revealed that the amount of austenite formed during intercritical annealing can be strongly influenced by the annealing temperature and the pre-annealing conditions of the hot band (coiling temperature) and cold band (% cold reduction). Current experiments have explored the combination of pre-annealing conditions and four annealing practices to help define the best practice to optimize the strength-formability balance in these higher strength DP steels. The steels used in these experiments contained (i) low carbon content for good spot weldability, (ii) the hardenability additions Mo and Cr for strength, and (iii) V for grain refinement, precipitation hardening and temper resistance. When processed correctly, these steels exhibited UTS levels up to 1000MPa, total elongation to 25%, reduction in area to 45%, and Hole Expansion Ratios to 50%. The results of this program will be presented and discussed.

Feedback of operating experience in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

The feedback of operating experience of nuclear facilities to the designers, manufacturers, operators and regulators is one important means of maintaining and improving safety. The Atomic Energy Control Board`s Advisory Committee on Nuclear Safety examined the means for feedback currently being employed, how effective they are and what improvements are advisable. The review found that the need for feedback of operating experience is well recognized within those institutions contributing to the safety of CANDU power reactors, and that the existing procedures are generally effective. Some recommendations, however, are submitted for improvement in the process.

Feedback of operating experience in nuclear power plants

International Nuclear Information System (INIS)

1995-06-01

The feedback of operating experience of nuclear facilities to the designers, manufacturers, operators and regulators is one important means of maintaining and improving safety. The Atomic Energy Control Board's Advisory Committee on Nuclear Safety examined the means for feedback currently being employed, how effective they are and what improvements are advisable. The review found that the need for feedback of operating experience is well recognized within those institutions contributing to the safety of CANDU power reactors, and that the existing procedures are generally effective. Some recommendations, however, are submitted for improvement in the process

Measurement of the muon-induced neutron yield in liquid scintillator and stainless steel at LNGS with the LVD experiment

International Nuclear Information System (INIS)

Persiani, R.; Garbini, M.; Sartorelli, G.; Selvi, M.

2013-01-01

We describe the measurement of the muon-induced neutron yield in liquid scintillator and stainless steel (SS) at the Gran Sasso National Laboratory (LNGS), with the LVD experiment. The Large Volume Detector (LVD) is located in Hall A of the LNGS and is made of 1000 t of liquid scintillator and 1000 t of SS. Using an independent measurement to evaluate the background and with the support of a full Monte Carlo simulation based on Geant4, we measured a neutron yield of (2.9±0.6)×10 −4 and (1.5±0.3)×10 −3 in liquid scintillator and in stainless steel, respectively

TSTA piping and flame arrestor operating experience data

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee C., E-mail: Lee.Cadwallader@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Willms, R. Scott [ITER International Organization, Cadarache (France)

2015-10-15

Highlights: • Experiences from the Tritium Systems Test Assembly were examined. • Failure rates of copper piping and a flame arrestor were calculated. • The calculated failure rates compared well to similar data from the literature. • Tritium component failure rate data support fusion safety assessment. - Abstract: The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility was operated with tritium for its research and development program from 1984 to 2001, running a prototype fusion fuel processing loop with ∼100 g of tritium as well as small experiments. There have been several operating experience reports written on this facility's operation and maintenance experience. This paper describes reliability analysis of two additional components from TSTA, small diameter copper gas piping that handled tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.

Experiments on different materials (polyamide, stainless & galvanized steel) influencing geothermal CaCO3 scaling formation: Polymorphs & elemental incorporation

Science.gov (United States)

Wedenig, Michael; Dietzel, Martin; Boch, Ronny; Hippler, Dorothee

2016-04-01

Thermal water is increasingly used for heat and electric power production providing base-load capable renewable and virtually unlimited geothermal energy. Compared to other energy sources geothermal facilities are less harmful to the environment, i.e. chemically and visually. In order to promote the economic viability of these systems compared to other traditional and renewable energy sources, production hindering processes such as corrosion and scaling of components arising from the typically high salinity thermal waters have to be considered as important economic factors. In this context, using proper materials being in contact with the thermal water is crucial and a playground for further improvements. Aim of the study presented, are basic experiments and observations of scaling and corrosive effects from hydrothermal water interacting with different materials and surfaces (stainless steel, polyamide, galvanized steel) and in particular the nucleation and growth effects of these materials regarding the precipitation of solid carbonate phases. The incorporation of Mg, Sr and Ba cations into the carbonate scalings are investigated as environmental proxy. For this purpose, hydrothermal carbonate precipitating experiments were initialized by mixing NaHCO3 and Ca-Mg-Sr-Ba-chloride solutions at temperatures ranging from 40 to 80 °C in glass reactors hosting artificial substrates of the above mentioned materials. The experiments show a strong dependence of the precipitation behaviour of calcium carbonate polymorphs on the particular material being present. Stainless steel and polyamide seem to restrict aragonite formation, whereas galvanized steel supports aragonite nucleation. Vaterite formation is promoted by polyamide surfaces. Importantly, vaterite is more soluble (less stable) compared to the other anhydrous calcium carbonate polymorphs, i.e. vaterite can be more easily re-dissolved. Thus, the use of polyamide components might reduce the amount and durability of

Operator training and the training simulator experience

International Nuclear Information System (INIS)

Mills, D.

The author outlines the approach used by Ontario Hydro to train operators from the day they are hired as Operators-in-Training until they are Authorized Unit First Operators. He describes in detail the use of the simulator in the final year of the authorization program, drawing on experience with the Pickering NGS A simulator. Simulators, he concludes, are important aids to training but by no means all that is required to guarantee capable First Operators

Operating Experience Report: Counterfeit, Suspect and Fraudulent Items. Working Group on Operating Experience. Proceedings and Analysis on an Item of Generic Interest

International Nuclear Information System (INIS)

2011-01-01

The NEA Committee on Nuclear Regulatory Activities (CNRA) believes that sharing operating experience from the national operating experience feedback programmes are a major element in the industry's and regulatory body's efforts to ensure the continued safe operation of nuclear facilities. Considering the importance of these issues, the Committee on the Safety of Nuclear Installations (CSNI) established a working group, PWG no.1 (Principle Working Group Number 1) to assess operating experience in the late 1970's, which was later renamed the Working Group on Operating Experience (WGOE). In 1978, the CSNI approved the establishment of a system to collect international operating experience data. The accident at Three Mile Island shortly after added impetus to this and led to the start of the Incident Reporting System (IRS). In 1983, the IRS database was moved to the International Agency for Atomic Energy (IAEA) to be operated as a joint database by IAEA and NEA for the benefit of all of the member countries of both organisations. In 2006, the WGOE was moved to be under the umbrella of the Committee on Nuclear Regulatory Activities (CNRA) in NEA. In 2009, the scope of the Incident Reporting System was expanded and re-named the International Reporting System for Operating Experience (although, the acronym remains the same). The purpose of WGOE is to facilitate the exchange of information, experience, and lessons learnt related to operating experience between member countries. The working group continues its mission to identify trending and issues that should be addressed in specialty areas of CNRA and CSNI working groups. The CSFI (Counterfeit, Suspect, and Fraudulent Items) issue was determined to be the Issue of Generic Interest at the April 2010 WGOE meeting. The Issue of Generic Interest is determined by the working group members for an in-depth discussion. They are often emerging issues in operating experience that a country or several countries would to the share

Operational safety experience reporting in the United States

International Nuclear Information System (INIS)

Hartfield, R.A.

1978-01-01

Licensees of nuclear power plants in the United States have many reporting requirements included in their technical specifications and the code of federal regulations, title 10. The Nuclear Regulatory Commisson receives these reports and utilizes them in its regulatory program. Part of this usage includes collecting and publishing this operating experience data in various reports and storing information in various data systems. This paper will discuss the data systems and reports on operating experience published and used by the NRC. In addition, some observations on operating experience will be made. Subjects included will be the Licensee Event Report (LER) Data File, the Operating Unit Status Report (Gray Book), Radiation Exposure Reports, Effluents Reports, the Nuclear Plant Reliability Data System, Current Events, Bulletin Wrapups and Annual Summaries. Some of the uses of the reports and systems will be discussed. The Abnormal Occurence Report to the US Congress will also be described and discussed. (author)

Diffusion zinc plating of structural steels

International Nuclear Information System (INIS)

Kazakovskaya, Tatiana; Goncharov, Ivan; Tukmakov, Victor; Shapovalov, Vyacheslav

2004-01-01

The report deals with the research on diffusion zinc plating of structural steels when replacing their cyanide cadmium plating. The results of the experiments in the open air, in vacuum, in the inert atmosphere, under various temperatures (300 - 500 deg.C) for different steel brands are presented. It is shown that diffusion zinc plating in argon or nitrogen atmosphere ensures obtaining the qualitative anticorrosion coating with insignificant change of mechanical properties of steels. The process is simple, reliable, ecology pure and cost-effective. (authors)

Proceedings of 2nd PHWR operating safety experience meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-04-01

Papers presented during the eight sessions of the meeting were devoted to the impact of PHWR operating experience on design of civil structures (reactor building integrity); operating experiences related to pressure tubes, nuclear steam supply system, plant stability; reactor maintenance and control systems, reactor operational safety. Some events concerned with reactor shutdown due to power failures are described, as well as action undertaken to prevent major damage.

Proceedings of 2nd PHWR operating safety experience meeting

International Nuclear Information System (INIS)

1991-04-01

Papers presented during the eight sessions of the meeting were devoted to the impact of PHWR operating experience on design of civil structures (reactor building integrity); operating experiences related to pressure tubes, nuclear steam supply system, plant stability; reactor maintenance and control systems, reactor operational safety. Some events concerned with reactor shutdown due to power failures are described, as well as action undertaken to prevent major damage

Experience of high-nitrogenous steel powder application in repairs and surface hardening of responsible parts for power equipment by plasma spraying

Science.gov (United States)

Kolpakov, A. S.; Kardonina, N. I.

2016-02-01

The questions of the application of novel diffusion-alloying high-nitrogenous steel powders for repair and surface hardening of responsible parts of power equipment by plasma spraying are considered. The appropriateness of the method for operative repair of equipment and increasing its service life is justified. General data on the structure, properties, and manufacture of nitrogen-, aluminum-, and chromium-containing steel powders that are economically alloyed using diffusion are described. It is noted that the nitrogen release during the decomposition of iron nitrides, when heating, protects the powder particles from oxidation in the plasma jet. It is shown that the coating retains 50% of nitrogen that is contained in the powder. Plasma spraying modes for diffusion-alloying high-nitrogenous steel powders are given. The service properties of plasma coatings based on these powders are analyzed. It is shown that the high-nitrogenous steel powders to a nitrogen content of 8.9 wt % provide the necessary wear resistance and hardness of the coating and the strength of its adhesion to the substrate and corrosion resistance to typical aggressive media. It is noted that increasing the coating porosity promotes stress relaxation and increases its thickness being limited with respect to delamination conditions in comparison with dense coatings on retention of the low defectiveness of the interface and high adhesion to the substrate. The examples of the application of high-nitrogenous steel powders in power engineering during equipment repairs by service companies and overhaul subdivisions of heat power plants are given. It is noted that the plasma spraying of diffusion-alloyed high-nitrogenous steel powders is a unique opportunity to restore nitrided steel products.

Corrosion of steel tanks in liquid nuclear wastes

International Nuclear Information System (INIS)

Carranza, Ricardo M.; Giordano, Celia M.; Saenz, Eduardo

2005-01-01

The objective of this work is to understand how solution chemistry would impact on the corrosion of waste storage steel tanks at the Hanford Site. Future tank waste operations are expected to process wastes that are more dilute with respect to some current corrosion inhibiting waste constituents. Assessment of corrosion damage and of the influence of exposure time and electrolyte composition, using simulated (non-radioactive) wastes, of the double-shell tank wall carbon steel alloys is being conducted in a statistically designed long-term immersion experiment. Corrosion rates at different times of immersion were determined using both weight-loss determinations and electrochemical impedance spectroscopy measurements. Localized corrosion susceptibility was assessed using short-term cyclic potentiodynamic polarization curves. The results presented in this paper correspond to electrochemical and weight-loss measurements of the immersed coupons during the first year of immersion from a two year immersion plan. A good correlation was obtained between electrochemical measurements, weight-loss determinations and visual observations. Very low general corrosion rates ( -1 ) were estimated using EIS measurements, indicating that general corrosion rate of the steel in contact with liquid wastes would no be a cause of tank failure even for these out-of-chemistry limit wastes. (author) [es

NPD Operating Experience

Energy Technology Data Exchange (ETDEWEB)

Horton, E. P. [Hydro-Electric Power Commission of Ontario, Rolphton, ON (Canada)

1968-04-15

NPD has demonstrated high-capacity factor operation and for the past three years has achieved an average net capacity factor of 98% for the ''winter-peak'' period. The net capacity factor for the year 1966 was 88% and for the period from the end of commissioning (October 1962) to the end of 1966 was 71%. The output of the station has been stretched from 22 MW(e) gross to 25 MW(e) gross. This was aided by the installation of an internal steam separator in the turbine but no basic modifications to the reactor-boiler systems were required. The turbine has also been modified by the installation of chrome steel diaphragms as a solution to an erosion problem. The station also continues as a test facility to develop new components and techniques. This includes the recent successful replacement of two reactor pressure tubes and the conversion of the reactor vault ventilation system to a ''dry'' atmosphere using a molecular sieve to collect heavy-water leakage and control the concentration of acidic oxides of nitrogen. Fuel performance has been excellent and the average burn-up in the core is now 84 MWh/kg U which is slightly above the equilibrium design value. Only three fuel bundles have been found with sheath failures and none of these was due to a deficiency in the fuel but was as a result of handling problems with the refuelling equipment. In spite of undesirably high maintenance time, the fuelling machines have now inserted over 1000 fuel bundles into the reactor ''on power''. Heavy-water loss rates have been acceptable and are improving. The average loss rate from leaks during 1966 was 210 g/h. A proposal to modify the NPD heavy-water heat transport system to allow boiling is under consideration. (author)

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Operating experience with the DRAGON High Temperature Reactor experiment

International Nuclear Information System (INIS)

Simon, R.A.; Capp, P.D.

2002-01-01

The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

International Nuclear Information System (INIS)

Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

1994-11-01

In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

The Clearing House on Operating Experience Feedback (CH-OEF)

International Nuclear Information System (INIS)

Tanarro Colodron, J.

2016-01-01

Full text: The Clearing House on Operating Experience Feedback (CH-OEF) is an online information system that contains three technical databases available only to registered users: 1) Operating Experience Feedback (OEF) records, containing information about events occurred at Nuclear Power Plants; 2) Nuclear Power Plant (NPP) records, containing technical details about NPPs; 3) Documents about operating experience, such as the Topical Operating Experience Reports (TOERs) and the quarterly reports on nuclear power plant events. The main objective of the information system is to develop communication, cooperation and sharing of operating experience amongst the national nuclear regulatory authorities participating in EU Clearinghouse network. The CH-OEF is essential for the preparation and dissemination of the quarterly reports on NPP events. These reports are published every three months and are intended to be complementary to other international reporting systems, containing mainly recent information publicly available. Only events that are considered to be likely to have lessons applicable to EU NPPs or with a real or potential impact on nuclear safety are addressed in the reports. The CH-OEF is a fundamental tool for their preparation, providing specific features for a more efficient sharing of information as well as for facilitating the related discussion and decision making. (author

Operating Experience Review of Tritium-in-Water Monitors

Energy Technology Data Exchange (ETDEWEB)

S. A. Bruyere; L. C. Cadwallader

2011-09-01

Monitoring tritium facility and fusion experiment effluent streams is an environmental safety requirement. This paper presents data on the operating experience of a solid scintillant monitor for tritium in effluent water. Operating experiences were used to calculate an average monitor failure rate of 4E-05/hour for failure to function. Maintenance experiences were examined to find the active repair time for this type of monitor, which varied from 22 minutes for filter replacement to 11 days of downtime while waiting for spare parts to arrive on site. These data support planning for monitor use; the number of monitors needed, allocating technician time for maintenance, inventories of spare parts, and other issues.

The experience of five years operation of Phenix

International Nuclear Information System (INIS)

Conte, F.; Lacroix, A.

1980-01-01

Two long periods of exceptional operation have satisfied the hopes of the designers and all parameters, power, efficiency, load factor, fuel behaviour, were better than was expected. The experience resulting from the only major incident provided a series of complementary data. Modern technology has need of sanction by experiment. The Phenix type reactor is a tool which is convenient to operate and to maintain. The two aspects of the demonstration, correct operation and ease of maintenance, take a concrete form in the harmlessness of Phenix on men and on the environment. There is no irradiation and few releases. (orig./DG)

Development of nuclear grade stainless steels at KCSSL

International Nuclear Information System (INIS)

Balachandran, G.; Dhere, M.; Mahadik, A.; Hinge, N.M.; Balasubramanian, V.

2011-01-01

Kalyani Carpenter Special Steels Ltd is an alloy steel plant, where a variety of alloy steel grades are produced for automotive, defence, nuclear and aerospace applications. The plant has developed expertise in processing of several alloy steel grades of superior quality that meets stringent specifications. Primary steel is processed through a combination of electric arc furnace, ladle furnace and vacuum degassing where stringent control over dephosphorisation, desulphurization, deoxidation is effected to get a refined high quality steel. The molten steel is cast through continuous casting of slabs or ingot casting. In grades specific to nuclear application, the primary cast products are further subjected to electroslag remelting to achieve further freedom from inclusions and to achieve a favourable solidification grain structure, which ultimately improve the hot workability of the alloy steel. Appropriate choice of slag and operating parameters are needed for realising the required ingot quality. The present study would examine the processing and quality aspects of some important grades of steels used in nuclear industry namely ferritic 9Cr-1Mo steel, martensitic stainless steels 403, 410, precipitation hardenable 17-4 PH stainless steel and austenitic 321, 316LN stainless steel, which were made and supplied for applications to Indian nuclear industry. The expertise developed in processing the steels in terms of melting, heat treatment and their relationship to structural features and mechanical properties would be highlighted. (author)

Update of operations with Westinghouse steam generators

International Nuclear Information System (INIS)

Malinowski, D.D.; Fletcher, W.D.

1978-01-01

Westinghouse commercial steam generators in operation now number 112, of which 98 are tubed with Inconel 600, the remainder with stainless steel. The implementation of all volatile treatment (AVT) was reported. It was noted that several plants had exhibited some tube corrosion during their initial periods using AVT; this observation indicated that the transition from phosphate chemistry control to AVT may have been subject to certain residual effects due to incomplete removal of phosphated deposits. As inspection results from steam generators operated on AVT became more generally available with the passage of time, a pattern of results emerged that seemed to correlate with the operating experience with phosphate chemistry control. Specifically, all the plants that experienced corrosion problems had from 1 to 8 yr of operational history using phosphates, while those with less than a year's experience using phosphates tended to be less affected by corrosion problems

Nuclear power plant operation experience - a feedback programme

International Nuclear Information System (INIS)

Banica, I.; Sociu, F.; Margaritescu, C.

1994-01-01

An effective high quality maintenance programme is required for the safe reliable operation of a nuclear power plant. To achieve the objectives of such a programme, both plant management and staff must be highly dedicated and motivated to perform high quality work at all levels. Operating and maintenance experience data collections and analysis are necessary in order to enhance the safety of the plant and reliability of the structures systems and components throughout their operating life. Significant events, but also minor incident, may reveal important deficiencies or negative trends adverse to safety. Therefore, a computer processing system for collecting, classifying and evaluating abnormal events or findings concerning operating-maintenance and for feeding back the results of the lessons learned from experience into the design and the operation of our nuclear power plant is considered to be of paramount importance. (Author)

Study Of The Wet Multipass Drawing Process Applied On High Strength Thin Steel Wires

Science.gov (United States)

Thimont, J.; Felder, E.; Bobadilla, C.; Buessler, P.; Persem, N.; Vaubourg, JP.

2011-05-01

Many kinds of high strength thin steel wires are involved in so many applications. Most of the time, these wires are made of a pearlitic steel grade. The current developments mainly concern the wire last drawing operation: after a patenting treatment several reduction passes are performed on a slip-type multipass drawing machine. This paper focuses on modeling this multipass drawing process: a constitutive law based on the wire microstructure evolutions is created, a mechanical study is performed, a set of experiments which enables determining the process friction coefficients is suggested and finally the related analytical model is introduced. This model provides several general results about the process and can be used in order to set the drawing machines.

Corrosion processes of alloyed steels in salt solutions

Energy Technology Data Exchange (ETDEWEB)

Kienzler, Bernhard [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Institut fuer Nukleare Entsorgung

2018-02-15

A summary is given of the corrosion experiments with alloyed Cr-Ni steels in salt solutions performed at Research Centre Karlsruhe (today KIT), Institute for Nuclear Waste Disposal (INE) in the period between 1980 and 2004. Alloyed steels show significantly lower general corrosion in comparison to carbon steels. However, especially in salt brines the protective Cr oxide layers on the surfaces of these steels are disturbed and localized corrosion takes place. Data on general corrosion rates, and findings of pitting, crevice and stress corrosion cracking are presented.

Simple experiments studying the catastrophic corrosion of stainless steel caused by sodium leaks

International Nuclear Information System (INIS)

Mathison, J.; Trevalion, P.A.; Hamer, A.N.

1971-01-01

Under certain conditions, small quantities of soditum escaping through a small hole may cause extensive corrosion and cratering on the outside of the containment in the vicinity of the leak and there is a possibility that this might lead to a major rupture of the containment. It is difficult to estimate such a corrosion rate by conducting post mortems after an incident because of the lack of precise information about times and temperatures. Simple sodium burning experiments have therefore been carried out in an attempt to provide rough quantitative data on the size of these corrosion rates. These showed that the average rate of corrosion of specimens of 18.8.1 stainless steel beneath a burning pool of sodium was of the order of 0.05 cm/hr at 600 ° C and 0.005 cm/hr at 400 ° C. Enhanced corrosion occurs at the periphery of the burning sodium. The rate of penetration will depend on the shape of the corrosion profile which exists in the affected surface. The times needed to penetrate different wall thicknesses of stainless steel pipework have been calculated for various corrosion profiles similar to those which have been observed after incidents in REML. (author)

Telescience testbed: Operational support functions for biomedical experiments

Science.gov (United States)

Yamashita, Masamichi; Watanabe, Satoru; Shoji, Takatoshi; Clarke, Andrew H.; Suzuki, Hiroyuki; Yanagihara, Dai

A telescience testbed was conducted to study the methodology of space biomedicine with simulated constraints imposed on space experiments. An experimental subject selected for this testbedding was an elaborate surgery of animals and electrophysiological measurements conducted by an operator onboard. The standing potential in the ampulla of the pigeon's semicircular canal was measured during gravitational and caloric stimulation. A principal investigator, isolated from the operation site, participated in the experiment interactively by telecommunication links. Reliability analysis was applied to the whole layers of experimentation, including design of experimental objectives and operational procedures. Engineering and technological aspects of telescience are discussed in terms of reliability to assure quality of science. Feasibility of robotics was examined for supportive functions to reduce the workload of the onboard operator.

Water requirements of the iron and steel industry

Science.gov (United States)

Walling, Faulkner B.; Otts, Louis Ethelbert

1967-01-01

Twenty-nine steel plants surveyed during 1957 and 1958 withdrew from various sources about 1,400 billion gallons of water annually and produced 40.8 million tons of ingot steel. This is equivalent to about 34,000 gallons of water per ton of steel. Fifteen iron ore mines and fifteen ore concentration plants together withdrew annually about 89,000 million gallons to produce 15 million tons of iron ore concentrate, or 5,900 gallons per ton of concentrate. About 97 percent of the water used in the steel plants came from surface sources, 2.2 percent was reclaimed sewage, and 1.2 percent was ground water. Steel plants supplied about 96 percent of their own water requirements, although only three plants used self-supplied water exclusively. Water used by the iron ore mines and concentration plants was also predominantly self supplied from surface source. Water use in the iron and steel industry varied widely and depended on the availability of water, age and condition of plants and equipment, kinds of processes, and plant operating procedures. Gross water use in integrated steel plants ranged from 11,200 to 110,000 gallons per ton of steel ingots, and in steel processing plants it ranged from 4,180 to 26,700 gallons per ton. Water reuse also varied widely from 0 to 18 times in integrated steel plants and from 0 to 44 times in steel processing plants. Availability of water seemed to be the principal factor in determining the rate of reuse. Of the units within steel plants, a typical (median) blast furnace required 20,500 gallons of water per ton of pig iron. At the 1956-60 average rate of pig iron consumption, this amounts to about 13,000 gallons per ton of steel ingots or about 40 percent of that required by a typical integrated steel plant 33,200 gallons per ton. Different processes of iron ore concentration are devised specifically for the various kinds of ore. These processes result in a wide range of water use from 124 to 11,300 gallons of water per ton of iron ore

LOFT instrumented fuel design and operating experience

International Nuclear Information System (INIS)

Russell, M.L.

1979-01-01

A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

Nuclear power plant operating experience, 1976

International Nuclear Information System (INIS)

1977-11-01

This report is the third in a series of reports issued annually that summarize the operating experience of U.S. nuclear power plants in commercial operation. Power generation statistics, plant outages, reportable occurrences, fuel element performance, occupational radiation exposure and radioactive effluents for each plant are presented. Summary highlights of these areas are discussed. The report includes 1976 data from 55 plants--23 boiling water reactor plants and 32 pressurized water reactor plants

Cryogenic system operating experience review for fusion applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1992-01-01

This report presents a review of cryogenic system operating experiences, from particle accelerator, fusion experiment, space research, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of cryogenic component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with cryogenic systems are discussed, including ozone formation, effects of spills, and modeling spill behavior. This information should be useful to fusion system designers and safety analysts, such as the team working on the International Thermonuclear Experimental Reactor design

Compression test of cold-formedsteel perforated profile with steel sheathing

Directory of Open Access Journals (Sweden)

Shamanin Aleksandr Yur’evich

2015-05-01

Full Text Available The subject of this paper is the stability and strength of cold-formed and perforated steel sigma-section columns with steel sheathing of different thickness. Ceilings with and without steel sheathing of different thickness are tested to failure in compression on a laboratory machine, which was based on a manual hydraulic jack. Series of 4 experiments with full-scale walls (2.5 m height were carried out. Also, for examination of the role of boundary conditions, the sheet in a ceiling is either left free or connected to base with screws.In civil engineering there are many experiments and methodologies for calculating the strength and buckling of ceiling with the sheathing of various materials, such as oriented strand board and gypsum board. However, for producing superstructures of ships the materials with high plastic properties and strength characteristics are required. For example steel possesses such properties. It was the main reason for conducting a series of experiments and studying the behavior of cold-formed steel columns with steel sheathing. During the experiments the deformation of the cross-section of three equally spaced cross sections was determined, as well as the axial deformation of the central column in the ceiling with steel sheathing.The test results showed the influence of the thickness of sheathing and boundary condition of a sheet on the strength and buckling of ceiling. According to the results of the tests it is necessary to evaluate the impact of the sheathing made of different materials and if necessary to carry out further tests.

Microstructure and elevated temperature stability of 9-12% Cr steels

Energy Technology Data Exchange (ETDEWEB)

Dogan, Omer N.; Hawk, Jeffrey A.

2005-02-01

Medium Cr steels have been used in fossil fired power plants for many years because of their excellent high temperature stability and mechanical properties. As the desire to increase the efficiency of power plants continues, the operating temperature (>650C) continues to go up. Currently available low and medium Cr containing steels will not withstand the new operating temperature and must be reassessed in terms of their solid-solution and precipitation strengthening schemes. Three medium Cr steels were developed to investigate high temperature alloy strengthening strategies: 0.08C-(9-12)Cr-1.2Ni-0.7Mo-3.0Cu-3.0Co-0.5Ti. The microstructure of the alloy will be described in the as-cast and thermo-mechanically worked states. In addition, the effect on microstructure from long-term high temperature exposure will also be discussed. Finally, the overall stability of these steels will be compared against currently available power plant steels.

Efeito do tempo de experiência de operadores de Harvester no rendimento operacional Effect of time experience of Harvester operators in operating yield

Directory of Open Access Journals (Sweden)

Elaine Cristina Leonello

2012-12-01

Full Text Available A mecanização da colheita de madeira permite maior controle dos custos e pode proporcionar reduções em prazos relativamente curtos. Além disso, tem um lugar de destaque na humanização do trabalho florestal e no aumento do rendimento operacional. O presente trabalho teve por objetivo avaliar o desempenho de operadores de harvester em função do tempo de experiência na atividade. Foram avaliados oito operadores do sexo masculino, com idade entre 23 e 46 anos. O estudo consistiu na análise do volume de madeira colhida pelo harvester. O tempo de experiência afeta significativamente o rendimento operacional dos operadores de harvester. Tal rendimento aumenta expressivamente nos primeiros 18 meses de experiência, mantendo-se em ascensão nos próximos 26 meses. Após os 44 meses de experiência, o rendimento dos operadores tende a reduzir, revelando as possíveis acomodações do cotidiano. Tais resultados permitem concluir que por volta dos 50 meses de experiência na atividade de operação de harvester, se faz necessária a adoção de medidas de reciclagem, motivação, entre outras, a fim de proporcionar aos operadores melhores condições de trabalho que os possibilitem continuar exercendo a atividade de forma eficiente e rentável à empresa.The mechanization of timber harvesting allows greater control of costs and can provide reductions in relatively short intervals. Moreover, it has a place in the humanization of the working forest and the increase in performance. This work provides comparisons of operating performance of different operator harvester according to the time of experience in the activity. The operators evaluated were eight males, aged between 23 and 46 years old. The study consisted of analysis of the volume of timber harvested by the harvester. The experience significantly affects the performance of harvesters operators. The performance increases significantly in the first 18 months of experience, and it remained on

Strain rate dependent tensile behavior of advanced high strength steels: Experiment and constitutive modeling

International Nuclear Information System (INIS)

Kim, Ji-Hoon; Kim, Daeyong; Han, Heung Nam; Barlat, F.; Lee, Myoung-Gyu

2013-01-01

High strain rate tensile tests were conducted for three advanced high strength steels: DP780, DP980 and TRIP780. A high strain rate tensile test machine was used for applying the strain rate ranging from 0.1/s to 500/s. Details of the measured stress–strain responses were comparatively analyzed for the DP780 and TRIP780 steels which show similar microstructural feature and ultimate tensile strength, but different strengthening mechanisms. The experimental observations included: usual strain rate dependent plastic flow stress behavior in terms of the yield stress (YS), the ultimate tensile strength (UTS), the uniform elongation (UE) and the total elongation (TE) which were observed for the three materials. But, higher strain hardening rate at early plastic strain under quasi-static condition than that of some increased strain rates was featured for TRIP780 steel, which might result from more active transformation during deformation with lower velocity. The uniform elongation that explains the onset of instability and the total elongation were larger in case of TRIP steel than the DP steel for the whole strain rate range, but interestingly the fracture strain measured by the reduction of area (RA) method showed that the TRIP steel has lower values than DP steel. The fractographs using scanning electron microscopy (SEM) at the fractured surfaces were analyzed to relate measured fracture strain and the microstructural difference of the two materials during the process of fracture under various strain rates. Finally, constitutive modeling for the plastic flow stresses under various strain rates was provided in this study. The proposed constitutive law could represent both Hollomon-like and Voce-like hardening laws and the ratio between the two hardening types was efficiently controlled as a function of strain rate. The new strength model was validated successfully under various strain rates for several grades of steels such as mild steels, DP780, TRIP780, DP980 steels.

Stack Monitor Operating Experience Review

International Nuclear Information System (INIS)

Cadwallader, L.C.; Bruyere, S.A.

2009-01-01

Stack monitors are used to sense radioactive particulates and gases in effluent air being vented from rooms of nuclear facilities. These monitors record the levels and types of effluents to the environment. This paper presents the results of a stack monitor operating experience review of the U.S. Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) database records from the past 18 years. Regulations regarding these monitors are briefly described. Operating experiences reported by the U.S. DOE and in engineering literature sources were reviewed to determine the strengths and weaknesses of these monitors. Electrical faults, radiation instrumentation faults, and human errors are the three leading causes of failures. A representative 'all modes' failure rate is 1E-04/hr. Repair time estimates vary from an average repair time of 17.5 hours (with spare parts on hand) to 160 hours (without spare parts on hand). These data should support the use of stack monitors in any nuclear facility, including the National Ignition Facility and the international ITER project.

Influence of Fretting on Flexural Fatigue of 304 Stainless Steel and Mild Steel

National Research Council Canada - National Science Library

Bill, Robert

1978-01-01

Fretting fatigue experiments conducted on 304 stainless steel using a flexural-fatigue test arrangement with bolted-on fretting pads have demonstrated that fatigue life is reduced by at least a factor...

14th Biennial conference on reactor operating experience plant operations: The human element

International Nuclear Information System (INIS)

Anon.

1989-01-01

Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

Service experience with AISI type 316 steel components in CEGB Midlands Region power plant

International Nuclear Information System (INIS)

Plastow, B.; Bagnall, B.I.; Yeldham, D.E.

1979-01-01

The service performance of AISI Type 316 steel components in sections up to 100 mm thick in Power Plant of the Midlands Region of the C.E.G.B. is reviewed. A comparison is drawn between the satisfactory performance of components whose dimensional stability is not critical and the difficulties experienced when rapid rates of change of temperature cause distortion in thick section components. Weldment manufacture and performance are reviewed and both are considered to be satisfactory. In general the material has performed well and the difficulties due to distortion have been overcome by imposing operating regimes which limit rates of temperature change. (author)

Steel containment buckling

International Nuclear Information System (INIS)

Bennett, J.G.; Fly, G.W.; Baker, W.E.

1984-01-01

The Steel Containment Buckling program is in its fourth phase of work directed at the evaluation of the effects of the structural failure mode of steel containments when the membrane stresses are compressive. The structural failure mode for this state of stress is instability or buckling. The program to date has investigated: (1) the effect on overall buckling capacity of the ASME area replacement method for reinforcing around circular penetrations; (2) a set of benchmark experiments on ring-stiffened shells having reinforced and framed penetrations; (3) large and small scale experiments on knuckle region buckling from internal pressure and post-buckling behavior to failure for vessel heads having torispherical geometries; and (4) buckling under time-dependent loadings (dynamic buckling). The first two investigations are complete, the knuckle buckling experimental efforts are complete with data analysis and reporting in progress, and the dynamic buckling experimental and analytical work is in progress

Steel containment buckling

International Nuclear Information System (INIS)

Butler, T.A.; Baker, W.E.

1986-01-01

Two aspects of buckling of a free-standing nuclear steel containment building were investigated in a combined experimental and analytical program. In the first part of the study, the response of a scale model of a containment building to dynamic base excitation is investigated. A simple harmonic signal was used for preliminary studies followed by experiments with scaled earthquake signals as the excitation source. The experiments and accompanying analyses indicate that the scale model response to earthquake-type excitations is very complex and that current analytical methods may require a dynamic capacity reduction factor to be incorporated. The second part of the study quantified the effects of framing at large penetrations on the static buckling capacity of scale model containments. Results show little effect from the framing for the scale models constructed from the polycarbonate, Lexan. However, additional studies with a model constructed of the prototypic steel material are suggested

US nuclear power plant operating cost and experience summaries

International Nuclear Information System (INIS)

Kohn, W.E.; Reid, R.L.; White, V.S.

1998-02-01

NUREG/CR-6577, U.S. Nuclear Power Plant Operating Cost and Experience Summaries, has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants. Cost incurred after initial construction are characterized as annual production costs, representing fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operating summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from annual operating reports submitted by the licensees, plant histories contained in Nuclear Power Experience, trade press articles, and the Nuclear Regulatory Commission (NRC) web site (www.nrc.gov)

US nuclear power plant operating cost and experience summaries

Energy Technology Data Exchange (ETDEWEB)

Kohn, W.E.; Reid, R.L.; White, V.S.

1998-02-01

NUREG/CR-6577, U.S. Nuclear Power Plant Operating Cost and Experience Summaries, has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants. Cost incurred after initial construction are characterized as annual production costs, representing fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operating summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from annual operating reports submitted by the licensees, plant histories contained in Nuclear Power Experience, trade press articles, and the Nuclear Regulatory Commission (NRC) web site (www.nrc.gov).

Simulation of Friction Stir Processing in 304L Stainless Steel

Directory of Open Access Journals (Sweden)

Miles M.P.

2016-01-01

Full Text Available A major dilemma facing the nuclear industry is repair or replacement of stainless steel reactor components that have been exposed to neutron irradiation. When conventional fusion welding is used for weld repair, the high temperatures and thermal stresses inherent in the process enhance the growth of helium bubbles, causing intergranular cracking in the heat-affected zone (HAZ. Friction stir processing (FSP has potential as a weld repair technique for irradiated stainless steel, because it operates at much lower temperatures than fusion welding, and is therefore less likely to cause cracking in the HAZ. Numerical simulation of the FSP process in 304L stainless steel was performed using an Eulerian finite element approach. Model input required flow stresses for the large range of strain rates and temperatures inherent in the FSP process. Temperature predictions in three locations adjacent to the stir zone were accurate to within 4% of experimentally measure values. Prediction of recrystallized grain size at a location about 6mm behind the tool center was less accurate, because the empirical model employed for the prediction did not account for grain growth that occurred after deformation in the experiment was halted.

European Clearinghouse for Nuclear Power Plants Operational Experience Feedback

International Nuclear Information System (INIS)

Martin Ramos, M.; Noel, M.

2010-01-01

In the European Union, in order to support the Community activities on operational experience, a centralized regional network on nuclear power plants operational experience feedback (European Clearinghouse on Operational Experience Feedback for Nuclear Power Plants) was established in 2008 at the EC JRC-IE, Petten (The Netherlands) on request of nuclear Safety Authorities of several Member States. Its main goal is to improve the communication and information sharing on OEF, to promote regional collaboration on analyses of operational experience and dissemination of the lessons learned. The enlarged EU Clearinghouse was launched in April 2010, and it is currently gathering the Regulatory Authorities of Finland, Hungary, Lithuania, the Netherlands, Romania, Slovenia, Switzerland, Bulgaria, Czec Republic, France, Germany, Slovak Republic, and Spain (these last six countries as observers). The OECD Nuclear Energy Agency, the IAEA, the EC Directorates General of the JRC and ENER are also part of the network. Recently, collaboration between some European Technical Support Organizations (such IRSN and GRS) and the EU Clearinghouse has been initiated. This paper explains in detail the objectives and organization of the EU Clearinghouse, as well as the most relevant activities carried out, like research work in trend analysis of events ocurred in NPP, topical reports on particular events, dissemination of the results, quarterly reports on events reported publicly and operational experience support to the members of the EU Clearinghouse. (Author)

Fort Saint Vrain operational experience

International Nuclear Information System (INIS)

Fuller, C.H.

1989-01-01

Fort St. Vrain (FSV), on the system of the Public Service Company of Colorado, is the only high temperature gas-cooled (HTGR) power reactor in the United States. The plant features a helium-cooled reactor with a uranium-thorium fuel cycle. The paper describes the experience made during its operation. (author). 2 refs, 4 figs, 2 tabs

Comparing PRAs with operating experience

International Nuclear Information System (INIS)

Picard, R.R.; Martz, H.F.

1998-01-01

Probabilistic Risk Assessment is widely used to estimate the frequencies of rare events, such as nuclear power plant accidents. An obvious question concerns the extent to which PRAs conform to operating experience--that is, do PRAs agree with reality? The authors discuss a formal methodology to address this issue and examine its performance using plant-specific data

The effect of cyclic and dynamic loads on carbon steel pipe

International Nuclear Information System (INIS)

Rudland, D.L.; Scott, P.M.; Wilkowski, G.M.

1996-02-01

This report presents the results of four 152-mm (6-inch) diameter, unpressurized, circumferential through-wall-cracked, dynamic pipe experiments fabricated from STS410 carbon steel pipe manufactured in Japan. For three of these experiments, the through-wall crack was in the base metal. The displacement histories applied to these experiments were a quasi-static monotonic, dynamic monotonic, and dynamic, cyclic (R = -1) history. The through-wall crack for the third experiment was in a tungsten-inert-gas weld, fabricated in Japan, joining two lengths of STS410 pipe. The displacement history for this experiment was the same history applied to the dynamic, cyclic base metal experiment. The test temperature for each experiment was 300 C (572 F). The objective of these experiments was to compare a Japanese carbon steel pipe material with US pipe material, to ascertain whether this Japanese steel was as sensitive to dynamic and cyclic effects as US carbon steel pipe. In support of these pipe experiments, quasi-static and dynamic, tensile and fracture toughness tests were conducted. An analysis effort was performed that involved comparing experimental crack initiation and maximum moments with predictions based on available fracture prediction models, and calculating J-R curves for the pipe experiments using the η-factor method

Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Fire-induced collapses of steel structures

DEFF Research Database (Denmark)

Dondera, Alexandru; Giuliani, Luisa

Single-story steel buildings such as car parks and industrial halls are often characterised by stiff beams and flexible columns and may experience an outward (sway) collapse during a fire, endangering people and properties outside the building. It is therefore a current interest of the research...... to investigate the collapse behaviour of single-story steel frames and identify relevant structural characteristics that influence the collapse mode. In this paper, a parametric study on the collapse a steel beam-column assembly with beam hinged connection and fixed column support is carried out under...... on the beam. By means of those tables, a simple method for the assessment and the countermeasure of unsafe collapse mode of single-story steel buildings can be derived....

Grain Refinement of Low Carbon Martensitic Steel by Heat Treatment

Directory of Open Access Journals (Sweden)

N. V. Kolebina

2015-01-01

Full Text Available The low-carbon steels have good corrosion and technological properties. Hot deformation is the main operation in manufacturing the parts from these steels. So one of the important properties of the material is a property of plasticity. The grain size significantly influences on the ductility properties of steel. The grain size of steel depends on the chemical composition of the crystallization process, heat treatment, and steel machining. There are plenty methods to have grain refinement. However, taking into account the large size of the blanks for the hydro turbine parts, the thermal cycling is an advanced method of the grain refinement adaptable to streamlined production. This work experimentally studies the heat treatment influence on the microstructure of the low-carbon 01X13N04 alloy steel and proposes the optimal regime of the heat treatment to provide a significantly reduced grain size. L.M. Kleiner, N.P. Melnikov and I.N. Bogachyova’s works focused both on the microstructure of these steels and on the influence of its parameters on the mechanical properties. The paper focuses mainly on defining an optimal regime of the heat treatment for grain refinement. The phase composition of steel and temperature of phase transformation were defined by the theoretical analysis. The dilatometric experiment was done to determine the precise temperature of the phase transformations. The analysis and comparison of the experimental data with theoretical data and earlier studies have shown that the initial sample has residual stress and chemical heterogeneity. The influence of the heat treatment on the grain size was studied in detail. It is found that at temperatures above 950 ° C there is a high grain growth. It is determined that the optimal number of cycles is two. The postincreasing number of cycles does not cause further reducing grain size because of the accumulative recrystallization process. Based on the results obtained, the thermal cycling

Industrial investigations of the liquid steel filtration

Directory of Open Access Journals (Sweden)

K. Janiszewski

2014-07-01

Full Text Available Hitherto existing investigations concerning the ceramic filter use in the steel making processes have given good results. The obtained results of filtration have proved that this method may be used as an effective and cheap way of steel filtration from non-metallic inclusions. Placing filters in the tundish is the best location considering the limitation of the possibility of secondary pollution of steel. Yet, the results presented in this paper, of an experiment prepared and carried out in the industrial environment, are the only positive results obtained, which are connected with so much quantities of liquid steel processed with use of the multi-hole ceramic filters.

Fatigue behaviour of friction welded medium carbon steel and austenitic stainless steel dissimilar joints

International Nuclear Information System (INIS)

Paventhan, R.; Lakshminarayanan, P.R.; Balasubramanian, V.

2011-01-01

Research highlights: → Fusion welding of dissimilar metals is a problem due to difference in properties. → Solid state welding process such as friction welding is a solution for the above problem. → Fatigue life of friction welded carbon steel and stainless steel joints are evaluated. → Effect of notch on the fatigue life of friction welded dissimilar joints is reported. → Formation of intermetallic is responsible for reduction in fatigue life of dissimilar joints. -- Abstract: This paper reports the fatigue behaviour of friction welded medium carbon steel-austenitic stainless steel (MCS-ASS) dissimilar joints. Commercial grade medium carbon steel rods of 12 mm diameter and AISI 304 grade austenitic stainless steel rods of 12 mm diameter were used to fabricate the joints. A constant speed, continuous drive friction welding machine was used to fabricate the joints. Fatigue life of the joints was evaluated conducting the experiments using rotary bending fatigue testing machine (R = -1). Applied stress vs. number of cycles to failure (S-N) curve was plotted for unnotched and notched specimens. Basquin constants, fatigue strength, fatigue notch factor and notch sensitivity factor were evaluated for the dissimilar joints. Fatigue strength of the joints is correlated with microstructure, microhardness and tensile properties of the joints.

Ventilation Systems Operating Experience Review for Fusion Applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1999-01-01

This report is a collection and review of system operation and failure experiences for air ventilation systems in nuclear facilities. These experiences are applicable for magnetic and inertial fusion facilities since air ventilation systems are support systems that can be considered generic to nuclear facilities. The report contains descriptions of ventilation system components, operating experiences with these systems, component failure rates, and component repair times. Since ventilation systems have a role in mitigating accident releases in nuclear facilities, these data are useful in safety analysis and risk assessment of public safety. An effort has also been given to identifying any safety issues with personnel operating or maintaining ventilation systems. Finally, the recommended failure data were compared to an independent data set to determine the accuracy of individual values. This comparison is useful for the International Energy Agency task on fusion component failure rate data collection

ABOUT RATIONING MAXIMUM ALLOWABLE DEFECT DEPTH ON THE SURFACE OF STEEL BILLETS IN PRODUCTION OF HOT-ROLLED STEEL

Directory of Open Access Journals (Sweden)

PARUSOV E. V.

2017-01-01

Full Text Available Formulation of the problem. Significant influence on the quality of rolled steel have various defects on its surface, which in its turn inherited from surface defects of billet and possible damage to the surface of rolled steel in the rolling mill line. One of the criteria for assessing the quality indicators of rolled steel is rationing of surface defects [1; 2; 3; 6; 7]. Current status of the issue. Analyzing the different requirements of regulations to the surface quality of the rolled high-carbon steels, we can conclude that the maximum allowable depth of defects on the surface of billet should be in the range of 2.0...5.0 mm (depending on the section of the billet, method of its production and further the destination Purpose. Develop a methodology for calculating the maximum allowable depth of defects on the steel billet surface depending on the requirements placed on the surface quality of hot-rolled steel. Results. A simplified method of calculation, allowing at the rated depth of defects on the surface of the hot-rolled steel to make operatively calculation of the maximum allowable depth of surface defects of steel billets before heating the metal in the heat deformation was developed. The findings shows that the maximum allowable depth of surface defects is reduced with increasing diameter rolled steel, reducing the initial section steel billet and degrees of oxidation of the metal in the heating furnace.

Metallurgy of steels for PWR pressure vessels

International Nuclear Information System (INIS)

Kepka, M.; Mocek, J.; Barackova, L.

1980-01-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

Metallurgy of steels for PWR pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

1980-09-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

77 FR 24200 - Clean Air Act Operating Permit Program; Petitions for Objection to State Operating Permits for...

Science.gov (United States)

2012-04-23

... Objection to State Operating Permits for Consolidated Environmental Management, Inc.--Nucor Steel Louisiana... Management, Inc.--Nucor Steel Louisiana (``Nucor'') in Saint James Parish, Louisiana. Pursuant to sections... Environmental Management, Inc.--Nucor Steel Louisiana (``Nucor'') is available electronically at: http://www.epa...

Microbial Corrosion and Cracking in Steel

DEFF Research Database (Denmark)

Hilbert, Lisbeth Rischel

1998-01-01

The aim of the report is to give a fundamental understanding of the response of different electrochemical techniques on carbon steel in a sulphide environment as well as in a biologically active sulphate-reducing environment (SRB). This will form the basis for further studies and for recommendati......The aim of the report is to give a fundamental understanding of the response of different electrochemical techniques on carbon steel in a sulphide environment as well as in a biologically active sulphate-reducing environment (SRB). This will form the basis for further studies...... will be based on results from the entire 3 year period, but only selected experimental data primarily from the latest experiments will be presented in detail here.Microbial corrosion of carbon steel under influence of sulphate-reducing bacteria (SRB) is characterised by the formation of both biofilm...... and corrosion products (ferrous sulphides) on the metal surface. Experiments have been conducted on carbon steel exposed in near neutral (pH 6 to 8.5) saline hydrogen sulphide environment (0 to 100 mg/l total dissolved sulphide) for a period of 14 days. Furthermore coupons have been exposed in a bioreactor...

austenitic steel corrosion by oxygen-containing liquid sodium

International Nuclear Information System (INIS)

Rivollier, Matthieu

2017-01-01

France is planning to construct the 4. generation of nuclear reactors. They will use liquid sodium as heat transfer fluid and will be made of 316L(N) austenitic steel as structural materials. To guarantee optimal operation on the long term, the behavior of this steel must be verified. This is why corrosion phenomena of 316L(N) steel by liquid sodium have to be well-understood. Literature points out that several corrosion phenomena are possible. Dissolved oxygen in sodium definitely influences each of the corrosion phenomenon. Therefore, the austenitic steel corrosion in oxygen-containing sodium is proposed in this study. Thermodynamics data point out that sodium chromite formation on 316L(N) steel is possible in sodium containing roughly 10 μg.g -1 of oxygen for temperature lower than 650 C (reactor operating conditions).The experimental study shows that sodium chromite is formed at 650 C in the sodium containing 200 μg.g -1 of oxygen. At the same concentration and at 550 C, sodium chromite is clearly observed only for long immersion time (≥ 5000 h). Results at 450 C are more difficult to interpret. Furthermore, the steel is depleted in chromium in all cases.The results suggest the sodium chromite is dissolved in the sodium at the same time it is formed. Modelling of sodium chromite formation - approached by chromium diffusion in steel (in grain and grain boundaries -, and dissolution - assessed by transport in liquid metal - show that simultaneous formation and dissolution of sodium chromite is a possible mechanism able to explain our results. (author) [fr

Sodium leak and combustion experiment-II report. Evaluation result of damage of mild steel liner

Energy Technology Data Exchange (ETDEWEB)

Aoto, K.; Hirakawa, Y.; Kuroda, T. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

Several material analyses on damage of the floor liner made of a mild steel which was in the test cell of the second sodium leak and combustion experiment (Test-2) performed in OEC/PNC on June 7 in 1996 were carried out to clarify the following issues. (1) Difference of the corrosion mechanism of Test-2 liner to that of the first sodium leak and combustion experiment (Test-1) liner. (2) The vital factor which can desides corrosion mechanism and damage location. The following analyses were accomplished. (1) Microstructure observation, (2) EPMA for cross-section of vicinity of corroded area, (3) X-ray diffraction (XRD) for the interface between corrosion product-liner (mild steel). The differences between the corrosion mechanism of Test-1 liner which is seemed to be the same that of `MONJU` liner and that of Test-2 liner is discussed based on the results of these material analyses. As the result, the Na-Fe double oxidization with mechanical/chemical removal of reaction product can be occurred on the Test-1 and `MONJU` liner. On the other hand, a hot-corrosion, that is the molten salt type corrosion is subject to be thinning of the Test-2 liner. All failures of Test-2 liner surround at the halfway up a convex. Considering the above corrosion mechanism, that fact leads that significant damage is occurred at the molten salt level. (author)

Welding stainless steels for structures operating at liquid helium temperature

International Nuclear Information System (INIS)

Witherell, C.E.

1980-01-01

Superconducting magnets for fusion energy reactors require massive monolithic stainless steel weldments which must operate at extremely low temperatures under stresses approaching 100 ksi (700 MPa). A three-year study was conducted to determine the feasibility of producing heavy-section welds having usable levels of strength and toughness at 4.2 0 K for fabrication of these structures in Type 304LN plate. Seven welding processes were evaluated. Test weldments in full-thickness plate were made under severe restraint to simulate that of actual structures. Type 316L filler metal was used for most welds. Welds deposited under some conditions and which solidify as primary austenite have exhibited intergranular embrittlement at 4.2 0 K. This is believed to be associated with grain boundary metal carbides or carbonitrides precipitated during reheating of already deposited beads by subsequent passes. Weld deposits which solidify as primary delta ferrite appear immune. Through use of fully austenitic filler metals of low nitrogen content under controlled shielded metal arc welding conditions, and through use of filler metals solidifying as primary delta ferrite where only minimum residuals remain to room temperature, welds of Type 316L composition have been made with 4.2K yield strength matching that of Type 304LN plate and acceptable levels of soundness, ductility and toughness

Hot cracking of welded joints of the 7CrMoVTiB 10-10 (T/P24) steel

Energy Technology Data Exchange (ETDEWEB)

Adamiec, J, E-mail: janusz.adamiec@polsl.pl [Department of Materials Science, Silesian University of Technology, Krasinskiego 8, 40-019 Katowice (Poland)

2011-05-15

Bainitic steel 7CrMoVTiB10-10 is one the newest steels for waterwalls of modern industrial boilers. In Europe, attempts have been made to make butt welded joints of pipes made of this steel of the diameter up to 51 mm and thickness up to 8 mm. Many cracks have been observed in the welded joint, both during welding and transport and storage. The reasons of cracking and the prevention methods have not been investigated. No comprehensive research is carried out in Europe in order to automate the welding process of the industrial boiler elements made of modern bainitic steel, such as 7CrMoVTiB10-10. There is no information about its overall, operative and local weldability, influence of heat treatment, as well as about resistance of the joints to cracking during welding and use. The paper presents experience of Energoinstal SA from development of technology and production of waterwalls of boilers made of the 7CrMoVTiB 10-10 steel on a multi-head automatic welder for submerged arc welding.

Safety review of experiments at Albuquerque Operations Office

International Nuclear Information System (INIS)

Elliott, K.

1984-01-01

The Department of Energy (DOE) Albuquerque Operations Office is responsible for the safety overview of nuclear reactor and critical assembly facilities at Sandia National Laboratories, Los Alamos National Laboratory, and the Rocky Flats Plant. The important safety concerns with these facilities involve the complex experiments that are performed, and that is the area emphasized. A determination is made by the Albuquerque Office (AL) with assistance from DOE/OMA whether or not a proposed experiment is an unreviewed safety question. Meetings are held with the contractor to resolve and clarify questions that are generated during the review of the proposed experiment. The AL safety evaluation report is completed and any recommendations are discussed. Prior to the experiment a preoperational appraisal is performed to assure that personnel, procedures, and equipment are in readiness for operations. During the experiment, any abnormal condition is reviewed in detail to determine any safety concerns

Effects of hydrogen on carbon steels at the Multi-Function Waste Tank Facility

International Nuclear Information System (INIS)

Carlos, W.C.

1995-01-01

Concern has been expressed that hydrogen produced by corrosion, radiolysis, and decomposition of the waste could cause embrittlement of the carbon steel waste tanks at Hanford. The concern centers on the supposition that the hydrogen evolved in many of the existing tanks might penetrate the steel wall of the tank and cause embrittlement that might lead to catastrophic failure. This document reviews literature on the effects of hydrogen on the carbon steel proposed for use in the Multi-Function Waste Tank Facility for the time periods before and during construction as well as for the operational life of the tanks. The document draws several conclusions about these effects. Molecular hydrogen is not a concern because it is not capable of entering the steel tank wall. Nascent hydrogen produced by corrosion reactions will not embrittle the steel because the mild steel used in tank construction is not hard enough to be susceptible to hydrogen stress cracking and the corrosion product hydrogen is not produced at a rate sufficient to cause either loss in tensile ductility or blistering. If the steel intended for use in the tanks is produced to current technology, fabricated in accordance with good construction practice, postweld heat treated, and operated within the operating limits defined, hydrogen will not adversely affect the carbon steel tanks during their 50-year design life. 26 refs

Tritiated Water Interaction with Stainless Steel

International Nuclear Information System (INIS)

Glen R. Longhurst

2007-01-01

Experiments conducted to study tritium permeation of stainless steel at ambient and elevated temperatures revealed that HT converts relatively quickly to HTO. Further, the HTO partial pressure contributes essentially equally with elemental tritium gas in driving permeation through the stainless steel. Such permeation appears to be due to dissociation of the water molecule on the hot stainless steel surface. There is an equilibrium concentration of HTO vapor above adsorbed gas on the walls of the experimental apparatus evident from freezing transients. The uptake process of tritium from the carrier gas involves both surface adsorption and isotopic exchange with surface bound water

Physical characterization of steel and stainless steel metal powders

International Nuclear Information System (INIS)

Lavilla, A.O.; Lucchesi, C.G.; Sandin, O.O.

1991-01-01

A methodology has been developed for the physical characterization of steel powders (obtained by atomization) for later sintering and for the construction of porous sheets and filtrating tubes, capable of operating at temperatures between 600 deg C and 800 deg C in corrosive atmospheres. This methodology was based on the equipment and methods used for the physical characterization of uranium oxide powders. (Author) [es

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

Science.gov (United States)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.

Operational experience with CMS Tier-2 sites

International Nuclear Information System (INIS)

Gonzalez Caballero, I

2010-01-01

In the CMS computing model, more than one third of the computing resources are located at Tier-2 sites, which are distributed across the countries in the collaboration. These sites are the primary platform for user analyses; they host datasets that are created at Tier-1 sites, and users from all CMS institutes submit analysis jobs that run on those data through grid interfaces. They are also the primary resource for the production of large simulation samples for general use in the experiment. As a result, Tier-2 sites have an interesting mix of organized experiment-controlled activities and chaotic user-controlled activities. CMS currently operates about 40 Tier-2 sites in 22 countries, making the sites a far-flung computational and social network. We describe our operational experience with the sites, touching on our achievements, the lessons learned, and the challenges for the future.

Japan steel mill perspective

Energy Technology Data Exchange (ETDEWEB)

Murase, K. [Kobe Steel Ltd., Tokyo (Japan)

2004-07-01

The international and Japan's steel industry, the coking coal market, and Japan's expectations from Canada's coal industry are discussed. Japan's steel mills are operating at full capacity. Crude steel production for the first half of 2004 was 55.8 million tons. The steel mills are profitable, but costs are high, and there are difficulties with procuring raw materials. Japan is trying to enhance the quality of coke, in order to achieve higher productivity in the production of pig iron. Economic growth is rising disproportionately in the BRICs (Brazil, Russia, India, and China), with a large increase in coking coal demand from China. On the supply side, there are several projects underway in Australia and Canada to increase production. These include new developments by Elk Valley Coal Corporation, Grande Cache Coal, Western Canadian Coal, and Northern Energy and Mining in Canada. The Elga Mine in the far eastern part of Russia is under development. But the market is expected to remain tight for some time. Japan envisions Canadian coal producers will provide a stable coal supply, expansion of production and infrastructure capabilities, and stabilization of price. 16 slides/overheads are included.

Vacuum system operating experience review for fusion applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1994-03-01

This report presents a review of vacuum system operating experiences from particle accelerator, fusion experiment, space simulation chamber, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of vacuum system component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with vacuum systems are discussed, including personnel safety, foreign material intrusion, and factors relevant to vacuum systems being the primary confinement boundary for tritium and activated dusts. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor

Aging degradation of cast stainless steel

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1985-10-01

A program is being conducted to investigate the significance of in-service embrittlement of cast-duplex stainless steels under light-water reactor operating conditions. Data from room-temperature Charpy-impact tests for several heats of cast stainless steel aged up to 10,000 h at 350, 400, and 450 0 C are presented and compared with results from other studies. Microstructures of cast-duplex stainless steels subjected to long-term aging either in the laboratory or in reactor service have been characterized. The results indicate that at least two processes contribute to the low-temperature embrittleent of duplex stainless steels, viz., weakening of the ferrite/austenite phase boundary by carbide precipitation and embrittlement of ferrite matrix by the formation of additional phases such as G-phase, Type X, or the α' phase. Carbide precipitation has a significant effect on the onset of embrittlement of CF-8 and -8M grades of stainless steels aged at 400 or 450 0 C. The existing correlations do not accurately represent the embrittlement behavior over the temperature range 300 to 450 0 C. 18 refs., 13 figs

Spacelab operations planning. [ground handling, launch, flight and experiments

Science.gov (United States)

Lee, T. J.

1976-01-01

The paper reviews NASA planning in the fields of ground, launch and flight operations and experiment integration to effectively operate Spacelab. Payload mission planning is discussed taking consideration of orbital analysis and the mission of a multiuser payload which may be either single or multidiscipline. Payload analytical integration - as active process of analyses to ensure that the experiment payload is compatible to the mission objectives and profile ground and flight operations and that the resource demands upon Spacelab can be satisfied - is considered. Software integration is touched upon and the major integration levels in ground operational processing of Spacelab and its experimental payloads are examined. Flight operations, encompassing the operation of the Space Transportation System and the payload, are discussed as are the initial Spacelab missions. Charts and diagrams are presented illustrating the various planning areas.

Medium carbon vanadium steels for closed die forging

International Nuclear Information System (INIS)

Jeszensky, Gabor; Plaut, Ronald Lesley

1993-01-01

This work analyses the medium carbon micro alloyed vanadium potential for closed die forged production. The steels reach the mechanical resistance requests during cooling after forging, eliminating the subsequent thermal treatment. Those steels also present good fatigue resistance and machinability. The industrial scale experiments are also reported

Heavy-Section Steel Technology program fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1989-10-01

Large scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL. 24 refs., 18 figs

Heavy-section steel technology program: Fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

Large-scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low-strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring-forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL

Effect of smelt aluminium on mechanical properties of steels

International Nuclear Information System (INIS)

Ryabov, V.R.; Dykhno, I.S.; Deev, G.F.; Karikh, V.V.

1987-01-01

Effect of smelt aluminium on mechanical properties of armco-iron and 12 Kh18N10T steel is studied. It is stated that in smelt aluminium and aluminium alloy contact with armco-iron the sample ductility is decreased. Corrosion effect of smelt alluminium on (18Kh15N5AM3) steel in the form of reinforced wire in aluminium-steel KAS-1A composite material is investigted. It is stated in experiment that during smelt alluminium-steel contact interaction of heterogeneous phases takes place

Operational safety experience feedback by means of unusual event reports

International Nuclear Information System (INIS)

1996-07-01

Operational experience of nuclear power plants can be used to great advantage to enhance safety performance provided adequate measures are in place to collect and analyse it and to ensure that the conclusions drawn are acted upon. Feedback of operating experience is thus an extremely important tool to ensure high standards of safety in operational nuclear power plants and to improve the capability to prevent serious accidents and to learn from minor deviations and equipment failures - which can serve as early warnings -to prevent even minor events from occurring. Mechanisms also need to be developed to ensure that operating experience is shared both nationally as well as internationally. The operating experience feedback process needs to be fully and effectively established within the nuclear power plant, the utility, the regulatory organization as well as in other institutions such as technical support organizations and designers. The main purpose of this publication is to reflect the international consensus as to the general principles and practices in the operational safety experience feedback process. The examples of national practices for the whole or for particular parts of the process are given in annexes. The publication complements the IAEA Safety Series No.93 ''Systems for Reporting Unusual Events in Nuclear Power Plants'' (1989) and may also give a general guidance for Member States in fulfilling their obligations stipulated in the Nuclear Safety Convention. Figs, tabs

Operational safety experience feedback by means of unusual event reports

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

Operational experience of nuclear power plants can be used to great advantage to enhance safety performance provided adequate measures are in place to collect and analyse it and to ensure that the conclusions drawn are acted upon. Feedback of operating experience is thus an extremely important tool to ensure high standards of safety in operational nuclear power plants and to improve the capability to prevent serious accidents and to learn from minor deviations and equipment failures - which can serve as early warnings -to prevent even minor events from occurring. Mechanisms also need to be developed to ensure that operating experience is shared both nationally as well as internationally. The operating experience feedback process needs to be fully and effectively established within the nuclear power plant, the utility, the regulatory organization as well as in other institutions such as technical support organizations and designers. The main purpose of this publication is to reflect the international consensus as to the general principles and practices in the operational safety experience feedback process. The examples of national practices for the whole or for particular parts of the process are given in annexes. The publication complements the IAEA Safety Series No.93 ``Systems for Reporting Unusual Events in Nuclear Power Plants`` (1989) and may also give a general guidance for Member States in fulfilling their obligations stipulated in the Nuclear Safety Convention. Figs, tabs.

Operation experience with elevated ammonia

International Nuclear Information System (INIS)

Vankova, Katerina; Kysela, Jan; Malac, Miroslav; Petrecky, Igor; Svarc, Vladimir

2011-01-01

The 10 VVER units in the Czech and Slovak Republics are all in very good water chemistry and radiation condition, yet questions have arisen regarding the optimization of cycle chemistry and improved operation in these units. To address these issues, a comprehensive experimental program for different water chemistries of the primary circuit was carried out at the Rez Nuclear Research Institute, Czech Republic, with the goal of judging the influence of various water chemistries on radiation build-up. Four types of water chemistries were compared: standard VVER water chemistry (in common use), direct hydrogen dosing without ammonia, standard VVER water chemistry with elevated ammonia levels, and zinc dosing to standard VVER water chemistry. The test results showed that the types of water chemistry other than the common one have benefits for the operation of the nuclear power plant (NPP) primary circuit. Operation experience with elevated ammonia at NPP Dukovany Units 3 and 4 is presented which validates the experimental results, demonstrating improved corrosion product volume activity. (orig.)

Operational experience feedback with precursor analysis

International Nuclear Information System (INIS)

Koncar, M.; Ferjancic, M.; Muehleisen, A.; Vojnovic, D.

2003-01-01

Experience of practical operation is a valuable source of information for improving the safety and reliability of nuclear power plants. Operational experience feedback (Olef) system manages this aspect of NPP operation. The traditional ways of investigating operational events, such as the root cause analysis (RCA), are predominantly qualitative. RCA as a part of the Olef system provides technical guidance and management expectations in the conduct of assessing the root cause to prevent recurrence, covering the following areas: conditions preceding the event, sequence of events, equipment performance and system response, human performance considerations, equipment failures, precursors to the event, plant response and follow-up, radiological considerations, regulatory process considerations and safety significance. The root cause of event is recognized when there is no known answer on question 'why has it happened?' regarding relevant condition that may have affected the event. At that point the Olef is proceeding by actions taken in response to events, utilization, dissemination and exchange of operating experience information and at the end reviewing the effectiveness of the Olef. Analysis of the event and the selection of recommended corrective/preventive actions for implementation and prioritization can be enhanced by taking into account the information and insights derived from Pasa-based analysis. A Pasa based method, called probabilistic precursor event analysis (PPE A) provides a complement to the RCA approach by focusing on how an event might have developed adversely, and implies the mapping of an operational event on a probabilistic risk model of the plant in order to obtain a quantitative assessment of the safety significance of the event PSA based event analysis provides, due to its quantitative nature, appropriate prioritization of corrective actions. PPEA defines requirements for PSA model and code, identifies input requirements and elaborates following

High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

International Nuclear Information System (INIS)

Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

1989-01-01

In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

Interface fracture behavior of zinc coatings on steel : Experiments and finite element calculations

NARCIS (Netherlands)

Song, G.M.; Sloof, W.G.; Pei, Y.T.; de Hosson, J.T.M.

2006-01-01

Hot-dipped galvanized steels are widely used in the automotive industry. The formability and damage resistance of zinc coatings depend strongly on their microstructure and adhesion to the steel substrate. In order to improve the mechanical performance of zinc coatings, the influence of their

Experience in melting of high-quality chromium-nickel-molybdenum steel in oxygen converter

Energy Technology Data Exchange (ETDEWEB)

Kosoi, L F; Yaburov, S I; Shul' kin, M L; Vedernikov, G G; Bragin, E D; Filork' yan, B K

1978-10-01

Technology of melting high-quality medium-carbon constructional chromium-nickel-molybdenum steel has been developed and tested in 130-t converters. The technology envisages metal refinement in a casting laddle using synthetic lime-aluminous slag and argon blowing, as well as liquid ferroallys (master alloys) for steel deoxidation and alloying. Due to a smaller content of sulfur, phosphorus, arsenic and sulphide inclusions, and to a smaller grain size (N 11-12), the steel, produced according to this technology possesses higher plastic properties and impact strength than conventional open-hearth furnace metal after heat treatment for the same strength.

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

Experiments on the Impact of language Problems in the Multi-cultural Operation of NPPs' Emergency Operation

International Nuclear Information System (INIS)

Kang, Seongkeun; Kim, Taehoon; Seong, Poong Hyun; Ha, Jun Su

2016-01-01

In 2010, The Korea Electric Power Corporation (KEPCO) was awarded a multi-billion dollar bid to construct the first nuclear power plant in Barakah, UAE. One must keep in mind however, that with technology transfer and international cooperation comes a host of potential problems arising from cultural differences such as language, everyday habitudes and workplace expectation. As of now, how problematic these potential issues may become is unknown. Of the aforementioned factors, communication is perhaps of foremost importance. We investigated UAE culture-related issues through analysis of operating experience reviews (OERs) and came to the conclusion that the language barrier needed utmost attention. Korean nuclear power plant operators will work in UAE and will operate the NPPs with operators and managers of other nationalities as well. The purpose of this paper is firstly to confirm that operators are put under mental stress, and secondly to demonstrate the decline in accuracy when they must work in English. Reducing human error is quite important to make nuclear power plants safer. As the mental workload of human operator is increased, the probability of a human error occurring also increases. It will have a negative influence on the plant’s safety. There are many factors which can potentially increase mental workload. We focused on communication problem which is a key factor of increasing mental workload because many Korean operators will work in UAE nuclear power plants and may work together with UAE operators. From these experiments we compared how performance of both Korean and UAE subjects were decreased when they use English. We designed experimental methods to be able to check this problem qualitatively and quantitatively. We analyzed four factors to find the communication problems from the experiments which are accuracy, efficiency, NASA-TLX, and brain wave. Accuracy, efficiency, brain wave are quantitative factors, and NASA-TLX is qualitative factor. To

Corrosion of low carbon steel by microorganisms from the 'pigging' operation debris in water injection pipelines.

Science.gov (United States)

Cote, Claudia; Rosas, Omar; Sztyler, Magdalena; Doma, Jemimah; Beech, Iwona; Basseguy, Régine

2014-06-01

Present in all environments, microorganisms develop biofilms adjacent to the metallic structures creating corrosion conditions which may cause production failures that are of great economic impact to the industry. The most common practice in the oil and gas industry to annihilate these biofilms is the mechanical cleaning known as "pigging". In the present work, microorganisms from the "pigging" operation debris are tested biologically and electrochemically to analyse their effect on the corrosion of carbon steel. Results in the presence of bacteria display the formation of black corrosion products allegedly FeS and a sudden increase (more than 400mV) of the corrosion potential of electrode immersed in artificial seawater or in field water (produced water mixed with aquifer seawater). Impedance tests provided information about the mechanisms of the interface carbon steel/bacteria depending on the medium used: mass transfer limitation in artificial seawater was observed whereas that in field water was only charge transfer phenomenon. Denaturing Gradient Gel Electrophoresis (DGGE) results proved that bacterial diversity decreased when cultivating the debris in the media used and suggested that the bacteria involved in the whole set of results are mainly sulphate reducing bacteria (SRB) and some other bacteria that make part of the taxonomic order Clostridiales. Copyright © 2013 Elsevier B.V. All rights reserved.

Low-temperature creep of austenitic stainless steels

Science.gov (United States)

Reed, R. P.; Walsh, R. P.

2017-09-01

Plastic deformation under constant load (creep) in austenitic stainless steels has been measured at temperatures ranging from 4 K to room temperature. Low-temperature creep data taken from past and unreported austenitic stainless steel studies are analyzed and reviewed. Creep at cryogenic temperatures of common austenitic steels, such as AISI 304, 310 316, and nitrogen-strengthened steels, such as 304HN and 3116LN, are included. Analyses suggests that logarithmic creep (creep strain dependent on the log of test time) best describe austenitic stainless steel behavior in the secondary creep stage and that the slope of creep strain versus log time is dependent on the applied stress/yield strength ratio. The role of cold work, strain-induced martensitic transformations, and stacking fault energy on low-temperature creep behavior is discussed. The engineering significance of creep on cryogenic structures is discussed in terms of the total creep strain under constant load over their operational lifetime at allowable stress levels.

Power ramping/cycling experience and operational recommendations in KWU power plants

International Nuclear Information System (INIS)

Jan, R. von; Wunderlich, F.; Holzer, R.

1980-01-01

The power cycling and ramping experience of KWU is based on experiments in test and commercial reactors, and on evaluation of plant operation (PHWR, PWR and BWR). Power cycling of fuel rods have never lead to PCI failures. In ramping experiments, for fast ramps PCI failure thresholds of 480/420 W/cm are obtained at 12/23 GWd/t(U) burn-up for pressurized PWR fuel. No failures occurred during limited exceedance of the threshold with reduced ramp rate. Operational recommendations used by KWU are derived from experiments and plant experience. The effects of ramping considerations on plant operation is discussed. No rate restrictions are required for start-ups during an operating cycle or load follow operation within set limits for the distortion of the local power distribution. In a few situations, e.g. start-up after refueling, ramp rates of 1 to 5 %/h are recommended depending on plant and fuel design

Operating procedures: Fusion Experiments Analysis Facility

Energy Technology Data Exchange (ETDEWEB)

Lerche, R.A.; Carey, R.W.

1984-03-20

The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility.

Operating procedures: Fusion Experiments Analysis Facility

International Nuclear Information System (INIS)

Lerche, R.A.; Carey, R.W.

1984-01-01

The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Microstructure and Mechanical Properties of Dissimilar Friction Stir Spot Welding Between St37 Steel and 304 Stainless Steel

Science.gov (United States)

Khodadadi, Ali; Shamanian, Morteza; Karimzadeh, Fathallah

2017-05-01

In the present study, St37 low-carbon steel and 304 stainless steel were welded successfully, with the thickness of 2 mm, by a friction stir spot welding process carried out at the tool dwell time of 6 s and two different tool rotational speeds of 630 and 1250 rpm. Metallographic examinations revealed four different zones including SZ and HAZ areas of St37 steel and SZ and TMAZ regions of 304 stainless steel in the weld nugget, except the base metals. X-ray diffraction and energy-dispersive x-ray spectroscopy experiments were used to investigate the possible formation of such phases as chromium carbide. Based on these experiments, no chromium carbide precipitation was found. The recrystallization of the weld nugget in the 304 steel and the phase transformations of the weld regions in the St37 steel enhanced the hardness of the weld joint. Hardness changes of joint were acceptable and approximately uniform, as compared to the resistance spot weld. In this research, it was also observed that the tensile/shear strength, as a crucial factor, was increased with the rise in the tool rotational speed. The bond length along the interface between metals, as an effective parameter to increase the tensile/shear strength, was also determined. At higher tool rotational speeds, the bond length was found to be improved, resulting in the tensile/shear strength of 6682 N. Finally, two fracture modes were specified through the fracture mode analysis of samples obtained from the tensile/shear test consisting of the shear fracture mode and the mixed shear/tensile fracture mode.

Effect of welding thermal cycles on the structure and properties of simulated heat-affected zone areas in X10CrMoVNb9-1 (T91) steel at a state after 100,000 h of operation

Energy Technology Data Exchange (ETDEWEB)

Łomozik, Mirosław, E-mail: miroslaw.lomozik@is.gliwice.pl [Instytut Spawalnictwa, Testing of Materials Weldability and Welded Constructions Department, 44-100 Gliwice, Bł. Czesława 16-18 (Poland); Hernas, Adam, E-mail: adam.hernas@polsl.pl [Silesian University of Technology, Faculty of Materials Engineering and Metallurgy, 40-019 Katowice, Krasińskiego 8 str. (Poland); Zeman, Marian L., E-mail: marian.zeman@is.gliwice.pl [Instytut Spawalnictwa, Testing of Materials Weldability and Welded Constructions Department, 44-100 Gliwice, Bł. Czesława 16-18 (Poland)

2015-06-18

The article presents results of structural tests (light, scanning electron and scanning transmission electron microscopy) of X10CrMoVNb9-1 (T91) creep-resisting steel after approximately 100,000 h of operation. It was ascertained that the parent metal of T91 steel is characterized by the microstructure of tempered martensite with M{sub 23}C{sub 6} carbide precipitates and few dispersive precipitates of MX-type niobium and vanadium carbonitrides. The most inconvenient change in T91 steel precipitate morphology due to long-term operation is the appearance of the Laves Fe{sub 2}Mo phase which along with M{sub 23}C{sub 6} carbide particles forms elongated blocks and conglomerates on grain boundaries. The article also presents results of tests related to the effect of simulated welding thermal cycles on selected properties of X10CrMoVNb9-1 (T91) grade steel at a state after approximately 100,000 h of operation. The tests involved the determination of the chemical composition of the steel tested as well as impact tests, hardness measurements and microscopic metallographic examination (based on light microscopy) of simulated heat-affected zone (HAZ) areas for a cooling time (t{sub 8/5}) restricted within a range between 3 s and 120 s, with and without heat treatment. The tests revealed that, among other results, hardness values of simulated HAZ areas in X10CrMoVNb9-1 (T91) steel do not guarantee cold crack safety of the steel at the state without additional heat treatment. It was also observed that simulated welding thermal cycles of cooling times t{sub 8/5}=3, 12, 60 and 120 s do not significantly affect the toughness and hardness of simulated HAZ areas of the steel tested.

Reduced-activation steels: present status and future development

International Nuclear Information System (INIS)

Klueh, R.L.

2007-01-01

Full text of publication follows: Reduced-activation steels for fusion reactor applications were developed in the 1980's to replace the commercial elevated- temperature steels first considered. In the United States, this involved replacing Sandvik HT9 and modified 9Cr-1Mo steels. Reduced-activation steels, which were developed for more rapid radioactivity decay following exposure in a fusion neutron environment, were patterned after the commercial steels they were to replace. The objective for the reduced-activation steels was that they have strengths (yield stress and ultimate tensile strength from room temperature to 600 deg. C) and impact toughness (measured in a Charpy test) comparable to or better than the steels they were replacing. That objective was achieved in reduced-activation steels developed in Japan, Europe, and the United States. Since the reduced-activation steels were developed in the 1980's, reactor designers have been interested designs for increased efficiency of future fusion plants. This means reactors will need to operate at higher temperatures-above 550 deg. C, which is the upper-temperature limit for the reduced-activation steels. Although the tensile and impact toughness of the reduced-activation steels exceed those of the commercial steels they were patterned after, their creep-rupture properties are inferior to some of the commercial steels they replaced. furthermore, they are much inferior to commercial steels that have been developed since the 1980's. Reasons for why the creep-rupture properties for the new commercial ferritic/martensitic steels are superior to the earlier commercial steels and the reduced-activation steels were examined. The reasons involve compositional changes that were made in the earlier commercial steels to give the new commercial steels their superior properties. Computational thermodynamics calculations were carried out to compare the expected equilibrium phases. It appears that similar changes in composition

Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

International Nuclear Information System (INIS)

Cave, L.; Clarke, A.W.

1984-01-01

The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

Stress Corrosion Cracking of Steel and Aluminum in Sodium Hydroxide: Field Failure and Laboratory Test

Directory of Open Access Journals (Sweden)

Y. Prawoto

2012-01-01

Full Text Available Through an investigation of the field failure analysis and laboratory experiment, a study on (stress corrosion cracking SCC behavior of steel and aluminum was performed. All samples were extracted from known operating conditions from the field failures. Similar but accelerated laboratory test was subsequently conducted in such a way as to mimic the field failures. The crack depth and behavior of the SCC were then analyzed after the laboratory test and the mechanism of stress corrosion cracking was studied. The results show that for the same given stress relative to ultimate tensile strength, the susceptibility to SCC is greatly influenced by heat treatment. Furthermore, it was also concluded that when expressed relative to the (ultimate tensile strength UTS, aluminum has similar level of SCC susceptibility to that of steel, although with respect to the same absolute value of applied stress, aluminum is more susceptible to SCC in sodium hydroxide environment than steel.

Experimental study of the effect of gas nature on plasma arc cutting of mild steel

International Nuclear Information System (INIS)

Kavka, T; Mašláni, A; Hrabovský, M; Křenek, P; Stehrer, T; Pauser, H

2013-01-01

This paper is devoted to the experimental investigation of arc cutting of mild steel using plasmas generated in gas and liquid media. Due to different chemical compositions, the examined media have different thermophysical properties, which affect the properties of the generated plasma and cutting performance. The experiments are performed on 15 mm mild steel plates using commercial equipment at 60 A to approach real operation conditions in application areas. The studied gases are chosen according to recommendations of the world's leading manufacturers of arc cutting equipment for mild steel. Specific differences between plasma gases are discussed from the point of view of properties of the gas and the generated plasma, amount of removed material, kerf shape and overall energy balance of the cutting process. The paper describes the role of exothermic reaction of iron oxidation for oxygen cutting and explains its neglect for liquid cutting. This paper explains the potential of facilitating the cutting process by modification of the plasma gas chemical composition and flow rate. (paper)

Medium carbon vanadium micro alloyed steels for drop forging

International Nuclear Information System (INIS)

Jeszensky, Gabor; Plaut, Ronald Lesley

1992-01-01

Growing competitiveness of alternative manufacturing routes requires cost minimization in the production of drop forged components. The authors analyse the potential of medium carbon, vanadium microalloyed steels for drop forging. Laboratory and industrial experiments have been carried out emphasizing deformation and temperature cycles, strain rates and dwell times showing a typical processing path, associated mechanical properties and corresponding microstructures. The steels the required levels of mechanical properties on cooling after forging, eliminating subsequent heat treatment. The machinability of V-microalloyed steels is also improved when compared with plain medium carbon steels. (author)

Heat treatments in a conventional steel to reproduce the microstructure of a nuclear grade steel

International Nuclear Information System (INIS)

Rosalio G, M.

2014-01-01

The ferritic steels used in the manufacture of pressurized vessels of Boiling Water Reactors (BWR) suffer degradation in their mechanical properties due to damage caused by the neutron fluxes of high energy bigger to a Mega electron volt (E> 1 MeV) generated in the reactor core. The materials with which the pressurized vessels of nuclear reactors cooled by light water are built correspond to low alloy ferritic steels. The effect of neutron irradiation on these steels is manifested as an increase in hardness, mechanical strength, with the consequent decrease in ductility, fracture toughness and an increase in temperature of ductile-brittle transition. The life of a BWR is 40 years, its design must be considered sufficient margin of safety because pressure forces experienced during operation, maintenance and testing of postulated accident conditions. It is necessary that under these conditions the vessel to behave ductile and likely to propagate a fracture is minimized. The vessels of light water nuclear reactors have a bainite microstructure. Specifically, the reactor vessels of the nuclear power plant of Laguna Verde (Veracruz, Mexico) are made of a steel Astm A-533, Grade B Class 1. At present they are carrying out some welding tests for the construction of a model of a BWR, however, to use nuclear grade steel such as Astm A-533 to carry out some of the welding tests, is very expensive; perform these in a conventional material provides basic information. Although the microstructure present in the conventional material does not correspond exactly to the degree of nuclear material, it can take of reference. Therefore, it is proposed to conduct a pilot study to establish the thermal treatment that reproduces the microstructure of nuclear grade steel, in conventional steel. The resulting properties of the conventional steel samples will be compared to a JRQ steel, that is a steel Astm A-533, Grade B Class 1, provided by IAEA. (Author)

Stainless Steel Foil with Improved Creep-Resistance for Use in Primary Surface Recuperators for Gas Turbine Engines

International Nuclear Information System (INIS)

Browning, P.F.; Fitzpatrick, M.; Grubb, J.F.; Klug, R.C.; Maziasz, P.J.; Montague, J.P.; Painter, R.A.; Swindeman, R.W.

1998-01-01

Primary surface recuperators (PSRs) are compact heat-exchangers made from thin-foil type 347 austenitic stainless steel, which boost the efficiency of land-based gas turbine engines. Solar Turbines uses foil folded into a unique corrugated pattern to maximize the primary surface area for efficient heat transfer between hot exhaust gas on one side, and the compressor discharge air on the other side of the foil. Allegheny-Ludlum produces 0.003 - 0.0035 in. thick foil for a range of current turbine engines using PSRs that operate at up to 660 degrees C. Laboratory-scale processing modification experiments recently have demonstrated that dramatic improvements can be achieved in the creep resistance of such typical 347 stainless steel foils. The modified processing enables fine NbC carbide precipitates to develop during creep at 650-700 degrees C, which provides strength even with a fine grain size. Such improved creep-resistance is necessary for advanced turbine systems that will demand greater materials performance and reliability at higher operating conditions. The next challenges are to better understand the nature of the improved creep resistance in these 347 stainless steel foil, and to achieve similar improvements with scale-up to commercial foil production

GNF2 Operating Experience

International Nuclear Information System (INIS)

Schardt, John

2007-01-01

GNF's latest generation fuel product, GNF2, is designed to deliver improved nuclear efficiency, higher bundle and cycle energy capability, and more operational flexibility. But along with high performance, our customers face a growing need for absolute fuel reliability. This is driven by a general sense in the industry that LWR fuel reliability has plateaued. Too many plants are operating with fuel leakers, and the impact on plant operations and operator focus is unacceptable. The industry has responded by implementing an INPO-coordinated program aimed at achieving leaker-free reliability by 2010. One focus area of the program is the relationship between fuel performance (i.e., duty) and reliability. The industry recognizes that the right balance between performance and problem-free fuel reliability is critical. In the development of GNF2, GNF understood the requirement for a balanced solution and utilized a product development and introduction strategy that specifically addressed reliability: evolutionary design features supported by an extensive experience base; thoroughly tested components; and defense-in-depth mitigation of all identified failure mechanisms. The final proof test that the balance has been achieved is the application of the design, initially through lead use assemblies (LUAs), in a variety of plants that reflect the diversity of the BWR fleet. Regular detailed surveillance of these bundles provides the verification that the proper balance between performance and reliability has been achieved. GNF currently has GNF2 lead use assemblies operating in five plants. Included are plants that have implemented extended power up-rates, plants on one and two-year operating cycles, and plants with and without NobleChem TM and zinc injection. The leading plant has undergone three pool-side inspections outages to date. This paper reviews the actions taken to insure GNF2's reliability, and the lead use assembly surveillance data accumulated to date to validate

Special stainless steels for sea water service

International Nuclear Information System (INIS)

Tomaselli, A.C.

1983-01-01

Very exacting demands are made on the corrosion resistance and mechanical properties of materials which in their service come into contact with seawater, and in many cases simultaneously with corrosive process solutions. The demand for higher alloy stainless steels for seawater application is rising in pace with the increasing requirements for safety and operation economy. The corrosion conditions in seawater and the resistance of stainless steels in this medium will be dealt with in the following. Sanicro 28 will then be compared with stainless steels, types AISI 304, 316 and 317, as well as with Alloy 20, Alloy 825 and SANDVIK 2RK65. (Author) [pt

Experiment and simulation analysis of roll-bonded Q235 steel plate

International Nuclear Information System (INIS)

Zhao, G.; Huang, Q.; Zhou, C.; Zhang, Z.; Ma, L.; Wang, X.

2016-01-01

Heavy-gauge Q235 steel plate was roll bonded, and the process was simulated using MARC software. Ultrasonic testing results revealed the presence of cracks and lamination defects in an 80-mm clad steel sheet, especially at the head and tail of the steel plate. There were non-uniform ferrite + pearlite microstructures and unbound areas at a bond interface. Through scanning electron microscopy analysis, long cracks and additional inclusions in the cracks were observed at the interface. A fracture analysis revealed non-uniform inclusions that pervaded the interface. Moreover, MARC simulations demonstrated that there was little equivalent strain at the centre of the slab during the first rolling pass. The equivalent centre increased to 0.5 by the fourth rolling pass. Prior to the final pass, the equivalent strain was not consistent across the thickness direction, preventing bonding interfaces from forming consistent deformation and decreasing the residual stress. The initial rolling reduction rate should not be very small (e.g. 5%) as it is averse to the coordination of rolling deformation. Such rolling processes are averse to the rolling bond. (Author)

Formability of dual-phase steels in deep drawing of rectangular parts: Influence of blank thickness and die radius

Science.gov (United States)

López, Ana María Camacho; Regueras, José María Gutiérrez

2017-10-01

The new goals of automotive industry related with environment concerns, the reduction of fuel emissions and the security requirements have driven up to new designs which main objective is reducing weight. It can be achieved through new materials such as nano-structured materials, fibre-reinforced composites or steels with higher strength, among others. Into the last group, the Advance High Strength Steels (AHSS) and particularly, dual-phase steels are in a predominant situation. However, despite of their special characteristics, they present issues related to their manufacturability such as springback, splits and cracks, among others. This work is focused on the deep drawing processof rectangular shapes, a very usual forming operation that allows manufacturing several automotive parts like oil pans, cases, etc. Two of the main parameters in this process which affect directly to the characteristics of final product are blank thickness (t) and die radius (Rd). Influence of t and Rd on the formability of dual-phase steels has been analysed considering values typically used in industrial manufacturing for a wide range of dual-phase steels using finite element modelling and simulation; concretely, the influence of these parameters in the percentage of thickness reduction pt(%), a quite important value for manufactured parts by deep drawing operations, which affects to its integrity and its service behaviour. Modified Morh Coulomb criteria (MMC) has been used in order to obtain Fracture Forming Limit Diagrams (FFLD) which take into account an important failure mode in dual-phase steels: shear fracture. Finally, a relation between thickness reduction percentage and studied parameters has been established fordual-phase steels, obtaining a collection of equations based on Design of Experiments (D.O.E) technique, which can be useful in order to predict approximate results.

LMFBR operational safety: the EBR-II experience

International Nuclear Information System (INIS)

Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

1978-01-01

The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

The SM and MIR reactors operation experience

International Nuclear Information System (INIS)

Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

1995-01-01

The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

Experimental Work at Harwell on the Injection of Sodium into Liquid Steel

International Nuclear Information System (INIS)

Asher, R.C.; Bradshaw, L.; Collett, R.; Davies, D.

1976-01-01

The equipment and experimental technique used to inject molten sodium beneath the surface of molten steel (and subsequently molten uranium dioxide) are described. The results of an exploratory experiment, in which 2 g of sodium at 380 deg. C were injected into 54 g of steel at 1530 deg. C, are outlined. A violent reaction occurred, apparently as a result of a vapour explosion, and a number of pulses were recorded by a force transducer. A preliminary examination of the steel debris is reported. In conclusion: The injection of ∼2 g of sodium at ∼380 deg. C beneath the surface of ∼54 g of steel at 1530 deg. C resulted in a vapour explosion. A series of three pulses (perhaps more) at intervals of ∼6 ms were observed; the peak amplitude was at least 600 N (equivalent to a pressure of ∼20 bar on the base of the crucible). The steel was highly dispersed giving material of predominant particle size ∼500 μ and calculated surface area ∼13.5 cm 2 g -1 ; the particles were almost spherical but some had a small indentation on their surface. The apparatus operated very well, especially considering its rudimentary nature and the speed with which it was assembled. Nevertheless experience of the first experiment showed that many improvements are desirable. (a) Interpretation of the pulses recorded by the force transducer was difficult and would be facilitated by the use of a strong crucible and by a redesigned pillar. (b) Instrumentation on the injector would enable t M and t I to be determined and would permit the injection timing to be related to the pulse record, thus giving a measure of the dwell time. (c) The closed circuit television needs to be linked to a video tape recorder. (d) The recording instrumentation should have a better resolution so that short pulses are not lost or inaccurately recorded; a multichannel tape recorder may be desirable. (e) Direct measurement of the surface area of the debris should be carried out using an existing technique (BET low

Recent operating experiences with steam generators in Japanese NPPs

International Nuclear Information System (INIS)

Yashima, Seiji

1997-01-01

In 1994, the Genkai-3 of Kyushu Electric Power Co., Inc. and the Ikata-3 of Shikoku Electric Power Co., Inc. started commercial operation, and now 22 PWR plants are being operated in Japan. Since the first PWR plant now 22 PWR plants are being operated in was started to operate, Japanese PWR plants have had an operating experience of approx. 280 reactor-years. During that period, many tube degradations have been experienced in steam generators (SGs). And, in 1991, the steam generator tube rupture (SGTR) occurred in the Mihama-2 of Kansai Electric Power Co., Inc. However, the occurrence of tube degradation of SGs has been decreased by the instructions of the MITI as regulatory authorities, efforts of Electric Utilities, and technical support from the SG manufacturers. Here the author describes the recent SGs in Japan about the following points. (1) Recent Operating Experiences (2) Lessons learned from Mihama-2 SGTR (3) SG replacement (4) Safety Regulations on SG (5) Research and development on SG

National steel tries wheeling

International Nuclear Information System (INIS)

Dudak, J.R.

1992-01-01

In 1989, National Steel felt the need to take the next step to make its Detroit-based division, Great Lakes Steel, more competitive in the world flat-rolled steel market. In 1988, Great Lakes Steel started flowing natural gas through the first fully litigated bypass (Competitive Sourcing Option) of a local distribution company. In 1989, the second connection with the new supply route for gas transportation, Panhandle Eastern had started flowing and the LDC, Michigan Consolidated Gas Co. (MichCon) had pulled out their piping previously serving the plants. Since we had been able to structure a fully reliable supply route, storage and balancing program for gas in the face of such strong opposition by the LDC, the author felt it was time to attack the next singularly sourced major commodity, electricity. Electricity, at this major integrated steel plant, represented approximately 7% of plant cost yearly. Yet being monopolized, Great Lakes Division (GLD) could not multiple source this commodity like it does with its other 93% of costs, except for labor (25% of the 93%). Multiple sourcing is done to bring competitive pressure to suppliers and to diversify supplies and protect plant operation in the event of failure by one supplier. This paper describes National Steel's strategy to reduce the cost of power, at the minimum of capital costs, the most expedient way possible, that does not sacrifice any major long-term potential cost improvements. The results show that competitively priced power is available across the mid-west, at prices well below many state regulated electric utilities, for at least 5 to 15 years, but with major obstacles in obtaining transmission access

Three integrated photovoltaic/sound barrier power plants. Construction and operational experience

International Nuclear Information System (INIS)

Nordmann, T.; Froelich, A.; Clavadetscher, L.

2002-01-01

After an international ideas competition by TNC Switzerland and Germany in 1996, six companies where given the opportunity to construct a prototype of their newly developed integrated PV-sound barrier concepts. The main goal was to develop highly integrated concepts, allowing the reduction of PV sound barrier systems costs, as well as the demonstration of specific concepts for different noise situations. This project is strongly correlated with a German project. Three of the concepts of the competition are demonstrated along a highway near Munich, constructed in 1997. The three Swiss installations had to be constructed at different locations, reflecting three typical situations for sound barriers. The first Swiss installation was the world first Bi-facial PV-sound barrier. It was built on a highway bridge at Wallisellen-Aubrugg in 1997. The operational experience of the installation is positive. But due to the different efficiencies of the two cell sides, its specific yield lies somewhat behind a conventional PV installation. The second Swiss plant was finished in autumn 1998. The 'zig-zag' construction is situated along the railway line at Wallisellen in a densely inhabited area with some local shadowing. Its performance and its specific yield is comparatively low due to a combination of several reasons (geometry of the concept, inverter, high module temperature, local shadows). The third installation was constructed along the motor way A1 at Bruettisellen in 1999. Its vertical panels are equipped with amorphous modules. The report show, that the performance of the system is reasonable, but the mechanical construction has to be improved. A small trial field with cells directly laminated onto the steel panel, also installed at Bruettisellen, could be the key development for this concept. This final report includes the evaluation and comparison of the monitored data in the past 24 months of operation. (author)

Welding stainless steels for structures operating at liquid helium temperature

Energy Technology Data Exchange (ETDEWEB)

Witherell, C.E.

1980-04-18

Superconducting magnets for fusion energy reactors require massive monolithic stainless steel weldments which must operate at extremely low temperatures under stresses approaching 100 ksi (700 MPa). A three-year study was conducted to determine the feasibility of producing heavy-section welds having usable levels of strength and toughness at 4.2/sup 0/K for fabrication of these structures in Type 304LN plate. Seven welding processes were evaluated. Test weldments in full-thickness plate were made under severe restraint to simulate that of actual structures. Type 316L filler metal was used for most welds. Welds deposited under some conditions and which solidify as primary austenite have exhibited intergranular embrittlement at 4.2/sup 0/K. This is believed to be associated with grain boundary metal carbides or carbonitrides precipitated during reheating of already deposited beads by subsequent passes. Weld deposits which solidify as primary delta ferrite appear immune. Through use of fully austenitic filler metals of low nitrogen content under controlled shielded metal arc welding conditions, and through use of filler metals solidifying as primary delta ferrite where only minimum residuals remain to room temperature, welds of Type 316L composition have been made with 4.2K yield strength matching that of Type 304LN plate and acceptable levels of soundness, ductility and toughness.

From EDI to Internet Commerce: The BHP Steel Experience.

Science.gov (United States)

Chan, Caroline; Swatman, Paula M. C.

2000-01-01

Discusses the issue of business-to-business electronic commerce implementation and the factors affecting it. Discusses electronic data interchange technology, describes the results of a case study of BHP Steel (Australia), and considers paradigm shifts in implementation issues related to electronic commerce that occur over time. (Author/LRW)

Commissioning and Operational Experience in Power Reactor Fuel Reprocessing Plant

Energy Technology Data Exchange (ETDEWEB)

Pradhan, S., E-mail: spradhan@barctara.gov.in [Tarapur Based Reprocessing Plant, Bhabha Atomic Research Centre, Tarapur (India)

2014-10-15

After completing design, construction, commissioning, operation and maintenance experience of the reprocessing plants at Tarapur, Mumbai and Kalpakkam a new reprocessing plant is commissioned and put into operation at BARC, Tarapur since 2011. Subsequent to construction clearance, commissioning of the plant is taken in many steps with simultaneous review by design and safety committees. In spite of vast experience, all the staff was retrained in various aspects of process and utility operations and in operation of innovative changes incorporated in the design. Operating personnel are licensed through an elaborate procedure consisting of various check lists followed by personnel interview. Commissioning systems were divided in sub-systems. Sub-systems were commissioned independently and later integrated testing was carried out. For commissioning, extreme operating conditions were identified in consultation with designers and detailed commissioning procedures were made accordingly. Commissioning was done in different conditions to ensure safety, smooth operation and maintainability. Few modifications were carried out based on commissioning experience. Technical specifications for operation of the plant are made in consultation with designers and reviewed by safety committees. Operation of the plant was carried out after successful commissioning trials with Deep Depleted Uranium (DDU). Emergency operating procedures for each design basis accident were made. Performance of various systems, subsystems are quite satisfactory and the plant has given very good capacity factor. (author)

Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

International Nuclear Information System (INIS)

James, L.A.

1977-01-01

The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

Corrosion in lithium-stainless steel thermal-convection systems

International Nuclear Information System (INIS)

Tortorelli, P.F.; DeVan, J.H.; Selle, J.E.

1980-01-01

The corrosion of types 304L and 316 austenitic stainless steel by flowing lithium was studied in thermal-convection loops operated at 500 to 650 0 C. Both weight and compositional changes were measured on specimens distributed throughout each loop and were combined with metallographic examinations to evaluate the corrosion processes. The corrosion rate and mass transfer characteristics did not significantly differ between the two austenitic stainless steels. Addition of 500 or 1700 wt ppM N to purified lithium did not increase the dissolution rate or change the attack mode of type 316 stainless steel. Adding 5 wt % Al to the lithium reduced the weight loss of this steel by a factor of 5 relative to a pure lithium-thermal-convection loop

Phase formation at bonded vanadium and stainless steel interfaces

International Nuclear Information System (INIS)

Summers, T.S.E.

1992-01-01

The interface between vanadium bonded to stainless steel was studies to determine whether a brittle phase formed during three joining operations. Inertia friction welds between V and 21-6-9 stainless steel were examined using TEM. In the as-welded condition, a continuous, polygranular intermetallic layer about 0.25 μm thick was present at the interface. This layer grew to about 50 μm thick during heat treatment at 1000 degrees C for two hours. Analysis of electron diffraction patterns confirmed that this intermetallic was the ω phase. The interface between vanadium and type 304, SANDVIK SAF 2205, and 21-6-9 stainless steel bonded by a co-extrusion process had intermetallic particles at the interface in the as-extruded condition. Heat treatment at 1000 degrees C for two hours caused these particles to grow into continuous layers in all three cases. Based on the appearance, composition and hardness of this interfacial intermetallic, it was also concluded to be ω phase. Bonding V to type 430 stainless steel by co-extrusion caused V-rich carbides to form at the interface due to the higher concentration of C in the type 430 than in the other stainless steels investigated. The carbide particles initially present grew into a continuous layer during a two-hour heat treatment at 1000 degrees C. Co-hipping 21-6-9 stainless steel tubing with V rod resulted in slightly more concentric specimens than the co-extruded ones, but a continuous layer of the ω phase formed during the hipping operation. This brittle layer could initiate failure during subsequent forming operations. The vanadium near the stainless steel interface in the co-extruded and co-hipped tubing in some cases was harder than before heat treatment. It was concluded that this hardening was due to thermal straining during cooling following heat treatment and that thermal strains might present a greater problem than seen here when longer tubes are used in actual applications

Improving the Sharing and Use of Operating Experience Among Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Llewellyn, Michael D.

1998-01-01

Effective use of operating experience is an essential and fundamental aspect of the business of improving safety and reliability of nuclear power plant. Operating experience is considered of such importance, that it is embedded as a fundamental element in the WANO mission statement: 'To maximise the safety and reliability of operation of nuclear power plants by exchanging information and encouraging communication, comparison, and emulation amongst its members'. The exchange of information on plant operating experience and lessons learned from events is at the core of our WANO mission and is an essential element of effective operating experience use. Recognizing this, WANO - AC has joined together with Canadian PHWR operators in a cooperative effort to further strengthen the sharing of the event information, and to facilitate communication of PHWR operating experience worldwide. The content of the paper is: 1. Discussion; 2. Expectation; 3. Improving use of operating experience; 4. Internalizing operating experience; 5. Summary; 6. Attachments. The three attachments deal with: - WANO event reporting guidelines; - Root cause investigation guidelines; - Example prevent events briefing sheet. The paper is completed with the five slides used in the oral presentation

Estimation of Fatigue Life of Laser Welded AISI304 Stainless Steel T-Joint Based on Experiments and Recommendations in Design Codes

DEFF Research Database (Denmark)

Lambertsen, Søren Heide; Damkilde, Lars; Kristensen, Anders Schmidt

2013-01-01

In this paper the fatigue behavior of laser welded T-joints of stainless steel AISI304 is investigated experimentally. In the fatigue experiments 36 specimens with a sheet thickness of 1 mm are exposed to one-dimensional cyclic loading. Three different types of specimens are adopted. Three groups...

Operational experience of extreme wind penetrations

Energy Technology Data Exchange (ETDEWEB)

Estanqueiro, Ana [INETI/LNEG - National Laboratory for Energy and Geology, Lisbon (Portugal); Mateus, Carlos B. [Instituto de Meteorologia, Lisboa (Portugal); Pestana, Rui [Redes Energeticas Nacionais (REN), Lisboa (Portugal)

2010-07-01

This paper reports the operational experience from the Portuguese Power System during the 2009/2010 winter months when record wind penerations were observed: the instantaneous wind power penetration peaked at 70% of consumption during no-load periods and the wind energy accounted for more than 50% of the energy consumed for a large period. The regulation measures taken by the TSO are presented in the paper, together with the additional reserves operated for added system security. Information on the overall power system behavior under such extreme long-term wind power penetrations will also be addressed. (org.)

Operational experience of the ATLAS accelerator

Energy Technology Data Exchange (ETDEWEB)

Den Hartog, P K; Bogaty, J M; Bollinger, L M; Clifft, B E; Craig, S L; Harden, R E; Markovich, P; Munson, F H; Nixon, J M; Pardo, R C; Phillips, D R; Shepard, K W; Tilbrook, I R; Zinkmann, G P [Argonne National Lab., IL (USA). Physics Div.

1990-02-01

The ATLAS accelerator consists of a HVEC model FN tandem accelerator injecting into a linac of independently-phased niobium superconducting resonators. The accelerator provides beams with masses 6 {le} A {le} 127 and with energies ranging up to 20 MeV/A for the lightest ions and 4 MeV/A for the heaviest ions. Portions of the linac have been in operation since 1978 and, over the last decade, more than 35000 h of operating experience have been accumulated. The long-term stability of niobium resonators, and their feasibility for use in heavy-ion accelerators is now well established. (orig.).

Preheat operating experiences at the FFTF

International Nuclear Information System (INIS)

Tucker, W.R.

1978-01-01

The rather extensive test program performed on the FFTF preheat control system resulted in successful sodium fill of one secondary heat transport loop on July 2, 1978. The data obtained during testing and the attendant operating experience gained resulted in some design changes and provided the information necessary to fully characterize system performance. Temperature excursions and deviations from preset limits of only a minor nature were encountered during preheat for sodium fill. The addition of the rate alarm feature was beneficial to operation of the preheat system and allowed early detection and correction of impending excursions

Operating experience on radiation reduction in the latest BWRs

International Nuclear Information System (INIS)

Ohsumi, K.; Uchida, S.; Aizawa, M.; Takagi, K.; Amano, O.; Yamashita, K.

1988-01-01

In Japan, BWR plants have been operated commercially since 1970, and the reduction of radiation exposure has been an important concern. The application of the procedure for reducing occupational exposure is incorporated in Japanese Improvement and Standardization Program for LWRs. No.2 and No.4 plants in Fukushima No.2 Nuclear Power Station were designed and constructed as the latest 1,100 MWe BWRs in conformity with the Improvement and Standardization Program. No.2 plant began the commercial operation in February, 1984, and experienced three times of the scheduled annual maintenance outage. No.4 plant began the commercial operation in September, 1987, and the first annual maintenance is scheduled from September, 1988. In this paper, discussion is focused on recent radiation reduction measures, that is the control of iron and nickel in primary coolant for reducing the radiation dose rate in primary systems, based on the experience with No.2 and No.4 plants. The design concept of a low radiation dose rate nuclear power plant, the experience on water chemistry in No.2 plant, the control of iron and nickel in No.4 plant operation and so on are reported. It is believed that these operation experiences contribute to the reduction of occupational exposure in BWR plants currently in operation and in future. (Kako, I.)

Large remote manipulator operating and maintenance experience at IEM cell

International Nuclear Information System (INIS)

Hicks, D.F.; McGuinness, P.W.

1985-01-01

The Interim Examination and Maintenance (IEM) Cell at the Fast Flux Test Facility (FFTF) has two large Electro-Mechanical Manipulators (EMM's). These manipulators are used for cell operations (processing of reactor core components) as well as general cell maintenance. From our eleven years of operation and maintenance experience with these large EMM's, we have learned many lessons concerning manipulator design. This paper describes the IEM Cell EMM design features and discusses operating and maintenance experience at the IEM Cell

Operation experience at the UWTF

International Nuclear Information System (INIS)

Ueno, Kazuhiro; Inada, Kameji; Ohmori, Kouji; Usui, Kazuya; Irinouchi, Sigenori; Asami, Makoto; Tohchi, Katsunori

2003-01-01

This report describes the operation experience on the volume reduction of metal wastes and used air filters contaminated with uranium at the Uranium contaminated Waste Treatment Facility (UWTF) in JNC Tokai Works. The UWTF consists of the metal waste treatment system and the filter-waste treatment system. The former treats metal wastes, the latter treats used air filters. Metal wastes are unpacked from drums, cut, and then compacted. Used air filters are separated into filter media and frames. Then the filter media are compacted and the frames are crushed. The operation of the UWTF was started in June 1998. The following volumes of wastes had been treated at the UWTF from the beginning of the operation to March 2003 (for about 5 years). (1) 1,524 drums of the metal wastes had been reduced to 410 drums. The volume reduction factor was 3.7. (2) 372 drums of the used air filters had been reduced to 39 drums. The volume reduction factor was 9.5. These systems have been operated without trouble for 5 years and have demonstrated to be able to reduce the volumes of the wastes to designed values. The volume reduction technologies for metal wastes and used air filters contaminated with uranium were successfully demonstrated at the UWTF. (author)

Summary of operating experience at Swedish nuclear power plants in 1984

International Nuclear Information System (INIS)

1985-01-01

The four owners on nuclear power plants in Sweden - The Swedish State Power Board, Forsmarks Kraftgrupp AB, Sydkraft AB and OKG AKTIEBOLAG - formed in 1980 the Nuclear Safety Board of the Swedish Utilities as a joint body for collaboration in safety matters. The Board participates in coordination of the safety work of the utilities and conducts its own safety projects, whereever this is more efficient than the utilities' working independently. The work of the Board shall contribute to optimizing safety in the operation of the Swedish nuclear power plants. The most important function of the Board is to collect, process and evaluate information on operational disturbances and incidents at Swedish and foreign nuclear power plants and then use the knowledge thus gained to improve the safety of the operation of the Swedish nuclear power plants (experience feedback). The work with Experience Feedback proceeds in three stages: Event follow-up, Fault analysis and Feedback of results. The Board runs a system for experience feedback (ERF). ERF is a computer-based information and communication system. ERF provides the Board with a daily update of operating experience in both Swedish and foreign nuclear power plants. Each Swedish nuclear power station supplies the ERF system with data on, among other things, operation and operational distrubances. Important experiences are thereby fed back to plant operation. Experience from foreign nuclear power stations can be of interest to the Swedish nuclear power plants. This information comes to RKS and is reviewed daily. The information that is considered relevant to Swedish plants is fed after analysis into the ERF system. Conversely, foreign nuclear power stations can obtain information from the operation of the Swedish plants. (author)

Plasma-induced surface degradation in 304 stainless steel used for TRIAM-1M limiter

International Nuclear Information System (INIS)

Tsukuda, N.; Kuramoto, E.; Tokunaga, K.; Muroga, T.; Yoshida, N.; Itoh, S.

1994-01-01

Surface degradation in a 304 stainless steel limiter of TRIAM-1M by long-pulse discharge during long period operation has been examined by means of X-ray diffraction, scanning electron microscopy and dynamical microindentation tests. Particular exfoliation and hardening of the surface of the electron drift side were observed. These result from the formation of α prime martensite induced by hydrogen in the plasma. The stability of the martensitic phase has been studied by annealing experiments on the cathodically hydrogen charged 316 stainless steel by X-ray diffraction. Both ε and α prime martensites were formed by 22 h cathodic charging. The former reverts to γ-phase and/or converts to α prime martensite below 723 K and the latter reverts to γ-phase below 923 K, repectively. ((orig.))

Evaluation of Cutting Fluids in Multiple Reaming of Stainless Steel

DEFF Research Database (Denmark)

Belluco, Walter; Zeng, Z.; De Chiffre, Leonardo

2001-01-01

subsequent reaming operations were carried out on austenitic stainless steel using high-speed-steel and solid carbide tools. The tested fluids were all significantly different from the reference fluid in at least some of the tested conditions. Significant differences down to 2 percent in cutting forces and 6...

Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

International Nuclear Information System (INIS)

Wu, Sujun; Jin, Huijin; Sun, Yanbin; Cao, Luowei

2014-01-01

The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σ F , of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors. - Highlights: • Critical cleavage fracture stress was calculated through experiments and FEM. • Effects of both grain and carbide particle sizes on σ F were discussed. • The SA508 Gr.4N steel is a better candidate for the next generation nuclear reactors

Analysis of interlocking performances on non-oriented electrical steels

Science.gov (United States)

Liu, Li-Hsiang; Liu, Lee-Cheng

2018-05-01

In order to reduce energy loss in motor, applications of high-efficiency non-oriented electrical steel sheets and optimal laminating process are both important elements. The motor core loss deterioration is influenced by a number of factors, such as flux distribution, stress and strain, space harmonics, temperature, and short circuits between lamination. In conventional clamping method, steel sheets are laminated via interlocking or welding in general manner. The measured energy loss by welding was much larger than that by interlocking. Therefore, interlocking is well known and usually employed with benefit of easy conducting. The protuberance shapes affected the fastening strength. Generally, the intensity of rectangular type is stronger than the circular counterparts. However, the circular interlocking has better magnetic characteristics. To clarify the method effectiveness, interlocking performances regarding fastened strength and magnetic deterioration by lamination were investigated. The key parameters of protuberance shape and forming depth were designed. Precisely manufacturing operation was applied to avoid interlocking failure. Magnetic properties largely influenced by clamping method are crucial to minimizing the magnetic deterioration during laminating procedure. Several experiments for various processing conditions were undertaken, and the quantification results showed the rectangular interlocking had better fastened strength but worsened iron loss comparing with the circular arrangement. To acquire the comprehensive mechanical and electrical identities for electrical steel lamination, deliberate producing conditions regarding minimizing the magnetic deterioration should be adopted prudently.

Experience from operating germanium detectors in GERDA

Science.gov (United States)

Palioselitis, Dimitrios; GERDA Collaboration

2015-05-01

Phase I of the Germanium Detector Array (GERDA) experiment, searching for the neutrinoless double beta (0νββ) decay of 76Ge, was completed in September 2013. The most competitive half-life lower limit for the 0νββ decay of 76Ge was set (T-0ν1/2 > 2.1 · 1025 yr at 90% C.L.). GERDA operates bare Ge diodes immersed in liquid argon. During Phase I, mainly refurbished semi-coaxial high purity Ge detectors from previous experiments were used. The experience gained with handling and operating bare Ge diodes in liquid argon, as well as the stability and performance of the detectors during GERDA Phase I are presented. Thirty additional new enriched BEGe-type detectors were produced and will be used in Phase II. A subgroup of these detectors has already been used successfully in GERDA Phase I. The present paper gives an overview of the production chain of the new germanium detectors, the steps taken to minimise the exposure to cosmic radiation during manufacturing, and the first results of characterisation measurements in vacuum cryostats.

Electrochemical aspects of stainless steel behaviour in biocorrosive environment

International Nuclear Information System (INIS)

Feron, D.

1990-01-01

Electrochemical measurements have been used to evaluate and follow, to understand and control microbial induced corrosion of stainless steels. Results include seawater loop tests and laboratory-based microbiological experiments. With natural flowing seawater, impedance spectroscopy measurements have been used to evaluate and follow biofilms on stainless steel tube-electrodes. With batch cultures of single bacterial strain (Sulphate Reducing Bacteria), open-circuit potential measurements and polarization curves performed on 316 L and 430 Ti stainless steels, have shown that the corrosion behaviour of these stainless steels is mainly dependent on the sulphide content of the culture media [fr

The VEPP-2000 Collider Control System: Operational Experience

CERN Document Server

Senchenko, A I; Lysenko, A P; Rogovsky, Yu A; Shatunov, P Yu

2017-01-01

The VEPP-2000 collider was commissioned and operated successfully in 2010-2013. During the operation the facility underwent continuous updates and experience in maintenance was acquired. Strong cooperation between the staff of the accelerator complex and the developers of the control system proved effective for implementing the necessary changes in a short time.

Energy-saving by the optimization of the operation conditions in the vinylcloride lining steel pipe baking furnace. Enbirainingu kokan yakitsukero no sogyo joken saitekika ni yoru sho energy

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Y. (Kawasaki Steel Corp., Tokyo (Japan))

1991-02-04

By the investigation of the operation conditions in the small experimental furnace, the improvement of the heat pattern of the steel pipes, and the remodeling and arrangement of the furnace, the progress of the yield and the productivity, and the reduction of the fuel unit were expected. In baking of a PVC pipe with a steel pipe, defects of PVC pipes as the blister of the inner surface occurred often. Accordingly the conditions when the blister occurred were investigated in the experimental furnace and optimized operation conditions when it does not occur were established. Next the flow rate of the hot air and the temperature distribution in the actual furnace were examined and the mechanism of occurrence of the defects of the blister of the inner surface was investigated. As a result, it was proved that the deviation of the hot air flux and the intrusion of the outer air cause it, and the places where the outer air intruded were closed by steel plates. After the improvements the rate of the defects of PVC pipes of the blister of the inner surface was decreased from 5% to 0.5%. Further, by the optimization of the operation conditions the fuel unit decreased and 31% of the rate of the energy-saving was achieved. The reduction of 10,414,000 yen a year was achieved. 29 figs., 4 tabs.

Multi Objective Optimization of Weld Parameters of Boiler Steel Using Fuzzy Based Desirability Function

Directory of Open Access Journals (Sweden)

M. Satheesh

2014-01-01

Full Text Available The high pressure differential across the wall of pressure vessels is potentially dangerous and has caused many fatal accidents in the history of their development and operation. For this reason the structural integrity of weldments is critical to the performance of pressure vessels. In recent years much research has been conducted to the study of variations in welding parameters and consumables on the mechanical properties of pressure vessel steel weldments to optimize weld integrity and ensure pressure vessels are safe. The quality of weld is a very important working aspect for the manufacturing and construction industries. Because of high quality and reliability, Submerged Arc Welding (SAW is one of the chief metal joining processes employed in industry. This paper addresses the application of desirability function approach combined with fuzzy logic analysis to optimize the multiple quality characteristics (bead reinforcement, bead width, bead penetration and dilution of submerged arc welding process parameters of SA 516 Grade 70 steels(boiler steel. Experiments were conducted using Taguchi’s L27 orthogonal array with varying the weld parameters of welding current, arc voltage, welding speed and electrode stickout. By analyzing the response table and response graph of the fuzzy reasoning grade, optimal parameters were obtained. Solutions from this method can be useful for pressure vessel manufacturers and operators to search an optimal solution of welding condition.

On the adsorption-induced fatigue of structural steels in the presence of alcohols

International Nuclear Information System (INIS)

Loboiko, V.I.; Karpenko, G.V.; Vasilenko, I.I.

1976-01-01

The purpose of the work was to study he effect of anhydrous alcohols on the cyclic fatigue of steels in the absence of contact of the alcohol with atmospheric moisture during the testing process. A vacuum was created in the operating space and then the vacuum annealing was carried out in a bath with the sample and through metal vacuum connection the bath was filled with anhydrous alcohol. Studies were made on several construction steels (20Kh, 40Kh, 50Kh, and ShKh15); steels 40Kh, 50Kh, and ShKh15 were quenched from 840-860 0 C in oil and then tempered at 200 0 C (2 h), steel 20Kh was studied in the as-received state. It was shown that with increase in the carbon content of steel with a martensite structure, the decrease in strength in the presence of anhydrous alcohol was greater than in dry air. Experiments showed that anhydrous alcohol causes an adsorption decrease in the strength both of samples with preliminarily formed cracks and V-shaped stress concentrators and of smooth samples. The greatest adsorption effect of alcohols in our case, as in static fatigue, was observed in samples with cracks. A dependence was shown between the length of the carbon chain and the fatigue limit. This dependence indicates the monotone nature of the decrease in the fatigue limit with transfer from methyl to butyl and then to octyl alcohol

Aluminum and stainless steel tubes joined by simple ring and welding process

Science.gov (United States)

Townhill, A.

1967-01-01

Duranel ring is used to join aluminum and stainless steel tubing. Duranel is a bimetal made up of roll-bonded aluminum and stainless steel. This method of joining the tubing requires only two welding operations.

Duke staff tap into a fund of operating experience

International Nuclear Information System (INIS)

Anon.

1997-01-01

The benefits of the Operating Experience Database (OEDB) allows employees of the nuclear industry to access a fund of operating experience data on-line in order to help them make appropriate management decisions concerning problem solving for various plant-based problems. Recent successful uses include fire protection system design, improving compliance with technical specifications, and reducing unnecessary trips in turbines. It is anticipated that OEDB may well be used in the Ukraine and Russia in the future. (UK)

The interaction between nitride uranium and stainless steel

Science.gov (United States)

Shornikov, D. P.; Nikitin, S. N.; Tarasov, B. A.; Baranov, V. G.; Yurlova, M. S.

2016-04-01

Uranium nitride is most popular nuclear fuel for Fast Breeder Reactor New Generation. In-pile experiments at reactor BOR-60 was shown an interaction between nitride fuel and stainless steel in the range of 8-11% burn up (HA). In order to investigate this interaction has been done diffusion tests of 200 h and has been shown that the reaction occurs in the temperature range 1000-1100 ° C. UN interacted with steel in case of high pollution oxygen (1000-2000 ppm). Also has been shown to increase interaction UN with EP-823 steel in the presence of cesium. In this case the interaction layer had a thickness about 2-3 μm. Has been shown minimal interaction with new ODS steel EP-450. The interaction layer had a thickness less then 2 μm. Did not reveal the influence of tellurium and iodine increased interaction. It was show compatibility at 1000 °C between UN and EP-450 ODS steel, chrome steel, alloying aluminium and silicium.

Evolution of mechanical properties of boron/manganese 22MnB5 steel under magnetic pulse influences

International Nuclear Information System (INIS)

Falaleev, A P; Meshkov, V V; Vetrogon, A A; Shymchenko, A V

2016-01-01

The boron/manganese 22MnB5 steel can be noted as the widely used material for creation of details, which must withstand high amount of load and impact influences. The complexity and high labor input of restoration of boron steel parts leads to growing interest in the new forming technologies such as magnetic pulse forming. There is the investigation of the evolution of mechanical properties of 22MnB5 steel during the restoration by means of magnetic pulse influence and induction heating. The heating of 22MnB5 blanks to the temperature above 900 0 C was examined. The forming processes at various temperatures (800, 900 and 950 0 C) were performed during the experiments. The test measurements allowed to obtain the relationships between the strain and the operation parameters such as induced current, pulse discharge time and the operation temperature. Based on these results the assumption about usage of these parameters for control of deformation process was made. Taking into account the load distribution and the plasticity evolution during the heating process, the computer simulation was performed in order to obtain more clear strain distribution through the processed area. The measurement of hardness and the comparison with the properties evolution during hot stamping processes confirmed the obtained results. (paper)

Innovative Concrete Repairing Technique Using Post Tensioning Steel Straps

Directory of Open Access Journals (Sweden)

Ma Chau-Khun

2017-01-01

Full Text Available In this paper, innovative technique using low-cost recycled steel straps confinement to repair load-induced damaged high-strength concrete (HSC columns were studied. This paper explains the effects of repairing technique using post tensioning steel straps. A series of experimental test was carried out to investigate the stress-strain relationships of such concrete. A total of 6 HSC columns were compressed 50% of their ultimate strength, then repaired by using steel straps. The proposed repairing technique significantly improved the performance of damaged concrete columns, in both strength and ductility. It was evidenced from this study that the steel strapping confining technique is effective in repairing of damaged HSC columns but ensured reasonable operating costs.

Energy efficiency technologies in cement and steel industry

Science.gov (United States)

Zanoli, Silvia Maria; Cocchioni, Francesco; Pepe, Crescenzo

2018-02-01

In this paper, Advanced Process Control strategies aimed at energy efficiency achievement and improvement in cement and steel industry are proposed. A flexible and smart control structure constituted by several functional modules and blocks has been developed. The designed control strategy is based on Model Predictive Control techniques, formulated on linear models. Two industrial control solutions have been developed, oriented to energy efficiency and process control improvement in cement industry clinker rotary kilns (clinker production phase) and in steel industry billets reheating furnaces. Tailored customization procedures for the design of ad hoc control systems have been executed, based on the specific needs and specifications of the analysed processes. The installation of the developed controllers on cement and steel plants produced significant benefits in terms of process control which resulted in working closer to the imposed operating limits. With respect to the previous control systems, based on local controllers and/or operators manual conduction, more profitable configurations of the crucial process variables have been provided.

Experiments on the Impact of language Problems in the Multi-cultural Operation of NPPs' Emergency Operation

Energy Technology Data Exchange (ETDEWEB)

Kang, Seongkeun; Kim, Taehoon; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Ha, Jun Su [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-10-15

In 2010, The Korea Electric Power Corporation (KEPCO) was awarded a multi-billion dollar bid to construct the first nuclear power plant in Barakah, UAE. One must keep in mind however, that with technology transfer and international cooperation comes a host of potential problems arising from cultural differences such as language, everyday habitudes and workplace expectation. As of now, how problematic these potential issues may become is unknown. Of the aforementioned factors, communication is perhaps of foremost importance. We investigated UAE culture-related issues through analysis of operating experience reviews (OERs) and came to the conclusion that the language barrier needed utmost attention. Korean nuclear power plant operators will work in UAE and will operate the NPPs with operators and managers of other nationalities as well. The purpose of this paper is firstly to confirm that operators are put under mental stress, and secondly to demonstrate the decline in accuracy when they must work in English. Reducing human error is quite important to make nuclear power plants safer. As the mental workload of human operator is increased, the probability of a human error occurring also increases. It will have a negative influence on the plant’s safety. There are many factors which can potentially increase mental workload. We focused on communication problem which is a key factor of increasing mental workload because many Korean operators will work in UAE nuclear power plants and may work together with UAE operators. From these experiments we compared how performance of both Korean and UAE subjects were decreased when they use English. We designed experimental methods to be able to check this problem qualitatively and quantitatively. We analyzed four factors to find the communication problems from the experiments which are accuracy, efficiency, NASA-TLX, and brain wave. Accuracy, efficiency, brain wave are quantitative factors, and NASA-TLX is qualitative factor. To

The corrosion and protection of less carbon containing steel in subsoil

International Nuclear Information System (INIS)

Kazimov, A. M; Mamedyarova, I. F; Selimkhanova, G. G; Bskhishova, D. A; Ibragimova, S. G.

2007-01-01

Full text: The protection and corrosion resistance of steel in subsoil waters of Baku subway were investigated. Kinetic curves were drawn. The results obtained from the experiment coincide with calculated results. There have been revealed and proposed hudron and fuel oil mixture protecting steel from corrosion in subsoil waters (97.8%) for the internal surface of steel pipes

Corrosion Behavior and Oxide Film Formation of T91 Steel under Different Water Chemistry Operation Conditions

Energy Technology Data Exchange (ETDEWEB)

Zhang, D. Q.; Shi, C.; Li, J.; Gao, L. X. [Shanghai University of Electric Power, Shanghai (China); Lee, K. Y. [Dalian University of Technology, Dalian (China)

2017-02-15

The corrosion behavior of a ferritic/martensitic steel T91 exposed to an aqueous solution containing chloride and sulfate ions is investigated depending on the stimulated all-volatile treatment (AVT) and under oxygenated treatment (OT) conditions. The corrosion of T91 steel under OT condition is severe, while the corrosion under AVT condition is not. The co-existence of chloride and sulfate ions has antagonistic effect on the corrosion of T91 steel in both AVT and OT conditions. Unlike to corrosion resistance in the aqueous solution, OT pretreatment provides T91 steel lower oxidation-resistance than VAT pretreatment. From scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction (XRD) analysis, the lower corrosion resistance in the aqueous solution by VAT conditions possibly is due to the formation of pits. In addition, the lower oxidation resistance of T91 steel pretreated by OT conditions is explained as follows: the cracks formed during the immersion under OT conditions accelerated peeling-off rate of the oxide film.

Carbon steel protection in G.S. (Girlder sulfide) plants. Pressure influence on iron sulfide scales formation. Pt. 5

International Nuclear Information System (INIS)

Delfino, C.A.; Lires, O.A.; Rojo, E.A.

1987-01-01

In order to protect carbon steel towers and piping of Girlder sulfide (G.S.) experimental heavy water plants against corrosion produced by the action of aqueous solutions of hydrogen sulfide, a method, previously published, was developed. Carbon steel, exposed to saturated aqueous solutions of hydrogen sulfide, forms iron sulfide scales. In oxygen free solutions evolution of corrosion follows the sequence: mackinawite → cubic ferrous sulfide → troilite → pyrrotite → pyrite. Scales formed by pyrrotite-pyrite or pyrite are the most protective layers (these are obtained at 130 deg C, 2MPa, for periods of 14 days). Experiments, at 125 deg C and periods of 10-25 days, were performed in two different ways: 1- constant pressure operations at 0.5 and 1.1 MPa. 2- variable pressure operation between 0.3-1 MPa. In all cases pyrrotite-pyrite scales were obtained. (Author) [es

Corrosion behaviour of dissimilar welds between martensitic stainless steel and carbon steel from secondary circuit of candu npp

International Nuclear Information System (INIS)

Popa, L.; Fulger, M.; Tunaru, M.; Velciu, L.; Lazar, M.

2015-01-01

Corrosion damages of welds occur in spite of the fact that the proper base metal and filler metal have been correctly selected, industry codes and standards have been followed and welds have been realized with full weld penetration and have proper shape and contour. It is not unusual to find that, although the base metal or alloy is resistant to corrosion in a particular environment, the welded counterpart is not resistant. In secondary circuit of a Nuclear Power Station there are some components which have dissimilar welds. Our experiments were performed in chloride environmental on two types of samples: non-welded (420 martensitic steel and 52.2k carbon steel) and dissimilar welds (dissimilar metal welds: joints beetween 420 martensitic steel and 52.2k carbon steel). To evaluate corrosion susceptibility of dissimilar welds was used electrochemical method (potentiodynamic method) and metallography microscopy (microstructural analysis). The present paper follows the localized corrosion behaviour of dissimilar welds between austenitic stainless steel and carbon steel in solutions containing chloride ions. We have been evaluated the corrosion rates of samples (welded and non-welded) by electrochemically. (authors)

Undergraduate surgical nursing preparation and guided operating room experience: A quantitative analysis.

Science.gov (United States)

Foran, Paula

2016-01-01

The aim of this research was to determine if guided operating theatre experience in the undergraduate nursing curricula enhanced surgical knowledge and understanding of nursing care provided outside this specialist area in the pre- and post-operative surgical wards. Using quantitative analyses, undergraduate nurses were knowledge tested on areas of pre- and post-operative surgical nursing in their final semester of study. As much learning occurs in nurses' first year of practice, participants were re-tested again after their Graduate Nurse Program/Preceptorship year. Participants' results were compared to the model of operating room education they had participated in to determine if there was a relationship between the type of theatre education they experienced (if any) and their knowledge of surgical ward nursing. Findings revealed undergraduates nurses receiving guided operating theatre experience had a 76% pass rate compared to 56% with non-guided or no experience (p nurses achieved a 100% pass rate compared to 53% with non-guided or no experience (p research informs us that undergraduate nurses achieve greater learning about surgical ward nursing via guided operating room experience as opposed to surgical ward nursing experience alone. Copyright © 2015 Elsevier Ltd. All rights reserved.

Overview of Steel Slag Application and Utilization

Directory of Open Access Journals (Sweden)

Lim J.W.

2016-01-01

Full Text Available Significant quantities of steel slag are generated as waste material or byproduct every day from steel industries. Slag is produced from different types of furnaces with different operating conditions. Slag contains Ferrous Oxide, Calcium Oxide, Silica etc. Physical and chemical properties of slag are affected by different methods of slag solidification such as air cooled, steam, and injection of additives. Several material characterization methods, such as X-ray Diffraction (XRD, Scanned Electron Microscopy (SEM and Inductive Coupled Plasma (ICP-OES are used to determine elemental composition in the steel slag. Therefore, slags can become one of the promising materials in various applications such as in transportation industry, construction, cement production, waste water and water treatment. The various applications of steel slag indicate that it can be reused and utilized rather than being disposed to the landfill. This paper presents a review of its applications and utilization.

Microstructural stability of 21-6-9 stainless steel

International Nuclear Information System (INIS)

Krenzer, R.W.; Sanderson, E.C.

1978-01-01

Two experiments were designed to better define parameters for thermomechanical processing of 21-6-9 stainless steel. This steel is one of the nitrogen-strengthened chromium, manganese, and nickel austenitic stainless steels having mechanical properties that can be improved by a combination of plastic deformation and heat treatments. By heat-treating coupons, the time-temperature relationship of the precipitate phase, and the solutionizing, recrystallizing, and stress-relieving temperature ranges in 21-6-9 were established. Secondly, mechanical properties and microstructure as a function of percent deformation and stress-relieving temperature are reported

Design and fabrication of stainless steel components for long life of spent fuel reprocessing plants

International Nuclear Information System (INIS)

Natarajan, R.; Ramkumar, P.; Sundararaman, V.; Kamachi Mudali, U.; Baldev Raj; Shanmugam, K.

2010-01-01

Reprocessing of spent nuclear fuels based on the PUREX process is the proven process with many commercial plants operating satisfactorily worldwide. The process medium being nitric acid, austenitic stainless steel is the material of construction as it is the best commercially available material for meeting the conditions in the reprocessing plants. Because of the high radiation fields, contact maintenance of equipment and systems of these plants are very time consuming and costly unlike other chemical process plants. Though the plants constructed in the early years required extensive shut downs for replacement of equipment and systems within the first fifteen years of operation itself, development in the field of stainless steel metallurgy and fabrication techniques have made it possible to design the present day plants for an operating life period of forty years. A review of the operational experience of the PUREX process based aqueous reprocessing plants has been made in this paper and reveals that life limiting failures of equipment and systems are mainly due to corrosion while a few are due to stresses. Presently there are no standards for design specification of materials and fabrication of reprocessing plants like the nuclear power plants, where well laid down ASTM and ASME codes and standards are available which are based on the large scale operational feedbacks on pressure vessels for conventional and nuclear industries. (author)

Compatibility of graphite with a martensitic-ferritic steel, an austenitic stainless steel and a Ni-base alloy up to 1250 C

International Nuclear Information System (INIS)

Hofmann, P.

1994-08-01

To study the chemical interactions between graphite and a martensitic-ferritic steel (1.4914), an austenitic stainless steel (1.4919; AISI 316), and a Ni-base alloy (Hastelloy X) isothermal reaction experiments were performed in the temperature range between 900 and 1250 C. At higher temperatures a rapid and complete liquefaction of the components occurred as a result of eutectic interactions. The chemical interactions are diffusion-controlled processes and can be described by parabolic rate laws. The reaction behavior of the two steels is very similar. The chemical interactions of the steels with graphite are much faster above 1100 C than those for the Ni-base alloy. Below 1000 C the effect is opposite. (orig.) [de

Chemistry in water reactors: operating experience and new developments. 2 volumes

International Nuclear Information System (INIS)

1994-01-01

These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

Operational experience with the Daresbury accelerator tube

International Nuclear Information System (INIS)

Aitken, T.W.; Eastham, D.A.; Joy, T.; Leese, J.M.; Tait, N.R.S.; Thorn, R.

1986-01-01

Operational experience with the Daresbury MKI accelerator tube is reviewed with particular attention to conditioning and high voltage performance. The effects of surges and transients on the tube are described and lines of future development are discussed. (orig.)

High Cycle Fatigue of Metastable Austenitic Stainless Steels

OpenAIRE

Fargas Ribas, Gemma; Zapata Dederle, Ana Cristina; Anglada Gomila, Marcos Juan; Mateo García, Antonio Manuel

2009-01-01

Metastable austenitic stainless steels are currently used in applications where severe forming operations are required, such as automotive bodies, due to its excellent ductility. They are also gaining interest for its combination of high strength and formability after forming. The biggest disadvantage is the difficulty to predict the mechanical response, which depends heavily on the amount of martensite formed. The martensitic transformation in metastable stainless steels can b...

Experience in startup and operation of fast flux facility

International Nuclear Information System (INIS)

Moffitt, W.C.

1980-01-01

The testing program was structured to perform all testing under formal testing procedures with a test engineer as the test director and the plant operators operating the systems and equipment. This provided excellent training and experience for the operators in preparation for eventual reactor operation. Operations preparations for the testing and operation activities has consisted of academic training, formal on-the-job training including systems operation and examinations by persons with an expert knowledge on that portion of the plant, training at EBR-II and the High Temperature Sodium Facility for selected senior operators, operating procedure preparation, training on an FFTF Control Room operator training simulator, and formal written, oral and operating examinations

Operational experience of the ATLAS accelerator

International Nuclear Information System (INIS)

Den Hartog, P.K.; Bogaty, J.M.; Bollinger, L.M.

1989-01-01

The ATLAS accelerator consists of a HVEC model FN tandem accelerator injecting into a linac of independently-phased niobium superconducting resonators. The accelerator provides beams with masses from 6≤A≤127 and with energies ranging up to 20 MeV/A for the lightest ions and 4 MeV/A for the heaviest ions. Portions of the linac have been in operation since 1978 and, over the last decade, more than 35,000 hours of operating experience have been accumulated. The long-term stability of niobium resonators, and their feasibility for use in heavy-ion accelerators is now well established. 11 refs., 3 figs., 1 tab

Operational experience at the Sludge Treatment Facility

International Nuclear Information System (INIS)

Sy, D.J.

1987-01-01

The Sludge Treatment Facility (STF) at the Oak Ridge Gaseous Diffusion Plant has been in operation since April 1987. The facility was designed to encapsulate hazardous sludge wastes in a cement matrix. Fixation will allow the waste to meet or exceed applicable compressive strength and leachability requirements. Thus, the grout mixture complies with the Resource Conservation and Recovery Act (RCRA) guidelines as a nonhazardous waste. The grout mixture is based upon a recipe formulation developed after several years of waste stream characterization and formulation studies. The wastes to be treated at the STF are wastes impounded in two ponds. The ponds have a combined capacity of 4.5 million gallons of sludge. The sludge is transferred from the ponds to a 15,000-gallon capacity storage tank by the use of a dredge. The grout mixture recipe dictates the amount of sludge, cement, fly ash, and admixture required for weighing per batch. All ingredients are weighed and then transferred to a tilt or high energy mixer for mixing. The grout mixture is then transferred to 89- or 96-gallon steel drums. The drums are placed in a storage yard designed for a point source discharge from the yard

Design and operating experiences with 50MW steam generator

International Nuclear Information System (INIS)

Kawara, M.; Yamaki, H.; Kanamori, A.; Tanaka, K.; Takahashi, T.

1975-01-01

The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

Design and operating experiences with 50MW steam generator

Energy Technology Data Exchange (ETDEWEB)

Kawara, M; Yamaki, H; Kanamori, A; Tanaka, K; Takahashi, T

1975-07-01

The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

Microstructure and Mechanical Behavior of Deep Drawing DC04 Steel at Different Length Scales

OpenAIRE

Schreijäg, Simone

2013-01-01

The deformation behavior of steels is strongly influenced by their microstructure which is a result of the alloying elements and thermal treatments. In this work, the microstructure and the deformation behavior of a non-alloyed deep drawing DC04 steel was investigated. The microstructure was analyzed during heat treatment by EBSD, then microcompression experiments were performed on selected microstructural units and then bulk steel samples were mechanically tested by tensile experiments.

PFR experience of bellows operating in sodium systems

International Nuclear Information System (INIS)

Hodgson, D.

1980-01-01

Although there was little operating experience with bellows in sodium systems available during the design phase of the UK 250 MWe prototype fast reactor, bellows were extensively utilised to seal valves and to overcome the problems of differential thermal movements in several important applications. However, because of this lack of operational experience in sodium and the non-availability of design codes it was considered prudent not to install bellows in situations where failure could possibly initiate events of unpredictable proportions, or in positions where replacement following failure would be extremely difficult i.e. involving complete shutdown of the reactor over an extended period (in excess of six months) and/or necessitate removal of large quantities of primary sodium from the reactor vessel. This paper describes some of the bellows units installed in the PFR and the performance achieved during six years of reactor operation

Flooding PSA with Plant Specific Operating Experiences of Korean PWRs

International Nuclear Information System (INIS)

Choi, Sun Yeong; Yang, Joon Yull

2006-01-01

The purpose of this paper is to update the flooding PSA with Korean plant specific operating experience data and the appropriate estimation method for the flooding frequency to improve the PSA quality. The existing flooding PSA used the NPE (Nuclear Power Experience) database up to 1985 for the flooding frequency. They are all USA plant operating experiences. So an upgraded flooding frequency with Korean specific plant operation experience is required. We also propose a method of only using the PWR (Pressurized Water Reactor) data for the flooding frequency estimation in the case of the flooding area in the primary building even though the existing flooding PSA used both PWR and BWR (Boiled Water Reactor) data for all kinds of plant areas. We evaluate the CDF (Core Damage Frequency) with the modified flooding frequency and compare the results with that of the existing flooding PSA method

Lesson Learned from the Recent Operating Experience of Domestic Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Chang-Ju; Kim, Min-Chull; Koo, Bon-Hyun; Kim, Sang-Jae; Lee, Kyung-Won; Kim, Ji-Tae; Lee, Durk-Hun

2007-01-01

According to the public concerns, it seems that one of the main missions of a nuclear regulatory body is to collect operational experiences from various nuclear facilities, and to analyze their follow-up information. The extensive use of lessons learned from operating experiences to back fit safety systems, improve operator training and emergency procedures, and to focus more attention on human factors, safety culture and quality management systems are also desired. Collecting operational experiences has been mainly done regarding the incidents and major failures of components (so called 'event'), which usually demands lots of regulatory resources. This paper concentrates on new information, i.e. lesson learned from recent investigation results of domestic events which contain 5 years' experience. This information can induce many insights for improving operational safety of nuclear power plants (NPPs)

Operating experience of Fugen-HWR in Japan

Energy Technology Data Exchange (ETDEWEB)

Yoshino, F [Reactor Regulation Division, Nuclear Safety Bureau, Science and Technology Agency, Tokyo (Japan)

1991-04-01

Fugen is a 165 MWe prototype heavy water reactor which mainly uses plutonium-uranium mixed oxide (MOX) fuel. Power Reactor and Nuclear Fuel Development Corporation (PNC) has taken responsibility for the advanced thermal reactor (ATR) project, with its name 'FUGEN' taken from the Buddhist God of Mercy. The project started in October 1967, to develop and establish the technology for this new type of reactor and to clarify MOX fuel performance in the reactor. Site construction began in December 1970 at Tsuruga and the plant commenced commercial operation on March 20, 1979. Since then, Fugen has been operated successfully for more than twelve years. The plant performance and reliability of this type of reactor has been demonstrated through the operation. All these operational experiences have contributed to the establishment of the ATR technology.

Operating experience of Fugen-HWR in Japan

International Nuclear Information System (INIS)

Yoshino, F.

1991-01-01

Fugen is a 165 MWe prototype heavy water reactor which mainly uses plutonium-uranium mixed oxide (MOX) fuel. Power Reactor and Nuclear Fuel Development Corporation (PNC) has taken responsibility for the advanced thermal reactor (ATR) project, with its name 'FUGEN' taken from the Buddhist God of Mercy. The project started in October 1967, to develop and establish the technology for this new type of reactor and to clarify MOX fuel performance in the reactor. Site construction began in December 1970 at Tsuruga and the plant commenced commercial operation on March 20, 1979. Since then, Fugen has been operated successfully for more than twelve years. The plant performance and reliability of this type of reactor has been demonstrated through the operation. All these operational experiences have contributed to the establishment of the ATR technology

The National Flood Interoperability Experiment: Bridging Resesarch and Operations

Science.gov (United States)

Salas, F. R.

2015-12-01

The National Weather Service's new National Water Center, located on the University of Alabama campus in Tuscaloosa, will become the nation's hub for comprehensive water resources forecasting. In conjunction with its federal partners the US Geological Survey, Army Corps of Engineers and Federal Emergency Management Agency, the National Weather Service will operationally support both short term flood prediction and long term seasonal forecasting of water resource conditions. By summer 2016, the National Water Center will begin evaluating four streamflow data products at the scale of the NHDPlus river reaches (approximately 2.67 million). In preparation for the release of these products, from September 2014 to August 2015, the National Weather Service partnered with the Consortium of Universities for the Advancement of Hydrologic Science, Inc. to support the National Flood Interoperability Experiment which included a seven week in-residence Summer Institute in Tuscaloosa for university students interested in learning about operational hydrology and flood forecasting. As part of the experiment, 15 hour forecasts from the operational High Resolution Rapid Refresh atmospheric model were used to drive a three kilometer Noah-MP land surface model loosely coupled to a RAPID river routing model operating on the NHDPlus dataset. This workflow was run every three hours during the Summer Institute and the results were made available to those engaged to pursue a range of research topics focused on flood forecasting (e.g. reservoir operations, ensemble forecasting, probabilistic flood inundation mapping, rainfall product evaluation etc.) Although the National Flood Interoperability Experiment was finite in length, it provided a platform through which the academic community could engage federal agencies and vice versa to narrow the gap between research and operations and demonstrate how state of the art research infrastructure, models, services, datasets etc. could be utilized

Fast and Accurate Prediction of Stratified Steel Temperature During Holding Period of Ladle

Science.gov (United States)

Deodhar, Anirudh; Singh, Umesh; Shukla, Rishabh; Gautham, B. P.; Singh, Amarendra K.

2017-04-01

Thermal stratification of liquid steel in a ladle during the holding period and the teeming operation has a direct bearing on the superheat available at the caster and hence on the caster set points such as casting speed and cooling rates. The changes in the caster set points are typically carried out based on temperature measurements at the end of tundish outlet. Thermal prediction models provide advance knowledge of the influence of process and design parameters on the steel temperature at various stages. Therefore, they can be used in making accurate decisions about the caster set points in real time. However, this requires both fast and accurate thermal prediction models. In this work, we develop a surrogate model for the prediction of thermal stratification using data extracted from a set of computational fluid dynamics (CFD) simulations, pre-determined using design of experiments technique. Regression method is used for training the predictor. The model predicts the stratified temperature profile instantaneously, for a given set of process parameters such as initial steel temperature, refractory heat content, slag thickness, and holding time. More than 96 pct of the predicted values are within an error range of ±5 K (±5 °C), when compared against corresponding CFD results. Considering its accuracy and computational efficiency, the model can be extended for thermal control of casting operations. This work also sets a benchmark for developing similar thermal models for downstream processes such as tundish and caster.

Technician support for operation and maintenance of large fusion experiments: the tandem mirror experiment upgrade (TMX-U) approach

International Nuclear Information System (INIS)

Mattson, G.E.

1983-01-01

As experiments continue to grow in size and complexity, a few technicians will no longer be able to maintain and operate the complete experiment. Specialization is becoming the norm. Subsystems are becoming very large and complex, requiring a great deal of experience and training for technicians to become qualified maintenance/operation personnel. Formal in-house and off-site programs supplement on-the-job training to fulfill the qualification criteria. This paper presents the Tandem Mirror Experiment-Upgrade (TMX-U) approach to manpower staffing, some problems encountered, possible improvements, and safety considerations for the successful operation of a large experimental facility

Operational experience of the Marcoule reactors; Experience d'exploitation des reacteurs de Marcoule

Energy Technology Data Exchange (ETDEWEB)

Conte, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1963-07-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [French] Les resultats atteints apres trois ans de fonctionnement des reacteurs G-2/G-3 permettent une accumulation considerable de l'experience d'exploitation de ces reacteurs. Les principales originalites: - caisson en beton precontraint - chargement en marche - surveillance automatique des temperatures sont largement justifiees par l'exploitation actuelle. L'auteur confirme l'interet de ces solutions d'avant-garde et en tire des conclusions pour les etudes de futures centrales nucleaires. (auteur)

Comparative study in the induced corrosion by sulfate reducing microorganisms, in a stainless steel 304L sensitized and a carbon steel API X65

International Nuclear Information System (INIS)

Diaz S, A.; Gonzalez F, E.; Arganis J, C.; Luna C, P.; Carapia M, L.

2004-01-01

In spite of the operational experience related with the presence of the phenomenon of microbiological corrosion (MIC) in industrial components, it was not but until the decade of the 80 s when the nuclear industry recognized its influence in some systems of Nuclear Generating Power plants. At the moment, diverse studies that have tried to explain the generation mechanism of this phenomenon exist; however, they are even important queries that to solve, especially those related with the particularities of the affected metallic substrates. Presently work, the electrochemical behavior of samples of stainless steel AISI 304L sensitized is evaluated and the carbon steel APIX65, before the action of sulfate reducing microorganisms low the same experimental conditions; found that for the APIX65 the presence of this type of bacteria promoted the formation of a stable biofilm that allowed the maintenance of the microorganisms that damaged the material in isolated places where stings were generated; while in the AISI 304L, it was not detected damage associated to the inoculated media. The techniques of Resistance to the Polarization and Tafel Extrapolation, allowed the calculation of the speed of uniform corrosion, parameter that doesn't seem to be influenced by the presence of the microorganisms; while that noise electrochemical it distinguished in real time, the effect of the sulfate reducing in the steel APIX65. (Author)

ATLAS Strip Detector: Operational Experience and Run1-> Run2 Transition

CERN Document Server

Nagai, Koichi; The ATLAS collaboration

2014-01-01

Large hadron collider was operated very successfully during the Run1 and provided a lot of opportunities of physics studies. It currently has a consolidation work toward to the operation at $\\sqrt{s}=14 \\mathrm{TeV}$ in Run2. The ATLAS experiment has achieved excellent performance in Run1 operation, delivering remarkable physics results. The SemiConductor Tracker contributed to the precise measurement of momentum of charged particles. This paper describes the operation experience of the SemiConductor Tracker in Run1 and the preparation toward to the Run2 operation during the LS1.

Operating experience and corrective action program at Ontario Hydro Nuclear

International Nuclear Information System (INIS)

Collingwood, Barry; Turner, David

1998-01-01

This is a slide-based talk given at the COG/IAEA: 5. Technical Committee Meeting on 'Exchange of operating experience of pressurized heavy water reactors'. In the introduction there are presented the operating experience (OPEX) program of OHN, and the OPEX Program Mission, ensuring that the right information gets to the right staff at the right time. The OPEX Processes are analysed. These are: - Internal Corrective Action; - Inter-site Lesson Transfer; - External Lesson Transfer; - External Posting of OHN Events; - Internalizing Operating Experience. Steps in solving the Corrective Action Program are described: - Identify the Problem; - Notify Immediate Supervision/Manager; - Evaluate the Problem; - Correct the Problem; Monitor/Report Status. The Internal Corrective Action is then presented as a flowchart. The internalizing operating experience is presented under three aspects: - Communication; - Interface; - Training. The following items are discussed, respectively: peer meetings, department/section meetings, safety meetings, e-mail folders, newsletters and bulletin boards; work planning, pre-job briefings, supervisors' briefing cards; classroom initial and refresher (case studies), simulator, management courses. A diagram is presented showing the flow and treatment of information within OHN, centered on the weekly screening meetings. Finally, the corrective action processes are depicted in a flowchart and analysed in details

Aging of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1984-10-01

A program is being conducted to investigate the significance of in-service embrittlement of cast duplex stainless steels under light-water reactor operating conditions. The existing data are evaluated to determine the expected embrittlement of cast components during the operating lifetime of reactors and to define the objectives and scope of the investigation. This presentation describes the status of the program. Data for the metallurgical characterization of the various cast stainless steels used in the investigation are presented. Charpy impact tests on short-term aged material indicate that CF-3 stainless steels are less susceptible to embrittlement than CF-8 or CF-8M stainless steels. Microstructural characterization of cast stainless steels that were obtained from Georg Fischer Co. and aged for up to 70,000 h at 300, 350, and 400 0 C reveals the formation of four different types of precipitates that are not α'. Embrittlement of the ferrite phase is primarily due to pinning of the dislocations by two of these precipitates, designated as Type M and Type X. The ferrite phase is embrittled after approx. 8 y at 300 0 C and shows cleavage fracture. Examination of the fracture surfaces of the impact-test specimens indicates that the toughness of the long-term aged material is determined by the austenite phase. 8 figures, 3 tables

High temperature oxidation in boiler environment of chromized steel

Science.gov (United States)

Alia, F. F.; Kurniawan, T.; Asmara, Y. P.; Ani, M. H. B.; Nandiyanto, A. B. D.

2017-10-01

The demand for increasing efficiency has led to the development and construction of higher operating temperature power plant. This condition may lead to more severe thickness losses in boiler tubes due to excessive corrosion process. Hence, the research to improve the corrosion resistance of the current operated material is needed so that it can be applied for higher temperature application. In this research, the effect of chromizing process on the oxidation behaviour of T91 steel was investigated under steam condition. In order to deposit chromium, mixture of chromium (Cr) powder as master alloy, halide salt (NH4Cl) powder as activator and alumina (Al2O3) powder as inert filler were inserted into alumina retort together with the steel sample and heated inside furnace at 1050°C for ten hours under argon gas environment. Furthermore, for the oxidation process, steels were exposed at 700°C at different oxidation time (6h-24h) under steam condition. From FESEM/EDX analysis, it was found that oxidation rate of pack cemented steel was lower than the un-packed steel. These results show that Cr from chromizing process was able to become reservoir for the formation of Cr2O3 in high temperature steam oxidation, and its existence can be used for a longer oxidation time.

Operating experience at CEBAF

International Nuclear Information System (INIS)

Legg, R.

1996-01-01

CEBAF, the Continuous Electron Beam Accelerator Facility, is a 5-pass, recirculating, superconducting rf linac designed to provide exceptional beam quality at 4 GeV up to 200 μA CW. It is made up of an injector, two 400-MeV linacs, and 9 recirculation arcs having a total beamline length of more than 4.5 km. On Nov. 5, 1995, CEBAF delivered a 4 GeV, 25-μA CW electron beam to the first of 3 experimental halls and the experimental physics program was started 10 days later. Accelerator availability during the first month of the experimental run exceeded 75%. Beam properties measured in the experimental hall to date are a one sigma momentum spread of 5x10 -5 and an rms emittance of 0.2 nanometer-radians, better than design specification. CW beam has been provided from all 5 passes at 800 MeV intervals. Outstanding performance of the superconducting linacs suggests a machine energy upgrade to 6 GeV in the near term with eventual machine operation at 8-10 GeV. Results from commissioning and operations experience since the last conference are presented

Production of Austenitic Steel for the LHC Superconducting Dipole Magnets

CERN Document Server

Bertinelli, F; Komori, T; Peiro, G; Rossi, L

2006-01-01

The austenitic-steel collars are an important component of the LHC dipole magnets, operating at cryogenic temperature under high mechanical stress. The required steel, known as YUS 130S, has been specifically developed for this application by Nippon Steel Corporation (NSC), who was awarded a CERN contract in 1999 for the supply of 11 500 tonnes. In 2005 - after six years of work - the contract is being successfully completed, with final production being ensured since October 2003 by Nippon Steel & Sumikin Stainless Steel Corporation (NSSC). The paper describes the steel properties, its manufacturing and quality control process, organization of production, logistics and contract follow-up. Extensive statistics have been collected relating to mechanical, physical and technological parameters. Specific attention is dedicated to measurements of magnetic permeability performed at cryogenic temperatures by CERN, the equipment used and statistical results. Reference is also made to the resulting precision of the...

Microstructural characterization of cermet-steel interface in rock drilling tool

International Nuclear Information System (INIS)

Ybarra, L.A.C.; Molisani, A.L.; Yoshimura, H.N.

2010-01-01

Rock drilling tools basically present a WC cermet bonded to a steel shank. The interface cermet-steel plays fundamental role during drilling operation, since the fracture of this interface is the main failure mode of the tools. In this work, the microstructure of this interface in crown samples (type A), prepared in an industrial like process, was evaluated. In this process, a WC-containing powder was infiltrated with a copper alloy at 1100 deg C in a graphite mold previously mounted with a 1020 steel tube. The powder was characterized by XRD analysis and the cross-section microstructure of cermet-steel was analyzed using SEM-EDS. It was observed that Ni and small amount of Cu from cermet matrix diffused into the superficial region of the steel, and the Cu alloy dissolved and penetrated along the steel grain boundaries, resulting in good metallurgical bonding of the interface.(author)

Experience from operating germanium detectors in GERDA

International Nuclear Information System (INIS)

Palioselitis, Dimitrios

2015-01-01

Phase I of the Germanium Detector Array (GERDA) experiment, searching for the neutrinoless double beta (0νββ) decay of 76 Ge, was completed in September 2013. The most competitive half-life lower limit for the 0νββ decay of 76 Ge was set (T- 0ν 1/2 > 2.1 · 10 25 yr at 90% C.L.). GERDA operates bare Ge diodes immersed in liquid argon. During Phase I, mainly refurbished semi-coaxial high purity Ge detectors from previous experiments were used. The experience gained with handling and operating bare Ge diodes in liquid argon, as well as the stability and performance of the detectors during GERDA Phase I are presented. Thirty additional new enriched BEGe-type detectors were produced and will be used in Phase II. A subgroup of these detectors has already been used successfully in GERDA Phase I. The present paper gives an overview of the production chain of the new germanium detectors, the steps taken to minimise the exposure to cosmic radiation during manufacturing, and the first results of characterisation measurements in vacuum cryostats. (paper)

The University of Colorado OSO-8 spectrometer experiment. IV - Mission operations

Science.gov (United States)

Hansen, E. R.; Bruner, E. C., Jr.

1979-01-01

The remote operation of two high-resolution ultraviolet spectrometers on the OSO-8 satellite is discussed. Mission operations enabled scientific observers to plan observations based on current solar data, interact with the observing program using real- or near real-time data and commands, evaluate quick-look instrument data, and analyze the observations for publication. During routine operations, experiments were planned a day prior to their execution, and the data from these experiments received a day later. When a shorter turnaround was required, a real-time mode was available. Here, the real-time data and command links into the remote control center were used to evaluate experiment operation and make satellite pointing or instrument configuration changes with a 1-90 minute turnaround.

Operations experience with the NAC-1 legal weight truck cask

International Nuclear Information System (INIS)

Viebrock, J.M.; Hoffman, C.C.

1978-01-01

The first three years of operation of Nuclear Assurance Corporation's (NAC) four (4) NAC-1 Casks have demonstrated that shipments of spent fuel, fuel rods and other highly irradiated reactor components can be moved routinely by legal weight truck transport. Shipments of these materials have involved some 800,000 miles of highway travel and cask handling at some fifteen different nuclear facilities. This paper presents details on NAC's operations experience with these casks including cask description, cask handling (loading and unloading), pre-shipment testing, facility turnaround and transit times, operator exposure, transport vehicles and shipper/carrier/cask owner responsibilities, actual experience with regard to facility interfacing requirements and operational procedures. Cask and equipment utilization is discussed together with the methods used to control operation costs and to improve the economics of truck transport

Application and validation of the notch master curve in medium and high strength structural steels

Energy Technology Data Exchange (ETDEWEB)

Cicero, Sergio; Garcia, Tiberio [Universidad de Cantabria, Santander (Spain); Madrazo, Virginia [PCTCAN, Santander (Spain)

2015-10-15

This paper applies and validates the Notch master curve in two ferritic steels with medium (steel S460M) and high (steel S690Q) strength. The Notch master curve is an engineering tool that allows the fracture resistance of notched ferritic steels operating within their corresponding ductile-to-brittle transition zone to be estimated. It combines the Master curve and the Theory of critical distances in order to take into account the temperature and the notch effect respectively, assuming that both effects are independent. The results, derived from 168 fracture tests on notched specimens, demonstrate the capability of the Notch master curve for the prediction of the fracture resistance of medium and high strength ferritic steels operating within their ductile-to-brittle transition zone and containing notches.

Low-temperature nitriding of austenitic steel in a vibrofluidized bed

Science.gov (United States)

Baraz, V. R.; Grachev, S. V.

1999-11-01

The prospects for use of a vibrofluidized bed (VFB) for low-temperature nitrogen saturation of high-strength austenitic steel based on Cr-Ni-Mn (12Kh17N8G2S2MF) are considered. The positive effect of preliminary plastic deformation on the intensity of nitriding is described. The temperature and time parameters of nitriding in a VFB for strain-aging austenitic steel 12Kh17N8G2S2MF are shown to be adequate for the regimes of the final heat-treatment operation of aging. This creates the possibility of combining the operations of surface alloying and strain aging into a single cycle. This combined treatment increases substantially the resistance of the steel to cyclic loads while preserving the strength parameters. It is shown that the presented method of low-temperature nitriding in a VFB is expedient for improving the service characteristics of austenitic steel 12Kh17N8G2S2MF used for production of force springs of automobile brake systems.

Oxidation Phenomena in Advanced High Strength Steels : Modelling and Experiment

NARCIS (Netherlands)

Mao, W.

2018-01-01

Galvanized advanced high strength steels (AHSS) will be the most competitive structural material for automotive applications in the next decade. Oxidation of AHSS during the recrystalization annealing process in a continuous galvanizing line to a large extent influences the quality of zinc coating

Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

International Nuclear Information System (INIS)

Chakraborty, Pritam; Biner, S.Bulent

2015-01-01

Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

Properties of Galvanized and Galvannealed Advanced High Strength Hot Rolled Steels

Energy Technology Data Exchange (ETDEWEB)

V.Y. Guertsman; E. Essadiqi; S. Dionne; O. Dremmailova; R. Bouchard; B. Voyzelle; J. McDermid; R. Fourmentin

2008-04-01

The objectives of the project were (i) to develop the coating process information to achieve good quality coatings on 3 advanced high strength hot rolled steels while retaining target mechanical properties, (ii) to obtain precise knowledge of the behavior of these steels in the various forming operations and (iii) to establish accurate user property data in the coated conditions. Three steel substrates (HSLA, DP, TRIP) with compositions providing yield strengths in the range of 400-620 MPa were selected. Only HSLA steel was found to be suitable for galnaizing and galvannealing in the hot rolled condition.

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Removal of oil films from stainless steel tubes

Energy Technology Data Exchange (ETDEWEB)

Yan, J.F.; Saez, A.E.; Grant, C.S. [North Carolina State Univ., Raleigh, NC (United States). Dept. of Chemical Engineering

1997-01-01

The contamination of metal surfaces with oil is a widespread problem in the chemical, metalworking, and automotive industries. The main source of oil fouling comes from the process fluids in various operations. For example, in a heat exchanger, the oil contaminates the equipment surface causing a lower heat-transfer efficiency. The fouled equipment leads to increased costs due to added heat-transfer area, maintenance, energy, and production losses caused by unit downtime. The removal of oil films from the inner surface of a stainless steel tube cell using aqueous cleaning solutions was studied. The two oils used in the cleaning experiments, Sunquench 1042 and heavy mineral oil, contained P{sup 32} labeled tributyl phosphate (TBP) as a radioactive tracer. The {beta}{sup {minus}} particles emitted from the radioactive TBP were detected by a CaF{sub 2} scintillator and used as a measure of the amount of oil remaining in the tube cell. Cleaning experiments performed at different flow rates, surface treatment, and surfactant concentrations indicated that initially the oil films were removed rapidly. At the end of the experiments, the oil removal rate reduced significantly, eventually becoming negligible. The stainless steel morphology affected oil removal significantly, and the rougher tube tended to retard the oil removal. The rate and extent of the decontamination were significantly increased in the presence of sodium dodecyl sulfate, a nonionic surfactant. Experimental data were compared to a hydrodynamic model based on the removal of a liquid contaminant from a solid surface by an immiscible fluid. The model deviated from the experimental data due to the presence of instabilities at the oil-water interface.

Impingement wastage experiments with 9Cr 1Mo steel

Energy Technology Data Exchange (ETDEWEB)

Kishore, S., E-mail: skishore@igcar.gov.in [IGCAR (India); Beauchamp, François; Allou, Alexandre [CEA (France); Kumar, A. Ashok; Chandramouli, S.; Rajan, K.K. [IGCAR (India)

2016-02-15

Highlights: • Sodium heated steam generators are crucial components of fast breeder reactors. • A leak in steam generator tube will cause sodium water reaction that damages the tubes. • A collaborative study by CEA and IGCAR was conducted to quantify the extent of damage on 9Cr 1Mo tube due to a steam/water leak. • It was compared against the predictions of PROPANA code. - Abstract: Steam Generator (SG) is one of the vital components of sodium cooled fast reactor (SFR). The main safety concern with SG is a probable sodium–water reaction. In case, one of its water/steam carrying tubes leaks, water/steam gets into contact with sodium causing sodium-water reaction, which is highly exothermic and producing corrosive NaOH and hydrogen. The ejecting reaction products at high temperature, impinges upon adjacent tubes by a process called impingement wastage. It could damage one of the neighboring tubes in a short time, if the detection and protection systems are failing. IGCAR and CEA carried out a collaborative study on impingement wastage of 9Cr 1Mo steel, which is one of the candidate materials for SFR SG tubes. The studies comprise of experimental works at IGCAR and simulation works with PROPANA code at CEA. This paper brings out the data and experience gained through this cooperative work.

Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

International Nuclear Information System (INIS)

Dybeck, P.; Gustafsson, B.

1990-01-01

This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

Canadian fuel development program and recent operational experience

International Nuclear Information System (INIS)

Cox, D.S.; Kohn, E.; Lau, J.H.K.; Dicke, G.J.; Macici, N.N.; Sancton, R.W.

1995-01-01

This paper provides an overview of the current Canadian CANDU fuel R and D programs and operational experience. The details of operational experience for fuel in Canadian reactors are summarized for the period 1991-1994; excellent fuel performance has been sustained, with steady-state bundle defect rates currently as low as 0.02%. The status of introducing long 37-element bundles, and bundles with rounded bearing pads is reviewed. These minor changes in fuel design have been selectively introduced in response to operational constraints (end-plate cracking and pressure-tube fretting) at Ontario Hydro's Bruce-B and Darlington stations. The R and D programs are generating a more complete understanding of CANDU fuel behaviour, while the CANDU Owners Group (COG) Fuel Technology Program is being re-aligned to a more exclusive focus on the needs of operating stations. Technical highlights and realized benefits from the COG program are summarized. Re-organization of AECL to provide a one-company focus, with an outward looking view to new CANDU markets, has strengthened R and D in advanced fuel cycles. Progress in AECL's key fuel cycle programs is also summarized. (author)

Study on application of operating experience to new nuclear power plant

International Nuclear Information System (INIS)

Hong, Nam Pyo

1991-01-01

From the standpoint of designing the nuclear power plant, nine operating units have been designed and constructed as turn-key base by foreign Nuclear Steam Supply System (NSSS) Suppliers or as component base by foreign Architect/Engineer companies. In case of the component base project, the owner of electric company generally has merits that owner's operational experiences can be effectively incorporated from the beginning stage of design by A/E. Even though six nuclear units, Kori Units 3 and 4, Yonggwang Units 1 and 2, and Ulchin Units 1 and 2, were designed as component base by foreign A/E's, operational experience feedback from Kori Unit 1, such as design improvement and system upgrade, could not be reflected, because the design process of the following units started well ahead before Kori Unit 1 operating experience is obtained enough to reflect on future nuclear power plant design. It can be stated that foreign A/E's used their experience in designing nuclear projects on very limited basis

Behaviour of F82H mod. stainless steel in lead-bismuth under temperature gradient

Science.gov (United States)

Gómez Briceño, D.; Martín Muñoz, F. J.; Soler Crespo, L.; Esteban, F.; Torres, C.

2001-07-01

Austenitic steels can be used in a hybrid system in contact with liquid lead-bismuth eutectic if the region of operating temperatures is not beyond 400°C. For higher temperatures, martensitic steels are recommended. However, at long times, the interaction between the structural material and the eutectic leads to the dissolution of some elements of the steel (Ni, Cr and Fe, mainly) in the liquid metal. In a non-isothermal lead-bismuth loop, the material dissolution takes place at the hot leg of the circuit and, due to the mass transfer, deposition occurs at the cold leg. One of the possible ways to improve the performance of structural materials in lead-bismuth is the creation of an oxide layer. Tests have been performed in a small natural convection loop built of austenitic steel (316L) that has been operating for 3000 h. This loop contains a test area in which several samples of F82Hmod. martensitic steel have been tested at different times. A gas with an oxygen content of 10 ppm was bubbled in the hot area of the circuit during the operation time. The obtained results show that an oxide layer is formed on the samples introduced in the loop at the beginning of the operation and this layer increases with time. However, the samples introduced at different times during the loop operation, are not protected by oxide layers and present material dissolution in some cases.

Behaviour of F82H mod. stainless steel in lead-bismuth under temperature gradient

International Nuclear Information System (INIS)

Gomez Briceno, D.; Martin Munoz, F.J.; Soler Crespo, L.; Esteban, F.; Torres, C.

2001-01-01

Austenitic steels can be used in a hybrid system in contact with liquid lead-bismuth eutectic if the region of operating temperatures is not beyond 400 deg. C. For higher temperatures, martensitic steels are recommended. However, at long times, the interaction between the structural material and the eutectic leads to the dissolution of some elements of the steel (Ni, Cr and Fe, mainly) in the liquid metal. In a non-isothermal lead-bismuth loop, the material dissolution takes place at the hot leg of the circuit and, due to the mass transfer, deposition occurs at the cold leg. One of the possible ways to improve the performance of structural materials in lead-bismuth is the creation of an oxide layer. Tests have been performed in a small natural convection loop built of austenitic steel (316L) that has been operating for 3000 h. This loop contains a test area in which several samples of F82Hmod. martensitic steel have been tested at different times. A gas with an oxygen content of 10 ppm was bubbled in the hot area of the circuit during the operation time. The obtained results show that an oxide layer is formed on the samples introduced in the loop at the beginning of the operation and this layer increases with time. However, the samples introduced at different times during the loop operation, are not protected by oxide layers and present material dissolution in some cases

SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 different elements plus a stainless steel

International Nuclear Information System (INIS)

Maerker, R.E.

1976-01-01

The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables

Accelerated development of advanced steels for nuclear applications

International Nuclear Information System (INIS)

Ghoniem, N.; Zinkle, S.

2009-01-01

Significant progress has been achieved in the operational performance and radiation resistance of ferritic-martensitic steels during the past few decades. Conventional high temperature steels, such as HT-9 and 2 1/4 Cr-1Mo have evolved into super Oxide Dispersion Strengthened (ODS) steels through successive optimization to meet strict performance and radiation-resistance constraints. Such evolution was possible through a combination of experimentation, modeling and empirical information. Further development and optimization of structural steels in nuclear applications will require full utilization of the available array of sophisticated experimental techniques and multiscale computational modeling, in addition to empirical data. We present here a systematic approach to the process of optimum steel development, by linking material fabrication to thermo-mechanical properties through a physical understanding of microstructure evolution. The optimization process is based on the application of design constraints (e.g. low activation, low DBTT, low swelling, creep resistance, and weldability) to describe the required microstructures, which in turn, can be controlled through material processing techniques. Prospects for future design of optimum structural steels in nuclear applications by utilization of the hierarchy of multiscale experimental and computational strategies will be described. (author)

Feedback of safety - related operational experience: Lessons learned

Energy Technology Data Exchange (ETDEWEB)

Elias, D [Commonwealth Edison Co. (United States)

1997-09-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues.

Feedback of safety - related operational experience: Lessons learned

International Nuclear Information System (INIS)

Elias, D.

1997-01-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues

Characterization of Al-coated and Uncoated Steel Slags in Flow-through Experiments: An Approach to Evaluate the Potential Efficiency of P Sorption Materials in P Removal Structures

Science.gov (United States)

Chagas, I. S. P.; Penn, C. J.; Huang, C. H.

2017-12-01

Excessive phosphorus (P) in surface waters is one of the key drivers of eutrophication. P removal structures are an emerging technology developed to reduce excessive dissolved P in runoff and drainage water, preventing or mitigating P delivery to water systems. One of the determining factors for the success of these structures is the type of P sorption material (PSM) being used. Steel slag, a residue of the steel industry, is an example of PSM proven to be efficient in sequestering dissolved P from water. However, its P sorption capacity can significantly vary, mostly because different steel-making processes generate this PSM. Aluminum-coating is a technology aiming to improve the P sorptive qualities of steel slag. In this study, we characterized eighteen different slag samples from different plants and steel-making processes. Safety, i.e., presence of trace metals, as well as chemical and physical properties were evaluated through digestions, metal-extractions and general chemical and physical characterization (e.g.: pH, buffer index, bulk density). We conducted flow-through experiments, a dynamic sorption approach, on coated and uncoated slag samples in order to evaluate differences in P removal efficiency and the effects of Al-coating. For the Al-coating, a solution of Al2(SO4)3 at two concentrations (94.5 or 66.2 g L-1) was used to coat the slag samples. After 48 hours in contact with the solution, flow-through experiments were performed. All samples were tested with an incoming P concentration of 0.5 mg L-1. Hydraulic residence time was regulated for each steel slag sample, alternating between 9.85 minutes or 0.28 minutes. This study will provide essential information about intrinsic differences in steel slag composition and its efficiency in sequestering P from flowing waters. Moreover, we explore the effects of the Al-coating technique, which can in turn enhance P removal structures efficacy and broaden its adoption.

The operation of a single-sided linear induction motor with squirrel-cage and solid-steel reaction rails

Science.gov (United States)

Eastham, A. R.; Katz, R. M.

1980-09-01

Two test programs have been conducted to evaluate the performance of a single-sided linear induction motor with a squirrel-cage reaction rail and with a solid steel reaction rail. A 1.73-m-long six-pole stator interacted with the rails mounted on the rim of a 7.6-m-diam wheel. A 64-channel data acquisition system allowed tests to be performed over a wide range of operating conditions at speeds up to 20 m/sec. Typical test results which compare and contrast the mechanical, electrical and magnetic behavior of the SLIMs are presented. The test data are being used to assess the SLIM as an integrated suspension/propulsion system and for other transportation applications.

Solidification behavior of austenitic stainless steel filler metals

International Nuclear Information System (INIS)

David, S.A.; Goodwin, G.M.; Braski, D.N.

1980-02-01

Thermal analysis and interrupted solidification experiments on selected austenitic stainless steel filler metals provided an understanding of the solidification behavior of austenitic stainless steel welds. The sequences of phase separations found were for type 308 stainless steel filler metal, L + L + delta + L + delta + γ → γ + delta, and for type 310 stainless steel filler metal, L → L + γ → γ. In type 308 stainless steel filler metal, ferrite at room temperature was identified as either the untransformed primary delta-ferrite formed during the initial stages of solidification or the residual ferrite after Widmanstaetten austenite precipitation. Microprobe and scanning transmission electron microscope microanalyses revealed that solute extensively redistributes during the transformation of primary delta-ferrite to austenite, leading to enrichment and stabilization of ferrite by chromium. The type 310 stainless steel filler metal investigated solidifies by the primary crystallization of austenite, with the transformation going to completion at the solidus temperature. In our samples residual ferrite resulting from solute segregation was absent at the intercellular or interdendritic regions

Spanish Power Exchange Market Concepts and Operating Experience

International Nuclear Information System (INIS)

Gonzalez, J. J.; Gamito, C.

2000-01-01

On January, 1st, 1998, the Spanish Electricity Market started operations. All generators, distributors, commercialization companies, and final consumers negotiate al power exchanges either through the spot market or using bilateral contracts. The Spanish Power Exchange Market Operator (Compania Operadora del Mercado Espanol de electricidad, OMEL) is responsible for the management of the market and for the economic settlement and billing of a transactions on the Power Exchange market, and the technical operational process handled by the System Operator. This paper describes in detail the Spanish market principles and the experience gathered through the design, installation and first two years of market operation. The paper presents also the Spanish market results from January 1998 up to December 1999 indicating each specific market results and aggregate statistics. (Author)

Investigation Of Adhesion Formation In New Stainless Steel Trim Spring Operated Pressure Relief Valves

Energy Technology Data Exchange (ETDEWEB)

Gross, Robert E. [Savannah River Site (SRS), Aiken, SC (United States); Bukowski, Julia V. [Villanova University, Villanova, PA (United States); Goble, William M. [exida, Sellersville, PA (United States)

2013-04-16

Examination of proof test data for new (not previously installed) stainless steel (SS) trim spring operated pressure relief valves (SOPRV) reveals that adhesions form between the seat and disc in about 46% of all such SOPRV. The forces needed to overcome these adhesions can be sufficiently large to cause the SOPRV to fail its proof test (FPT) prior to installation. Furthermore, a significant percentage of SOPRV which are found to FPT are also found to ''fail to open'' (FTO) meaning they would not relief excess pressure in the event of an overpressure event. The cases where adhesions result in FTO or FPT appear to be confined to SOPRV with diameters < 1 in and set pressures < 150 psig and the FTO are estimated to occur in 0.31% to 2.00% of this subpopulation of SS trim SOPRV. The reliability and safety implications of these finding for end-users who do not perform pre-installation testing of SOPRV are discussed.

Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1992-09-01

This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal

Steel-fabricated butterfly valves for condenser circulating water system

International Nuclear Information System (INIS)

Kawase, Hiroshi; Yasuoka, Masahiro; Nanao, Teruaki.

1979-01-01

The steel-fabricated butterfly valves, which are large in general, and gave rubber linings inside to prevent the corrosion due to sea Water, are utilized for the condenser circulating water systems of thermal and nuclear power plants. Cast iron butterfly valves, having been used hitherto, have some technical irrationalities, such as corrosion prevention, the techniques for manufacturing large castings, severe thermal transient operation. On the contrary, the steel plate-fabricated butterfly valves have the following advantages; much superior characteristics in strength, rigidity and shock resistance, the streamline shape of valve plates, the narrow width between two flanges, superior execution of works for rubber lining, the perfect sealed structure, safety to vibration, light weight and easy maintenance. The structural design and the main specifications for the steel plate butterfly valves with the nominal bore from 1350 mm to 3500 mm are presented. Concerning the design criteria, the torque of operating butterfly valves and the strength of valve bodies, valve plates and valve stems are explained. The performance tests utilizing the mock-up valve were carried out for the measurements of stress distribution, the deformation of valve body, the endurance and the operating torque. In the welding standards for steel plate butterfly valves, three kinds of welded parts are classified, and the inspection method for each part is stipulated. The vibration of the valves induced by flow vortexes and cavitation is explained. (Nakai, Y.)

Study of irradiation damage structures in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hamada, Shozo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-08-01

The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs.

Study of irradiation damage structures in austenitic stainless steels

International Nuclear Information System (INIS)

Hamada, Shozo

1997-08-01

The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs

Steel story founded on coal

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

Paper reports on an iron and steel plant in New Zealand which uses non-coking subbituminuous coal to produce the sponge iron. The transport of the ironsand and the coal to the site and the operation of the kiln in which the ironsand is reduced by the coal is described.

Operating experience with nuclear power plants 2015. Pt. 1